ML13032A185

Initial Exam 2012-301 Draft RO Written Exam

ML13032A185
Person / Time
Site:	Brunswick
Issue date:	01/29/2013
From:	NRC/RGN-II
To:	Duke Energy Corp
References
50-324/12-301, 50-325/12-301
Download: ML13032A185 (193)

Text

RO Written Exam Reference Index

1. OEOP-01-UG, Users Guide, Attachment 5, Figure 3, Heat Capacity Temperature Limit

2. QEOP-Ol-UG, Users Guide, Attachment 5, Figure 5, Core Spray NPSH Limit

3. OEOP-O1-UG, Users Guide, Attachment 5, Figure 6, RHR NPSH Lmit

4. OEOP-O1-UG, Users Guide, Attachment 5, Figure 7, Pressure Suppression Pressure

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1. 201001 1 With Unit Two operating at rated power which one of the following choices completes the statements below concerning the ROD DRIFT annunciator for a selected control rod?

The ROD DR/FTannunciator will be received when an (1) reed switch position is detected with no motion signal demanded.

The appropriate action to take lAW the APP for one drifting control rod is to (2)

A. 1) odd-numbered

2) insert a manual reactor scram B. 1) odd-numbered

2) attempt to arrest the drift and latch the control rod C. 1) even-numbered

2) insert a manual reactor scram D. 1) even-numbered

2) attempt to arrest the drift and latch the control rod Answer: B K/A: 201001 Control Rod Drive Hydraulic System G2.04.50 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual. (CFR: 41.10/43.5/45.3)

RO/SRO Rating: 4.2/4.0 Objective: LOI-CLS-LP-302--B, Obj. 4 Given plant conditions, determine the required supplementary actions lAW OAOP-02, Control Rod Malfunction/Misposition.

Reference:

None Cog Level: High Explanation: The rod drift alarm is actuated when no motion is demanded and an odd numbered reed switch is detected. If a rod is drifting the APP attempts to latch the rod by giving it an insert or withdraw signal as appropriate. If more than one rod is drifting then a reactor manual scram is inserted.

Distractor Analysis:

Choice A: Plausible because the first part of the answer is correct and if more than. one rod is drifting then a reactor manual scram is required.

Choice B: Correct Answer, see explanation Choice C: Plausible because control rods are latched on even numbered positions and a reactor scram is inserted if more than one rod is drifting.

Choice D: Plausible because control rods are latched on even numbered positions and the second part of the answer is correct.

SRO Basis: N/A

IflJa 103 %laaa . Aa S Unit I APP A-OS 3-2 Page 1 of 4 ROD DRIFT CAUSE

1. Rod in uneven position due tot Leaking scram valve

- High cooling water pressure

- Failure of directional control valves

- Slow to settle due to fuel bundle channel bow

2. Malfunction in alarm circuit ACTIONS

1. Determine if affected control rod(s) is drifting or if rod(s) has scrammed using full core display, RPIS, and RWM.

2. If more than one control rod is drifting, manually scram the reactor and refer to IEDP-Ol-RSP.

3. If more than one control rod has scrammed, then perform the followingt

a. If reactor power is less than or equal to 25t, then perform the following:

(1) Manually scram reactor.

2) Refer to IEOP-Ol-RSP,

b. If reactor power is greater than 25%, and nine or more rods V have scrammed, then perform the following:

(1) Manually scram reactor.

2) Refer to IEOP-Ol-RSP.

c. If the sum of scrammed and inoperable control rods is greater than eight, then refer to Technical Specification 3.1.3 for shutdown requirements.

d. If reactor power is greater than 25% and the sum of scrammed and inoperable control rods is less than nine, then refer to OAOP-02 .0, Supplementary Actions.

4, Select the drifting rod and determine direction of drift.

S. Attempt to arrest the drift and latch rod by giving appropriate insert or withdrawal signals to the rod using RMCS controls and bypassing RWM if necessary.

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2. 203000 1 Which one of the following identifies the power supply to 2C RHR Pump?

A.E1 B. E2 C.E3 D.E4 Answer: A K/A: 203000 Residual Heat Removal /Low Pressure Coolant Injection: Injection Mode K2.01 Knowledge of electrical power supplies to the following: Pumps (CFR: 41.7)

ROISRO Rating: 3.5 / 3.5 Objective: LOI-CLS-LP-01 7-A Obj. 1 7a List the normal and emergency power sources for the following:

RHR Pumps

Reference:

None Modified question 230000_4 which was used on the 10-2 exam. Changed pumps to 2C which changed the answer and removed the second question of the power supply to the e bus.

Cog Level: Low Explanation: 2C RHR pump is a Div I pump with a power supply from El Distractor Analysis:

Choice A: Correct Answer, see explanation Choice B: Plausible because E2 is a Unit One bus that supplies power to Unit One and Unit Two loads.

RHR Pumps 1 D and 2D are supplied from this bus.

Choice C: Plausible because E3 is a Unit Two bus that supplies power to Unit One and Unit Two loads.

RHR Pumps 1A and 2A are supplied from this bus.

Choice D: Plausible because E4 is a Unit Two bus that supplies power to Unit One and Unit Two loads.

RHR Pumps lB and 2B are supplied from this bus.

SRO Basis: N/A

LA jI UNIT 2 LOW PRESSURE EGGS NOTE INJECTION FLOW PATH AND POWER SUPPLIES SHOWN LOGIC & OTIER FLOW PAThS NOT SHOWN.

SD-17 Rev. 14 Page99ofl27j

3. 205000 1 RHR Loop A is operating in the Shutdown Cooling mode of operation. It becomes desired to reduce the reactor cooldown rate.

Which one of the following identifies the action necessary to reduce the cooldown rate lAW 20P-17, Residual Heat Removal System Operating Procedure?

A. Throttle open El 1-FOO3A, HX 2A Outlet Valve.

B. Throttle closed El l-F047A, HX 2A Inlet Valve.

C. Throttle open El l-F048A, HX 2A Bypass Valve.

D. Throttle closed Ell-FO17A, Outboard Injection Valve.

Answer: C K/A: 205000 Shutdown Cooling System (RHR Shutdown Cooling Mode)

Al .10 Ability to predict and/or monitor changes in parameters associated with operating the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) controls including: Throttle valve position (CFR: 41.5 / 45.5)

RO/SRO Rating: 3,0 / 2.9 Objective: LOI-CLS-LP-017, Obj. 15 Describe how the reactor cool down rate is controlled when the RHR system is in the Shutdown Cooling mode.

Reference:

None Cog Level: Low Explanation: The procedure allows throttling closed the F003 or F068 and throttling open the F048.

Throttling open the F048 will reduce the flow through the heat exchanger thereby bypassing the water preventing it from being cooled down.

Distractor Analysis:

Choice A: Plausible because the F003 would be throttled closed to reduce the cooldown.

Choice B: Plausible because throttling this valve closed would reduce the cooldown, but it is not allowed lAW the procedure.

Choice C: Correct Answer, see explanation Choice D: Plausible because throttling this valve closed would reduce the cooldown, but it is not allowed lAW the procedure.

SRO Basis: N/A

4 s a 7 JiL awe aJbt faLJW4LeOWS&W From OP-17 Section 8.12 4 IF a lower cooldown rate is desired, THEN PERFORM the following, as necessary, for each operating RHR loop while maintaining desired flow rate, NOT to exceed 10,000 gpm per loop:

CAUTION IF El 1-FOO3A(B) is closed, THEN RHR HEAT EXCHANGER 2A(23) inlet temperature, located on E41-TR-R605, Point 1(2), is NOT a valid indication of reactor coolant temperature

a. SLOWLY THROTTLE CLOSE HX2A(2B) fl OUTLET VLV, E11-FUO3A(B), as necessary.

b. THROTTLE CLOSED HX2A(2B) SWD!SCH LI VLV El 1-PD V-FOGGA (B). as necessary. to reduce RHRSW flow rate.

c. SLOWLY THROTTLE OPEN HK 2A(28) LI BYPASS VLV, El l-FO4OA (B), as necessary, maintaining RHRflowrate greater than 4500 gpm.

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4. 205000 2 Unit Two is in Mode 3 when a loss of shutdown cooling occurs. The operating crew is performing alternate shutdown cooling with the SRVs lAW OAOP-1 5.0, Loss of Shutdown Cooling.

Reactor pressure is 205 psig above torus pressure.

Which one of the following identifies the status of cooldown rate and the operator actions required lAW OAOP-1 5.0?

Cooldown rate is (1) and the operator is required to (2)

A. (1) inadequate (2) start additional RHR pumps B. (1) excessive (2) secure the running RHR pump(s)

C. (1) inadequate (2) open an additional SRV D. (1)excessive (2) close SRVs Answer: C K/A: 205000 Shutdown Cooling System (RHR Shutdown Cooling Mode)

A2.12 -Ability to (a) predict the impacts of the following on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Inadequate sysem flow (CFR: 41.5 / 45.6)

RO/SRO Rating: 2.9/3.0 Objective: LOI-CLS-LP-302-L, Obj. 07c State the reason(s) for the following actions taken during the use of Alternate Shutdown Cooling: Opening an additional SRV if reactor pressure rises to >164 psig above Suppression Chamber pressure.

Reference:

None Cog Level: High Explanation: Loss of shutdown cooling is addressed in OAOP-1 5 which provides for multiple methods of providing decay heat removal. Using alternate shutdown cooling with the SRVs requires raising reactor water level to the steam lines and opening an SRV. A low pressure pump (preferably RHR) is then used to inject into the reactor with the coolant flowing to the torus via the open SRV. RHR pumps will deadhead at approximately 165 psig. lAW OAOP-15, if pressure reaches shutoff head, an additional SRV must be opened. This relieves vessel pressure, allowing additional flow and preventing deadheading of the RHR pumps thereby maintaining decay heat removal.

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Choice A: Plausible because high pressure will deadhead the RHR pump and result in inadequate cooling. If pumps were not deadheaded, starting an additional pump would increase flow and cooldown rate.

Choice B: Plausible because high pressure will increase the flow through the SRV. If cause of pressure increase was due to to excessive RHR pump flow, and shutoff head were not reached, then excessive cooldown could be achieved and stopping an RHR pump might be the appropriate action.

Choice C: Correct answer, see explanation.

Choice D: Plausible because high flow could increase cooldown rate. Closing the SRV would stop cooling flow.

SRO Basis: N/A AOP-1 5.0 This procedure addresses a loss of normal decay heat removal capability during shutdown conditions. The procedure provides contingencies for the following methods of decay heat removal:

RHRSW loop failure

- RHRloopfailure Condenser cooling failure

- Feed and bleed combinations

- Alternate shutdown coaling with SRVs 3.2.14 IF ALL of the above methods can NOT maintain reactor vessel coolant temperature less than 212°F, THEN INITIATE alternate shutdown cooling MAth the SRVs as follows:

ILIII1IIII1IL hL I_£ JL RHR NC RHR BID CS A CS B HIGHEST 621-FO13F B21-FO13A B21-FO13E COOLDOWN 821-FOl 3K 821-FQ13H B21-FO13B B21-FQ13L B21-F013G B21-F0i 3C 821-FOl 3G B21 -FOl 3C B21-FOI3J 521-F013D 821-FOI3J B21-F013D B21-FO13E 821-FQ13E B21-FOI3A B21-FO13A 621-FOI3F 621-FO13F 621-FOI3B B21-FO13B B21-FO13H B21-FOI3H B21-FQ13K B21-FO13K 821-FD13L B21-F013L 821-FO13C B21-FO13G B21-FOI3C B21-FO13G 621-FO13D B21-FO13J B21-FOI3D 621-FO13J LOWEST 621-FO13E B21-FO13A 821-FO13F COOLDOWN 821-FO13K 821-FO13L 621-FO13B B21-FO13H

9. PLACE the control switch for the desired SRV to OPEN. n

10. RAISE AN MAINTAIN reactor water level greater than 254 inches. El

13. IF reactor pressure can NOT be maintained less than El I 64 psig above Suppression Chamber pressure, THEN PLACE another SRV control switch to OPEN.

5. 206000 1 Unit Two is operating at rated power when the circuit breaker on 1 20V Distribution Panel 32A to the Div I Steam Leak Detection Numacs trips.

Which one of the following identifies the effect of this condition on the HPCI system?

A. E41-F002, HPCI Inboard Steam Line Isolation Valve, will isolate.

B. E41-F003, HPCI Outboard Steam Line Isolation Valve, will isolate.

C. E41-F002, HPCI inboard Steam Line Isolation Valve, area temperature isolation signal is disabled.

D. E41-F003, HPCI Outboard Steam Line Isolation Valve, area temperature isolation signal is disabled.

Answer: D K/A: 206000 High Pressure Coolant Injection System K6.10 Knowledge of the effect that a loss or malfunction of the following will have on the HIGH PRESSURE COOLANT INJECTION SYSTEM: PCIS: BWR-2,3,4 (CFR: 41.7 /45.7)

RO/SRO Rating: 3.8/4.0 Objective: LOI-CLS-LP-01 9, Obj. 03q Given plant conditions, predict how the HPCI System will respond to the following events: PCIS failure

Reference:

None Cog Level: High Explanation: The Group 4 isolation logic contains two isolation trip systems. Trip System A provides isolation signals to the Outboard steam line isolation and torus suction valve; trip system B operates the Inboard isolation valves. The NUMACs provide indication and trip channels for RWCU, HPCI and RCIC leak detection. The NUMACs are powered from 120 Vac E bus power; Div. I from 31A (32A) and Div. II from 31B (32B). On a loss of power, the NUMAC output relays for RWCU are de-energize to trip and will initiate a RWCU isolation. The HPCI/RCIC output relays are energize to trip and will not cause a HPCI or RCIC isolation on a loss of AC power.

Certain ASSD procedures and the SBO procedure requires HPCI be maintained available for safe shutdown under conditions of elevated reactor building temperature due to fire and/or loss of ventilation. The steam leak detection inputs to the isolation logic are defeated by turning off the 120 Vac power supply breaker, 31A (32A) and 31 B (32B) to the NUMAC steam leak detection modules.

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• Distractor Analysis:

Choice A: Plausible because if the HPCI isolation was de-energize to actuate (as is the RWCU system) then this would be correct. 32A provides to outboard not inboard.

Choice B: Plausible because if the HPCI isolation was de-energize to actuate (as is the RWCU system) then this would be correct.

Choice C: Plausible because if the loss was 32B then this would be correct.

Choice D: Correct Answer, see explanation.

SRO Basis: N/A The isolation system is divided into two logic systems, Logic Bus A and Logic Bus B. Logic Bus A isolates the outboard valves, E41-F003 and E41-F041, while Logic Bus B isolates the inboard valves, E41 -F002 and E4 1 -F042. A Manual Isolation pushbutton is provided on Panel P601 to permit the operator to insert an isolation should plant conditions require it. The Manual Isolation only initiates a Logic A isolation and is only in effect when a Reactor Low Level Two or a High Drywell Pressure signal is present Power supplies are as follows:

Inboard Isolation Logic 125 VDC PnL 3B(U1),4B(U2) -

Outboard Isolation Logic 125 VDC Pnl. 3A(U1 ),4A(U2)

( Leak Detection Logic Div. I 120 VAC Pnl. 31A(U1),32A(U2)

Leak Detection Logic Div. II 120 VAC Pnl. 31B(U1),32B(U2)

The Seam Leak Detection System for RWCU, HPCI and RCIC consists of four NUMAC microprocessor units located in the Control Room back panel area (additional information on the operation of the NUMAC microprocessors is in Section 3). The NUMAC5 provide indication and trip channels for RWCU, HPCI and RCIC leak detection. In addition the NUMAC output signals interface with the ERFIS and Process Computer for various calculations and monitoring of plant parameters (Le., Heat Balance). The NUMAC5 are powered from 120 Vac E bus power; Div I from 31A (32A) and Div. II from 31 B (328). On a loss of power, the NUMAC output relays for RWCU are de-energize to trip and will initiate a RWCU isolation. The HPCIIRCIC output relays are energize to trip and will not cause a HPCI or RCIC isolation on a loss of AC power Steam leak detection for the Unit 1 Main Steam system is accomplished by temperature switches located along the main steam lines. A total of 16 temperature switches monitor the main steam lines for steam leaks in the Reactor Building (MSIV pit) and Turbine Building steam tunnels. Four SD-12 Rev.1O I Pagei5of208

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6. 209001 1 A transient has occurred on Unit One causing the following plant conditions:

Drywell pressure 12 psig Reactor water level 65 inches Reactor pressure 360 psig Which one of the following choices completes the statement below?

Core Spray A Inboard Injection Valve (E21-FOO5A) is (1) , and Mm Flow Bypass Valve (E21-FO3IA) is (2)

A. (1) open (2) open B. (1) open (2) closed C. (1) closed (2) open D. (1) closed (2) closed Answer: A K/A: 209001 Low Pressure core Spray System K4.08 Knowledge of LOW PRESSURE CORE SPRAY SYSTEM design feature(s) and/or interlocks which provide for the following: Automatic system initiation (CFR: 41.7)

RO/SRO Rating: 3.8/4.0 Objective: LOI-CLS-LP-01 8, Obj. 07 Given plant conditions, determine if the Core Spray System should automatically initiate.

Reference:

None Modified question from the 07 NRC exam. (209001_i) C)

Cog Level: High Explanation: Initiation signal present due to low RPV pressure and high DW pressure. Injection valves are open <410 psig, but RPV press is above 300 psig, the shutoff head of the pump, therefore the pump will be running on mm flow.

Distractor Analysis:

Choice A: Correct Answer, see explanation Choice B: Plausible because if reactor pressure was below 300 psig Core Spray would be injecting (discharge pressure greater than shutoff head) and the minimum flow valve would be closed.

Choice C: Plausible because if reactor pressure was above 410 psig the discharge valve F005A would be closed and the minimum flow valve FO31A would be open.

Choice D: Plausible because if an initiation signal was not present the injection valve would be closed.

SRO Basis: N/A FIGURE 1:3-7 Core Spray Initiation Logic ALL i7 TC NIT:A11C)3 LOCIcc3LC)OPj K1 CO AY 2Pi 5k1 CNT 3,IR. ijr FLL CDO rC4. CWEN LCNA1. CPLN N3 a:KuP VJbS PWCLGCNIT 3 &

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Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# I KIA# 209001 K1.09 Importance Rating 3.2 Knowledge of the physical connections andlor cause- effect relationships between LOW PRESSURE CORE SPRAY SYSTEM and the following: Nuclear boiler instrumentation Proposed Question: Common 6 A feedwater line rupture has occurred on Unit Two (2) and the following conditions exist:

Drywell pressure 8.8 psig Reactor water level - 35 inches RPV pressure 500 psig Which ONE of the following describes the configuration of both loops of Core Spray?

A. No Core Spray pumps are running.

Both mm flow valves 2E21-FO3IA(B) are open.

Both inboard injection valves 2E21-FOO5A(B) are closed.

Both outboard injection valves 2E21-FOO4A(B) are open.

B. Both Core Spray pumps are running with flow to the vessel.

Both mm flow valves 2E21-FO3IA(B) are closed.

Both inboard injection valves 2E21-FOO5A(B) are open.

Both outboard injection valves 2E21-FOO4A(B) are closed.

C. Both Core Spray pumps are running.

Both mm flow valves 2E21-FO3IA(B) are open.

Both inboard injection valves 2E21-FOO5A(B) are closed.

Both outboard injection valves 2E21-FOO4A(B) are open.

D. Both Core Spray pumps are running with flow to the vessel.

Both mm flow valves 2E21-FO3IA(B) are open.

Both inboard injection valves 2E21-FOO5A(B) are open.

Both outboard injection valves 2E21-FOO4A(B) are open.

Proposed Answer: C Explanation (Optional):

A. Incorrect Core Spray will be running due to Low Level signal.

B. Incorrect No injection flow because F005 valves (not the F004) are closed and pressure is> pump shutoff head.

C. Correct Response the FO3IA(B) will be open for mm flow protection of the pump, FOO4A(B) are normally open and the FOO5A(B) are closed because the low pressure permissive of 410 psig has not been met.

D. Incorrect F005 valves are closed. Rx pressure is > pump shutoff head. Mm flow valves FO3IA(B) are open because the pumps are running with no flow to the vessel.

Per SD-18, Step 3.1.3 Core Spray Inboard and Outboard Injection Valves The Core Spray Inboard and Outboard Injection Valves, E21-FOO5B(A) and E21-FOO4B(A), respectively, have automatic and manual control functions. Normally, the valves are in the automatic mode as dictated by the Control Switches, E21 -SI B/A and S2BIA (CLOSE-AUTO-OPEN, spring return to AUTO), for each valve.

Both valves will automatically open provided the following conditions are met:

El An initiation signal present (Low reactor water level #3 or low reactor vessel pressure coincident with a high drywell pressure).

(KI0)

Low reactor pressure permissive satisfied (K20)

El 10 second start timer relay timed out (timer starts once the E-bus is energized) (K3/K4)

If the inboard (E21-FOO5BIA) valve control switch is turned to CLOSE while an initiation signal is present, the valve automatic opening function will be disabled. The manually initiated CLOSE signal will override the automatic OPEN signal thus allowing the valve to be throttled as the Operator desires. The valves automatic opening function will remain disabled until the system initiation signal is cleared (initiation logic is reset, reactor pressure increases above 410 psig, or power is lost to the associated Emergency Bus). A white light will illuminate CLOSE SIC SEALED VLV E21-F005 on P601 which indicates that the automatic opening is disabled.

Technical Reference(s): OP-I 8, 1.5.5.2 /R1 9 (Attach if not previously provided)

i!11IIIIIIIIIi11IL FIGURE 18-5 Injection Valve Interlock Logic L

KI3B SI B AUTO OPEN SHUT WHEN CIS IN OPEN

:S1B SHUT WHEN MCC CIS IN CLOSE ZK2OB FQO4B KI 4B -

RLOW - LSG OPEN SHUT WHEN OVERRIDE PRESS.

FULL CLOSED I

420 42C FOO5 çHROTTLEABLE)

INBOARD FIGURE 18-4 Minimum Flow Valve Interlock Logic CORE SPRAY FS#1 FROM FS-NOQ6B PUMP MOTOR B CLOSES WHEN FLOW BREAKER CLOSED <LOW FLOW SETPONT FS2 FROM FSN)Q6F3 CLOSES WHEN FLOW LOW FLOW SETPONT F031 B L1iJ1JW1iIIt

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Core Spray Minimum Flow Bypass Valve, E21-F03tB(A), is operated by a Flow Switch (E21-FS-N006B/A) installed on the pump discharge ilne. This valve is normally open when the associated Core Spray pump is not running. If the flow switch senses a flow greater than 1500 gpm increasing (operational value, nominal and the corresponding Core Spray pump breaker is closed, indicating that the pump is running, the associated minimum flow valve will close and remain closed as long as flow remains greater than 500 gpm decreasing. If pump flow drops below 500 gpm decreasing jjØ the pump breaker is closed, the flow switch opens the corresponding minimum flow valve.

The minimum flow valves may be operated manually with Control Switches E21-S3B/A (CLOSE-AUTO-OPEN, spring return to AUTO);

however, if the pump breaker is closed, the valve will return to the position desired by the flow instrumentation, If the Core Spray pump breaker is open, indicating that the pump is not running, the Operator may close the valve using the control switch to allow for the isolation of primary containment.

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7. 2110001 An ATWS has occurred on Unit One and reactor water level deliberately lowered lAW LPC. The following conditions exist:

Reactor Water Level maintained between LL4 and -rAF Reactor Power 9%

Reactor Pressure 960 psig SLC Tank Level 2800 gallons SLC Pumps Both operating Which one of the following choices completes the statements below?

Adequate mixing of the boron with reactor water (1) assured at this level.

Under the current conditions the time for the SLC tank to reach 0% would be approximately (2) minutes.

A. (1) is (2) 32 to 34 B. (1) is (2) 65 to 68 C. (1) is NOT (2) 32to34 D. (1) is NOT (2) 65 to 68 Answer: C K/A: 211000 Standby Liquid Control System K5.02 Knowledge of the operational implications of the following concepts as they apply to STANDBY LIQUID CONTROL SYSTEM: Chugging (as it pertains to boron mixing)

(CFR: 41.5 / 45.3)

• • • • Question to be written to address boron mixing vice chugging per discussion with Chief Examiner (Bruno Caballero) 2/27112***

RO/SRO Rating: 2.8/3.0 Objective: LOI-CLS-LP-300-E, Obj.14C Given plant conditions and the Level/Power Control Procedure, determine the following: When Boron Mixing is required after being injected into the reactor.

Reference:

None Cog Level: HIGH Explanation: From SD-05, Section 2.11:

Adequate mixing of the solution with the reactor water should occur if the solution is injected when natural circulation exists with normal reactor water level. In this case the sodium pentaborate remains in the lower plenum of the reactor until the reactor water level is raised.

The injection rate of the SLC pumps isper design 43 gpm while the PT acceptance criteria is 41.2 gpm. In the case both pumps are operating so the numbers are calculated with both numbers.

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Distractor Analysis:

Choice A: Plausible because adequate mixing would occur with forced circulation, however during an ATWS level is deliberately lowered and recirculation pumps are not operating. 32 to 34 minutes would be the injection time of both pumps.

Choice B: Plausible because adequate mixing would occur with forced circulation, however during an ATWS level is deliberately lowered and recirculation pumps are not operating. 65 to 68 minutes would be the injection time of one pump.

Choice C: Correct answer, see explanation.

Choice D: Plausible because adequate mixing has not occurred and 65 to 68 minutes would be the injection time of one pump.

SRO Basis: N/A CAN -

iEACTOR POWt-.

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LESS THAN 2 .-

NO I RCIRC PUMPS

- a 1W JIWSL WSaWW From 001-37.5, Level/Power Control Procedure Basis Document:

Tripping the recirculation pumps from high reactor power effects a prompt reduction in power. If boron injection is later required, three-dimensional model tests have demonstrated that forced recirculation need not be maintained because natural circulation flow provides adequate boron mixing. Tripping the recirculation pumps is allowed in this step because the operator will have already run the speeds back, if necessary, and the resulting changes will be significantly reduced.

Maintaining reactor water level below the normal operating range suppresses reactor power by reducing natural circulation core flow. If reactor water level is lowered to and maintained near LL-4, little if any natural circulation flow exists within the reactor vessel.

F Three-dimensional scale model tests have been conducted which confirm that little boron mixing occurs under these conditions if the boron is injected into the lower plenum. The injected boron concentrates in the lower plenum region of the reactor vessel and does not contribute to reactor shut down until in-core distribution (mixing) is achieved. When an amount of boron sufficient to shut down the reactor has been injected into the reactor vessel, mixing is accomplished by raising reactor water level thereby increasing natural circulation flow through the vessel.

From SD-OS, Standby Liquid Control System:

2.11 SLC Injection Piping (Figure 05-5)

The solution enters the reactor vessel through the SLC/core differential pressure line penetration. The penetration consists of a pipe within a pipe, with the outer pipe welded to the reactor pressure vessel nozzle, providing an annular space between the nozzle and the inner line, used for SLC injection, to minimize thermal shock to the vessel if SLC is used. The outer pipe connects to the core support plate and senses the above core plate pressure. The inner pipe penetrates (sealed) the outer pipe inside the reactor vessel and ends below the core plate. The SLC solution is injected through the inner pipe and is dispersed beneath the lower core plate.

Adequate mixing of the solution with the reactor water should occur if the solution is injected when natural circulation exists with normal reactor water level.

However, if natural circulation does not exist, during cases of extreme low water level, no mixing will occur. In this case the sodium pentaborate remains in the lower plenum of the reactor until the reactor water level is raised allowing normal natural circulation be established. In addition to providing the SLC solution injection point, this penetration provides the following functions:

24 SLC Pumps There are two full capacity triplex piston positive displacement pumps that can inject the solution into the reactor at a rate of 43 gpm at 1190 psig.

When both pumps are operating simultaneously, the flow rate is increased to approximately 86 gpm.

One pump will inject the tank contents in 58 to 81 minutes and two pumps will inject the solution into the reactor in 29 to 40 minutes.

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6.0 ACCEPTANCE CRITERIA NOTE: A condition report shall be initiated for any ott normal condition observed during testing.

This test may be considered satisfactory when the following criteria are met:

6.1 SLC solution is recirculated through the SLC pumps to the storage tank.

6.2 Pump Tests 6.21 Each Standby Liquid Control Pump develops a flowrate of greater than or equal to 41.2 gpm with a pump discharge pressure greater than or equal to 1190 psig.

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8. 212000 1 Unit Two is in an ATWS with the following plant conditions:

Reactor power 31%

Mode Switch RUN Which one of the following choices will prevent the operator from resetting RPS prior to the LEP-02, Section 3 jumper installation with the given conditions?

A. Scram discharge volume Hi Hi level RPS trip sealed in.

B. Reactor water level is controlling at the setdown setpoint.

C. IRMs upscale Hi Hi due to being inserted but not ranged up.

D. Inboard MSIV B21-F022A and Outboard MSIV B21-F028D closed.

Answer: A K/A: 212000 Reactor Protection System A4.04 Ability to manually operate and/or monitor in the control room: Bypass SCRAM instrument volume high level SCRAM signal (CFR: 41.7 /45.5 to 45.8)

RO/SRO Rating: 3.9/3.9 Objective: LOI-CLS-LP-003, Obj. 15, Describe the method to reset a scram, including the conditions that must be met.

Reference:

None Cog Level: High Explanation: SDV hi hi trip can be bypassed only if the Mode switch is in Shutdown or Refuel. Setdown is 170, 4 above scram setpoint, IRM trips bypassed in RUN, MSIV input to RPS for scram is not made up.

Distractor Analysis:

Choice A: Correct answer, see explanation.

Choice B: Plausible because reactor water level sets down to 170 following a scram to prevent overfeeding of the vessel following the scram. This setdown, however, is above the low level scram setpoint of 166 Choice C: Plausible because misranging of IRMs with the reactor critical could cause a reactor scram but the IRM scram input is bypassed with the mode switch in RUN.

Choice D: Plausible because the right combination of MSIVs closed with the mode switch in RUN would result in a reactor scram.

SRO Basis: N/A

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SD-03:

3.1.4 Scram Discharge Volume High Level (Figures 03-11, 03-12 arid 03-13)

The Scram Discharge Volume High Level initiates a Scram while adequate volume is still available to receive Scram discharge water to assure that all operable Control Rods will fully insert.

Since this Scram signal will be generated following any Scram, it must be bypassed to allow for resetting a Scram condition. A two position NORMAL/BYPASS keylock switch (S4), on Panel 603, must be in BYPASS coincident with the Reactor Mode Switch (Si) in Refuel or Shutdown in order to initiate bypass of this Scram parameter.

3.1.5 Main Steam Isolation Valves Closure, Setpoint 10 % Closed (Figures 03-11, 03-12 and 03-14)

A Scram signal is initiated when specific combinations of Main Steam Line Isolation Valves (MSIV5) close. This scram acts to mitigate the positive reactivity addition transient resulting from sudden MSIV closure at high power and precludes high power reactor operation under low pressure conditions such as may occur following a main steam line rupture while at 100% power. This scram is bypassed when the Reactor Mode Switch is in RUN.

As with the Turbine Stop Valves, two Steam Lines may be isolated

( 10% closed) without causing a reactor scram.

• Isolating one steam line does not cause a half-Scram.

• Isolating Main Steam Unes A and D or B and C does not cause a half-Scram.

• Isolating any other combination of two Main Steam Lines will cause a half-Scram.

• Isolating any combination of three Main Steam Lines will cause a Reactor Scram.

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n. aL ar tira 3.1.7 Reactor Low Water Level, Setpoint 166 Inches (Figure 03-16)

The Reactor water level trip point was chosen far enough below the normal operating level to avoid spurious Scrams but high enough above the fuel to assure that there is adequate water to account for evaporation losses and displacement of coolant following the most severe transients. Signals are from four (4) level indicating switches.

Since the backup Scram valve logic requires a full Scram signal, the logic is used to isolate the Scram Discharge Volume. The Backup Scram Solenoid logic (K21 relays) is used to by Digital Feedwater Control System (DFCS) to set the DFWLC setpoint down to 17O, transfer to Single Element Control, and bypass the Single Element Control Noise Filters following a scram. These DFCS post-SCRAM functions are intended to improve Reactor level control post-SCRAM.

These relays are also used in the Scram Reset logic.

Only one IRM per RPS Trip System may be manually bypassed due to physical arrangement of the bypass switch. The IRM5 are automatically bypassed when the Reactor Mode Switch is in Run.

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9. 215001 1 A drywell entry is being made with the reactor at 10% power and rated pressure. TIPs are currently at the indexer position for decay following a TIP core scan.

Which one of the following TIP system manipulations is required by 001-01.03, Non Routine Activities, Attachment 11, Drywell Entry Requirements, to reduce radiation levels in the drywell?

TIPs shall be relocated from the indexer position to the:

A. core bottom limit with the TIP machine mode switch in Off.

B. core bottom limit with the TIP machine mode switch in Manual.

C. in-shield position with the TIP machine mode switch in Off.

D. in-shield position with the TIP machine mode switch in Manual.

Answer: C K/A: 215001 Traversing In-Core Probe Al .01 Ability to predict and/or monitor changes in parameters associated with operating the TRAVERSING IN-CORE PROBE controls including: Radiation levels: (Not-BWRI)(CFR: 41 .5/45.5)

RO/SRO Rating: 2.8/2.9 Objective: None

Reference:

None Bank question last used on the 08 NRC exam.

Cog Level: Low Explanation: 001-01.03 attachment 11 requires that TIPs be in the stored position (in shield) in the TIP Room and a clearance on each TIP ball valve (closed) and the TIP machine Auto Manual Switch (Off)

41 IITJIL!LcY Distractor Analysis:

Choice A: Plausible because the TIP must be stored in the TIP Room, but storing the TIP in the vessel would provide shielding from the reactor vessel and internals and the biological shield, and the off position would prevent withdraw (Note this would not be acceptable as an equivalent boundary since the PCIS function would be defeated)

Choice B: Plausible because the TIP must be stored in the TIP Room, but storing the TIP in the vessel would provide shielding from the reactor vessel and internals and the biological shield, and the manual position would maintain primary containment isolation capability (Note this would not be acceptable as an equivalent boundary since the TIP could automatically retract through the drywell raising the radiation levels)

Choice C: Correct answer, see explanation Choice D: Plausible because since manual position is available on the switch and would prevent any automatic operation of the machine except for the primary containment isolation function which is already met.

SRO Basis: N/A 001-01.03 4.0 The TIPS shall be in the stored position (in shield) in the TIP Room, and a clearance placed on each TIP ball valve (Closed) and each TIP machine AUTO-MANUAL mode switch (OFF).

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10. 2150031 Which one of the following indicates the purpose for performing the IRM range 6/7 overlap determination?

IRM range 6/7 overlap determination is required to be performed in order to ensure that the IRM:

A. voltage preamplifier has properly transitioned from more noise immune to more sensitive operation.

B. voltage preamplifier has properly transitioned from more sensitive to more noise immune operation.

C. pulse height discriminator circuitry has properly transitioned from more sensitive to more noise immune operation.

D. pulse height discriminator circuitry has properly transitioned from more noise immune to more sensitive operation.

Answer: B K/A: 215003 Intermediate Range Monitor System K4.04 Knowledge of INTERMEDIATE RANGE MONITOR (IRM) SYSTEM design feature(s) and/or interlocks which provide for the following: Varying system sensitivity levels using range switches (CFR: 41.7)

RO/SRO Rating: 2.9/2.9 Objective: LOl-CLS-LP-009-A, Obj. 15e Explain the basis or precautions associated with the following:

Performing IRM range 6 to 7 overlap test.

Reference:

None Cog Level: LOW Explanation: The IRMs utilize a preamplifier to boost the output signal and routes this signal through range switches to provide indication over varying power levels. This preamplifier operates over two discreet frequency ranges. As power increases, the output signal of the detectors increases and noise increases. When range six is exceeded, the frequency of the preamplifiers is shifted higher. In order to ensure accurate indication for the control operator, the indications resulting from this frequency shift are verified to be correct by the performance of OMST-IRM25R, IRM Channels Range Correlation Adjustment.

Distractor Analysis:

Choice A: Plausible because overlap determination is performed when IRMs change from low frequency (more sensitive) to high frequency (more noise immune)(range six to seven). If ranging was from range seven to six, the opposite relationship would be true.

Choice B: Correct Answer, see explanation Choice C: Plausible because the pulse height discriminator is part of the SRM circuit that provides for the removal of the noise in the circuit.

Choice D: Plausible because the pulse height discriminator is part of the SRM circuit that provides for the removal of the noise in the circuit.

a p IW Cf Lala SI I aJaiWSStII, SRO Basis: N/A SD-09.1 2.4.1 Voltage Preamplifier (IRM)

Each IRM instrument channel utilizes a voltage preamplifier which is used in conjunction with a mean square voltage vide range monitor to measure the mean square value of a current or voltage signal over a range of three decades in each of two bandwidths.

The voltage preamplifier raises the output of the detector so mat it can overcome the noise picked up in the long cable to the control room. The voltage preamplifier

1. Provides amplification of the low level signal from the detector

2. Passes only those portions of the detector signal V necessary for proper operation of the IRM signal by isolating the HVPS and detector signal from one another.

The preamplifier is located in Cabinet P030 a chassis just outside the drywell. Relays are in cabinet P008. The preamplifier passes only one of two sets of frequencies depending on the detector range selected by the IRM range selector switch as follows:

Range Frequencies Passed 1-6 0.8-16Khz 7-10 300-600Khz The reason for selection of frequency bands is because these frequency bands exhibit the best proportionality to power. Correct overlap of IRM range 6 to 7 is verified during a reactor startup.

The lower band (0.8 16 Khz) is more sensitive. This frequency band has signals from both electron and ion collection pulses. while the upper band sees signals from only electron collection, which has a shorter (higher frequency) pulse. This lower band frequency is utilized for ranges 1-6, where neutron flux levels are relatively low.

The upper band has greater immunity to noise and develops a signal based solely on the change in frequency as a result of electron pulses from neutron interaction within the detector.

fljlrnWjfflr Si OGPM2 5.2.26 IF this is the initial startup following a refuel outage, THEN PERFORM the following:

1. OPT-14.3.1 (if OPT-14.3 was NOT completed prior to startup).

2. OMST-IRM25R, IRM Range 6 and 7 Correlation adjustments for any IRM Channel(s) with open PMT requirements prior to reaching range 7 (SR 3.3.1.1.13).

OMST-I RM25R 1.0 PURPO5E 1.1 Range correlation adjustment of IRM channels ensures high and low frequency ranges will provide similar readings when the operator switches between Range 6 and Range 7 during reactor startup or shutdown. This procedure partially fulfills the channel calibration requirements of Technical Specification Tables 3.3.1.1-i Item 1 .a, SR 3.3.1.1.13.

TEST DESCRIPTION Test ensures high and low frequency ranges of IRM channels agree to provide similar readings for operator when switching betten Range 6 and Range 7 during reactor startup or shutdown.

11. 2150041 Unit Two is in the process of a reactor startup lAW OGP-02, Approach to Criticality and Pressurization of the Reactor, following a refueling outage.

The following SRM readings are indicated:

SRM Channel A 8.0E4 SRM Channel B 7.0E4 SRM Channel C 2.0E5 SRM Channel D 6.0E5 All lRMs are on range 4.

Which one of the following identifies the expected plant response?

A. Alarm ONLY.

B. Alarm and rod block ONLY.

C. Alarm, rod block and 1/2 scram.

D. Alarm, rod block and full scram.

Answer: B K/A: 215004 Source Range Monitor System A4.03 Ability to manually operate and/or monitor in the control room: CRT displays: Plant-Specific (CFR: 41.7 / 45.5 to 45.8)

RO/SRO Rating: 2.9/2.7 Objective: LOl-CLS-LP-009-A, Obj. 05 - Describe the interrelationships between the SRM and IRM systems.

Reference:

None Cog Level: High Explanation: A non-coincidence reactor scram would occur with any SRM greater than 5.0 E5 IF shorting links are removed. SRM rod block occurs if any SRM greater than 2.0 E5. SRM rod block is bypassed if mode switch is in RUN or if all IRMs are greater than range 7.

Candidate must also recognize that the mode switch would be not be in RUN while executing OGP-02.

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Distractor Analysis:

Choice A: Plausible because rod block and upscale alarms would both occur.

Choice B: Correct Answer, see explanation Choice C: Plausible because an alarm and rod block would occur. If shorting links were removed a scram would also occur. Candidate must have knowledge of the non-coincidence scram function of SRMs without shorting links installed. Most scrams from nuclear instrumentation are divisionalized with the exception of SRMs Choice D: Plausible because an alarm and rod block would occur. If shorting links were removed a scram would also occur. Candidate must have knowledge of the non-coincidence scram function of SRMs without shorting links installed. Most scrams from nuclear instrumentation are divisionalized with the exception of SRMs SRO Basis: N/A SD-09.1 TABLE 09.1-1 INSTRUMENT AND CONTROL SETPOINTS STARTUP RANGE NEUTRON MONITORING SYSTEM NSTR,MENT CESIG4AT ON TRIP SETPOIT AJD FNCION, ADDITIONAL CONDITIONS AND AND RIP FU4CTION FNO ON COMNENTS SRM InopTrip HUFSS- 10% +/- 1% Initiaesa rodblockiftheidhwirg ccntii.cnsaie met C1-SRM-K00A-Dl wi1nin OERATE cr *ReactcrMODESWITCisri nRUN Mnuniair SRM Mdule rpluged ANY di. shnal IRM Range 8 an bssed.

SRM UPSOALEINOPIA-052-3I Nce:8.paE faldwsicnallRMsare a:eRare7 SRM Downecale Trip +/-1 .

Iniliaes a rod block ifthe fcl wng ccnd cne are met C51-SRM-KO0 AD) .Reactcc MODE SWITCI is n RUN Munciar SRM DOWNSC.cLE ANY di shnal IRM < Rangs 3 andNObipsssed.

A-O5 14 Nc.e: passed faIdisicnalIRMsare ece Rare2 SRM Refract Permissive ape :io ic InitieleE a rod block if Ihe fdwhg ccndi cns aie met:

C51-SRM-K00 A-Di Reactar MODE SVITCR Ia n. n RUN Mn unr.ia tar SRM deec1or nc: FULL N SRM RETRACT NOT ANY di shnal IRM < Range 3 an NO bypassed.

PERMITTED A25 4-3: Nc.e: paaed f at disicrraI IRMs ace Rarge 2 Nce: SRM Reract Perm esie is 3YPASSED hen te Made Sitch is in RUN. Reference draing OFP.05852-S REV F SRM Upscale Alarm 2 X 10 qps Intiises a rod block if the fclhwhg ccndi cns are met C51-SRM-KO0IA-Di U :1.3 X 103.OX 1O ReCtOrMODESWITCHiSnO1 nRUN Mn uncia tar SRM ANY dhrshnal IRM < Range 8 and j,,Qbypassed.

UPSOALEINOPA.052-3 Nce.3passea fal dWsiccral IRMeare ate Rarge 7 SRM UpscaleTrip 5 X 1D cps Full Scram if refeIing shortng hike removed C1SRMK500(AD1 S.3 X 107.X andReaGrMcde3witahisfltinRUN SRM Period 50 seccnds -10, se Anunc,iaor 3RM PERIOD IA-05 33 C51-SRM-K0O A-D1 Technral Requemen vaniai (rRr: related SRM Inab-mntatn iaTecrn cal S dfisic rela:ed hceverlrps tsed are ii lire TRMi HVPSS s the hh c tags pawer spply settg (35E-&O Vth range and tie percentages e of this vaLe. : Acan ete CSS CF pcsrwl I produ an appa-ent Irip CFaIl :rip unr.s (e. pull scram f sbiin I nlcs are rerraedae Ia RM Lpscae TrpI

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12. 2150051 The quantity of operating LPRM detectors in the flux average is 15 for APRM 1.

Which one of the following identifies the impact of this condition?

A. Rod Out Block Annunciator ONLY B. APRM Upscale Trip! mop Annunciator ONLY C. APRM Trouble and Rod Out Block Annunciators.

D. APRM Trouble and APRM Upscale Trip! lnop Annunciators.

Answer: C K/A: 215005 Average Power Range Monitor/Local Power Range Monitor System A3.04 Ability to monitor automatic operations of the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM including: Annunciator and alarm signals (CFR: 41.7 / 45.7)

RO/SRO Rating: 3.2/3.2 Objective: LOl-CLS-LP-09.6, Obj. 14 Given PRNMS settings for abnormal conditions or operation, use the Annunciator Panel Procedures (APP) to determine the probable cause(s) for the following alarms: e. A-6 3-7, APRM TROUBLE, m. A-5 2-2, ROD OUT BLOCK

Reference:

None Cog Level: Low Explanation: An APRM with too few LPRM inputs will cause a trouble alarm and rod block. In this case, there are too few detectors since there are only 15 inputs in average and a minimum of 17 inputs in average are required.

Distractor Analysis:

Choice A: Plausible because a Rod Out Block annunciator would be received but would not be the only alarm.

Choice B: Plausible APRM Upscale Trip/Inop is one alarm that could be received in association with APRM Trouble.

Choice C: Correct answer, see explanation Choice D: Plausible APRM Upscale Trip/Inop is one alarm that could be received in association with APRM Trouble.

SRO Basis: N/A

zaSfl( (LII S tL ni Unit 2 APP A-06 3-7 Page 1 of 2 APRM TROUBLE AUTO ACTIONS

1. Rod Withdrawal Block if alarm initiated by too few LPRM detectors per level or too few LPRM detectors in flux average.

2. If alarm is due to a 2/4 Logic Module RPS Voter) power supply failure, the associated RPS Trip Channel trips CAUSE

1. The quantity of operating LPRM detectors at any given reactor level is less than three.

2. The quantity of operating LPRJ4 detectors in the flux average is less than 17.

3. A 2/4 Igic Module RPS Voterl power supply failure

4. Any self-test fault.

OBSERVATIONS

1. ROD OUT BLOCK A-0S 2-2) alarm,

2. The Rod Withdrawal Permissive indicating light will be off.

3. On APR14 BARGP.APH display at P608 and PPC Displays 882-885, LPR?4s in average is less than 17, if this condition caused the alarm,

4. If a 2/4 Logic Module power supply failure caused the alarm, the following indications can be observeth

- Home APR14 for the 2/4 Logic Module (RPS Voter), which experienced the power sply , diaplays POWER SUPPLY ERROR REMOTE in its alarm summary.

- REACTOR AUTO SCRAM 515 ACE: A-OS 1-7 (2-7) alarm

( - NEUTRON MON SYS TRIP (A-OS 4-7)alarm

_sS V.L& tJt:mwaitet Unit 2 APP A-OS 2-2 Pagelof 2 ROD OUT BLOCK ATJTO ACTIONS

1. Rod withdrawal prohibited.

CAUSE

1. South SD? not drained.

2. North SD? not drained.

3. SRM downacale and any lEN is below Range 3.

4. IRM downacale and affected IRM channel is not on Range 1.

5. SEN upscale/inoperative and any IRN channel is below Range 8.

6. IRN upscale and the reactor system mode switch is not in the RUN posit ion.

7. IRN A upscale/inoperative and the reactor system mode switch is not in the RUN position.

8. SRN detector not fully inserted and log count rate is less tban or equal to 100 cps Wypassed when all IRN channels are above Range 2 or the reactor system mode switch is in the RUN position).

9. IRN B upscale/inoperative and the reactor system mode switch is not in the RUN position.

10. APR14 downscale and the reactor systen mode switch is in the RUN position.

11, APR14 TJPSCALE alarm.

12. APR14 UPSCALE TRIP/INOP alarm.

13. Less than 17 LPRM inputs to any APR14 or less than 3 LPRM5 per axial level for any APR[4.

14, REM downacale and reactor system mode switch is in the RUN position

15. RBI4 upscale/inoperative.

16. Recirc flow signal to any APR14 greater than or equal to 110%.

17. Discharge Volume Hi Water Level Trip Bypass switch in Bypass with the Reactor Systen Mode Switch in Shutdown or Refuel.

18. Reactor System Mode Switch in Refuel with a second rod selected and r another rod not full in.

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13. 2170001 Unit One is operating at rated power when Division I DC Switchboard is lost.

Which one of the following identifies the impact of this power loss on Reactor vessel level control using RCIC?

RCIC (1) automatically initiate on valid low level signal.

RCIC (2) shutdown on a valid high level signal.

A. (1) will (2) will B. (1) will (2) will NOT C. (1) will NOT (2) will D. (1) wilINOT (2) will NOT Answer: B K/A: 217000 Reactor Core Isolation Cooling System K3.01 Knowledge of the effect that a loss or malfunction of the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) will have on following: Reactor water level (CFR: 41.7 / 45.4)

RO/SRO Rating: 3.7/3.7 Objective: LOl-CLS-LP-016-A, Obj. 15e Given plant conditions, predict the RCIC System response to the following conditions: DC power failure

Reference:

None Bank question that was modified from the 07 NRC Exam. (217000_i)

Cog Level: High Explanation: The majority of the RCIC System components are powered from Division 11125/250 Vdc Electrical Distribution via MCC 1-XDB. A loss of Division I DC power will make the inboard isolation logic (Isolation Logic A) inoperative, and result in the failure of the Turbine Steam Supply Valve to automatically close on a high vessel level condition. RCIC operation would be otherwise unaffected. A loss of Division II DC power will render the RCIC System totally inoperative for normal use.

Distractor Analysis:

Choice A: Plausible because loss of Div II power would make this correct.

Choice B: Correct answer, see explanation Choice C: Plausible because this is a common misapplication of knowledge of power supplies to RCIC logic (impact of the power loss is the opposite of this).

Choice D: Plausible because RCIC will not shutdown on high water level with a loss of Div I DC power.

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SRO Basis: N/A SD-16 The RCIC Relay Logic A, which includes Isolation Logic A and one of the required high level inputs to the high vessel level closure of the RCIC Turbine Steam Supply Valve, E51-F045, is powered from 125 Vdc Distribution Panel 3A (4A). Control power to the Condensate Pump Discharge Outboard Drain Valve (E51-F005) and the Supply Drain Pot Inboard Drain Valve (E51-F025) is from 125 Vdc Distribution Panel 3A (4A). The Remote Shutdown Panel RCIC Turbine ESM Control Box is powered from 125 Vdc Distribution Panel lB (28).

A loss of Division I DC power will make the inboard isolation logic (Isolation Logic A) inoperative, and result in the failure of the Turbine Steam Supply Valve to automatically close on a high vessel level condition. RCIC operation would be otherwise unaffected.

A loss of Division II DC power v.111 render the RCIC System totally inoperative for normal use.

OAOP-39.O ATTACHMENT 2 Page lot 1 Plant Effects from Loss of DC Panel 3A(4A)

RCIC: Will not shutdown on reactor high water level, inboard isolation logic inoperable (F51-F007, -FQ31, and -F062 will not auto close). Valves E51-F005 and -F025 fail closed.

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14. 218000 1 A LOCA has occurred on Unit One resulting in the start of all low pressure ECCS pumps and rapidly lowering reactor water level.

Auto Depress Relays Energized annunciator is received and immediately clears.

Auto Depress Control Pwr Failure annunciator alarms.

Which one of the following choices completes the statement below?

The RO will be directed to open seven ADS valves (1) after the receipt of the (2) annunciator.

A. (1) immediately (2) Reactor ADS Lo Water Level (A-03 4-2)

B. (1) 83 seconds (2) Reactor ADS Lo Water Level (A-03 4-2)

C. (1) immediately (2) Reactor Low Wtr Level Initiation (A-03 6-9)

D. (1) 83 seconds (2) Reactor Low Wtr Level Initiation (A-03 6-9)

Answer: D K/A: 218000 Automatic Depressurization System G2.04.50 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual. (CFR: 41.10 /43.5/45.3)

RO/SRO Rating: 4.2/4.0 Objective: LOI-CLS-LP-020, Obj. 11 - Given plant conditions, determine if an automatic initiation of ADS should occur.

Reference:

None Cog Level: High Explanation: Low level 1 (LL1) occurs when reactor water level reaches 166 inches and Annunciator ReactorADS Lo Water Levelalarms. At 45 inches, LL3 reactor water level, Annunciator Reactor Low Wtr Level Initiation alarms. Once the above low level conditions are met, a time delay begins (83 seconds). If one Core Spray pump or one loop of RHR pumps is running ADS will initiate after the 83 second timer times out. If control power is lost, Auto Depress Control Pwr Failure will annunciate and power to actuate the ADS valves will be lost. Manual action will be required to open the ADS valves.

Distractor Analysis:

Choice A: Plausible because this alarm is needed for initiation and the student may get the level alarms backwards. The pumps are running but the logic is not made up without LL3 alarm and the 83 second timer.

Choice B: Plausible because this alarm is needed for initiation and the student may get the level alarms backwards.

Choice C: Plausible because the logic is made up with the exception of the timer.

Choice D: Correct Answer, see explanation SRO Basis: N/A Unit 1 1APP-A-03 3-2 Page 1 of 3 AUTO DEPRESS RELAYS ENERGIZED AUTO ACTIONS 1 Energizes one half of the ADS valve logic to allow opening of the seven ADS valves when tinier elapses at 83 seconds Unit 2 APP A-03 4-2 Page 1 of 1 REACTOR ADS liD WATER LEVEL AUTO ACTIONS 1 Prc.vides confirmatory low water level permissive for ADS initiation

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15. 2190001 Unit Two is operating at rated power with B Loop Suppression Pool Cooling in service with cooling maximized when a spurious LOCA signal occurs.

Which one of the following choices completes the statement below?

In order to re-establish Suppression Pool Cooling lAW the SPC hard card (1) must be restarted,and the use of n over-ride (2 be required. /

A. (1) 2BID RHR and 2B/D RHR SW pumps (2) will B. (1) 2BID RHR and 2BID RHR SW pumps (2) will NOT C. (1) ONLY the 2BID RHR SW Pumps (2) will D. (1) ONLY the 2BID RHR SW Pumps.

(2) will NOT Answer: C K/A: 219000 RHR/LPCI: Torus/Suppression Pool Cooling Mode A4.12 -Ability to manually operate and/or monitor in the control room: Suppression pool temperature (CFR: 41.7 / 45.5 to 45.8)

RO/SRO Rating: 4.1/4.1 Objective: LOI-CLS-LP-01 7, Obj. I 8o Given plant conditions, determine how the following will affect the RHR system: Loss of Coolant Accident (LOCA)

Reference:

None Modified question from 10-1 exam (295026_i 0). This question does not specifically ask which override is needed and asks if pumps were tripped. The previous question asked what override was needed to establish SPC.

Cog Level: High Explanation: With the initial conditions both RHR pumps and both RHR SW pumps will be running with the HX bypass valve (F048) full closed. Based on the initial conditions no over-rides will be required at this time for Suppression Pool Cooling lineup. Upon receipt of the LOCA initiation signal, the HX bypass valve will receive an auto open signal and the torus cooling valves will auto close. Suppression pool cooling suction valves will remain in the current alignment.

Running RHR pumps will continue to run (no loss of suction trip). RHR SW pumps will receive a trip signal. To re-establish cooling)he RHR SW pumps will have to be restarted using a LOCA override switch and the torus cooling valves will require the Think switch to be manipulated in order to re-open them.

Distractor Analysis:

Choice A - Plausible because the RHR SW pumps will trip.

Choice B - Plausible because the RHR SW pumps will trip, Initial conditions did not require use of overrides.

Choice C - Correct Answer, see explanation Choice D Plausible because the RHR SW Pumps will trip. Initial conditions did not require use of overrides SRO Basis: N/A

3.6.2 Containment Cooling/Spray Discharge Valves (Figures 17-22, 17-23. 17-24, 17-25, and 17-26)

The Suppression Pool spray and cooling valves (FO28AIB and FQ24A/B and FO27NB) and the Containment spray valves (FOl GA/B and F02IAfB), which automatically close on a LPCI initiation, can be opened manually when satisfying the Containment Spray Permissive Logic. The containment spray valve control logic requires Drywell pressure to be gçr than LpJ,g AND the reactor vessel water level inside the core shroud be above the level equivalent to 2/3 of the cores height as proven by B2i-LTM-N036-I or N037-l with a LPCI initiation signal present, OR the 213 Core Height LPCI Initiation Manual Override keylock switch placed in OVERRIDE. Positioning the Containment Spray Valve Control (THINK) switch, in MANUAL will complete the spray permissive. The manual positioning (OPEN) command to the valves originating from the associated control switch can then be executed.

With no accident signal present, F027A(B) and F028A(B) are interlocked with each other such that one of the valves must be closed in order to open the other. Torus isolation valve F028A(S) also requires both Shutdown Cooling suction valves FOO6A(B) and FOO6C(D) to be full closed prior to opening. Drywell Spray valves FO1GA(B) and FQ21A(B) have the same interlock arrangement with no accident signal as F027A(B) and FO2BA(B).

Containment-side discs for Eil-F028A(B) have been drilled with a vent hole to prevent thermally induced pressure-locking when required to open.

The suppression pool cooling valves which automatically close during a LPCI initiation can be opened manually with the following permissives satisfied: The water level inside the core shroud is above the level equivalent to 213 of the core height, as proven by B21-LTM-N036-t or N037-1 with a LPCI initiation signal present. If it is absolutely necessary to provide cooling (e.g., to lower the suppression pool temperature), the two-thirds of the core height water level inhibit and LPCI signal can be overridden by placing 2/3 core height LPCI initiation manual override keylock override switch El l-518A(B) to MANUAL OVERRIDE position. Think switch El i-517A(B) must also be activated to MANUAL before the manual positioning (OPEN) command to the valves associated control switch can be effected and executed.

SD-17 Rev. 18 Page43ofl28 IF rr nhInaar\a

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4.2.2 LPCI Initiation During Suppression Pool Cooling If an RHR LPCI initiation signal is received while in Suppression Pool cooling, the following actions will occur in addition to the normal system response:

• FIX Bypass valve F048A(B) opens automatically and cannot be closed for 3 minutes.

• Torus isolation valves F024A(B}, and 1028A(B) would all close automatically if open, and cannot be reopened until the Containment Cooling Spray permissives are satisfied.

• RHR pump in operation vould continue to run as long as no bus undervoltage condition exists.

• RHRSW pumps in operation trip and cannot be restarted unless the RI-IRSW pump LOCA Override switch is placed in Manual OVERRIDE.

SD-17 Rev. 18 Page51of128

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16. 223002 1 A LOCA has occurred on Unit Two. Subsequently a steam line leak occurs on the RCIC system. The following plant conditions are present 10 minutes after the steam line leak occurred:

Reactor water level 95 inches Drywell pressure 3.5 psig Reactor pressure 900 psig RCIC Steam Line Tunnel Ambient Temp 170°F The CRS orders the RO to manually isolate RCIC.

Which one of the following describes the effect on RCIC when the manual isolation pushbutton is depressed?

A. Inboard and Outboard Steam Supply Isolation valves F007 AND F008 close and-the cR&[S1urhine-trips.

B. ONLY the Outboard Steam Supply Isolation valve F008 closes andlhe-RGIC--turbine tnps.

C. ONLY the Inboard Steam Supply Isolation valve F007 closes and-the-RCIC tuthine-trips

( D. No effect on RCIC Isolation Valves.

Answer: B KJA: 223002 Primary Containment Isolation System /Nuclear Steam Supply Shut-Off K1 .07 Knowledge of the physical connections and/or cause effect relationships between PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF and the following:

Reactor core isolation cooling; Plant-Specific (CFR: 41.2 to 41.9 / 45.7 to 45.8)

RO/SRO Rating: 3.4/3.6 Objective: LOI-CLS-LP-016-A, Obj. 12c Given plant conditions with RCIC controlled from the RTGB, determine if the following automatic actions should occur: RCIC System isolation.

LOl-CLS-LP-016-A, Obj. 4b Describe the function of the following: Manual isolation pushbutton.

Reference:

None Bank question that was last used on the 03 NRC exam.

Cog Level: High Explanation: The Manual Isolation pushbutton is effective only if a low reactor water level signal (LL2) is present, it will close the outboard isolation valves, E51-F008 and F029. If the low reactorwater level condiUon does not exist, the outboard isolation valves will not close. There is an isolation signal present on RCIC Steam Line Tunnel Ambient Temperature but it has a 27 minute time delay when temp is 165°F and this is 10 minutes into the event.

JIZIfLJF I I4 IItIIIIJIIII Distractor Analysis:

Choice A: Plausible because only the F008 valve is affected by the Manual Isolation pushbutton while there is an initiation signal present, but the F007 is not because this manual isolation pushbutton only affects the B logic valves.

Choice B: Correct Answer, see explanation Choice C: Plausible because the F007 valve is an isolation valve but it is not not affected by the Manual Isolation pushbutton because this pushbutton only affects the B logic valves.

Choice D: Plausible because there is an isolation signal present on RCIC Steam Line Tunnel Ambient Temperature but it has a 27 minute time delay when temp is 165°F and this is 10 minutes into the event. The Isolation pushbutton has no effect on RCIC if an initiation signal is not present.

SRO Basis: N/A From SD-16 In addition to the automatic isolation, the RCIC System may be manually isolated. This manual isolation may be accomplished in two separate and independent ways; one is only applicable during RCICs automatic initiation, and the other is applicable during RCIC non-automatic operation.

The first manual isolation function is accomplished by depressing the STEAM ISOLATION pushbutton. This isolation, effective only if a low reactor water level signal (LL2) is present, closes the outboard isolation valves, E51-F008 and F029. If the low reactor water level condition does not exist, the outboard isolation valves will not close The second manual isolation function is the regular close and open feature of the isolation valves. This function is initiated through the use of the isolation valves individual control switches.

K15 SEAL-IN CIRCUIT (THIS SHEET) KSS - CLOSES SUPPRESSION POOL OUTBOARD SUCTiON VALVE (F029)

CLOSES STEAM SUPPLY OUTBOARD ISOLATIOF4 VALVE WDOS)

ENEROFZESTURBINETRIP AUXILIARY RELAY 1,

CLOSES ON ITURBINE EXHAUS DIAPHRAGM HIGH PRESSURE C-)

0-Wa, rn 0

Di r ¶ 9 1i IF-

17. 226001 1 A steam line break has occurred in the Unit Two Drywell. The CRS has directed drywell sprays be placed in service on B loop RHR lAW SEP-02, Drywell Spray Procedure. The following plant conditions exist:

Recirc Pumps Secured Drywell Coolers Secured Reactor level 100 inches Reactor pressure 400 psig Drywell pressure 15 psig The RO has momentarily placed the Containment Spray Valve Control (Think) Switch to Manual.

Which one of the following choices completes the statement below?

The Ctmt Spr Ovrd light is (1) and the minimum action(s) the operator must take to initiate Drywell Spray lAW SEP-02 is to (2)

A. (1) on (2) place the 2/3 Core Height LPCI Initiation Override switch to Manual Overrd, then open E11-FOI6B and E11-FO2IB B. (1) on (2) openE11-F016BandE11-F021Bi C. (1) off (2) place the 2/3 Core Height LPCI Initiation Override switch to Manual Overrd, then open E11-FO16B and E11-FO2IB D. (1) off (2) open E11-FO16B and E11-FO2IB Answer: B K/A: 226001 RHR/LPCI: Containment Spray System Mode K4.01 Knowledge of RHR/LPCI: CONTAINMENT SPRAY SYSTEM MODE design feature(s) and/or interlocks which provide for the following: Testability of all operable components (CFR: 41.7)

RO/SRO Rating: 2.6/2.8 Objective: LOl-CLS-LP-017, Obj. 24- Given plant conditions, determine if any of the white lights associated with the RHR system should be illuminated.

LOl-CLS-LP-017, Obj. 25 Given plant conditions, ensure that all permissives are met to spray the containment.

Reference:

None Cog Level: High

g;___

Explanation: Since a LOCA signal exists (>1.7 psig DW pressure concurrent with <410 psig reactor pressure) and reactor water level is greater than 2/3 core height, use of the 2/3 Core Height LPCI Initiation Override switch is unnecessary. In this condition, operation of the THINK switch in conjunction with DW pressure greater than 2.7 psig is sufficient to make up the spray logic and illuminate the Ctmt Spr Ovrd light. With the logic made up, the operator only has to open the spray valves El l-FOI6B and El l-F021 B.

Distractor Analysis:

Choice A: Plausible because there are two lights associated with the cooling/spray logic. If drywell pressure was below 2.7 psig, the spray light would not be illuminated but the cooling light would be illuminated. If a LOCA signal were not present or level was below 2/3 core height, the use of the 2/3 Core Height LPCI Initiation Override switch would be required.

Choice B: Correct Answer, see explanation Choice C: Plausible because there are two lights associated with the cooling/spray logic. If drywell pressure was below 2.7 psig, the spray light would not be illuminated but the cooling light would be illuminated. If a LOCA signal were not present or level was below 2/3 core height, the use of the 2/3 Core Height LPCI Initiation Override switch would be required.

Choice D: Plausible because there are two lights associated with the cooling/spray logic. If drywell pressure was below 2.7 psig, the spray light would not be illuminated but the cooling light would be illuminated.

SRO Basis: N/A SD-17

• Containment Cooling/Spray Valve Control (THINK): Two lights, one that is illuminated for Suppression Pool Cooling permissive (from K69B), the other for containment spray.

• If the Containment Spray or Suppression Pool Cooling permissive white lights are illuminated with the Think 1 switch in MANUAL, then the 2/3 Core Height permissive is satisfied and placing the 2/3 Core Height LPCI Initiation Override Switch to OVERRIDE is NOT necessary.

WL J/WV1IL F[GURE 1712 CoolingiSpry Permissive Logic I

r c1cl1I n APTER MANUAL IN POEE I THINk SWITCI4 I MAr4uAL SEAL IN)T lIUl)j,,

I 1 OR [!EIT PANEL I AL, IN) 23 CORE T if 3C4B I-IEIQHT I

ØVRb ON LPCI INITIATION IONAL T

RIiOA 1 I

L1 WT N I

IJRYWCL,L kEE I FO4FO KLIIA T CONTAINMINT PRE8S2.7P8LO I

SPRAY PERMISSIVE tF1, FII. F0211 FO21B & FO16B logic is similar for spray logic.

FIGURE 7-25 FOI6B Control Circuit 4 4II 1E4 ,iFIBFULCc En PLM4$WE I L

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)IL24 4

L RMJ4IdI!

44(j) rA(

I I ii

.RILSIx.t4 R4jL1,A4n 4f44LI- 44ITITI4 4,ZS C044T4JNkIEEIT 3PRnY PERuI4SE

s,-pi aaflflaca -r 3.7 213 Core Height LPCI Initiation Override Switch If it is absolutely necessary to provide spray or cooling to the primary containment (to lower the drywell pressure), the 2/3 core height water level and LPCI initiation inhibit can be overridden by placing the Keylocked override switch, CS-S18A(S), to MANUAL O/ERRIDE position.

3.8 Containment Spray Valve Control (THINK) Switch A Manual switch. CS-S 17A(B), that requires operator action to allow overriding the close signal sent to the containment spray and suppression pool cooling valves (FQ16A/B, FO21AJS, F027A/S, F028AB and F024A16) during a LPCI initiation.

SEP-02 RO: 2.4.6 IF necessary, THEN PLACE Loop A(S) 2/3 CORE [I HEIGHT LPCI INITIATION OVERRIDE switch, E11-CS-SI8A(SIBS), to MANUAL OVERRD.

RO: 2.4J IF the CTMT SPR OVRD light for Loop A(S) LI CONTAINMENT SPRAY VALVE CONTROL switch, E11-CS-SI7A(5178), is NOT on, THEN MOMENTARILY PLACE Loop A(S)

CONTAINMENT SPRAY VALVE CONTROL switch, El 1-CS-517A (SuB), to MANUAL.

V 1

IIIIIIIFIIIIW I1IIJf i JZ  !

18. 230000 1 Unit Two is operating at rated power when a pipe break occurs inside primary containment. A small break has also occurred in the downcomer. -

Which one of the following choices completes the statements below?

lAW PCCP, Suppression Pool Spray must be initiated before (1) pressure reaches 11 .5 psig.

Failure of Suppression Pool Spray may result in exceeding (2)

A. (1) drywell (2) RHR/Core Spray vortex limits B. (1) drywell (2) Pressure Suppression Pressure limit C. (1) suppression chamber (2) RHR/Core Spray vortex limits D. (1) suppression chamber (2) Pressure Suppression Pressure limit Answer: D K/A: 230000 RHR/LPCI: Torus/Suppression Pool Spray Mode K3.04 Knowledge of the effect that a loss or malfunction of the RHR/LPCI: TORUS/SUPPRESSION POOL SPRAY MODE will have on following: Suppression chamber air temperature (CFR: 41.7/ 45.4)

RO/SRO Rating: 3.7/3.8 Objective: LOI-CLS-LP-300-L, Obj. 4e State the effect on Primary Containment if the following limits are exceeded: Pressure Suppression Pressure Limit.

Reference:

None Cog Level: High Explanation: The Brunswick plant does not utilize suppression chamber airspace temperatures OPS,)

therefore this correlation is utilized to meet the K/A. Suppression pool spray is uicflo attempt to reverse a rising trend of primary containment pressure and is initiated before reaching 11.5 psig in the suppression chamber (PCCP PCIP-06). Pressure suppression pressure is a function of suppression pool level and suppression chamber pressure. The limit assumes no steam in the airspace. The question stem provides a crack in the downcomer in the suppression chamber area thereby bypassing the pressure suppression function of the suppression pool, introducing steam into the air space. Use of Suppression Pool sprays will mitigate this condition by condensing the steam and removing heat by evaporative and convective cooling. Loss of Suppression pool spray will result in uncondensed steam in the suppression chamber airspace, exceeding pressure suppression pressure. Rising pressure in the air space can be attributed to steam in the airspace which can be correlated to air space temperature.

Distractor Analysis:

Choice A: Plausible because drywell pressure is utilized in the same PCCP leg when determining actions for securing H2/02 monitors and for determining approach to Primary Containment Pressure Limit A. Vortex limits are to be considered but are a function of suppression pool level and pump flow, not pool or airspace temperature.

Choice B: Plausible because drywell pressure is utilized in the same PCCP leg when determining actions for securing H2/02 monitors and for determining approach to Primary Containment Pressure Limit A.

Choice C: Plausible because vortex limits are to be considered but are a function of suppression pool level and pump flow, not pool or airspace temperature.

Choice D: Correct answer, see explanation.

SRO Basis: N/A EOP-02-PCCP 1t5 5EFOE PSIG SUPPRESSION CH.i1 PRESS REACHES 115 P510 INITIATE SUPPRESSION POOL SPRAY PER EOP-CiSEPO3 ECCEPT RKRPIJMPS REQUIRED FOR ADEQUATE CORE COOLING BY CONTINUOUS OPERATION 1N LPCI MODE 001-37.8 STEP BASES:

Operation of suppression pool sprays reduces primary containment pressure by condensing steam that may be present in the suppression chamber airspace, and by absorbing heat energy from the enclosed atmosphere through the processes of evaporative and convective cooling.

OEOP-01 -UG PRESSURE SUPPRESSION PRESSURE The lesser of either (1) the highest suppression chamber pressure which can occur without steam in the suppression chamber air space or (2) the highest suppression chamber pressure at which initiation of reactor depressurization vill not result in exceeding Primary Containment Pressure Limit A before reactor pressure drops to the Minimum Reactor Flooding Pressure (MRFP), or (3) the highest suppression chamber pressure which can be maintained without exceeding the suppression chamber boundary design load if SRVs are opened.

This pressure is a function of primary containment water level, and is utilized to assure the pressure suppression function of the containment is maintained while the reactor is at pressure (Figure 7).

ILf TflI ! Ibd1I1!2!INI1!

19. 239001 1 During Unit Two power operation, a power supply loss results in a reactor scram. The operator notes the following MSIV indications immediately after the scram:

Inboard DC solenoid white light OUT Inboard AC solenoid white light LIT Outboard DC solenoid white light OUT Outboard AC solenoid white light OUT Which one of the following identifies the power supply that has been lost?

A. Division I AC Power B. Division I DC Power C. Division II AC Power D. Division II DC Power Answer: D K/A: 239001 Main and Reheat Steam System K2.01 Knowledge of electrical power supplies to the following: Main steam isolation valve solenoids (CFR: 41.7)

RO/SRO Rating: 3.2/3.3 Objective: LOl-CLS-LP-025, Obj. 5 List the power supplies (division and voltage) for the MSIV Solenoids

Reference:

None Cog Level: Low Explanation: Each MSIV operator contains two AC solenoids and one DC solenoid. One of the AC solenoids is used for valve stroke testing at power and is called the Slow Closure Test Solenoid. The other two solenoids (one AC and one DC) determine the position of the MSIV by porting or venting the pneumatic source to or from the operator. Both of these solenoids must be deenergized for the MSIV to be closed. The AC solenoids are powered from the Reactor Protection System and the DC solenoids are powered from the Station Battery System

Distractor Analysis:

Choice A: Plausible because this is a power supply to the MSIV solenoids but not the one that is de-energized for this example.

Choice B: Plausible because this is a power supply to the MSIV solenoids just but the one that is de-energized for this example.

Choice C: Plausible because this is a power supply to the MSIV solenoids but not the one that is de-energized for this example.

Choice D: Correct Answer, see explanation SRO Basis: N/A TABLE 25-3, MsW ISOLATION SIGNAL STATUS Light INBDDC INBDAC OUTBDDC OUTBDAC Solenoid 125 VDC RPS A 125 VDC RPS B Power A B PCISLogic B A A B

/

20. 239002 1 Which one of the following describes the effect that a loss of MCC 1 XDB will have on the Unit One SRVs, if needed for pressure control operations from the RSDP?

SRVs B, E, and (1) will lose (2)

A. (1) F (2) position indication ONLY B. (1) F (2) ALL control and indications C.(1) G (2) position indication ONLY D.(1) G (2) ALL control and indications Answer: D K/A: 239002 Relief/Safety Valves K3.01 Knowledge of the effect that a loss or malfunction of the RELIEF/SAFETY VALVES will have on following: Reactor pressure control (CFR: 41.7 / 45.4)

RO/SRO Rating: 3.9/4.0 Objective: CLS-LP-204. State which SRVs may be controlled from the Remote Shutdown Panel.

I 5d. Given plant conditions, predict how ADS/SRVs will be affected by the following: Loss of DC power

Reference:

None Cog Level:High Explanation: SRVs B,E,& G require 125 VDC to operate manually from the Remote Shutdown Panel.

Unit One RSDP control & indication for these SRVs is powered from MCC 1XDB (in RB).

This a Unit difference in that Unit Two RSDP SRV control & indication is powered form DP 2B (in DG BIdg).

Distractor Analysis:

Choice A: Plausible because SRV F is a non-ADS valve which has an alternate ASSD power supply but can only be controlled from the main control room.

Choice B: Plausible because SRV F is a non-ADS valve which has an alternate ASSD power supply but can only be controlled from the main control room.

Choice C: Plausible because SRV G has control on the RSDP, but unlike the control room, only one DC power supply is available to provide both control and indication.

Choice D: Correct Answer, see explanation SRO Basis: N/A

g From 01-50 AACHMENT 18 Page 23 o MCC IXOB LOCATION: NORMAL SUPPLY: ALTERNATE SUPPLY:

Reference Drawing: F-3OCO& Unit 1 Reactor Building 2C Switchboard 18 N/A Souiheast REAKEF LOAD EFIECT RErc:e Shhl0o1 a1el naIrurlerlalle, 1. or sr to l-E$1-IC-32, CC SDF Ficjw Co9:cIIe.

2. of i-!,:-FT-NO3, CC SD ow ndloalicr.

3. or 1-5D1-LS1-N01O-3, 9GIC SDP h;1- ewel:p 4 of 1-!1-FL-333, RD0 -1 ns:e,1 row.

5. of 1C3 SOP Reaclor FresE ire Ildicaurn

5. ..035 of 1-521-Ll-e977, liDI Rn water eve iruicalor.

7. of -CAC-Li-312. RSDP :cus water end lri.caror.

5. or i-CAC-Fi-34, RSP Oywe presaLe 1 051101.

9. of 1-521-Ll-Rf24-BX, RSD 9a waler ewel na: 0,.

10. of i-AC-R-778, SD tenperatire recore 1 . of 1-!5I-F045 poslior l9pu1 :0 Ihe EGM for :fle -amp fir oUcn 1ILaor ald rene:.

12. 0DP SRV control ard mdoalion. (59V , EQ

1. ofAODfur0Hor

14. of 1-CC-T-775, ROOF Peretator Cooing iernperaLure rerorer.

ATACHMENT 2A Pae7 o26 PANEL 4A LOCATION: NORMAL SUPPLY: ALTERNATE SUPPLY:

Reference Drawmg: LL-03024-6 Conirol Buldirrg 4 East Switchboard 2A N/A CT C LOAIO EIIECT l RHR Dw Low °rCSnLe °erniAn!e 1. -i Div 1 Low Frene:lre Fermissle 40P. Channe A1-5

/ccrIdl f41t 2. !ilher L002 can ILncio, i:fl C1. i_ow Pre55ie lnnlrumen:3tr.

S REolrc Pump A ALxary Ecupment 1. of al:erna:e 00n:rol powerlo ArO Trp o CA

?.1:erna:e Ccn:rol °iyger 2. Norma power in rori Pardl iDA Cli:. 3.

li Eacup OcraIr aIe, 2-C12-FIIOA 1. 5aoIcup Scram e all 101 epnton or aluii ncam. D U Eacup Z.cra,r yatre wn stii Ldn.

Dl Eac4Lp Ocrarr LogIc 1. Scra!r Disc1arge Vo ime Vent nnd Dm1 Vaes wil rot receie cose 5gm moir Dlv all J3en w1 cliii finolior #o!r Dlv 1

2. DLC0 cli nd reOeV nib net cai ron DV I DitnI eew3tnr all 101 set row1 :0 70.

3. Ten seconi lime Selay prIor 10 scar rese:, a ii 101 ncIon for A 7 2pan Spare E Spare Spare Reacxr Nner -i gr Level Trp C I. De-energzen for IrIp furclioq,

2. t 0 hgh eve trIp Is sealnci 1. requres 2 ccl of 5 10 ge: fill t!cp.

3. r pt Ien amber OFt w not umlrate.

10 Spare Spare 11 ADS E a:nmale power 1. or a:erna:e power, 5 oc will sti i,o:Ion wtli ,omrra! rower.

ADS1 09?) aiberrate solerold power 1. of aI:nrna:e power, all alve5 i%Ii fUl i, 9c:iorl a :fl norlra sower.

2. of ASSD power to SRV F.

IIIII!IdIII iW

21. 2390022 Unit Two is operating at rated power with no activities in progress.

A leaking SRV has resulted in slowly rising Suppression Pool temperature Which one of the following identifies the required actions to be performed lAW OAOP-30, Safety/Relief Valve Failures?

A. Place Suppression Pool Cooling in service. When Suppression Pool temperature exceeds 95° F, enter PCCP.

B. Place Suppression Pool Cooling in service. When Suppression Pool temperature exceeds 105°F, enter PCCP.

j.

5 C. When Suppression Pool temperature exceeds 95°F, SCRAM the reactor and enter PCCP.

D. When Suppression Pool temperature exceeds 105°F, SCRAM the reactor and enter PCCP.

Answer: A K/A: 239002 Relief/Safety Valves G2.04.02 Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions. (CFR: 41.7 /45.7 / 45.8)

RO/SRO Rating: 4.5/4.6 Objective: LOI-CLS-LP-300-L, Obj. 02 Given plant conditions, determine if the Primary Containment Control Procedure should be entered.

Reference:

None Cog Level: High Explanation: A leaking SRV requires entry into 0AOP-30, Safety/Relief Valve Failures, which directs placing Suppression Pool Cooling in service if pool temperatures are increasing due to a leaking SRV.

f Suppression Pool temperature reaches 95°F and no testing which could add heat to the torus is in progress, (1 05°F if testing is being performed) then EOP-02-PCCP would be entered. A reactor SCRAM is not required in this condition until Suppression Pool Temperature reaches 110°F (ref EOP-02-PCCP, SP/T-07)

Distractor Analysis:

Choice A: Correct answer, see explanation Choice B: Plausible because 105°F is a valid entry value for PCCP entry if testing is in progress which would add heat to the suppression chamber.

Choice C: Plausible because 95° is a valid entry value for PCCP if no testing in progress which adds heat to the torus but SCRAM is not required until 110°F.

Choice D: Plausible becausel 05° is a valid entry value for PCCP if no testing in progress which adds heat to the torus but SCRAM is not required until 110°F SRO Basis: N/A

SWWSY I iSV AOP-30 3.2.4 IF a safety/relief valve is leaking, THEN PERFORM the following:

1. MONITOR tailpipe temperatures LI

2. MONITOR primary containment parameters. LI

3. REFER to OAOP-14.0.

4. IF leakage is causing the suppression pool temperature LI to increase, THEN PLACE Suppression Pool Cooling in service in accordance with I (2)OP-1 7, as necessary.

3.2.1 IF a safety/relief valve is stuck open, THEN REDUCE reactor power in anticipation of a reactor scram.

3.22 IF suppression pool temperature increases to 110°F, THEN PERFORM the following:

1. INSERT a manual reactor SCRAM.

OEOP-02-PCCP

/

/PRIMARY CONTAINMENT CONTROL

// ENTRY CONDITIONS:

• SUPPRESSION POOLTEMP ABOVE WF 0&ABOVE IOVF WHEN DUE TO TESTING

• DRtWELLAVERAGE AIRTEMPABOVE 1SF

• DRYWELL PRESS ABOVE 1.7 PSIG

• SUPPRESSION POOL WATER LEVELABOVE- 27 INCHES

(-2 FEET&3 INCHES)

• SUPPRESSION POOL WATER LEVEL BELOW-31 INCHES

(-2 FEET ST INCHES)

• PRIMARYCTMTH2 CONCENTRATION ABOVE

\1L%

PCCP-2

LILft I fIJI1IIfIIIII

22. 245000 1 Unit Two is at 20% power with main turbine roll in progress lAW 20P-26, Turbine System Operating Procedure. The following turbine journal bearing vibration readings are observed on TSI-XR-640:

Bearing #1 5 mils Bearing #6 10 mils Bearing #2 5 mils Bearing #7 11 mils Bearing #3 6 mils Bearing #8 13 mils Bearing #4 7 mils Bearing #9 11 mils Bearing #5 8 mils Bearing #10 10 mils Turbine speed is 900 RPM and rising.

Which one of the following idenllfiesjhe impact of the vibration read ingson turbine operatknrari&what operator action i&-req-wred lAW 2OP-26?

A. The turbine should have automatically tripped.

Trip the main turbine ONLY.

B. The turbine should have automatically tripped.

Scram the reactor and then trip the main turbine.

C. The turbine should NOT have automatically tripped.

Trip the main turbine ONLY.

D. The turbine should NOT have automatically tripped.

Scram the reactor and then trip the main turbine.

Answer: A K/A: 245000 Main Turbine Generator and Auxiliary Systems A2.09 Ability to (a) predict the impacts of the following on the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS ; and (b) based on those prediotions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:-Turbine vibration (CFR: 41.5 /45.6)

RO/SRO Rating: 2.5/2.8 Objective: LOl-CLS-LP-026, Obj. 28n Given plant conditions, predict the effect that the following will have on the Main Turbine, Gland Seal, and Moisture Reheater System: Main Turbine Hi Vibration

Reference:

None Cog Level: High Explanation: TSI is normally disarmed during turbine operation (>23% power with the turbine online), but is armed for turbine roll. When turbine RPM is between 801 1400 RPM, the trip setpoints are 12 mils for bearings 1-8 and 10 mils for 9 & 10. This requires an immediate turbine trip per OP-26, Section 5.4.2. If reactor power is greater than 26% then a reactor scram will occur when the turbine is tripped. Operator actions call for scramming the reactor first, then tripping the turbine. In the described conditions, a scram is not required since power is below 26%.

Distractor Analysis:

Choice A: Correct Answer, see explanation Choice B: Plausible because if power was greater than 26% then a reactor scram would be required.

Choice C: Plausible because TSI is normally bypassed (but in this case it is not for the startup) and a turbine trip only is required.

Choice D: Plausible because TSI is normally bypassed (but in this case it is not for the startup) and a scram is not required at less than 26% power.

SRO Basis: N/A From OP-26, Turbine Startup:

17. IF any of the following conditions occur while turbine LI speed is increasing, THEN DEPRESS the EMERGENCY TRIP SYSTEM push button to trip the turbine AND ENSURE the following turbine valves close:

All four Stop Valves LI All four Control \alves El All four Intermediate Stop Valves El All four Intercept Valves LI 110P-26 Rev.84 Page41of153 5.42 Procedural Steps

a. Turbine shell to rotor differential expansion LI indicating in red band, as indicated on TSI-TXR-638 Point 1.

b Turbine journal bearing high vibration, as LI indicated on TSJ-XR-640:

Turbine Speed vs. Vibration Less than or Greater than 8 mils.

equal to 800 rpm 801 to 1400 rpm Greater than 12 mils for bearings V1-V&.

or greater than 10 mils for bearingsV9 and V10.

From the APP:

ACTIONS MAIN TURBINE:

H NOTE: The main turbine high vibration trip is disabled when operating at or above L 23% rated thermal power with the generator online.

b. IF vibration is at or above 12 mils on bearings 1-8 or 10 mils on bearing 9 & 10 AND an adjacent bearing has also exhibited a significant increase in vibration, THEN PERFORM the following:

(1) Scram the reactor (2) Trip the turbine (3) IF directed by the Unit SCO, THEN BREAK condenser vacuum.

(4) ENTER 1EOP-01-RSP AND EXIT this procedure.

1APP-UA-23 Rev. 65 Page 90 of 105

23. 256000 1 Unit One is operating at 30% power with the following plant conditions:

Hotwell temperature 118°F Condensate Pump A Standby Condensate Pump B/C Running Debris in the Off-Gas System begins to plug the Off-Gas Filter.

Which one of the following choices completes the statements below if debris continues to build up on the Off-Gas Filter?

The plant effect is that cnenserizacum will lower and (1) lAW OAOP-37.0, Loss of Condenser Vacuum, efficiency of the operating Steam Jet Air Ejector (SJAE) can be improved by throttling the SJAE Condensate Recirculation Valve, CO-FV-49, open as long as condensate flow does NOT exceed (2) gpm.

A. (1) Hotwell temperatures will increase (2) 14,400 B. (1) Hotwell temperatures will increase (2) 16,000 C. (1) Off Gas system will auto bypass (2) 14,400 D. (1) Off Gas system will auto bypass (2) 16,000 Answer: A K/A: 256000 Reactor Condensate System K6.09 Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR CONDENSATE SYSTEM: Offgas system (CFR: 41.7/45.7)

RO/SRO Rating: 2.6/2.6 Objective: LOI-CLS-LP-032-A, Obj. 20 Given plant conditions predict the effects that a loss or malfunction of the following will have on the Feed and/or Condensate System: Off-Gas System.

Reference:

None Cog Level: High Explanation: Lowering condenser vacuum is from a malfunction of the Off Gas system. As condenser vacuum decrease hotwell temperature will increase. AOP-37 provides guidance for maintaining the efficiency of the SJAE by throttling of the CO-FV-49 valve open and flow is limited by the table according to the pump arrangement. With B&C pump flow is limited to 14400 gpm.

AII1I!ALIIIII Distractor Analysis:

Choice A: Correct answer, see explanation Choice B: Plausible because hotwell temperatures will increase and if a different pump arrangement was in affect then 16000 gpm would be correct.

Choice C: Plausible because 14400 is correct but off gas will bypass on high flow conditions not low flow.

Choice D: Plausible because 16000 would be correct for a different pump arrangement and off gas will bypass on high flow conditions not low flow SRO Basis: N/A AOP-37 NOTE: SJAE efficiency at low reactor power may be improved by throttling open SJAE CONDENSATE RECIRCULATION VALVE, CO-FV-49, while maintaining condensate pump discharge header pressure greater than 190 psig.

3.2.6 IF desired, THEN PERFORM the following:

1. ENSURE VALVE CO-FV-49 INLET ISOLATION VALVE, CO-VilO, is open.

3.0 OPERATOR ACTIONS

2. THROTTLE OPEN SJAE CONDENSATE LI REC1RCULATJON VALVE, CO-F V-4, NOT to exceed the flow limits in the following table.

Condensate Pumps Average Hotwell Condensate Flow Limit Operating Temperature(Note 1) (GPM) <Note 2)

B&C NIA 14,400 A&BorA&C 115F 16.000 A&BorA&C <115°F 17,400 A,B,&C N/A 18,200 NOTE 1: Average hotwell temperature is available from process computer point Ui(U2) CO_C099 AVG HOTWELLTEMP U1(U2). IF U1(U2) CO C099 is NOT available, THEN use the highest functional hotwell temperature process computer point.

NOTE 2: Condensate flow through each of the CONDENSATE PUMP RECIRC VLVs, CO-FV-147, (CO-FV-i48, (CO-F V-149 of 1500 gpm will NOT be indicated on AIR EJECTOR COND FLOW RECIRC CTL, CO-F!C-49, and must be considered in the following system flow limits. In addition, total condensate flow may be influenced by hotwell reject flow, RFP or condensate booster pump minimum 110w valves opening, heater drain tank level controller positions, feedwater heaters out of service, etc.

24. 259002 1 Unit Two is operating at 65% power with Reactor Feed Pump (RFP) 2A running and RFP 2B unavailable. The operator observes the following:

RFP A Control Trouble alarm is received RFP A Manual/DFCS selector switch is in DFCS DFCS Control light for RFP A on XU-1 is out Which one of the following choices completes the statements below?

RFP 2A speed will (1)

The operator can control RFP speed by (2)

A. (1) drop to the idle sped.setpoint (2) operating the RFP Raise/Lower control switch on XU-1 B. (1) drop to the idle speed setpoint (2) placing the RFP A Speed Controller in Manual and adjusting the output demand C. (1) remainat the last known demand (2) operating the RFP Raise/Lower control switch on XU-1 D. (1) remain at the-last known demand (2) placing the RFP A Speed Controller in Manual and adjusting the output demand Answer: C K/A: 259002 Reactor Water Level Control System A3.1 0- Ability to monitor automatic operations of the REACTOR WATER LEVEL CONTROL SYSTEM including: TDRFP lockup: TDRFP (CFR: 41.7 /45.7)

RO/SRO Rating: 3.1/3.0 Objective: LOl-CLS-LP-032.2, Obj. 08j Given the following plant conditions, predict the response of the RFPT and Speed Control System: Loss of signal from the DFCS

Reference:

None Cog Level: High Explanation: For Brunswick a TDRFP lockup is being met by the loss of signal locking the RFP controls at the last known signal. IF RFPT B(A) MAN/DFCS selector switch is in DFCS, AND the DFCS control signal subsequently drops below 2450 rpm, OR increases to greater than 5450 rpm, THEN Woodward 5009 digital controls will automatically assume RFPT speed control and maintain current speed. In this condition, the RFPT will only respond to LOWER/RAISE speed control switch commands

Distractor Analysis:

Choice A: Plausible because the Woodward manual control signal automatically tracks the DFCS output signal. Although this failure mechanism no longer exists, a drop in speed to the idle speed setpoint is a failure that used to be associated with a loss of hydraulic oil pressure.

Choice B: Plausible because the DFCS control signal has failed. With the DFCS Control light out, the REP is under manual control of the Woodward governor and adjusting the output of the individual REP Speed Controller will have no effect. A drop in speed to the idle speed setpoint is a failure that used to be associated with a loss of hydraulic oil pressure. This failure mechanism no longer exists.

Choice C: Correct Answer, see explanation Choice D: Plausible because the Woodward manual control signal automatically tracks the DFCS output signal. With the DFCS Control light out, the RFP is under manual control of the Woodward governor and adjusting the output of the individual REP Speed Controller will have no effect.

SRO Basis: N/A 3.0 OPERATOR ACTIONS 3.1 Immediate Actions 3.1.1 IF automatic level control will NOT restore normal reactor vessel level. THEN MANUALLY CONTROL reactor feed pumps to restore normal level.

OAOP-23.O Rev. 37 Page 3 of ID NOTE: IF RFPT B(A) MAN/DFCS selector switch is in DFCS, and the DFCS control signal subsequently drops below 2450 rpm, or increases to greater than 5450 rpm, Woodward 5009 digital controls ll automatically assume RFPT speed control and maintain current speed. In this condition, the RFPT will only respond to LOWERJRA1SE speed control stch commands until the MANJDFCS selector switch is placed in MAN, DFCS CTRL RESET pushbutton is depressed, and the MAN/DFCS selector switch returned to DFCS.

20P-32 Rev. 180 Page 49 of 330 4.2.4 DFCS Control Signal Failure It the 5009 control system detects that the Remote Speed Setpoint (RSS) from the DFCS is outside the failure limits, an RSS signal failure condition is set and, if the 5009 control system was in the DFCS mode, an automatic transfer to the manual mode will occur.

The RFPT speed setpoint (and hence RFPT speed) will be maintained at the last good value and can be controlled using the Panel XU-1 RAISE I LOWER switch (Figure 32.3-14). The MANUAL / DFCS switch should be placed in the MANUAL position.

I SD-32.3 Rev. 5 I Page 66 of 123

LIP!I1IGTL1 IN4I1IIIIII

25. 2610001 Venting of the suppression chamber is being performed lAW lOP-I 0, Standby Gas Treatment System Operating Procedure. SBGT system valve status:

i-CAC-V172, Supp Pool Purge Exh Vlv Open i-CAC-V22, Torus Purge Exh Vlv Open 1-VA-i D-BFV-RB, Reactor Building SBGT Train 1A Inlet Valve Closed 1-VA-i H-BFV-RB, Reactor Building SBGT Train lB Inlet Valve Closed A transient occurs which causes Drywell pressure to rise to 1 .5 psig and Reactor water level to lower to 160 inches before being recovered.

Which one of the following choices completes the statements below concerning the expected response, if any, of Suppression Chamber Purge and SBGT system valves?

The 1-CAC-V172 and 1-CAC-V22 (1)

The 1-VA-I D-BFV-RB and 1-VA-I H-BFV--RB (2)

A. (1) close (2) remain closed B. (1) close (2) open C. (I) remain open (2) remain closed D. (1) remain open (2) open Answer: A K/A: 261000 Standby Gas Treatment System Ki .03 Knowledge of the physical connections and/or cause effect relationships between STANDBY GAS TREATMENT SYSTEM and the following: Suppression pool (CFR: 41.2 to 41.9 / 45.7 to 45.8)

RO/SRO Rating: 2.9/3.1 Objective: LOl-CLS-LP-004.1, Obj. 5 List the signals and setpoints that will cause a Secondary Containment isolation LO!-CLS-LP-012, Obj. 06 Given plant conditions, determine if a Group Isolation should occur.

Reference:

None Cog Level: High

Explanation: The suppression pool vent valves close with a group 6 isolation signal while the SBGT RB suction valves open on a Secondary Containment Isolation signal. Based on the level indication a LL1 signal is present which would actuate a Group 6 isolation. A SCI signal occurs at LL2 (105 inches). Both actuate with DW pressure greater than 1.7 psig. DW pressure of 1.5 psig actuates an alarm only.

Distractor Analysis:

Choice A: Correct Answer, see explanation Choice B: Plausible because there is a group 6 isolation due to low water level. If level were below LL2 (105) then this would be correct.

Choice C: Plausible because if level was above 166 inches then this would be correct. There is no isolation signal from DW pressure.

Choice D: Plausible because there is a group 6 signal but not a SCI signal. If drywell pressure were greater than 1.7 psig or level were below 105, an SCI would occur.

SRO Basis: N/A TABLE 12-2 Primary Containment Isolation System Group IsoIaton Instrumentation Setpoints ISOLATION ISOLATION TRIP SETPOINT NOTES GROUP SIGNAL Tech Spec. Actual Allowable (Note 1)

Group S High Steam Flow 275% 220% Note S Low Steam Pressure 53psig 70 psig High Turb Exh Pressure 6 psig 5 psig Steam Line Area Hi Temp 175°F 165°F Note 4 Steam Line Tunnel High 200°F 165°F/190°F Note 4 Amb Temp Steam Line Tunnel cIT High 50°F 47°F Note 4 EquipAreaHighTemp 175°F 165°F Equip Area dT High 50°F 47°F Group 6 Low Level #1 153 166 High Drywell Pressure 1.8 psig 1.7 psig Rx Bldg Exhaust Hi Rad 16 mR/hr 4 mR/hr Rx Bldg Exhaust Hi Temp N/A F*

0 135 Note 6 High Main Stack Rad ODCM ODCM Note 2

SD-i 2 The following CAC valves (Unit 1/2 power supply listed in parentheses) close on a Group 6 Isolation (Figure 12-1 1):

DMsion I AC Powered CAC-V5 Suppression Pool Nitrogen Inlet 31A8132A6 CAC-V6 Drvwell Nitrogen Inlet 31AB/32A8 CAC-V7 Suppression Pool Purge Exhaust 31A8/32A8 CAC-V9 Drvwell Purge Exhaust 3 16/328 CAC-V162 Suppression Pool CAD N 2 Inlet 31AB/S2AB CAC-V163 DrqweII Nitrogen Inlet 31AB/32AB CAC-V172 Suppression Pool Purge Exhaust 31 AB/32AB DC Powered CAC-V49 Drywell Head Purge Exhaust 1 1A/12A CAC-V160 Suppression Pool CAD N2 Inlet 1 1A/12A CAC-V161 Drywell Nitrogen Inlet 1 1A/12A Division II AC Powered (

CAC-V4 CAC Nitrogen Inlet 31A8/32A8 CAC-V8 Suppression Pool Purge Exh Byp 31 B /328 CAC-V10 Drvwell Purge Exh Backup 318/ 328 CAC-V15 Prim Cont Air Purge Inlet 31A8/32A8 CAC-V22 Suppression Pool Exh MCC 1XD/2XD CAC-V23 DW Purge Exh Backup Byp MCC 1XF/2XF

t 2.6 Standby Gas Treatment SBGT provides a means of minimizing the release of radioactivity from Secondary Containment by filtering and exhausting containment air. Two trains are provided, each consisting of a fan and filtration devices which remove particulate and iodine prior to exhausting to the main plant stack.

The units draw a suction on the 50 elevation of the Reactor Building into which all areas of the Reactor Building communicate. This system can be placed into service manually or will enter service automatically under the same conditions that RB HVAC will isolate. The parameters monitored for SBGT initiation and RB isolation are:

PARAMETER Low Reactor Water Level #2 High Drywell Pressure Main Stack High Radiation Reactor Building Ventilation Exhaust High Temperature SD-04.i Rev.5 I Page12of24 Reactor Building Ventilation Exhaust High Radiation Any of these signals initiate the following sequence of events:

• Closes the isolation dampers which stops the fans in the Reactor Building Ventilation System.

• Starts both SBGT fans simultaneously.

• Closes the isolation valves and stops the fans in the Purge System.

• Opens the SBGT inlet and outlet isolation valves (U2 only).

• Opens the SBGT Reactor Building suction valves.

• Closes the SBGT Primary Containment suction valves.

I.

a

26. 2610002 Which one of the following requires the SBGT Train A to be manually reset in order to restore auto start capability?

A. When the fan electrical power is lost and then restored.

B. When the control switch is repositioned from STBY to SYST A PREF.

C. ONLY when the inlet temperature (TS-7) reaches 180°F and-subsequently 1

lowers to <1-80°F D. ONLY when the prefilter temperature (TS-1) reaches 210°F and subsequently lowers to <210°F.

Answer: D K/A: 261000 Standby Gas Treatment System K4.04 Knowledge of STANDBY GAS TREATMENT SYSTEM design feature(s) and/or interlocks which provide for the following: Radioactive parbculate filtration (CFR: 41.7)

RO/SRO Rating: 2.7/2.9 Objective: LOI-CLS-LP-010, Obj. 06 Describe the function of the temperature switches under abnormal operations.

Reference:

None Cog Level: Low Explanation: If the 210 degree prefilter temperature is reached the SBGT Fan will shutdown and in order to reset the temperature must have reduced below the setpoint.

Distractor Analysis:

Choice A: Plausible because logic power failure may require a reset but the fan itself does not.

Choice B: Plausible because this action will place the system in auto initiation mode, but no manual reset is required.

Choice C: Plausible because inlet temperature greater than 180 would initiate the emergency operation logic for the SBGT train.

Choice D: Correct Answer, see explanation SRO Basis: N/A

X

2. Pretilter Compartment (TS-1)

The Pretilter compartment is equipped with two switches per train.

Switches VA-TS-5301 (VA-TS-5296) actuate at 21 OF to secure the Fan and Heater if the train is operating. Local and remote lights indicate actuation of the temperature switches.

The system starter circuit must be manuafly reset after actuation of these switches and after the temperature has dropped to less than the set point by momentarily placing the train selector switch to the RESET position.

(TS-2)

Switches VA-TS-5300 (VA-TS 5295) actuate at 180 F to control the heater to regulate the train inlet temperature when the Fan is running.

Local and remote lights indicate temperature switch actuation.

SD-iC Rev. 7 Page 17 of 381 3.2.1 Control Room: RTGB XU-51 The four-position, RESET? SYST A(B) PREF I STBY? ON, selector switch controls the status of the SBGT train. The RESET position is a momentary contact position with spring-return to the SYST A (B)

PREF position. The other three positions are maintained contact positions. The RESET position allows resetting the Starter Circuit when a manual reset is required following actuation of the Prefilter Compartment high temperature switches at 21 OF.

The SBGT A (B) PREF position enables the train to start automatically if an initiation signal is present. Placing the switch in the ON position will start the Fan provided all other start permissives are met.

The STANDBY position is misleading in that, with the switch in this position, the train is tiQI placed in the condition where it is ready to automatically start. Instead, the automatic start feature is defeated.

I SD-IC I Rev.7 Pagei4of38

1 iiiiiiiw&iiaii i iiiiir ii

27. 262001 1 Unit Two is operating at rated power when a Transformer Bus Lockout occurs.

Which one of the following identifies the equipment that will remain de-energized with no operator actions?

A. Recirc VFDs 2A and 2B.

B. Demin Water Xfer Pmps 2A, 2B, and 2C.

C. Circulating Water Discharge (CWOD) Pumps 2C and 2D.

D. ISFSI Drain Collection Pumps (2-DST-ISFSI-P1-PMP and 2-DST-ISFSI-P2-PMP).

Answer: A K/A: 262001 A.C. Electrical Distribution K2.01 Knowledge of electrical power supplies to the following: Off-site sources of power (CFR: 41.7)

ROISRO Rating: 3.3/3.6 Objective: LOl-CLS-LP-050-A, Obj. 08c Given plant conditions predict the changes in Unit 1 and/or 2 parameters associated with the operation of the following equipment: Transformer bus lockout relay

Reference:

None Cog Level: Low Explanation: A Transformer Bus overcurrent trip will result in the SAT being de-energized and the Feed to Caswell Beach Bus B. The SAT feeds Bus 2B and Common Bus B. Common Bus B would auto crosstie to Common Bus A and remain energized. Caswell Beach Bus B can be transferred to Caswell Beach Bus A, but this is a dead bus transfer (manual, not automatic).

Bus 2B feeds the VFDs which would require the plant to insert a manual scram due to no recirc pumps.

Distractor Analysis:

Choice A: Correct Answer, see explanation Choice B: Plausible because these are powered from the SAT feed but have auto-crosstie feature to Ui.

Even though these are U2 designated pumps they are normally fed from Common A through Common C to WTA.

Choice C: Plausible because these are powered from an offsite power source (Ui Transformer Bus to Caswell Beach Bus A).

Choice D: Plausible because this is an offsite power source from the Southport Feeder.

SRO Basis: N/A

p

_:_._a,adw-I 1

4 t 1 --_rpz p 1.3.1 System Components and Configuration (Figures 1, 2, and 3)

The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. 4160 VAC power is divided into Balance Of Plant (BOP Bus), and Emergency (E-Bus) distribution.

The BOP Buses consist of the Conventional Buses (Common A/B Buses; Buses 1BI2B, 1C/2C, 1D/2D) and Caswell Beach Buses A/B. The Emergency switchgear are Buses E1/E3 (Division I) and E2/E4 (Division II). (Se Figure 50.1-1)

The BOP Common Buses NB are powered from the respective unirs SAT and have auto dead bus crosstie capability should one lose its normal source. The Caswell Beach Buses NB are powered from the respective units Caswell Beach transformers and have manual dead bus crosstie capability. The B/C/D Buses can be powered from the respective units UAT or SAT. Buses B, C and D have their source breakers interlocked to prevent parallel operation of the power sources. Buses B, C and D are provided with manual bus transfer schemes that allow momentary parallel operation of the power I sources while transferring power supplies. Buses C/D are normally SD-50.1 Rev.22 I Page8of143 powered from the respective units UAT during power operations and have an auto dead bus fast transfer to the SAT on loss of the UAT (any generator lockout). The B Bus has no auto transfer capability, therefore is normally supplied from the respective units SAT.

Load: 480V Motor Control Center WTA Location: Water Treatment Building 20 Drawing

Reference:

F-3052 Upstream Power Source: 480V Substation Common C COMPT LOAD DESCRIPTION EFFECTS ON LOSS OF POWER C73 Paint and Sandblasting Shop Loss of power to 480V panel HA (OP-52.1) in Sandblasting Shop.

C78 Demin. Water XFER Pump 2A Loss of load.

(OP-31 .2)

C79 Demin. Water XFER Pump 28 Loss of load.

(OP-31 .2)

C80 Oemin. Water XFER Pump 2C Loss of load.

(OP-31 .2)

Load: Caswell Beach 4160V Bus A Location: Caswell Beach Pumping Station Drawing

Reference:

F-03025 Upstream Power Source: Unit 1, Caswell Beach Transformer No I COMPT LOAD DESCRIPTION EFFECTS ON LOSS OF POWER AN1 Circulating Water Discharge Loss of load. if Pump 2C (20P-29)

AN2 Circulating Water Discharge Loss of load.

Pump 2D (20P-29)

11!IIIIi1 M1IILUIII1PIIIVI!IIII The affects of a loss of the Southport Feeder:

6.10 Loss of powerto the ISFSI Drain Collection Pump Station 6:10.1 Loss of power to ISFSI Drain Collection Pumps (2-DST-ISFSI-P1-PMP and 2-DST-ISFSI-P2-PMP) 001-50.15 Rev. 15 Page 9 of 66

28. 262002 1 Unit Two is operating at rated power when a Primary UPS Inverter malfunction results in an overvoltage on the inverter output.

The following annunciators are in alarm:

UPS Primary Power Con vtr Trouble UPS Transfer to Reseive Which one of the following choices completes the statement below?

lAW the above APPs, the required operator action is to transfer UPS loads from (1) to (2)

A. (1) the standby UPS inverter (2) MCC2CA B. (1) the standby UPS inverter (2) MCC2CB C. (1) MCC2CA (2) the standby UPS inverter D. (1) MCC 2CB (2) the standby UPS inverter Answer: D K/A: 262002 Uninterruptable Power Supply (A.C./D.C.)

A2.02 Ability to (a) predict the impacts of the following on the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Over voltage (CFR: 41.5 /45.6)

RO/SRO Rating: 2.5/2.7 Objective: LOI-CLS-LP-052-A, Obj. 7a Predict the impact(s) of the following on the UPS System:

- Over Voltage.

Reference:

None Modified question from the NRC 10-2 exam. (262002_14) The previous question asked the power supply while this question asks the action required.

Cog Level: High

Explanation: Overvoltage or undervoltage on the UPS inverter output will cause an automatic transfer of the UPS loads. In normal alignment, the primary UPS will be in service supplying loads and the standby UPS will be energized. However, the standby UPS alignment is such that it will not automatically pick up the loads if the primary unit fails. This protects the standby unit from automatically tying onto a faulted bus. The transfer scheme instead automatically transfers the loads to the Division II AC source (hard source). The standby inverter must be manually placed in service. This configuration protects the standby UPS but results in vulnerability in that if the Division II AC source is lost, the UPS loads would be deenergized. APP-UA-06 6-9, UPS Transfer to reserve, provides direction to place the UPS loads on an operable inverter. Since the primary inverter is inoperable, the standby inverter would be placed in service.

Distractor Analysis:

Choice A: Plausible because the original design of the system was to have the standby inverter automatically pick up the loads. OP-52 provides a procedure section to transfer UPS loads to the alternate source.

Choice B: Plausible because the original design of the system was to have the standby inverter automatically pick up the loads. OP-52 provides a procedure section to transfer UPS loads to the alternate source. Inverter power supplies are from 2CA and 2CB.

Choice C: Plausible because MCC 2CA is the normal (and only) AC power supply to the primary inverter.

However, there is no hard connection or procedure instructions to align 2CA directly to the loads. APP-UA-06 6-9 directs transferring UPS loads to an operable inverter.

Choice D: Correct answer, see explanation SRO Basis: N/A SD-52 I .&3. Uriinterruptible Power Supplies Each units Vital UPS System consists of a vital bus (Distribution Panel 1A!2A), a primary unit, and a standby unit. The primary and standby units are each capable of supplying the total UPS load. Only one UPS unit will be in service at a time with the other unit being available as a backup. If the primary unit fails, the UPS loads are automatically transferred to the alternate AC source. If desired, the loads can then be manually transferred to the standby UPS unit.

_rt-i +/- LLar -.

The UPS system is normally aligned as follows:

The primary unit is in service with its output connected to the UPS distribution system. Its rectifier receives 480 VAC power from a Division I emergency distribution panel. A 250 VDC from DC Switchboard 1A (2A) is supplied in parallel with the rectifier output to power the inverter should the normal AC source be lost. The alternate AC source from the standby unit is available at the static transfer switch to pick up the loads if the inverter output is lost.

I The standby unit is also energized with its 480 VAC input supplied from a Division II emergency distribution panel and its 250 VDC supplied from DC Switchboard 16 (2B); but, its output is bypassed by its manual bypass switch and its alternate AC input is being supplied directly to the primary unit. The standby unit receives its alternate AC source of power from the same Division II distribution panel as its rectifier AC input through a 480-120/208 VAC transformer. The standby units alternate AC input is also referred to as the hard source.

TABLE 52-1 Vital UPS Power Supplies UNIT1 UNIT2 480 \AC SUPPLY PRIMARY UNIT MCC 1CA (ES) MCC 2CA (Efl STANDBY UNIT MCC 1CB (E6) MCC 2CB (B) 250 VDC SUPPLY PRIMARY UNIT DC SWBD IA DC SWBD 2A V STANDBY UNIT DC SWBD lB DC SWBD 2B ALTERNATE AC MCC 1CB (E6) MCC 2CB (E8)

The static transfer switch provides a means of switching either manually or automatically between two, three-phase, four wire power sources without an interruption of the load power. The switch performs a transfer between the power sources using solid state components; allowing a faster transfer than could be accomplished by a mechanical switch. The switching action itself is practically instantaneous, and the time involved in the operation is mainly the sensing time required to determine that a transfer is necessary. normally a small fraction of a cycle. The switch receives an inverter input through the inverter output breaker, CB1 02, and a input from a alternate AC source. The switch is normally aligned such that the inverter output is supplying the system loads. Should the inverter output voltage drop below or increase above a preset level the static transfer switch will automatically transfer to the alternate AC source. Both a high or a low output voltage could affect equipment operating characteristics and lead to equipment damage.

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C flETDI O.SPLT vsnei hr 114 (.Dfjr 2APP-UA-06 Unit 2 APP UA-O6 6-9 Page 1 of 1 UPS TRANSFER TO RESERVE ACTIONS

1. If UPS loads have automatically transferred to the alternate source, transfer UPS loads back to an operable UPS unit per 20P-S2, Section 8, Infrequent Operatione IL

29. 263000 1 Which one of the following completes the statements below regarding 125/250 VDC Station Distribution?

During an equalize charge, the charger output to the battery will be at a (1) voltage than when in the float mode.

The 125 VDC batteries are sized to supply emergency power at a 150 amp rate for (2) hours.

A. (1) lower (2) 8 B. (1) lower (2) 10 C. (1) higher (2) 8 D. (1) higher (2) 10 Answer: C K/A: 263000 D.C. Electrical Distribution Al .01 Ability to predict and/or monitor changes in parameters associated with operating the D.C.

ELECTRICAL DISTRIBUTION controls including: Battery charging/discharging rate (CFR: 41.5 / 45.5)

RO/SRO Rating: 2.5/2.8 Objective: LOl-CLS-LP-051, Obj. 13- Describethe location and operation of Battery Chargers lB-i, 1B-2, 2B-1, and 2B-2 AC Power Transfer Switches.

Reference:

None Cog Level: Low Explanation: The float mode voltage for the 125 VDC battery charger is 135 volts while in equalize the charger output is 140 volts. The design of the batteries is for 150 amps for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Distractor Analysis:

Choice A: Plausible because the student may have a knowledge deficiency on which (float vs equalize) value is for equalize and the 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is correct.

Choice B: Plausible because the Caswell Beach batteries are rated for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> and the student may have a knowledge deficiency on which (float vs equalize) value is for equalize.

Choice C: Correct answer, see explanation.

Choice D: Plausible because higher is correct and the Caswell Beach batteries are rated for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

SRO Basis: N/A

K 2.0 COMPONENT DESCRIPTION:DESIGN DATA 2.1 Battery Capacity Ratings All of the baiter, systems (with the exception of the Caswell Beach Microwave) have a design Ampere-Hour capacity rating which defines the batteries expected lifetime, in hours, based upon a given continuous loading.

in amperes. It should be noted that This IS merely a reference number and that hatter, lifetime is shortened if it is discharged at a higher rate or lengthened if discharged at a lower rate. The individual hat:ery capacties are:

BATTERY SYSTEM AMP-HOUR RATING 12550 VDC Station (each 1200 AMP-HOURS at a 150 amp division) rate for & hours 24148 VDC Station 600 AMP-HOURS at a 75 amp rate (each division) for S hours 125 VDC Caswell Beach 200 AMP-HOURS at a 20 amp rate for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> SD-Si Rev. 10 Page ii of84 There is no direct indication of the status of the battery charger: i.e.,

whether it is in the float charge or equalizer charge mode. If in the float charge mode the volt meter should read approximately 135 VDC. If in the equalizer charge mode the meter should read approximately 140 VDC.

SD-Si Rev. 10 Page 14 of 84 L 1

_____r

30. 264000 1 DGI has been paralleled to 4160kv Bus ID for the performance of 0PT-12.2A, No. I Diesel Generator Monthly Load Test. The current DG parameters are:

DIESEL GENERATOR #1 XU)))

]

LxI ] if il f r KILOWAITS X1000 -

LI LI_LLJJ_t1JImIItHLL1LtL1I 2 0 J ltiIhLLii 2 3 4 KlL0VAJS Xl 000 Which one of the following choices identifies the result of momentarily placing the DGI Voltage Adjusting Rheostat to Raise? (Assume system load remains constant)

Real load will (1) and reactive load will (2)

A. (1) rise (2) remain the same B. (1) rise (2) rise C. (1) remain the same (2) rise D. (1) rise (2) lower Answer: C K/A: 264000 Emergency Generators (Diesel/Jet)

K5.05 Knowledge of the operational implications of the following concepts as they apply to EMERGENCY GENERATORS (DIESEL/JET): Paralleling A.C. power sources (CFR: 41.5/ 45.3)

RO/SRO Rating: 3.4/3.4 Objective: AOl-CLS-LP-39, Obj. 03e - Describe the operation of the below listed EDG components:

Voltage adjust rheostat.

Reference:

None Cog Level: Low

wa s Explanation: Diesel Generator KW and KVAR is controlled locally using the Governor Motor Control switch and the Voltage Adjusting Rheostat switch, respectively. Once the DG is paralleled with AC sources, the Voltage Adjusting Rheostat switch will affect generator excitation which will in turn control KVAR output. Generator load is adjusted using the Governor Motor Control switch which, prior to synchronization, would be used to adjust generator speed and frequency.

Distractor Analysis:

Choice A: Plausible because raising voltage adjustment would raise KVAR but would have no effect on real load (KW).

Choice B: Plausible because raising voltage adjustment would raise KVAR but would have no effect on real load (KW)

Choice C: Correct answer, see explanation Choice D: Plausible because raising voltage adjustment would raise KVAR but would have no effect on real load (KW)

SRO Basis: N/A OPT-i 2.2A

1. ADJUST diesel generator load, at the allowed load rate per Table 4:

RAISE KW to greater than or equal to 2800 KW -

and less than or equal to 3000 1KW with GOVERNOR MOTOR CONTROL RAISE kvarsto between +1 700 and +1900 with -

VOLTAGE ADJUSTING RHEOSTAT OOP-50.1 prior to synchronization:

8.72 Procedural Steps

5. ADJUST generator stator voltage to equal to or slightly greater than emergency bus voltage with VOLTAGE ADJUSTING RHEOSTAT NOTE: Normal emergency bus voltage is 4160V.

6. PLACE the SYNHRONIZ?NG GENERATOR El synchroscope switch in ON at the appropriate generator gaugeboard.

7. ADJUST speed of the diesel generator with the LI GOVERNOR MOTOR CONTROL switch until the synchroscope is rotating slowly in the FAST direction (clockwise).

-. - p OOP-50.1 following synchronization:

12. MAINTAIN generator kvars approximately one-halt the LI 1KW load, Mth VOLTAGE ADJUSTING RHEOSTAT, while raising diesel generator load.

13. RAISE diesel generator load by momentarily placing the El GOVERNOR MOTOR CONTROL switch in RAiSE thus decreasing the normal supply amperage as reported by the Auxiliary Operator.

(

r fr 44 IaLsa4v iz . aaJ1Z-C

FIGURE 39-9 Diesel Engine RTGB Controls CONTROL LOCAL AUTO ROOM MANUAL MANUAL I MAN. REG. 1 NOT LOW HIGH LOADED AVAILABLE LIMIT LIMIT AVAILABLE NORMAL NO LOAD VOLTAGE

\oWAISRAISE AUTO REG.

AUTO MAN. LOW HIGH LIMIT LIMIT AUTO TRANSFER NORMAL GOVERNOR MANUAL VOLTAGE LOWER RAISE STOP START LOWER RAISE AUTO MODE AUTO MODE STOP SWITCH START SWITCH SD-39 Rev. 15 Page 114 of 125

1_I_II1__

31. 2640002 Unit Two is operating at rated power when a loss of offsite power occurs.

Subsequently a phase overcurrent occurs on Bus E3. The CRS has directed cross-tying electrical buses lAW OAOP-36.l, Loss of Any 4160V Buses or 480V E-Buses.

Which one of the following indicates the status of DG3 and the actions required by OAOP-36.1?

DG3 is (1) and actions to cross-tie bus (2) must be performed.

A. (1) tripped (2) El to E3 B. (1) tripped (2) L8 to E7 C. (I) running unloaded (2) El to E3 D. (1) running unloaded (2) E8 to E7 Answer: B K/A: 264000 Emergency Generators (Diesel/Jet)

A2.06 Ability to (a) predict the impacts of the following on the EMERGENCY GENERATORS (DIESEL/JET) and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Opening normal and/or alternate power to emergency bus (CFR: 41.5 /45.6)

RO/SRO Rating: 3.4/3.4 Objective: LOl-CLS-LP-050-B, Obj. 13c Given plant conditions, determine if the following breakrs could be closed: El to E3 (or E2 E4) cross-tie breakers.

Reference:

None Cog Level: High Explanation: A phase overcurrent condition on the E-bus will result in tripping of the associated DG output breaker and will also result in a trip of the DG. lAW OAOP-36.l, cross-tying of electrical buses should occur at the 4160V level unless evidence of a fault on the 4160V bus exists as determined by Attachment 5 of the procedure. In the conditions cited, cross-tying at the 480V (E8 E7) level would be required.

Distractor Analysis:

Choice A: Plausible because a trip of the DC would occur. However, cross-tying of E-buses must occur at the 480V (E8 to E7) level under these conditions.

Choice B: Correct answer, see explanation.

Choice C: Plausible because the DC output breaker would trip on the fault condition. During normal shutdown sequence, the output breaker would be tripped, then the DG manually secured.

However, in this condition a trip of the DG would result. El to E3 crosstie not allowed due to the fault condition.

Choice D: Plausible because the DC output breaker would trip on the fault condition. During normal shutdown sequence, the output breaker would be tripped, then the DC manually secured.

However, in this condition a trip of the DC would result. E8 to E7 would be the desired cross tie performed.

SRO Basis: N/A SD-39, Section 3.2.7 An automaUc trip of the Diesel output breaker will result from any of the following electrical faults:

• phase clifferenUal overcurrent

• phase overcurrent

• reverse power

• Loss of field These trips will also shut down and lockout the engine controls. The breaker will also trip if the Diesel Engine trips. (Figure 39-17)

OAOP-36. I 3.2.11 Emergency Bus Cross-Tie NOTE: IF cross-tying E-buses THEN Technical Specifications 3.8.7 OR 3.8.8 should be consulted.

I. IF desired to cross-tie a 4160V E bus, THEN PERFORM the following:

a. CONFIRM indications of a lockout OR phase El overcurrent trip do NOT exist as identified on Attachment 5.

ATTACHMENT 5 Page 1 of I 41 GOV E Bus Lockout Indications NOTE: If a lockout indication exists, a cross-tie for the associated bus should NOT be performed until the fault is analyzed.

2. IF 41 60V E buses are NOT cross-tied AND it is desired to cross-tie 480V E buses, THEN PERFORM the following:

a. OBTAIN permission from both Units CRS to close El the 480V E bus cross-tie breakers.

32. 271000 1 Unit One was operating at rated power with AOG-HCV-1 02, AOG System Bypass Valve, control switch in AUTO.

Subsequently, several annunciators began alarming, including:

UA-03 4-2, Process Off-Gas Rad Hi-Hi UA-03 5-4, Process OG Vent Pipe Rad Hi-Hi UA-48 5-2, AOG System Disch Rad High UA-48 6-2, AOG Building Radiation High Which one of following annunciators is also triggered by the same radiation monitor that will prevent the AOG System Bypass Valve from being manually opened?

A. UA-03 4-2 B. UA-03 5-4 C. UA-48 5-2 D. UA-48 6-2 Answer: A K/A: 271000 Offgas System Ki .11 Knowledge of the physical connections and/or cause effect relationships between OFFGAS SYSTEM and the following: tStation radioactive release rate (CFR: 41.2 to 41.9 / 45.7 to 45.8)

RO/SRO Rating: 3.1/3.6 Objective: LOl-CLS-LP-030-A, Obj. 7c Describe the interrelationships between the following systems and the Condenser Air Removal/Augmented Offgas system: Process Radiation Monitoring.

Reference:

None Modified question from 10-1 NRC exam. (272000_i) The previous question focused on the time delay while this question focuses on the signal.

Cog Level: Low Explanation: Steam Jet Air Ejector offgas is monitored for radiation levels and the output is provided to the off-gas system logic for alarms and automatic actuation signals. A number of signals, including off-gas radiation levels, will cause off-gas system actuations in the form of bed bypass, bypass closure, and bed isolation. If SJAE offgas radiation reaches a preset value, a 15 minute timer is initiated. If the condition still exists after the timer times out, the AOG Bypass Valve (HCV-1 02) will receive a closed signal to ensure all offgas flow is being routed through the AOG filters and charcoal beds. This ensures proper filtration of the off-gas and prevents the release of radioactive materials to the environment.

iI::r Distractor Analysis:

Choice A: Correct Answer, see explanation Choice B: Plausible because this is a radiation alarm associated with the Off Gas system.

Choice C: Plausible because this is a radiation alarm associated with the Off Gas system.

Choice D: Plausible because this is a radiation alarm associated with the Off Gas system.

SRO Basis: N/A SD-33, Sect 3.1.4 HCV-1 02 will automatically close on:

Any combination of Hi-Hi, downscale, or mop on both of the Off-Gas line radiation monitors (15-minute Time Delay)

SD-Il 2.2 Condenser Off-Gas Radiation Monitoring System, also called the Steam Jet Air Ejector Radiation Monitoring System (Figure 11-2)

The log radiation monitors (D12-RM-K6O1A, B, and K602) are located in the process radiation monitoring instrument rack (Hi 2-P604) in the electronic equipment room. The instrument provides personnel with a front panel display indication of the detected radiation level and supplies this radiation level signal to the system recorder for permanent record. The log radiation monitor also drives four trip circuits which actuate annunciators, initiates the off-gas timer which, after 15 minutes, initiates closure of 1(2)AOG-HCV-i 02 if the radiation level becomes excessively high.

Unit 1 APP UA-034-2 Page 1 of 2 PROCESS OFF-GAS RAD HI-HI AUTO ACTIONS

1. Process off-gas timer is initiated if both channels are affected

2. WHEN process off-gas timer has timed out (15 minutes). if open, the following valves close:

- AOG SYSTEM BYPASS VALVE AOG-HCV-102

- OFF-GAS FILTER HOUSE LOOP SEAL RESERVOIR DRAIN VALVE, 1-OG-Sv-4go7

J$

Unit 1 APP UA-03 5-4 Page 1 of 2 PROCESS OG VENT PIPE RAD HI-H AUTO ACTIONS

1. Reactor Building Ventilation System trips and isolates

2. Standby gas treatment trains start

3. Group 6 isolation valves close Unit 1 APP UA-48 5-2 Page 1 of 2 AOG SYSTEM DISCH RAD HIGH NOTE: Inoperability of this annunciator will result in an ODCM Required Compensatory Measure.

AUTO ACTIONS NONE Unit 1 APP UA-48 6-2 Page 1 of 2 AOG BUILDING RADIATION HIGH AUTO ACTIONS NONE

33. 272000 1 Which one of the following identifies the power supply to the Main Stack Radiation Monitor?

A. Powered from Unit One UPS ONLY.

B. Powered from Unit Two UPS ONLY.

C. Normally powered from Unit One UPS, can be transferred to Unit Two UPS.

D. Normally powered from Unit Two UPS, can be transferred to Unit One UPS.

Answer: D K/A: 272000 Radiation Monitoring System K2.03 Knowledge of electrical power supplies to the following: Stack gas radiation monitoring system (CFR: 41.7)

RO/SRO Rating: 2.5/2.8 Objective: CLS-LP-1 1.0, Obj. 2e State the purpose of the following major components of the Process Radiation Monitoring System: Main Stack Radiation Monitor

Reference:

None Cog Level: Low Explanation: From 01-50.5, U2 UPS is the normal power supply while Ui UPS is the alternate power supply Distractor Analysis:

Choice A: Plausible because UI is the alternate power supply to the Main Stack Radiation Monitor.

Choice B: Plausible because U2 is the normal power supply to the Main Stack Radiation Monitor.

Choice C: Plausible because Ui is the alternate power supply and U2 is the normal.

Choice D: Correct Answer, see explanation SRO Basis: N/A

From 01-50.5 ATTACHMENT 3 Page 2 of 8 120V UPS Distribution Panel 2-2A Load Summary Load: 120V UPS Distribution Panel 2-2A-UPS (HG4)

Location: Control Building 23 SW Drawing

Reference:

F-03027 Upstream Power Source: Primary 5OKVA UPS 2A Power Supply DC Source: 2SQVDC Distribution Swbd 2A, Compt GJ5 AC Source: 480V Motor Control Center 2CAI Compt C07 Alt Source: Standby 5OKVA UPS 28 Power Supply Standby 5OKVA UPS 28 Power Supply DC Source: 25OVDC Distribution Swbd 28, Compt GL5 AC Source: 48W Motor Control Center 2GB, Compt C65 Alt Source: 46W Motor Control Center 2CB, Compt C79 CKT LOAD DESCRIPTION EFFECTS ON LOSS OF POWER 6 Main Stack Radiation Monitor Will receive a full Group & isolation on 2-D12-RM-30S and Sample Detection Skid Unit I and Unit 2 (ar Group 6 valves wi I Normal Power Supply (TS 3.3.6.1, TRM 3A close, reactor buiding will isolate, SBGT ODCM 7.3,2) tr&ns will start due to stack rad monitor trip signa. 1/2IJA-03-6-3 and 1-UA-OS-6-1O will alarm. The stack rad monitor trip signals can be overridden for each Unit by taking 12-CAC-CS-5519 to OVERRIDE. Transfer switches 2-UPS-TRF-LOS (Control Building U-2 Cable Spread area) and 2-UPS-TRF-LDT (Stack Bad Monitor Building) may be used to restore power to the stack rad monitor from UPS Distribution Panel 1-lA-UPS, circuit #6. if desired.

ATTACHMENT 2 Page 1 of 7 120V UPS Distribution Panel 1-IA Load Summary Load: 120V UPS Distribution Panel 1-IA-UPS ([1G4)

Location: Control Building 23 NW Drawing

Reference:

F-90098 Upstream Power Source: Primary 5OKVA UPS IA Power Supply DC Source: 25OVDC Distribution Swbd IA, Compt GJS AC Source: 48W Motor Control Center ICA, Compt C07 Alt Source: Standby 5OKVA UPS lB Power Supply Standby 5OKVA UPS 18 Power Supply DC Source: 25OVDC Distribution Swbd 18, Compt GL5 AC Source: 48W Motor Control Center ICB, Compt C55 Alt Source: 480V Motor Control Center ICB, Compt C69 CKT LOAD DESCRIPTION EFFECTS ON LOSS OF POWER 6 Main Stack Radiation Monitor No effect if transfer switches 2-D12-RM-8OS and Sample Detection Skid 2-UPS-TRF-LOS (Control Building U-2 Alternate Power Supply Cable Spread area) and 2-UPS-TRF-LOT (Stack Bad Monitor Building) are selected to Normal. See Unit 2 UPS Distribution Panel 2-2A-UPS, circuit #6 for effects of loss of normal power to the stack radiation monitor

34. 288000 1 Unit Two is operating at rated power when an unisolable steam leak occurs in the turbine building.

Which one of the following choices completes the statement below?

A required action lAW RRCP is to (1) to (2)

A. (1) place Turbine Bldg Ventilation in the recirculation line-up (2) maintain the Turbine Building at a negative pressure B. (1) place Turbine Bldg Ventilation in the recirculation line-up (2) provide filtration of the release C. (1) start an additional Turbine Building Ventilation Exhaust Fan (2) maintain the Turbine Building at a negative pressure D. (1) start an additional Turbine Building Ventilation Exhaust Fan (2) provide filtration of the release Answer: B K/A: 288000 Plant Ventilation Systems K5.01 Knowledge of the operational implications of the following concepts as they apply to PLANT VENTILATION SYSTEMS: Airborne contamination control (CFR: 41.7 / 45.4)

RO/SRO Rating: 3.1/3.2 Objective: LOI-CLS-LP-300N, Obj. 19 Given plant conditions and OEOP-04-RRCP, determine the following: Required actions to be taken

Reference:

None Modified question from the 10-2 NRC exam (Radiation Control_20) The previous question asked if the AOP and EOP were performed together and if once through needed to be secured. This question asks the required action and why.

Cog Level: Low Explanation: lAW RRCP if once through is in service TB Vent should be placed in Recirc mode to be able to monitor the release. Starting an additional fan may help with keeping the TB at a negative pressure to limit the release but is not an action in the procedure.

Distractor Analysis:

Choice A: Plausible because this is an operating mode of TB Ventilation and although the recirc mode will keep the TB at a negative pressure during normal operation it is not assured under these conditions.

Choice B: Correct Answer, see explanation Choice C: Plausible because starting an additional exhaust fan would increase the negative pressure in the TB but would also increase the amount of unfiltered exhaust from the building.

Choice D: Plausible because starting an additional exhaust fan would increase the negative pressure in the TB. In once through ventilation, the exhaust is not filtered.

SRO Basis: N/A TURB BLDG)

MOIITORAND CONTROL RAtOAC1IVTTV RE LEASED FROMThETLMB BLDG VENT ma TIJRB BLDG VENTILATION

.NONCE.THICUGN UNEUP..

_--- RRFrB-o3 IVES EIISRE1IJRRBISG VEW1LATION IN SERVICE IN ICIRcIJL&TION UNEIJP

• 1 RRflBO4 I IFWNILEEXEcIJTINGThE N FOLLOWING STEPSTIJRB BLDG VENTILATION IS SHUTDOWN TIfl RESTART TURE BLDG VENTILATION IN REQRCUL&TION IJILUP j RRITD.05 Operation of the Turbine Building Ventilation in the recirculation lineup helps to improve Turbine Building accessibility. In addition, since both units share a common Turbine Building airspace, if the building is intact, removing Turbine Building Ventilation from the once through lineup will terminate a large unfiltered volume discharge flow path for a leak on either unit. Due to the normal Turbine Building Air Filtration Unit and WRGM operational requirements when in once through lineup, at least one Turbine Building Air Filtration Unit and WRGM will be in service providing a monitored and filtered discharge flowpath

35. 290001 1 A microburst thunderstorm with high wind speeds has caused Reactor Building static pressure to lower.

The control room receives annunciator Rx Bldg Static Press Duff-Low.

Which one of the following identifies the response and or actions necessary to maintain normal Reactor Building differential pressure?

Reactor Building ventilation:

A. supply and exhaust fans will trip requiring a manual start of the SBGT system.

B. supply and exhaust fans will trip and the SBGT system will AUTO start.

C. exhaust fan vortex dampers will throttle open.

D. supply fan vortex dampers will throttle closed.

Answer: D K/A: 290001 Secondary Containment A3.02 Ability to monitor automatic operations of the SECONDARY CONTAINMENT including:

Normal building differential pressure: Plant-Specific (CFR: 41.7 I 45.7)

ROISRO Rating: 3.5/3.5 Objective: LOl-CLS-LP-037.1, Obj. 7-Describe how the Reactor Building Ventilation System maintains a negative differential pressure between the Reactor Building and outside atmosphere.

Reference:

None Question is modified from an 04 NRC exam. (288000_2)

Cog Level: Low Explanation: With high wind conditions d/p can drop, the supply fans have vortex vanes that will throttle to maintain d/p. The exhaust fans have the vortex vanes disabled. The alarm point for duff-low is 0.1 inches of water. If static pressure increased to 4 inches of water the RB fans would trip.

Several of the RB trip signals are also auto start signals to SBGT but building static pressure is not one.

Distractor Analysis:

Choice A: Plausible because if the static pressure increased to four inches water, this would be correct.

Choice B: Plausible because if the static pressure was lower the fans would trip and several of the RB trip signals are also auto start signals to SBGT.

Choice C: Plausible because the exhaust fans do have vortex dampers but they are disabled.

Choice D: Correct Answer, see explanation SRO Basis: N/A

w -

The fan discharge dampers are operated by the fan control switch and are a permissive for fan start. The fan intake vortex vanes are automatically positioned as required by a differential pressure controller to maintain a preset negative pressure between Reactor Building and outside atmospheric pressure. To start the supply or exhaust fans the Reactor Building supply fl exhaust isolation dampers must be full open. The following damper SD-37.1 Rev.13 Page13of7O The fan discharge dampers are operated pneumatically by the fan control switch and are a permissive for fan start. The exhaust fans have intake vortex vanes but they have been disabled at the full open position.

Unit 2 APP tTh-OS 6-7 Page 1 of 1 P.R BLDG STATIC PRESS DIFF-LOW AUTO ACTIONS NONE CAUSE 1.. bw negative pressure differential in the Reactor Building.

2. High wind speeds 3, Circuit iaalfunction.

DEVIE/SETPOINrS Pressure Differential Switch 0.1 inches water YA-PDS-1SOB I

36. 295001 1 Unit One is at rated power.

Unit Two is at 48% power in single recirculation loop operation.

Which one the following choices completes the statement below concerning the Minimum Critical Power Ratio (MCPR) safety limit for Unit One and Unit Two?

MCPR shall be greater than or equal to (1) for Unit One and (2) for Unit Two.

A. (1) 1.11 (2) 1.12 B. (1) 1.11 (2) 1.13 C. (1) 1.12 (2) 1.12 D. (1) 1.12 (2) 1.13 Answer: B K/A: 295001 Partial or Complete Loss of Forced Core Flow Circulation G2.02.22 Knowledge of limiting conditions for operations and safety limits. (CFR: 41.5/43.2/45.2)

RO/SRO Rating: 4.0/4.7 Objective: LOI-CLS-LP-200-B, Obj. 3 - State each TS Safety Limit and discuss the basis for each of the Safety Limits

Reference:

None Cog Level: Low Explanation: The MCPR safety limit for both Units in two loop operation is 1.11. The MCPR safety limits for Single Loop Operations (SLO) for Ui is 1.12 and for U2 is 1.13.

Distractor Analysis:

Choice A: Plausible because 1.12 is the SLO safety limit for Unit 1.

Choice B: Correct Answer, see explanation Choice C: Plausible because 1.12 is the SLO safety limit for Unit 1 and the TLO limit is the same for both Units.

Choice D: Plausible because 1.12 is the SLO safety limit for Unit 1 and 1.13 is the SLO safety limit for Unit 2

SRO Basis: N/A

Unit 1 TS:

2.1 SLs 2.1.1 Reactor Core SLs 2.1 .1.1 With the reactor steam dome pressure <785 psig or core flow < 10%

rated core flow:

THERMAL POWER shall be 23% RTP.

2.1.1.2 With the reactor steam dome pressure 785 psig and core flow 10%

rated core flow:

MCPR shall be 1.11 for two recirculation loop operation or 1.12 for single recircufation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

Unit 2 TS:

2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10%

rated core flow:

THERMAL POWER shall be 23% RTP.

2.1.1.2 With the reactor steam dome pressure 785 psig and core flow 10%

rated core flow:

MCPR shall be 1.11 for two recirculation loop operation or 1.13 for single recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be 1326 psig.

37. 295002 1 Unit Two is at rated power with the following plant conditions:

SJAE Train A is in full load operation CWlPs,_BancLtLare in operation Circ Isol Valve Mode Selector Switch is in D position The following alarms are received:

CW Screen A Duff-High or Stopped CW Screen Duff Hi-Hi CW Pump A Trip Exhaust Hood A Vacuum Low Exhaust Hood B Vacuum Low Which one of the following identifies the action that is required lAW OAOP-37.0, Low Condenser Vacuum?

A. Start the Mechanical Vacuum Pumps.

B. Place SJAE A and B Trains in half load operation.

C. Starting of CWIP C and it is limited to two consecutive attempts D. Restarting of CWIP A and it is limited to two consecutive attempts Answer: C K/A: 295002 Loss of Main Condenser Vacuum G2.04.45 Ability to prioritize and interpret the significance of each annunciator or alarm.

(CFR: 41.10 /43.5 / 45.3 I 45.12)

ROISRO Rating: 4.1/4.3 Objective: LOI-CLS-LP-026, Obj. 4j Given the plant conditions and one of the following events use plant procedures to determine actions required to control and/or mitigate the consequences of the event: Loss of Vacuum.

Reference:

None Cog Level: High Explanation: A trip of a running CWIP wouldresult in degraded vacuum. AOP-37.0 directs start of an available pump. Standby pumps (at ambient) are allowed two consecutive starts.

la

-p Distractor Analysis:

Choice A: Plausible because mechanical vacuum pumps are used to establish condenser vacuum during startup.

Choice B: Plausible because placing additional steam jets in operation could help with restoration of vacuum if the cause were a malfunctioning steam jet. AOP-37 Choice C: Correct answer, see explanation.

Choice D: Plausible because a restart of a tripped CWIP could be performed but a running pump is limited to one restart.

SRO Basis: N/A OAOP-37.0 CAUTION Only two consecutive starts with no time interval between starts is permitted for each CWIP with the pump motor at ambient temperature. With the pump motor at rated temperature. only one consecutive start is allowed.

3.2.3 IF a CWIP has tripped with reactor power greater than El 80%, AND a pump can be started within 5 minutes, THEN START an available CWIP as needed to maintain condenser vacuum.

3.2.11 IF one SJAE train is in service in FULL LOAD AND the El train appears to have malfunctioned, THEN PLACE standby SJAE train in service in FULL LOAD AND SHUTDOWN malfunctioning train in accordance with I (2)OP-30.

3.2.12 IF two SJAE trains are in service in HALFLOAD AND El one train appears to have malfunctioned, THEN TRANSFER the other SJAE train to FULL LOAD AND SHUTDOWN malfunctioning train in accordance with 1 (2)OP-30.

38. 295003 1 Unit Two was operating at 50% power when an electrical transient occurred.

The BOP operator observes the following electrical indicationsçffr the transient:

SAT TO BUS 2!)

2C4 ___J I [D TO SUD

__ T u D TO E3 UAT TO BUS 2D 2mt 1

I 1ç ISw;H1.tfI I- I iRc I IIRoc I 11 r)

II H I 2C TO S SAT TO BUS 2C -

2

- jI BUS 20 TO E4 1 tJAT flt7 TO BUS 20 I I il I I Ita(] I I V I 5AT TD Which one of the following identifies the cause of the Unit Two electrical transient?

A. A UAT Lockout B. A SAT Lockout C. A Generator Lockout D. A 2C BOP Bus Lockout Answer: D

IiJJ A £ JiTtJIIIILJ1IIII!I K/A: 295003 Partial or Complete Loss of A.C. Power AA2.01 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: Cause of partial or complete loss of A.C. power (CFR: 41.10 /43.5 / 45.13)

RO/SRO Rating:t3.4/3.7 Objective: LOI-CLS-LP-050-B, Obj. 20 Given plant conditions, determine if lockouts will occur for the following supply breakers:

Reference:

None Cog Level: High Explanation: While operating)the typical lineup would be for the UAT to supply 2C and 2D busses while the SAT will supply 2B Bus. The 2C and 2D busses normally would transfer to the SAT feed on a scram/loss of the UAT. On a 2C bus lockout)the bus will not auto transfer to the SAT feed.

Distractor Analysis:

Choice A: Plausible because the normal feeder breaker from the UAT to 2C bus is open.

Choice B: Plausible because the SAT feeder breaker to 2C/2D Busses are open.

Choice C: Plausible because the normal feeder breaker from the UAT to 2C bus is open.

Choice D: Correct Answer, see explanation SRO Basis: N/A SD-SO. 1 Figure 10 shows the basic protective relaying scheme for the C1D Buses. Each incoming line from the SAT and UAT is monitored for overcurrent and undervoltage. Actuation of the overcurrent relay will trip the associated bus incoming supply breaker by actuation of a lockout relay. The lockout relay Will also prevent closure of the other bus supply breaker in the event of a bus overcurrent. Actuation of the undervoltage relay will trip the associated bus incoming feeder breaker but does not lock out the bus from being supplied from the other feeder.

IIPIJhIIIIIIIILIW1LLi $KL1iA fL

39. 295004 1 Unit One is operating at rated power when a loss of 125 VDC distribution panel 1 B occurs. No operator action has been taken.

Which one of the following choices completes the statement below?

The lB CRD pump can be shutdown (1)

If the CRD pump has an overcurrent condition, the 1 B CRD pump breaker (2) trip.

A. (1) from the RTGB or locally at the breaker (2) will B. (1) from the RTGB or locally at the breaker (2) will NOT C. (1) locally at the breaker ONLY (2) will D. (1) locally at the breaker ONLY (2) will NOT Answer: D K/A: 295004 Partial or Complete Loss of D.C. Power AAI .03 Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: A.C. electrical distribution (CFR: 41.7 / 45.6)

RO/SRO Rating: 3.4/3.6 Objective: LOI-CLS-LP-051, Obj. 7b Given plant conditions, determine the effect that a loss of DC power will have on the following: AC Electrical Distribution System

Reference:

None Cog Level: High Explanation: DC provides the control power source for 4kV. Electrically operated 480V emergency bus breakers cant be remotely operated without control power, but willtrip on overcurrent. 4kV breakers cannot be operated except manually, and will not trip even on fault conditions without control power. /

1L17MkI1Wii&iiLJ ILiL11III1iIII!IIVI Distractor Analysis:

Choice A: Plausible because DC control power is required for operation of 4KV breakers from the RTGB and it can be operated locally. This is different from 480V emergency bus breakers which would still trip on overcurrent if DC is lost.

Choice B: Plausible because DC control power is required for operation of 4KV breakers from the RTGB and it can be operated locally.

Choice C: Plausible because with a loss of DC control power, the breaker can only be operated locally.

Only the 480V emergency Bus breakers will trip on an overcurrent condition without control power.

Choice D: Correct Answer, see explanation SRO Basis: N/A SD-51

2. AC Electrical Distribution DC provides the control power source for 480 VAC Substations and larger sMtchgear Electrically operated 480 VAC breakers cannot be remotely operated without control power, but will trip on overcurrent.

4160 VAC and larger breakers cannot be operated except manually, and will not trip even art fault conditions without control power.

Several buses have an alternate control power source, but require manual transfer.

001-50 ATTACHMENT lB Page 3 o26 PANEL lB LOCATION: NORMAL SUPPLY: ALTERNATE SUPPL1 Referenoe Drawmg: LL-30024-5 Dieeel Building, SO, E-2 Cell Swllchboard lB N/A CKT LOAD E19ECT

1. or tesno CdD I5, soars sore 3 450 VAC 1:ctgear 56 Cor:rcl Power 1. ..oss or rorma cor:rl powelo . ra, ano orcas e ore3eers Altenlate e itorr i5, L 5. Cc:rcI powe s marual t-araerred.

2. or eaker :ndbalil an co:rol rcm oIl ur :s

3. or lcuI A0O ocitol.

& OYer-curret trips will ELU lLnc.:rcn.

5. o (lie na, breaker tp tu,ceor ,fla., :ie .:ctec 41(V breaker orer s or Icr a

ransformer r:gr ripea1ure IT:.

5. Ope.ailcn treakers car be cerrorrrel narualic.

Al AhMNi 1 Page 4 cr26 PANEL lB LOCATION: NORMAL SUPPLY: ALTERNATE SUPPLY:

Reference Drawing: LL-30024-5 Dieeei Building, 50. E-2 Cell Swi.chboard lB N/A CCT LOAD EFFECT 17 swccrearE2controiFower I. AflelaIeielrclr Fane: scet. I.

2. AlLernatepwe-rruslberna1uyagrea.

.3. or oper. cJoe ard :r fLflccn ol a: afTecteu 4V breakers. ee OC-SO.2 r

& of Iccal ani remote breaker pca:l, nains. Mear positon ri c.ars are aiafable:

5. AOD regu:renents rr attecca.1 bale i rojrolici.

JLJJIIIII&L

40. 295005 1 Unit Two is operating at rated power.

At 12:15:00 the following annunciators are received:

Stat Coolant Inlet Flow-Low Loss of Stat Coolant Trip Ckt Ener Reactor power has been lowered lAW OENP-24.5, Reactivity Control Planning.

At 12:16:00 Main Generator amperes are 17,814 amps.

Assuming no further operator action, which one of the following indicates the expected plant response?

The Main Generator will trip at:

A. 12:17:00.

B. 12:18:00.

C. 12:18:30..

D. 12:fr30.

Answer: C K/A: 295005 Main Turbine,Trip AK2.04 Knowledge of the interrelations between MAIN TURBINE GENERATOR TRIP and the following: Main generator protection (CFR: 41.7 / 45.8)

RO/SRO Rating: 3.3/3.3 Objective: LOI-CLS-LP-027.2 Obj. 05- Given plant conditions, determine if the Loss of Stator Cooling trip circuitry should be energized.

Reference:

None Cog Level: High Explanation: Generator amps above 6018 amps arms the 3.5 minute timer. With generator amps above 19, 814 amps, both timers (2.0 minute and 3.5 minute) are armed and will initiate at 12:15:00 on the low flow condition and timer energized annunciator. Since both timers start at the same time, the elapsed times are not cumulative (i.e. total time will be 3.5 minutes NOT 5.5 minutes). Lowering generator amps to below 18,823 amps will drop out the two minute timer before it can actuate. The 3.5 minute timer continues to run unless amperes are lowered below 5717 amps. With the conditions described and no further operator action, the 3.5 minute timer will initiate a generator trip at 12:18:30.

WW rIEE

J1 -r - . sILJLtt.. M& aS&.L Distractor Analysis:

Choice A: Plausible because the 2.0 minute timer would initiate a generator trip at 12:17:00 if generator amps were above the timer dropout setpoint of 18,823 amps.

Choice B: Plausible because the calculated trip time from the 2.0 minute timer would be at 12:18:00 if generator amps were above the timer dropout setpoint of 18,823 amps AND the candidate used 12:16:00 as the calculation start time.

Choice C: Correct answer, see explanation.

Choice D: Plausible because the calculated trip time from the 3.5 minute timer would be at 12:19:30 if generator amps were above the timer dropout setpoint of 5717 amps AND the candidate used 12:16:00 as the calculation start time.

SRO Basis: N/A SD-27.2 A trip circuit is installed on both units to automatically protect the generator on a loss of stator cooling. Additionally stator cooling water conductivity limits are provided for continued operation of the generator. On a loss of stator water cooling flow, the main turbine must be tripped after 60 minutes if initial conductivity was less than 0.5 pmhosfcm. If initial conductivity was greater than 0.6 iimhos/cm, then the turbine must be tripped within 3 minutes.

The automatic Stator Coolant Trip Circuit initiates whenever generator armature amperage is greater than 6018 amps, and any of the following conditions occur:

• System flow is less than 423 444 gpm

• System pressure is less than 37.6 (U-I), 36.3 (U-2) +/-2 psig

• Coolant temperature exceeds82-84C Once actuated, the trip circuit relays will energize causing actuation of a 2 minute timer if amps are above 19814 and a 3 1/2 minute timer if amps are above 6018. The trip circuit relays, when energized, will also actuate an annunciator, UA-02 1-9, LOSS OF STATOR COOLANT TRIP CIRCUIT ENER, and open a contact in the EHC load reference motors increase circuitry to prevent load increase.

The 2 minute timer will trip the turbine when timed out if stator amps are not <18823 amps. The 3.5 minute timer will trip the turbine when timed out if stator amps are not <5717 amps. The trip circuit, once energized, will remain energized until the timer(s) actuate, until the amperage reduction is achieved, or until the degraded condition is corrected.

rr r

a...-.. I .W 1LtCSA aar;ntsa p 2APP-UA-02 Unit 2 APP-t-02 1-9 Page 1 of 3 LOSS OF STAT COOLMT TRIP cR7 ENER AUTO ACTIONS

1. Two timers are actuated which will initiate a turbine trip.

CAUSE 1, Stator coolant outlet temperature is high when armature current is above 6018 amps.

2. Stator coolant flow is low.

3. Stator coolant pressure is low.

4. Circuit malfunction.

ACTIONS (Continued)

CAUTION If etator current is not less than 18,823 amps after 2.0 minutes or less than 5,717 amps after 3,8 minutes, the turbine will trip.

3. Reduce generator load per OENP-24.S and OQP-l2. If it is expected that the load reduction rate will not satisfy the 2.0 and 3.5 minute timers, then scram the reactor and trip the turbine.

Enter 2EOP-RSP, Reactor Scram Procedure.

JLA1I.

41. 2950061 Following a scram, which one of the following conditions meet the definition of Shutdown Under All Conditions Without Boron? j

_AtLrods inserted except:

A. nine rods at position 02.

B. two rods which are at position 04.

C. one rod at position 02 and one rod at position 24.

D. ten rods at position 02 and one rod at position 48.

Answer: A K/A: 295006 SCRAM AK1 .02 Knowledge of the operational implications of the following concepts as they apply to SCRAM:

Shutdown margin (CFR: 41.8 to 41.10)

ROISRO Rating: 3.4/3.7 Objective: LOl-CLS-LP-300-C, Obj. 07 Given plant conditions and the Reactor Scram Procedure, determine if branching into the Level/Power Control Procedure is required.

Reference:

None Cog Level: Low Explanation: From 001-37.3, Reactor Scram Procedure Basis Document: Positive confirmation that the reactor will remain shut down under all conditions is best obtained by determining that no control rod is withdrawn beyond the Maximum Subcritical Bank Withdrawal Position, of position 00. Table 1 has been added to provide a listing of those conditions for the reactor being shutdown under all conditions without boron. This was added specifically for condition where 10 control rods could be withdrawn to position 02 as long as no control rod is withdrawn beyond position 02.

Distractor Analysis:

Choice A: Correct answer, see explanation Choice B: Plausible because choice provides a possible configuration of control rods following a reactor scram.

Choice C: Plausible because choice provides a combination of allowable circumstances but together they do not satisfy the criteria.

Choice D: Plausible because choice provides a combination of allowable circumstances but together they do not satisfy the criteria.

SRO Basis: N/A

LtLIW aCt a LaaJ dTIEsafl TABLE I SNUTDOWN WITNOUT BORON QNLY ONE CONTROL ROD NOT rULLY INSERTED NO MONt THAN IC CONTNOL NODS WITHONAWN TO POSITION 42 AND NO CONTROL ROD WITHDRAWN BEYOND POSITION 02 AS DETERMINED BY REACTOR ENGINEERING C

(ERFDRM RERCTQR VESSEL 3

GO TO I CONrROL PROCEDURE LEVELIPOWER CONTROL I EOP RVCP) AND (EOP-4l. LPC$

I EXECUTE IT CONCURRENTLY WITH ThYS PROCEDURE From 001-37.3, Reactor Scram Procedure Basis Document:

Positive confirmation that the reactor will remain shut down under all conditions is best obtained by determining that no control rod is withdrawn beyond the Maximum Subcritical Bank Withdrawal Position, of position 00. Table I has been added to provide a listing of those conditions for the reactor being shutdown under all conditions without boron. This was added specifically for condition where 10 control rods could be withdrawn to position 02 as long as no control rod is withdrawn beyond position 02.

V in

I, i !IIlIIIIlIlIIIIIf

42. 295009 1 Unit Two is at rated power when the following conditions are observed:

Reactor water level is 180 inches Steam Flow indicates 12.76 Mlbs/hr Feed Flow indicates 12.05 Mlbs/hr RFPT speed indications exist at P603 are:

RFP A Speed REP B Speed L_2RFA517525 J 2RFBSI732 RFPA RFPB With the conditions continuing to degrade, which one of the following actions is required to restore and maintain level in-band lAW OAOP-23.0, Condensate/Feedwater System Failures?

A. The master control 2-C32-SIC-R600 has failed in auto, place the Master Feedwater Controller in MANUAL and restore normal level.

B. Place the A RFP Feedwater Controller in MANUAL and restore normal level.

C. Place the B RFP MANUAL/DFCS switch in MANUAL and restore normal level.

D. Place both RFP Feedwater Controllers in MANUAL and balance the RFP speed.

Answer: B

I1LIIIIIIIII!2d1I7 L JaUI A_ I!IICIPIIIIIIIU K/A: 295009 Low Reactor Water Level AA2.02 Ability to determine and/or interpret the following as they apply to LOW REACTOR WATER LEVEL: Steam flow/feed flow mismatch (CFR: 41.10 /43.5 / 45.13)

RO/SRO Rating: 3.6/3.7 Objective: LOI-CLS-LP-032.2, Obj. 11 Describe how feed flow is balanced if a flow mismatch exists and one RFPT will not respond to electrical control

Reference:

None 4

Modified question from 10-2 NRC Exam. (259002_I 3). this question was modified to provide pictures of the indications and had a failure downward in speed of a REP instead ofupwad.

Cog Level: High Explanation: With the given conditions steam flow is greater than feed flow resulting in a lowering level.

The A REP has failed in automatic operation causing the low level condition and the B RFP is trying to raise flow due to compensate. Manual control of the A REP is necessary to arrest the trend and restore level. If B RFP had failed high and A REP was compensating, the level would high, not low.

Distractor Analysis:

Choice A: Plausible because Master level control is in control now, placing it in MANUAL will not correct the failure of A RFP to control.

Choice B: Correct Answer, see explanation Choice C: Plausible because B REP is operating correctly attempting to raise flow to compensate for the low level.

Choice D: Plausible because placing both REPs in Manual and balancing rpms will NOT control water level.

SRO Basis: N/A

fl----r- wttW 43 S -s-- - g tt JtJa ES-From AOP-23:

3.0 OPERATOR ACTIONS 3.1 Immediate Actions 3.11 IF automatic level control will NOT restore normal reactor fl vessel level. THEN MANUALLY CONTROL reactor feed pumps to restore normal leveL 4.0 GENERAL DISCUSSION High or low water level during power operation is an abnormal condition that could result in major damage to the HPCI, RCIC, RFP, and main turbines, jet pumps, and recirculation pumps.

High water level may result in excessive moisture carryover, causing erosion of turbine blading. Low water level may cause steam carryunder that can lead to cavitation in recirculation and jet pumps and excessive core internal vibration.

L Automatic level control is preferred. Manual control should be taken only if operation in automatic is unsafe or would cause unnecessary transient& If two feed pumps are in operation and the feed pumps have opposite demand signals, take manual control of the pump whose demand signal coincides with the direction of the level change and attempt to control water level.

t

43. 2950101 A LOOP has occurred on Unit Two with the following plant conditions:

HPCI Failed RCIC Under Clearance SLC Injecting with Demin Water CRD Flow maximized lAW SEP-09 ADS Inhibited Reactor Water Level 36 inches and stable Reactor Pressure 950 psig and stable Suppression Pool Level -26 inches Drywell Pressure 2 psig and rising slowly Drywell Temperature 180°F and rising slowly Which one of the following choices identifies the required action lAW PCCP?

A. Initiate Drywell Sprays ONLY lAW SEP-02.

B. Start all available DW Coolers lAW SEP-10.

C. Perform Emergency Depressurization lAW RVCP.

D. Initiate Suppression Pool Sprays ONLY lAW SEP-03.

Answer: B K/A: 295010 High Drywell Pressure AKI .03 Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL PRESSURE: TemperatUre increases (CFR: 41 .8 to 41.10)

RO/SRO Rating: 3.2/3.4 Objective: LOI-CLS-LP-300L, Obj. 11 Given PCCP, which steps have been completed and plant parameters, determine the required operator actions. (LOCT)

Reference:

None bank question last used on 07 NRC Exam.

Cog Level: High Explanation: lAW SEP-I 0 if a low water condition is present and no actual loca is present then DW coolers can be jumpered and restarted. PCCP also has direction to initiate sprays but under these conditions it cannot be procedurally performed. ED would be required if cannot maintain less than 300 degrees in the DW.

&  ? LJLJItLIFI1E1 Distractor Analysis:

Choice A: Plausible because DW sprays can be performed in the temperature leg of PCCP, although with pressure at 2 psig it can not be performed..

Choice B: Correct answer, see explanation Choice C: Plausible because ED can be performed if temperature can not be restored and mantained less than 300 degrees.

Choice D: Plausible because the pressure leg of PCCP directs torus sprays prior to exceeding 11.5 psig although the conditions will not allow the procedure to be performed.

SRO Basis: N/A From SEP-10 24 IF:

RO: - Directed to defeat the drywell cooler LOCA lockout logic due to low reactor water level, AND RO: - Actual LOCA conditions do NOT exist in the drywell, AND RO: - RBCCW is operating and supplying the drywell, THEN RO: - PERFORM Section 4 on page 8 of this procedure. Lj From PCCP Temperature leg:

DWJT-*3 ow/i-150°F NOTE DRWELL COOLERS TRIPAND LOCKOUT ON A LOCA SIGNAL 01 DVWT-06 START ALL AVAILABLE -

DRYWELL COOLERS, DEFEATING DRYWELL COOLER INTERLOCKS IF NECESSARY PER CIRCUITALTERATION PROCEDuRE (EOP-01-SEP-10)

iUII1JII1!ILIII4II IIJ

44. 295016 1 Which one of the following is the reason for inserting a Scram during the performance of OAOP-32.0, Plant Shutdown From Outside Control Room?

A Scram is inserted prior to (1) to ensure (2)

A. (1) evacuating the Control Room ONLY (2) the reactor is placed in a hot shutdown condition B. (1) evacuating the Control Room ONLY (2) control of engineered safeguards systems from a backup control center can be executed C. (1) OR following Control Room evacuation (2) the reactor is placed in a hot shutdown condition D. (1) OR following Control Room evacuation (2) control of engineered safeguards systems from a backup control center can be executed Answer: C K/A: 295016 Control Room Aband ent AK3.01 Knowledge of th easons for the following responses as they apply to CONTROL ROOM ABANDONMENT: Reactor SCRAM(CFR: 41.5/ 45.6)

RO/SRO Rating: 4.1/4.2 Objective: LOI-CLS-LP-302-E, Obj. 7 Given plant conditions and entry into OAOP-32.0, Plant Shutdown From Outside Control Room, explain the basis for a specific caution, note, or series of procedure steps.

Reference:

None Cog Level: Low Explanation: The reason for a reactor scram during a control room abandonment if not performed prior to evacuating is part of Control Room design which incorporates the capability for prompt hot shutdown of the reactor per Criterion 19 of UFSAR.

Distractor Analysis:

Choice A: Plausible because AOP-32 immediate actions require inserting a manual Scram prior to evacuation, but also provides additional actions for opening RPS EPA breakers to Scram the reactor if not performed from the control room and hot shutdown is correct.

Choice B: Plausible because AOP-32 immediate actions require inserting a manual Scram prior to evacuation, but also provides additional actions for opening RPS EPA breakers to Scram the reactor if not performed from the control room and control of engineered safeguards equipment from the remote shutdown panel will NOT be required is an initial condition assumption and not the reason for inserting a Scram.

Choice C: Correct Answer, see explanation Choice D: Plausible because a scram is inserted prior to or following evacuation and control of engineered safeguards equipment from the remote shutdown panel will NOT be required is an initial condition assumption and not the reason for inserting a Scram SRO Basis: N/A

1 L J JCadttaafP2LWn 4.0 GENERAL DISCUSSION The following conditions are assumed to exist as the main Control Room becomes uninhabitable:

1. The event causing the main Control Room to become uninhabitable is assumed to be such that all operators are able to leave the Control Room and assemble at the remote shutdown panel to receive necessary keys:

headsets: and procedures, which are maintained in the shutdown panel.

(Security will maintain an extra set of keys): and to assist in the remote shutdown.

2. The Shift Manager will order a manual reactor scram, turbine trip, and main steam isolation valves closure prior to Control Room evacuation or go to the battery room and open EPA breakers for RPS MG5 and alternate power supply.

3. The emergency AC buses are energized by the preferred off-site power supply. No loss of off-site power is considered after the start of the emergency in this procedure.

4. No accident occurred concurrent with the event which required evacuation of the Control Room so that control of engineered safeguards systems from a backup control center will not be required.

5. DC services are supplied from at least one plant DC power system for each essential system or equipment required for remote shutdown.

6. All support systems are in normal lineups, operating normally, and continue to operate normally throughout the emergency.

7. Nothing abnormal in the plant between the time of Control Room evacuation and reporting to assigned stations occurs which would deteriorate conditions of the reactor to the point that immediate local action is required on reaching the station.

OAOP-32.O Rev. 48 Page 36 of 75

t LSfl ant -t,j.

4.0 GENERAL DISCUSSION This procedure provides instructions to place the plant in cold shutdown condition from outside the Control Room. The manual reactor scram and main steam isolation will automatically bring the RCIC, H PCI, and reactor relief valves into operation. This will place the reactor in hot shutdown condition. During this phase of shutdown: the suppression pool will be cooled as required by manually placing the RHR system in the suppression pool cooling mode. Reactor pressure will be controlled and core decay and sensible heat rejected to the suppression pool by steam flow to HPCI and RCIC and by dumping steam through the relief valves.

Reactor water inventory will be maintained initially by both RCIC and HPCI systems and later by the RCIC system during the pressure reduction period. This procedure will cool the reactor and reduce its pressure at a controlled rate until reactor pressure becomes so low that the relief valves will close or RCIC system will discontinue operation. This condition will be reached at 50-1 00 psig reactor pressure. One control rod drive pump will be used to maintain rod drives cool and augment the RCIC system: and till be available to maintain level after the RCIC system has been secured. At this time: RHR is manually placed in shutdown cooling to bring the reactor to a cold shutdown condition. The manual control of the equipment required for the above operation (refer to Table 1 for Units 1 and 2) is achieved by operating the control switches located on the individual breaker compartment doors. The key locked NORMAL/LOCAL selector switch is placed in LOCAL position and its START/STOP or OPEN/CLOSE local control switch is operated to control equipment required for remote shutdown

L &JIIJL1IIII

45. 2950171 On SPDS Screen 500, Radioactivity Release Control, the Off-Site Whole Body Dose Rate limit will first be exceeded (a red Alarm condition) at which one of the following values?

A. 450 mRem I Yr B. 500 mRem / Yr C. 2700 mRemlYr D. 3000 mRem/Yr Answer: B K/A: 295017 High Off-Site Release Rate AK2.08 Knowledge of the interrelations between HIGH OFF-SITE RELEASE RATE and the following: SPDS/ERIS/CRIDS/GDS (CFR: 41.7 /45.8)

RO/SRO Rating: 2.8/3.3 Objective: LOI-CLS-LP-060, Obj. 06 Given plant conditions, determine expected ERFIS/SPDS indications.

Reference:

None Cog Level: Low Explanation: The Off-Site Whole Body Dose Rate goes into prealarm (yellow) at 450 mR/Yr and goes into alarm (red) state at 500 mR/Yr. The other values are the prealarm and alarm conditions for Off-Site Skin Dose Rate.

Distractor Analysis:

Choice A: Plausible because this is the prealarm condition for Off-Site Whole Body Dose Rate Choice B: Correct Answer, see explanation Choice C: Plausible because this is the prealarm condition for Off-Site Skin Dose Rate Choice D: Plausible because this is the alarm condition for Off-Site Skin Dose Rate SRO Basis: N/A

1!LYflWAaMLAas WIII1IS

49. OFF-SITE SKIN DOSE RATE [500,826)

This graph fncorporates the Radioactive Gaseous Release Calculation whch determines off-site dose rates. If any of the values used in the equation are not available, the computer wil not substitute values as recommended n 0E&RC-2020, Section 10.3, Nob e Gas Instantaneous Release Rate Deternination The trend clot consists of the latest hstori. a bar graph ref ecting the current value, a current digital readout and limit tags. The limit tags and avalable scales are isted below.

Limit Tag Prealami Alarni TECH SPEC 2700 mremtvr 3000 mremivr SD-GO Rev. B Page 81 01 107

50. OFF-SITE WHOLE BODY DOSE RATE [500,825]

This graph ncorporates the Radioactive Gaseous Release Calculation whoh determines off-site evels. If any of the values used in the equation are not avaiab e. the computer wiF not substitute values as recommended in OE&RC-2020, Section 10.3. Noble Gas Instantaneous Release Rate Determination.

The trend olot consists of the latest hstory, a bar graph ref ecting the current value, a current digital readout and limit tags. The limit tags and available scales are t isted below.

- Limit Tag - - Alarm TECH SPEC 450 mremIyr so: mremtyr r i2

$35

4IAL L Z L1

46. 295018 1 Unit One is performing a reactor startup.

The following events occur prior to rolling the main turbine:

Bus I D experiences a fault and trips -

Unit One NSW header ruptures in the Service Water Building Unit One Service Water pumps supplying the NSW Header are manually tripped lAW OAOP-18.0, Nuclear Service Water System Failure.

Which one of the following identifies the status of the Diesel Generators?

A. ONLY DG1 is running with cooling water supplied from the Unit Two NSW header.

B. ONLY DGI is running with cooling water supplied from the Unit One CSW header.

C. DGs 1 & 3 are running with cooling water supplied to both DGs I & 3 from the Unit Two NSW header.

D. DGs I & 3 are running with cooling water supplied to DGI from the Unit One CSW header and to DG3 from the Unit Two NSW header.

Answer: C K/A: 295018 Partial or Complete Loss of Component Cooling Water AA1 .01 Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Backup systems (CFR: 41.7 145.6)

RO/SRO Rating: 3.3/3.4 Objective: AOl-CLS-LP-043, Obj. 6c Discuss the automatic functions/interlocks associated with the Service Water System: Diesel Generator Cooling Water Supply Valves

Reference:

None Cog Level: High Explanation: Divisional start signal will auto start both DG I and 3. If service water pressure upstream of the jacket water heater exchanger remains below 5.6 psig for 30 seconds then the alternate unit supply valve (in this case Unit 2) will open and the normal supply valve will close. Since DG 3 service water is normally from Unit 2, and Unit 2 service water system is intact, the normal supply from Unit 2 will remain in service. For DG 1, the Unit I service water header is depressurized due to the rupture, therefore cooling water for DG 1 will align to the Unit Two Nuclear Service Water header.

1II1Z IL I1 I L1WIIIIRIIIIILi Distractor Analysis:

Choice A: Plausible because if 1D 4160 deenergizes while UAT is energized (Unit online or in UAT backfeed), then only DG I would start on loss of E-Bus voltage. Unit two nuclear service water will automatically supply the diesel due to loss of Unit one nuclear service water.

Choice B: Plausible because if 1D 4160 deenergizes while UAT is energized (Unit online or in UAT backfeed), then only DG 1 would start on loss of E-Bus voltage. Without a casualty, conventional service water may be available to supply the nuclear header if aligned manually or aligned for auto start on the nuclear header.

Choice C: Correct answer, see explanation Choice D: Plausible because UAT is deenergized during startup prior to synchronizing the generator to the grid, If ID 4160 deenergizes while UAT is deenergized then a divisional DG start would result (DG 1 & 3).

SRO Basis: N/A From SD-39.0, Emergency Diesel Generators 2.7 Diesel Generator Service Water (Figure 39-7)

Two service water supply lines provide service water to tile tube side of each EDG set jacket water cooler. Each units Nuclear Service VIater (NSW)

System provides an independent source to all four Diesels. Diesel generator sfart and speed increase above 500 rpm opens the valve from the respective units NSW header. Should the service water pressure upstream of the jacket water heat exchanger remain below 5.6 psig for 30 seconds wrien the valve is open the alternate unit supply valve will open, then the normal supply valve will close. When the engine is shutdown and speed drops below 500 rpm the open valve will close. This switching sequence is initiated any time service water flr is lost when an EDG set is operating.

Return flow of service water from all four jacket water coolers is routed to a common return line which discharges to SW Outfall Collection Tank via an 18 CPVC line.

3.2.4 Automatic Start

3. A EDG auto start signal will be generated for EDGS I and 3 (2 and 4) if any one of the following conditions exists (Figure 39-13):

• Loss of 1C or 2C 4160 BUS will cause EDGs. 2 & 4 to start

• Loss of 1 D or 2D 41 60V BUS will cause EDGs 1 & 3 to start The loss of BOP bus is sensed by undervoltage on the secondary side of the UAT with the UAT to D(C) bus breaker open OR undervoltage on the secondary side of the SAT with the SAT to D(C) bus breaker open AND BOP bus undervoltage.

IWZL_ aj ZSCftSca&g awttafl%:S.

From OAOP-18.O, Nuclear Service Water System Failure S.O OPERATOR ACTIONS

&25 IF NSW Header pressure is low from a suspected leak or pipe rupture at an unknown location OR decreased service water pump avallabilit,, THEN PERFORM the folloving:

4. IF isolation of the above listed components did NOT increase NSW Header to greater than or equal to 40 psig, THEN PERFORM the following:

a. ASSUME NSW Header rupture has occurred. LI

b. STOP all service water pumps supplying the NSW LI Header.

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IIIIIII1IiIILJIALLII kL &1&ILiL4

47. 2950191 Which one of the following identifies an alarm signal that will initiate the Backup Nitrogen System, including the reason for initiating Backup Nitrogen System?

A. Instr Air Press-Low Ensures operabiIity-otA13S-valves and Inboard MSIVs.

B. RB Instr Air Receiver IA Press Low Ensures operability-of-ADS-valves and Inboard MSIVs.

C. Instr Air Press-Low Ensures operabil[tyoLADS-vaLves and the Hardened Wetwell Vent Valves.

D. RB Instr Air Receiver IA Press Low Ensures operability orADS-valvL.es and the Hardened Wetwell Vent Valves.

Answer: D K/A: 295019 Partial or Complete Loss of Instrument Air G2.01 .32 Ability to explain and apply system limits and precautions. (CFR: 41.10 /43.2 /45.12)

ROISRO Rating: 3.8/4.0 Objective: LOI-CLS-LP-046-A, Obj. 7d Given plant conditions, determine if the following automatic actions should occur: Nitrogen Backup Initiation LOl-CLS-LP-046-A, Obj. 14 Predict the effect that a loss or malfunction of the Pneumatic System would have on plant operation.

Reference:

None Modified from a Bank question that was used on 08 NRC Exam. (29501 8_9) modified to provide alarms instead of conditions.

Cog Level: High Explanation: The Backup Nitrogen System would supply pneumatics to SRV Accumulators, the Reactor Building to Suppression Chamber Vacuum Breaker Isolation Valves, and the Hardened Wetwell Vent Isolation Valves. RB lnstr Air Receiver pressure low is received at 95 psig which is the isolation setpoint for backup nitrogen.

1IiV Distractor Analysis:

Choice A: Plausible because alarm indicates a low air pressure condition but not the setpoint for BU Nitrogen. The inboard MSIVs are supplied from PNS during full power operations and from non-interruptible instrument air, BU Nitrogen does not supply the inboard MSIVs.

Choice B: Plausible because alarm does initiate BU Nitrogen. The inboard MSIVs are supplied from PNS during full power operations and from non-interruptible instrument air, BU Nitrogen does not supply the inboard MSIVs.

Choice C: Plausible because alarm indicates a low air pressure condition but not the setpoint for BU Nitrogen.

Choice D: Correct answer, see explanation.

SRO Basis: N/A Unit 1 APP UA-O1 1-1 Page 1 of 2 RB INSTR AIR RECEIVER IA PRESS LOW AUTO ACTIONS I. RNA-SV-5482. Hugh Preswre Bottle Rack lsoaton Valve, opens, supp yng SRVs and CAC-16 with a pneuniatic source.

Unit APP UA-Di -4 Page 1 of 2 INSTR AIR PRESS-LOW AUTO ACTIONS NONE

IAiJ Zi!L LJII3I11IIIHI

48. 295021 1 Unit Two had just been placed in Cold Shutdown when offsite power is lost.

All group isolations occur as expected.

The operators are executing OAOP-15.0, Loss of Shutdown Cooling, but are having difficulty opening inboard suction isolation valve (E11-F009).

Reactor water level is being maintained between 200-220 inches.

Which one of the following parameters must be monitored for determination of a mode change to Hot Shutdown?

A. Reactor vessel pressure.

B. Reactor bottom head temperature.

C. Reactor recirculation loop temperature.

D. RHR heat exchanger inlet temperature.

Answer: A K/A: 295021 Loss of Shutdown Cooling AA2.04 Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLI NG: Reactor water temperature (CFR: 41.10 I 43.5 I 45.13)

ROISRO Rating: 3.6/3.6 Objective: LOI-CLS-LP-307-B, Obj. ig - GP-05, Unit Shutdown: Given plant conditions, monitor cooldown rate per PT-01 .7.

LOl-CLS-LP-302-L, Obj. 03 Given plant conditions and AOP-1 5.0, Loss of Shutdown Cooling, determine the required supplementary actions.

Reference:

None Cog Level: Low Explanation: Natural circulation cannot be depended on to provide adequate flow through the bottom head region or the recirculation loops. The recirculation loop suction temperatures and bottom head temperatures therefore cannot be utilized for vessel coolant temperature monitoring for indication of boiling. Under natural circulation conditions, reactor vessel pressure must be monitored for coolant temperature determination. (AOP-15.0)

I IPL UJI1IIi6IIII11IL Distractor Analysis:

Choice A: Correct Answer, see explanation Choice B: Plausible because bottom head temperature is used in PT-01 .7, but is inaccurate for these conditions.

Choice C: Plausible because recirc loop temperatures are used in PT-O1 .7, but is inaccurate for these conditions.

Choice D: Plausible because if SDC was in service this would be an option but the group isolation has occurred.

SRO Basis: N/A OAOP-1 5 3.0 OPERATOR ACTIONS CAUTION Natural circulation can NOT be depended on to provide adequate flow throuqh the bottom head region or the recirculation loops. The recirculation loop suction temperatures and bottom head temperatures therefore can NOT be utilized for vessel coolant temperature monitoring for indication of boiling. Under natural circulation conditions, reactor vessel pressure must be monitored for coolant temperature determination, If coolant temperature was initially less than 212cF, pressure must be closely monitored for indications of a trend of rising pressure. If this trend is established, it must be assumed that 212F has been exceeded, boiling is occurring, and a mode change has taken place.

3.2.4 MONITOR reactor coolant heatup/cooldown in accordance with 1(2)PT-O1 .7 for any unexpected trends.

49. 295022 1 Which one of the following choices completes the statement below lAW OAOP-02.0, Control Rod Malfunction/Misposition, for a loss of CRD pumps?

If reactor pressure is (I) with charging header pressure less than 940 psig, immediately insert a manual reactor scram upon the receipt of the (2) HCU low pressure alarm.

A. (1) less than 950 psig (2) first B. (1) less than 950 psig (2) second C. (1) greater than or equal to 950 psig (2) first D. (1) greater than or equal to 950 psig (2) second Answer: A K/A: 295022 Loss of Control Rod Drive Pumps AK1 .02 Knowledge of the operational implications of the following concepts as they apply to LOSS OF CRD PUMPS: Reactivity control (CFR: 41.8 to 41.10)

RO/SRO Rating: 3.6/3.7 Objective: LOI-CLS-LP-008-B, Obj. 10 Given plant conditions, determine proper operator actions if no CRD pumps are operating.

Reference:

None Modified question from 10-1 NRC exam. (Equipment control_33). Question was modified to ask the conditions instead of given the conditions.

Cog Level: High Explanation: The CRD system is designed so that primary coolant pressure assists in driving control rods into the core upon receipt of a reactor scram. If low reactor pressure exists then accumulator pressure alone may not be sufficient to ensure control rods will insert. HCU low pressure alarms in the control room indicate either low accumulator pressure or high water level in the accumulator. AOP-02 conservatively assumes low pressure and directs a scram upon receipt of the alarm if reactor pressure is below 950 psig and CRD pressure cannot be restored to 940 psig or greater with either CRD pump.

.a t-SjLaa Ct JAIiS L SLI Distractor Analysis:

Choice A: Correct answer, see explanation Choice B: Plausible because if reactor pressure was greater than 950 psig and two accumulator alarm were received, action to restore pressure is required.

Choice C: Plausible because if reactor pressure was greater than 950 psig and two accumulator alarm were received, action to restore pressure is required.

Choice D: Plausible because if reactor pressure was greater than 950 psig and two accumulator alarm were received, action to restore pressure is required.

SRO Basis: N/A SD-08 4.5 Abnormal Operation 4_S. 1 Control Rod Malfunction!Misposition (AOP-02.O)

A malfunction in the Control Rod Drive System may result in the inability to move individual control rods. The system is designed so that primary coolant system pressure or accumulator pressure is available to scram the reactor AOP-02.0 should be consulted for specific actions to be performed. The AOP requires a manual scram if no CRD pump is operating, reactor pressure is less than 950 psig, and an HCU accumulator has low pressure. This is to ensure that the control rods can be inserted before accumulators depressurize since a scram on reactor pressure alone cannot be assured if reactor pressure is low.

AOP-02

2. IF reactor pressure is less than 950 psig (e.g., during LI startup OR shutdown evolutions), AND CRD pressure CANNOT be restored to greater than OR equal to 940 psig with either CRD pump, THEN upon receipt of the first 1-ICU low pressure alarm (A-07. 6-1, confirmed by amber light on Full Core Display), IMMEDIATELY INSERT a manual reactor SCRAM.

r Ia

3.0 OPERATOR ACTIONS 3.2.4 CONTACT the Reactor Engineer for further control rod movement instructions.

3.2.5 MONITOR off-gas radiation AND NOTIFY E&RC to take coolant samples if fuel element failure is suspected.

3.2.6 IF the CRD Hydraulic system has malfunctioned, THEN PERFORM the following:

1. IF the operating CRD Pump has failed, THEN RESTART the CRD Hydraulic system following loss of a CRD pump in accordance withl(2)OP-08 Section 8.17.

2. IF reactor pressure is less than 950 psig (e.g., during startup OR shutdown evolutions), AND CRD pressure CANNOT be restored to greater than OR equal to 940 psig with either CRD pump, THEN upon receipt of the first HCU low pressure alarm (A-07, 6-1, confirmed by amber light on Full Core Display), IMMEDIATELY INSERT a manual reactor SCRAM.

3. IF reactor pressure is greater than OR equal to 950 psig, AND two OR more HCU low pressure alarms (A-07, 6-1, confirmed by amber light on Full Core Display), THEN ENSURE CRD pressure is restored to greater than OR equal to 940 psig within 20 minutes.

4. REFER to Tech Spec 3.1.5 for any control rod scram accumulator required actions.

5. CHECK CRD pump suction AND drive water filters for high differential pressure.

6. MONITOR the following CRD system parameters for possible system leakage OR flow control valve failures:

CRD Drive Water Pressure, C11(C12)-PDI-R602. D CRD Cooling Water Pressure, C11(C12)-PDI-R603.

CRD Drive Temperature, C11(C12)-TR-R018.

CRD Charging Water Header Pressure, LI Cl l(C12)-PI-R601.

OAOP-02.0 Rev. 22 Page 4 of 9

50. 295023 1 Which one of the following choices completes the statement below lAW the Technical Specifications and Bases for LCO 3.9.1, Refueling Equipment Interlocks?

Refueling Equipment Interlocks (1) with the mode switch in the (2) position.

A. (1) prevents an iodine gas release (2) REFUEL B. (1) prevents an iodine gas release (2) SHUTDOWN C. (1) prevent inadvertent criticality (2) REFUEL D. (1) prevent inadvertent criticality (2) SHUTDOWN Answer: C K/A: 295023 Refueling Accidents AK3.02 Knowledge of the asonsfor the following responses as they apply to REFUELING ACCIDENTS: Interlocks associated with fuel handling equipment (CFR: 41.5 / 45.6)

RO/SRO Rating: 3.4/3.8 Objective: LOl-CLS-LP-058.1, Obj. 21 - State the function of the refueling interlocks

Reference:

None Cog Level: Low Explanation: In accordance with the TS! Bases! UFSAR, the purpose of the interlocks is to prevent an inadvertent criticality and the LCO and Applicability is when the mode switch is in Refuel.

Iodine gas would be released if the fuel was damaged. The interlock will prevent entry into an AOP or EOP.

Distractor Analysis:

Choice A: Plausible because the mode switch is correct but the reason is incorrect in that the interlocks do not prevent damage to the fuel which would cause a release of the iodine gas.

Choice B: Plausible because the interlocks do not prevent damage to the fuel which would cause a release of the iodine gas. For refueling the mode switch can be in Shutdown or refuel position in accordance with TS Table 1.1-1 Choice C: Correct Answer, see explanation Choice D: Plausible because for refueling the mode switch can be in Shutdown or refuel position in accordance with TS Table 1.1-1 SRO Basis: N/A

6 rasnais a IL SIC Refueling Equipment Interlocks 3.9.1 3.9 REFUELING OPERATIONS 3.9.1 Refueling Equipment Interlocks LCO 3.9.1 The refueling equipment interlocks associated with the refuel position of the reactor mode switch shall be OPERABLE.

APPLICABILITY: During in-vessel fuel movement vAth equipment associated with the interlocks when the reactor mode switch is in the refuel position.

Bases:

Refueling Equipment Interlocks B 3.9.1 B 19 REFUELING OPERATIONS B 3.9.1 Refueling Equipment Interlocks BASES BACKGROUND Refueling equipment interlocks restrict the operation of the refueling equipment or the withdrawal of control rods to reinforce unit procedures that prevent the reactor from achieving criticality during refueling. The refueling interlock circuitry senses the conditions of the refueling equipment and the control rods. Depending on the sensed conditions.

interlocks are actuated to prevent the operation of the refueling equipment or the withdrawal of control rods.

UPDATED FSAR 21 c&i r

r 9 nr INTRODUCTION AND

SUMMARY

age OT I .2.2.6.4 Refueling Interlocks A system of interlocks is provided to prevent an inadvertent criticality during refueling operations by restricting the movements of refueling equipment and control rods when the reactor is in the refuel mode.

The interlocks back up procedural controls that have the same objective. The interlocks affect the refuerng bridge, the refuelng bridge hoists, the fuel grapple, the control rods, (RefueFng interlocks do not affect the Asea-Erown Boveri hoist.)

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51. 2950241 During an accident, Unit One plant conditions are:

Reactor pressure 500 psig Drywell pressure 20 psig Suppression chamber pressure 19 psig Suppression pool level -42 inches Suppression pool temperature 160°F (Reference provided)

Which one of the following is the reason emergency depressurization is required?

A. Steam exists in the suppression chamber air space.

B. Prevent exceeding suppression chamber design temperature.

C. Prevent exceeding suppression chamber boundary design load.

D. Suppression chamber level is at the elevation of the downcomers.

Answer: A K/A: 295024 High Drywell Pressure EK3.04 Knowledge of theeasons for the following responses as they apply to HIGH DRYWELL PRESSURE: fEmergency depressurization (cFR: 41.5 /45.6)

RO/SRO Rating: 3.7/4.1 Objective: CLS-LP-300-L, Obj. 3E, Define the following terms: Pressure Suppression Pressure Limit CLS-LP-300-L, Obj. 4E, State the effect on Primary Containment if the following limits are exceeded: Pressure Suppression Pressure Limit

Reference:

Pressure Suppression Pressure graph (OEOP-01-UG Attachment 5 Figure 7)

Bank question last used on 08 NRC exam.

Cog Level: High Explanation: The PSP curve contains 4 segments. The bottom horizontal line is the elevation of the downcomers. The top horizontal line is the elevation of the bottom of the ring header. The diagonal line sloping up to the right starting at low suppression pool level is the highest pressure that could be in the suppression chamber air space without steam in the air space.

At the current suppression pool level this segment of the curve applies and is being exceeded.

The last segment toward the high suppression pool level limit is for design loading of the suppression chamber. Suppression chamber design temperature is the reason for emergency depressurization if Heat Capacity Temperature Limit is Unsafe; which it is not.

I JILJILF IIIIILIIIIL Distractor Analysis:

Choice A: Correct Answer, see explanation Choice B: Plausible since this would be correct for exceeding HCTL and suppression pool temperature is very elevated (would be correct at a higher reactor pressure)

Choice C: Plausible since this is part bases for the Pressure suppression Pressure limit (based on a higher suppression pool level and pressure)

Choice D: Plausible since this is part bases for the Pressure suppression Pressure limit (based on a lower suppression pool level)

SRO Basis: N/A ATTACHMENT 5 Page 22 of 27 FIGURE 7 Pressure Suppression Pressure

+2

+1 I

L1 a

uJ

> -1 LU

-J J

SAFE

-2 4

0 0 -3

+/-

• z 1

0 -4 (1)  ::UNSAFE::

C, w -5 a -6 D

0 0 10 20 30 40 SUPPRESSION CHAMBER PRESSURE (PSIG)

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52. 295025 1 Given the following plant conditions with RCIC in pressure control mode:

RCIC controller output 70%

Bypass to CST Vlv, E51-F022 Throttled RCIC Flow 300 gpm RPV pressure 990 psig, slowly rising RCIC controller Automatic set 300 gpm Which one of the following identifies two independent actions that will stabilize RPV pressure?

Throttle the E51-F022 in the (1) direction, or by (2) the RCIC Flow Controller auto setpoint.

A. (1) open (2) lowering B. (1) open (2) raising C. (1) closed (2) lowering D. (1) closed (2) raising Answer: D K/A: 295025 High Reactor Pressure EA1 .05 Ability to operate and/or monitor the following as they apply to HIGH REACTOR PRESSURE: RCIC: Plant-Specific (CFR: 41.7/ 45.6)

RO/SRO Rating: 3.7/3.7 Objective: CLS-LP-016-A Obj. 17b Describe how the following evolutions are performed during operation of the RCIC system: Adjusting RCIC flow in the reactor pressure control mode.

Reference:

None Cog Level: High Explanation: There are two ways to reduce the RPV pressure with the conditions given. One way is to close the E51-F022 valve, thereby decreasing the size of the hole and forcing the turbine to work harder to deliver the same flowrate. The second is to raise the controller setpoint thereby causing the turbine to work harder by forcing more flow through the same size hole.

an - WSaaAa hi.- ISa?JJLS&araw, Distractor Analysis:

Choice A: Plausible because these are the opposite of the actual answers and if the operator was trying to raise RPV pressure this would be correct.

Choice B: Plausible because raising is correct and the operator could have a misconception about the F022 valve.

Choice C: Plausible because closing the F022 is correct and the operator could have a misconception about the flow controller.

Choice D: Correct Answer, see explanation.

SRO Basis: N/A From RCIC Hard Card:

RCIC PRESSURE CONTROL (1OP-16 SECTION 8.2)

1. ENSURE THE FOLLOWING VALVES ARE OPEN: ES1-V8 (VALVE POSITION),

E51-V6 (ACTuATOR POSITION), AND E51-V9.

fl

2. OPEN E51-F0&6

3. START VACUUM PUMP AND LEAVE SWITCH IN START.

4. ENSURE ES1-F013 IS CLOSED L

5. ENSURE E41-F011 IS OPEN U

6. THROTTLE OPEN E51-F022 UNTIL DUAL INDICATION IS CBTAINED U

7. OPEN E51-F05 U

6. THROTTLE OPEN E51-F022 OR ADJUST RCIC FLOW CONTROL.

E51-FIc-R600. TO OBTAIN DESiRED SYSTEM PARAMETERS AND REACTOR fl PRESSURE.

9. ENSURE ES1-F019 IS CLOSED WITH FLCW GREATER THAN 80 GPM. U

10. ENSURE THE FOLLOWING VALVES ARE CLOSED: E51-F25, E51-F028.

E51-F004, AND E51-FCO5.

U

11. START SBGT(1OP-IW U

12. ENSURE BAROMETRiC CNDSR CONDENSATE PUMP OPERATES U FOR SHUTDOWN: REFER TO lOP-IS FOR TRANSFER BETWEEN PRESSURE AND LEVEL CONTROL: REFER TO IOP-I6 2 1/1086 110P-16 Rev.77 I Pa9e88ot921

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53. 295026 1 SUPPRESSION POOL TEMP REACHES 150F ESTABLISH CTMT COOLING REOUIRMENTS PER TABLE 3, REDUCING LPCF INJECTION FLOW IF NECESSARY lAW 01-37.4, Reactor Vessel Control Procedure Basis Document, which one of the following identifies why the step above is performed?

A. To prevent exceeding the Heat Capacity Temperature Limit.

B. To prevent exceeding primary containment design temperature.

C. To maintain long term operation of the Core Spray and RHR Pumps.

D. To minimize off-site releases per Alternative Source Term calculations.

Answer: C K/A: 295026 Suppression Pool High Water Temperature EK2.01 Knowledge of the interrelations between SUPPRESSION POOL HIGH WATER TEMPERATURE and the following: Suppression pool cooling (CFR: 41.7 /45.8)

RO/SRO Rating: 3.9/4.0 Objective:

Reference:

None Cog Level: Low Explanation: The calculation for the NPSH to the Core Spray and RHR pumps specify establishing cooling at ten minutes into the design basis LOCA. The calculation also assumes that the temperature of the suppression pool will be at approximately 169°F at ten minutes. If containment cooling is not established, then it is possible that the Core Spray or RHR pumps will be lost due to inadequate NPSH. To prevent having a ten minute action statement the temperature of 150°F was chosen.

Distractor Analysis:

Choice A: Plausible because high torus temperature does affect HCTL but at this point in RVCP, injection to the vessel is irrespective of NPSH or Vortex Limits Choice B: Plausible because long term containment cooling will limit primary containment temperature rise.

Choice C: Correct Answer, see explanation definition.

Choice D: Plausible because table 5 of RVCP has actions concerning Alternative Source Term.

SRO Basis: N/A

bJ!4tS4S1, U. sC / SSS STEPS RC!L-44 TABLE 3 MINIMUM CONtAINMENT COOLING REQUIREMENTS MINIMUM aow BEFORE PUMP OPM)

A EQUIREMENTS SUPPRLSSIDN POOL TEMP I - RHR PUMP 7700,tOVP REACI4LiS IIIF PER LOOP WIll-I HX ESTABLISH CTMT COOLING BYPASS VALVE REQUIREMENTS PER CLOSED TABLE U. REDUCING LPCI iNJECTION FLOW I - RHR SW PUMP 4flOEtOOP IF NECESSARY PER LOOP 2 ANN PUMPS RC.t-44 -

WITH HX BYPASS VALVE CLOSCO S - RHA SW frUMPS SUES STEP BASES:

This step provides guidance on establishing cooling to the suppression pool during a design basis LOCA. The calculation for the NPSH to the Core Spray and RHR pumps specify establishing cooling at ten minutes into the design basis LOCA. The calculation also assumes that the temperature of the suppression pool will be at approximately 1 69T at ten minutes. If containment cooling is not established at a suppression pool temperature of 169°F, then it is possible that the Core Spray or RI-IR pumps will be lost due to inadequate NPSH. The EPGs specify injecting irrespective of NPSH and vortex limits if reactor vessel water level is below Minimum Steam Cooling Reactor Water Level. This step provides guidance to reduce injection into the reactor vessel to establish the desired cooling for the containment. A value of 150°F has been selected for use in the step. This provides a margin of at least 19°F, to the limit used in the calculation for NPSH. These actions are incorporated to provide assurance that the unit can remain in the EOPs and not be required to go to primary containment flooding prematurely. The selection of a suppression pooi temperature limit precludes establishing a specific time limit in the procedure.

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54. 295028 1 Conditions on Unit Two have degraded to where the Drywell Air Temperature is 340°F.

Which one of the following identifies the components whose environmental qualification is affected by this temperature lAW 001-37.8, Primary Containment Control Procedure Basis Document?

A. SRV solenoids B. Inboard MSIV solenoids C. Torus to Drywell Vacuum Breakers D. CAC 4409 and 4410 Hydrogen Analyzers Answer: A K/A: 295028 High Drywell Temperature EK1 .02 Knowledge of the operational implicationsof the following concepts as they apply to HIGH DRYWELL TEMPERATURE: Equipment environmental qualification (CFR: 41.8 to 41.10)

ROISRO Rating: 2.9/3.1 Objective: CLS-LP-300L Obj. 4h, State the effect on Primary Containment if the following limits are exceeded: Drywell Design Temperature Limit.

Reference:

None Cog Level: Low Explanation: From the Bases document: Temperature should not be allowed to exceed the SRV maximum qualification temperature of 340°F.

Distractor Analysis:

Choice A: Correct Answer, see explanation Choice B: Plausible because this is a piece of equipment in the Drywell.

Choice C: Plausible because this is a piece of equipment in the Drywell that has a water level limit of +6 inches in the suppression pool.

Choice D: Plausible because the H2/02 analyzers were designed for pressures up to 30 psig. To preclude damage to these sample pumps and the subsequent radioactive release to secondary containment, the sample pumps are isolated when drywell pressure exceeds 30 psig.

SRO Basis: N/A

I11IIIIII1LII!hi W ZLLKJII!3:

Consistent with the definition of restore, emergency depressurization is riot required until it has been determined that drywell sprays (initiated in Step DWIT-1 6) are ineffective in reducing drywell temperature. This determination may be made when, before, or after the temperature actually reaches 300°F. It is not expected that either the containment integrity (300°F) or SRV operability (340°F) will be immediately challenged when the respective temperature limits are reached. If drywell temperature is already above 300F when Step DWIT-1 6 and DWIT-i 9 are reached, drywen sprays may still be used, it available, in preference to emergency depressurization. If sprays are effective in reducing the drywell temperature, emergency depressurization need not be pertormed. Extended operation above 300°F is not permitted and the temperature should not be allowed to exceed the SRV maximum qualification temperature of 340°F.

001-37.8 Rev. 4 Page 22 of 58

IfiIIZI LJIII!1!?IIIII!

55. 295029 1 A LOCA has occurred concurrently with a LOOP on Unit Two.

The following conditions exist:

Drywell Pressure 4.5 psig Reactor Water Level 95 inches and rising Reactor Pressure 280 psig Suppression Chamber level 5.5 inches CST level 20 feet Which one of the following identifies an injection source that must be secured AW PCCP? (Assume no circuit alterations have been performed)

A. CRD B. RCIC C. HPCI D. Core Spray Answer: B K/A: 295029 High Suppression Pool Water Level EA1 .04 Ability to operate and/or monitor the following as they apply to HIGH SUPPRESSION POOL WATER LEVEL: RCIC: Plant-Specific (CFR: 41.7 / 45.6)

RO/SRO Rating: 3.4/3.5 Objective: LOI-CLS-LP-016-A Obj. 15g Given plant conditions, predict the RCIC system response to the following conditions: High/low Suppression Pool water level.

Reference:

None Cog Level: High Explanation: lAW PCCP, if torus level cannot be maintained below 6 inches and adequate core cooling is assured, injection from sources external to containment are terminated (SP/L-1 0). Reactor water level of 185 inches rising assures adequate core cooling. HPCI suction path will automatically align to the torus IF a low CST level exists OR a high torus level exists. In the conditions described, HPCI suction will have aligned to the torus so its injection is from an internal source. RCIC suction path will automatically align to the torus IF a low CST level exists but will NOT align to the torus on a high torus level so its injection supply is from external sources (CST) and must be secured lAW SP/L-1 0. Drywell spray and torus spray are running and would not be secured by procedure until drywell pressure lowers to 2.5 psig.

Drywell spray would be secured if suppression pool level cannot be maintained below 21 inches.

1IIt!1II1L!1!1LQ L Distractor Analysis:

Choice A: Plausible because SP/L-15 of PCCP requires securing injection sources external to containment. Systems being used for boron injection and CRD exception to this.

Choice B: Correct Answer, see explanation Choice C: Plausible because HPCI injection is one of the sources of level control. SPIL-28 requires securing of HPCI if torus level cannot be maintained above minus 6.5 feet.

Choice D: Plausible because these systems are providing input to containment and the vessel.

SRO Basis: N/A OEOP-02-PCCP I

MAINTAiN SUPPRESSION POOL WATER LEVEL BELOW

+6 INCHES SPIL-06 INCHES CAN (5 SUPPRESSION POOL WATER LEVEL BE MAINTAINED BELOW

+6 INCHES SPIL-GY NO INITIATEA REACTOR SCRAM AND ENTER EOP-.O1 SP!L-O8

/ EMERGENCY CONSIDERANTICIPM1ONOF \

DEPRESSURIZATION

\ PER RCP SECTION OF /

\ REACTORVESSEL CONTROL /

PROCEDURE(EOP-O1- RVCP)

/

SP!L-09 iF ADEQUATE CORE COOUNG IS ASSURED THEN TERMINATE INJECTION INTO THE REACTOR FROM SOUES EXTERNAL TO PRIMARYCTMT EXCEPT FROM BORON INJECTION SYSTEMSANDCRD )

ZZZJE5b0 YES IS SUPPRESSION

-<POOL WATER LEVEL BELOW>

+6 INCHES ...

SPJL-11

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MAINTAIN SUPPRESSION POOL WATER LEVEL BELOW

+21 INCHES SPIL-13

+21 ---

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CAN SUPPRESSION POOL WATER YES

<. LEVEL BE MAINTAINED BELOW -

+21 INCHES

-.. . SPFL-14 fNO ITERMINATE DRVWELLSPRAYS PEREOP-O1-SEP-02 L

SPIL-15 CAN ---.

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LEVEL BE MAINTAINED ABOVE.>

- 65 FEET .

- SPJL-27 4 N SECURE HPCI IRRESPECTIVE OF ADEQUATE CORE COOLING SPJL-28 OEOP-O 1-SEP-02 21 WHEN drywell pressure drops below 15 psig OR IF directed to terminate drywell spray, THEN PERFORM the following:

RO: 2.7i CLOSE Loop A(S) DRYWELL SPRAY OTED ISOL LI VLV, El l-FO16A(F0i68,).

RO: 21.2 CLOSE Loop A(S) DRYWELL SPRAY JNOD ISOL LI VLV, El 1-FO2IA(F0218).

SD-16 1.3.1 Water Flow Path (Figure 16-2)

The primary water supply to the RCIC System is the Condensate storage tank (CST) through the normally open Condensate Storage Tank Suction Valve, E51-F0i 0. In the event the CST level decreases to a predetermined level, the RCIC suction will automatically transfer to the suppression pool through normally closed Suppression Pool Suction Valves, ES 1 -F029 and E51-F031.

f

1 asAr AL - b a* S t SD-19 HPCI Pump suction is established from the Suppression Pool by opening the normally closed Suppression Pool Suction Valves, E41-F042 and E41-F041. These valves automatically open, it no HPCI System isolation signal is present on a low CST level or high Suppression Pool level. Upon receiving an isolation signal (PCIS Group 4), these valves automatically close.

Table 19-6 - [(PCI Suppression Pool Suction Transfer Signals Signal Setpoint Tech Spec CST Level Low 23S elev. 234 elev.

(3S tank level) (34 tank level)

Suppression Pool Level -25 2APP-A-O1 Unit 2 APP A-Ol 1-5 Page 1 of 2 STJPPEESSION CHAMBER LVL HI-HI AUTO ACTIONS

1. If closed, Torus Suction Vlv, E41-P042, opens

2. If closed, Torus Suction Vlv, E41-F041, opens 3, If open, CST Suction Vlv, E41-F004, closes CAUSE

1. Suppression pool water level high (-25 inches)

2. Circuit malfunction Unit 2 APP A-Ol 6-4 Page 1 of 2 HPCI COND STORACE TNK WTR LVL LO AUTO ACTIONS

1. If closed, Torus Suction Vlv, E41-F04l, opens

2. If closed, Torus Suction Vlv, E41-F042, opens

3. If open, CST Suction Vlv, E4l-F004, closes CAUSES

1. I,ow level in CST due to usage or leaks (23 feet, 5 inches)

2. Loss of power to 125V DC Distribution Panel 4A

3. Circuit malfunction

1LZ 2APP-A-02 WHITE 3-8 RCIC SUCT XFR CST LO LVL Page 1 of 1 1.0 OPERATOR ACTIONS:

1.1 CONFIRM CST level approximately 3 feet or less.

1.2 OBSERVE Automatic Functions:

IF closed, THEN TORUS SUCTION VLV. E51-F029, opens (if no RCIC isolation signal is present)

IF closed, THEN TORUS SUCTION VLV E51-F031, opens (if no RCIC isolation signal is present)

IF open. THEN CST SUCTION VLV, E51-FO1O, closes

56. 295030 1 Following a DBA LOCA on Unit Two, plant conditions are as follows:

Reactor water level 55 inches and rising Reactor pressure 150 psig Torus temperature 220°F Suppression Chamber pressure 10.5 psi 9

Torus level -43 inches 2A Core Spray pump flow 5000 gpm 2B Core Spray pump flow 2000 gpm 2A RHR pump flow 8000 gpm 2B RHR pump flow 6000 gpm (reference provided)

Which one of the following identifies the ECCS pump(s) that is/are operating within the associated NPSH limit(s)?

A. 2B CS Pump ONLY B. 2A CS and 2A RHR pumps ONLY C. 2B CS and 2B RHR pumps ONLY D. 2A CS, 2A RHR and 2B RHR ONLY Answer: A K/A: 295030 Low Suppression Pool Water Level EA2.02 Ability to determine and/or interpret the following as they apply to LOW SUPPRESSION POOL WATER LEVEL: Suppression pool temperature (CFR: 41.10 / 43.5 / 45.13)

RO/SRO Rating: 3.9/3.9 Objective: CLS-LP-300B, Obj. 17. Given plant condition and the NPSH and vortex limit graphs for the RHR and CS, determine if the NPSH and/or vortex limits have been exceeded for either of the two systems.

Reference:

OEOP-01-UG, Attachment 5, Figures 5 & 6, Core Spray & RHR NPSH Limit.

Bank question that was last used on the 08 NRC exam.

Cog Level: High EXPLANATION: The student will need to plot each point on NPSH limit graph.

Torus pressure must be corrected down 0.5 psig to obtain the proper restriction line.

The correct torus pressure is 10.5 psig -0.5 psig = 10 psig.

This correction must be performed for both the RHR and CS graphs.

_F_1 L iI!IIIII1IIYIII Distractor Analysis:

Choice A: Correct answer, see explanation.

Choice B: Plausible because if the student fails to adjust torus pressure on RHR and CS graphs and thinks that above the line is the safe region then this answer would be correct.

Choice C: Plausible because if the student fails to adjust torus pressure on RHR and CS graphs, this answer would be correct.

Choice D: Plausible because if the student believes that above the line is the safe region then this answer would be correct.

SRO Basis: N/A OEOP-O1-UG ATTACHMENT 5 Page 14 of 28 Definitions For the NPSH graphs, adequate net positive suction head is available when the How rate and Suppression Pool temperature combination is below the adjusted Suppression Pool pressure curve. The indicated Suppression Pool pressure must be reduced by 0.5 psig for every foot of water level less than -2.6 feet to determine the correct Suppression Pool pressure curve to be used in evaluating NPSH.

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57. 295031 1 The plant has experienced a level transient which has caused the Reactor Recirc Pumps to trip. (Level stabilized when the Reactor Recirc Pumps tripped.)

Which one of the following choices completes the statement below?

Under these conditions, primary loop natural circulation (1) occurring because reactor water level is (2)

A. (1) is (2) above the jet pump suction B. (1) is (2) above the steam separator return to the downcomer region C. (1) is NOT (2) below the jet pump suction D. (1) is NOT (2) below the steam separator return to the downcomer region Answer: D K/A: 295031 Reactor Low Water Level EK1 .02 Knowledge of the operational implications of the following concepts as they apply to REACTOR LOW WATER LEVEL: Natural circulation: Plant-Specific (CFR: 41.8 to 41.10)

RO/SRO Rating: 3.8/4.1 Objective: LOl-CLS-LP-002, Obj. 16 Given plant conditions, determine if the conditions for natural circulation to occur in the reactor vessel are met.

Reference:

None Cog Level: Low Explanation: In order to establish natural circulation water must reach the turnaround point in the steam separators (200 220 inches in the core). The student must equate the RR pump trip to a Low Level Two condition (105 inches).

Distractor Analysis:

Choice A: Plausible because level is above the jet pump suction which is where the water would travel through to the bottom head region (part of the flowpath) but this does not establish natural circulation conditions.

Choice B: Plausible because the student must know the relationship between LL2 (105 inches) and steam separator turnaround point (200-220 inches). If it is not recognized that a recirc pump trip occurs below the steam separator turnaround point, this choice may be picked.

Choice C: Plausible because natural circulation conditions are not established but jet pump suction level is not a requirement for natural circulation.

Choice D: Correct Answer, see explanation

S lSØ R4ZJ*2 16?LT SaL. - -

SRO Basis: N/A

2. Natural Circulation Natural circulation is important to core cooling following transient or accident conditions when forced circulation by way of the reactor recirculation pumps is not available. Natural circulation not only occurs following transients and accidents, but is a major contributor to the flow rate through the core during startup, full power operation and shutdown.

In order for natural circulation to occur, there must be communication between the heat source and the heat sink. The reactor internals are designed to ensure communication at all times. Several flow paths exist for natural circulation and a loss of one vill tend to promote the others.

Primary Loop Water flow through the core removes heat from the fuel and boils to produce steam. The water and steam mixture passes through the separator and dryer for moisture removal. Steam leaves the vessel and passes to the turbine, while the water stays behind and is recirculated to the downcomer region to mix vith the cooler feedwater. This flow continues downward through the jet pumps to pass back up through the core to repeat the cycle.

SD-O1 I Rev.7 Page8of8l t

C

iJ

58. 295032 1 Which one of the following statements identifies the reason SCCP directs emergency depressurization based on temperature lAW 001-37.9, Secondary Containment Control Procedure Basis Document?

A. To prevent an unmonitored release.

B. To ensure ODCM site boundary dose limits are not exceeded.

C. To prevent damage to equipment required for safe shutdown.

D. To preserve personnel access into the reactor building.

Answer: C K/A: 295032 High Secondary Containment Area Temperature EK3.01 Knowledge of the reasons for the following responses as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE: Emergency/normal depressurization (CFR: 41.5/45.6)

RO/SRO Rating: 3.5/3.8 Objective: LOI-CLS-LP-300-M Obj. 1 3a, Given plant conditions and the SCCP, determine the required actions if the following limits are exceeded: Maximum Safe operating values with a primary system discharging into secondary containment.

Reference:

None Bank questions that was last used on the 08 NRC exam.

Cog Level: Low Explanation: The Maximum Safe Operating Temperature values are the area temperatures above which equipment necessary for the safe shutdown of the plant will fail. These area temperatures are utilized in establishing the conditions which reactor depressurization is required. The criteria of more than one area specified in this step identifies the rise in reactor building parameters as a wide spread problem which may pose a direct and immediate threat to secondary containment integrity, equipment located in the RB, and continued safe operation of the plant.

Distractor Analysis:

Choice A: Plausible because not mitigating rad release.

Choice B: Plausible because other things, RRCP, deal with rad release concerns.

Choice C: Correct answer, see explanation.

Choice D: Plausible because this is the reason for Radiation limitations.

SRO Basis: N/A

III 001-37.9 The Maximum Safe Operating Values are the area temperatures above which equipment necessary for the safe shutdown of the plant will fall. These area temperatures are utilized in establishing the conditions under which reactor depressurization is required. Separate temperatures are provided for each Secondary Containment area.

$11 IIJ LL1LIItLJk z jJ) 1LIII11

59. 295037 1 LPC is being executed on Unit One due to an ATWS. Plant conditions are:

Reactor power is 21%

Reactor pressure is 940 psig, controlled by EHC Reactor water level is 170 inches, controlled by feedwater Which one of the following identifies why LPC directs deliberately lowering reactor water level to 90 inches?

A. Raise feedwater subcooling.

B. Increase the resulting boron concentration in the core.

C. Reduce heat input to primary containment if the MSIVs close.

D. Minimize the possibility of large scale core oscillations.

Answer: D KJA: 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown EK1 .02 Knowledge of the operational implications of the following concepts as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN:

Reactor water level effects on reactor power (CFR: 41.8 to 41.10)

RO/SRO Rating: 4.1/4.3 Objective: LOI-CLS-LP-300-E, Obj. 06 Explain the reason for lowering reactor water level while performing the Level/Power Control Procedure.

Reference:

None Cog Level: Low Explanation: (lAW Oi-37.5) To prevent or mitigate the consequences of any large irregular neutron flux oscillations induced by neutronic/thermal-hydraulic instabilities, reactor water level is lowered sufficiently below the elevation of the feedwater sparger nozzles. This places the feedwater spargers in the steam space providing effective heating of the relatively cold feedwater and eliminating the potential for high core inlet subcooling. For conditions that are susceptible to oscillations, the initiation and growth of oscillations is principally dependent upon the subcooling at the core inlet; the greater the subcooling, the more likely oscillations will commence and increase in magnitude.

a_1 __at .w w,ezaj.t a Distractor Analysis:

Choice A: Plausible because lowering level will reduce (not raise) feedwater subcooling by introducing the feedwater into the steam space below the feedwater spargers.

Choice B: Plausible because reducing the core water inventory would result in a higher boron concentration (per unit volume).

Choice C: Plausible because LPC would continue to lower level below 90 inches if conditions of Table 3 are met. The purpose of lowering level below 90 inches is to reduce heat input to containment.

Choice D: Correct Answer, see explanation SRO Basis: N/A STEP BASES:

To prevent or mitigate the consequences of any large irregular neutron flux oscillations induced by neutronicitherrnal-hydraulic instabilities, reactor water level is initially lowered sufficiently below the elevation of the feedwater sparger nozzles. This places the feedwater spargers in the steam space providing effective heating of the relatively cold feedwater and eliminating the potential for high core inlet subcooling. For conditions that are susceptible to oscillations, the initiation and growth of oscillations is principally dependent upon the subcooling at the core inlet; the greater the subcooling, the more likely oscillations will commence and increase in magnitude.

If reactor power is at or below the APRM downscale trip setpoint (2%). it is highly unlikely that the core bulk boiling boundary would be below that which provides suitable stability margin for operation at high powers and low flows. (A minimum boiling boundary of 4 ft above the bottom of active fuel has been shown to be effective as a stability control because a relatively long two-phase column is required to develop a coupled neutronic/ thermal-hydraulic instability.) Furthermore, flow/density variations would be limited with reactor power this low since the core has a relatively low average void content.

001-37.5 Rev. 11 The combination of high reactor power (above the APRM downscale trip), high Page 19 of 90 Suppression Pool temperature (above the Boron Injection Initiation Temperature), and r

an open SRV or high drywell pressure (above the scram setpoint), are symptomatic of heat being rejected to the Suppression Pool at a rate in excess of that which can be removed by the Suppression Pool Cooling system. Unless mitigated, these conditions ultimately result in loss of NPSI-1 for ECCS pumps taking suction on the Suppression Pool, containment overpressurization, and (ultimately) loss of Primary Containment integrity, which in turn could lead to a loss of adequate core cooling and uncontrolled release of radioactivity to the environment If the Suppression Pool heatup continues and heat being rejected to the Suppression Pool remains above decay heat levels, the reactor power reduction achieved by lowering reactor water level to TAF retards Suppression Pool heatup, thus preserving Suppression Pool heat capacity and delaying Emergency Depressurization as long as possible.

I r w - 1W rrr

60. 295038 1 Which one of the following alarms is an entry condition into RRCP?

A. Service Wtr Effluent Had High.

B. Area Rad Turbine Building High.

C. RBCCW Liquid Process Rad High.

D. Process Reactor Building Vent Had High.

Answer: A K/A: 295038 High Off-Site Release Rate EK2.06 Knowledge of the interrelations between HIGH OFF-SITE RELEASE RATE and the following: Process liquid radiation monitoring system (CFR: 41.7 / 45.8)

RO/SRO Rating: 3.4/3.7 Objective: LOl-CLS-LP-300-N, Obj. 2 Given plant conditions, determine if OEOP-04-RRCP should be entered.

Reference:

None Modified question from the 10-1 NRC exam. (Emergency Procedure_30) Changed the distractors to make a different answer correct.

Cog Level: Low Explanation: A number of conditions associated with pathways outside containment for the transport of radioactive materials are listed for entry to RRCP. Distractors are monitored areas and processes with valid alarms but which are enclosed processes or area rads which are not used for determination of off-site doses.

Distractor Analysis:

Choice A: Correct Answer, see explanation Choice B: Plausible because Turbine Building Vent Rad High annunciator is an entry condition for RRCP.

Area Rad is not but is similar in description.

Choice C: Plausible because this alarm requires actions up to and including entry into AOPs. However, EOP entry is not required directly from the alarm.

Choice D: Plausible because Reactor Building Roof Vent Hi and Process Off-Gas Vent Pipe Rad Hi alarms are both entry conditions for RRCP. Process Reactor Building Vent Rad High is not but is similar in description.

SRO Basis: N/A

LJSAj aSSLzr a RRCP I

<I RADIOACTNIFY RELEASE \

CONTIOL RR I

/ ENTRY CONDITIONS:

• MAIN STEAM LINE RAD IIANNUNSEFPOINT EXCEEDED 4UA-23,24)

• PROCESSOFF.GAS R&D NIANIJJN SETPOIM EXCEEDED 4UA-O3,5.2) 45.JAE)

• RXSLDG ROOF VENT RAn HIGH ANIIJN SETPOINT EXCEEDED 1UAO3,2.3)

• TIJRBBLDGVENTRAD HIGH ANN(JN SEFPOINT EXCEEDED UA.O3,34)

• PROCESSOGVENTpIPE RADII ANNUN SEIPOINT EXCEEDED lIJAO3,6.4)

STACK)

• SERVICE WTR EFFLUENT RAt I4GN ANNUN SETPCINT EXCEEDED (UA.e3S.

• ANY UNMOII1ORED OFF Slit RADIOACTIVITY RELEASE

• CALCUlATED DOSE RATE UMITOF IJOGLE GAS INSTANTANEOUS RELEASE RATE DETERMNATIOtr

\ 4E&RCdO2 EXCEEDED RR2 neea

JI iIF ?&1I1I11IIII1

61. 3000001 The service air compressors have failed causing the following plant conditions:

200 200 RB INSTR AIR RECEIVER 2A Not in Alarm PRESS LOW mr-E100

! 1150 E-1oo N

0 p

RB INST AIR RECEIVER 2B PRESS LOW Not in Alarm 5o 50 GE InAlarm 0

SERVICE AIR 1NS1UMENT 4R HE.AOER SR1CE R 1EADER PRESSLOW In Alarm PRES5JRE PRESSURE 24i Which one of the following identifies the operator action that must be taken lAW OAOP-20.0, Pneumatic (Air/Nitrogen) System Failures?

A. Insert a manual reactor scram B. Close Service Air Isol Vlvs, SA-PV-706-1 & 706-2 C. Open Div 1(11) Backup N2 Rack Isol Vlvs, RNA-SV-5482(5481)

D. Open Serv Air Dryer IA Bypass Pressure Control Valve, 1-SA-PV-5067 Answer: B K/A: 300000 Instrument Air System G2.02.44 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

(CFR: 41.5 /43.5 /45.12)

RO/SRO Rating: 4.2/4.4 Objective: LOl-CLS-LP-046, Obj. 11 b State the reason(s) for the following actions occurring during operation of the Pneumatic Systems: Service Air Header is isolated at 105 psig LOI-CLS-LP-046, Obj. 07a Given plant conditions, determine if the following automatic actions should occur:

t smu

Reference:

None Modified question from the 10-2 NRC exam. (300000_i 4) Previous question asked what auto action should occur while this question has the student determine what auto action that has failed.

Cog Level: High Explanation: The service air and instrument air systems use the same air compressors and are essentially the same air system. As pressure decreases the service air system will auto isolate at 105 psig. When the isolation occurs the pressure indication will go to zero because the isolation valves are upstream of the tap off for the indication. With system leakage it goes to zero fairly quickly. The alarm for low service air comes in at 105 psig and low instrument air at 100 psig. This is an auto action that should have occurred so the operator will make it happen per our procedures.

Distractor Analysis:

Choice A: Plausible because if Unable to maintain at least one division noninterruptible instrument air pressure above 95 psig then this is an operator action.

Choice B: Correct Answer, see explanation Choice C: Plausible because if RB air pressure alarm is in then this is an operator action Choice D: Plausible because if pressure was <98 psig this would be an operator action.

SRO Basis: N/A Unit 1 APP UA-01 4-4 Page 1 of 2 INSTR AIR PRESS-LOW AUTO ACTIONS NONE CAUSE

1. Abnormal supply to system due to:

a. Loss of power to compressor(s).

b. Fadure of the Air Compressor 15 or 25 to start or oad autonaticaly on decreasing pressure.

c. Service Air Isolation Valves. SA-PV-706-1 and SA-PV-706-2. did not close on decreasing header pressure (105 psig).

th Heavy leak in the supply system.

e. lde compressor check valve leaking.

f. Service air dryer high differential pressure or malfunction.

2. Abnormal distribution system due to:

a. Heavy usage of service air or instrument air.

b. Leaks in the service air or instrument air piping.

3. Circut nt&function.

OBSERVATIONS

1. Instrument air header pressure decreases as shown by Confrol Room and Loca Indicators IA-P 1-724-i and lA-P 1-724-2.

2. Service air header pressure decreases as shown by Control Room and Local Indicators SA-Pl-765-1 and SA-Pl-765-2.

3. Air compressor is and 25 running when pressure falls to approximately 110 psig.

4. SERVICE AIR PRESS-LOW (UA-0l 5-4 alarms.

5. AIR DRYER TROUBLE (112 APP-UA-01 5-3, 1APP-UA-01 6-3:1.

6. Serwce air header is isolated at less than 105 psig.

7. Overal unit operation for any abnomlaiti due to low instrument &r system pressure.

St a a. 1 it

7. Written procedures are not necessary for situations where:

a Prompt action is necessary to minimize personnel injury, damage to the facility, and to protect the health and safety of the public.

b Prompt action is necessary to prevent the deterioration of plant conditions or components to a possibly unsafe or unstable level.

If time permits, approval from the SM/CRS shall be obtained.

Factors to consider included: complexity of action, potential for damage from common cause. etc.

c. Conditions exist which may require timely actions due to failure of automatic control systems or uncertain equipment status (i.e.,

taking manual control of hand/auto stations, position of selector switches. etc.).

OPS-NGGC-i000 Rev. 6 Page 94 of 147 rssk

iE3III1&I JiL LL!IIIV

62. 400000 1 Which one of the following choices completes the statements below?

The Service Water to TBCCW MOVs (SW-V3 and V4) are powered from (1)

The Service Water to RBCCW MOVs (SW-Vl 03 and Vi 06) are powered from (2)

A. (1) BOP distribution (2) BOP distribution N B. (1) BOP distribution (2) E Bus distribution C. (1) E Bus distribution (2) BOP distribution D. (1) E Bus distribution (2) E Bus distribution Answer: D K/A: 400000 component Cooling Water System K2.02 Knowledge of electrical power supplies to the following: CCW valves (CFR: 41.7)

ROISRO Rating: 2.9/2.9 Objective: LOl-CLS-LP-043, Obj. 8d State the power supply (bus and voltage) for the following Service Water System components: MOVs

Reference:

None Cog Level: Low Explanation: Conventional Header to TBCCW Heat Exchanger Outboard Valve SW-V4 1XA (2XA)

Conventional Header to TBCCW Heat Exchanger Inboard Valve SW-V3 1XB (2XB)

SW Header to RBCCW Heat Exchangers Isolation Valve SW-V106 1)(A (2XA)

SW Header to RBCCW Heat Exchangers Isolation Valve SW-V1 03 1 XB (2XB)

Distractor Analysis:

Choice A: Plausible because other SW system valves are powered from BOP.

Choice B: Plausible because other SW system valves are powered from BOP.

Choice C: Plausible because other SW system valves are powered from BOP.

Choice D: Correct Answer, see explanation SRO Basis: N/A Example of SW valves powered from BOP sources (From SD-43):

TBCCW Heat Exchanger 1TG (2TG) EG3 (EG3) 1 B (2B) SV) Outlet Vave SW-V6

i I

63. 4000002 Unit Two is at rated power with the following alignment for RBCCW:

2A RBCCW control switch in ON 2B RBCCW control switch in ON 2C RBCCW control switch in AUTO The 2D to E3 master breaker tripped spuriously and DG3 failed to auto start.

Which one of the following identifies the expected response of the RBCCW pumps?

A. No RBCCW pumps are running.

B. ONLY the 2B RBCCW pump is running.

C. ONLY the 2C RBCCW pump is running.

D. 2B and 2C RBCCW pumps are running.

Answer: B K/A: 400000 Component Cooling Water System K6.07 Knowledge of the effect that a loss or malfunction of the following will have on the CCWS:

Breakers, relays, and disconnects (CFR: 41.7 I 45.7)

R0/SRO Rating: 2.7/2.8 Objective: L0l-CLS-LP-021, Obj. 9a Given plant conditions, determine how RBCCW is affected by the following: Loss of AC Power

Reference:

None Cog Level: High Explanation: Failure of power to E3 renders A and C pumps without power. (Plausibility of no pumps running) Failure of power to E3 also disables the auto start pressure switch, RCC-PS-672 (DP-2A 120V) therefore B RCC pump would not receive an AUTO start signal but it is already running.

Distractor Analysis:

Choice A: Plausible because the A and C pumps lose power and if the B pump was not running then it would not get a start signal.

Choice B: Correct Answer, see explanation Choice C: Plausible if the student has the power supplies wrong as A and B from E3 instead of A and C.

Choice D: Plausible if the student thinks the power supply is lost only to A pump and considers the low pressure starting the C pump.

SRO Basis: N/A

St1PS3 4f__ L LiStLe t ALt 0 I

Load: 480V Motor Control Center 2-2XE Location: Reactor Building 50 N Drawing

Reference:

F-03049 Upstream Power Source: 480V Substation E7 COMPT LOAD DESCRIPTION EFFECTS ON LOSS OF POWER EM RBCCW Pump 24 RCC-2A-PMP Loss of load EM RBCCW Pump 2C RCC-2C-PMP Loss of load Load: 120V Distribution Panel 2-2A-120V (H06)

Location: Control Building 23 SE Drawing

Reference:

LL-0934I-4 Upstream Power Source: 120V Emergency Distribution Panel 2E7 OCT LOAD DESCRIPTION EFFECTS ON LOSS OF POWER 24 RBCCW Discharge Header Pressure Switch Auto stat of RBCCW pumps will be defeated RCC-PS-672 for low discharge header pressure From SD-21:

Power suppiles ae as follows:

PUMP POWER SUPPLY RGCCW Pump 1A MCC 1XE R6CCW Pump 16 MCC 1XF RSCCW Pump 1C MCC 1XE RBCCW Pump 2A MCC 2XE RBCCW Pump 26 MCC2XF RBCCWPump2C MCC2XE

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64. 600000 1 The CRS has declared the Control Room Envelope Boundary inoperable.

A fire external to the control room has resulted in smoke intrusion into the control room.

The CRS has determined that control room evacuation is not necessary and has ordered the donning of SCBAs by control room personnel.

Which one of the following indicates the approximate length of time each SCBA bottle will last lAW 001-01.01, BNP Conduct of Operations Supplement?

A. 10 minutes B. One hour C. Two hours D. Four hours Answer: B K/A: 600000 Plant Fire On Site AA2.10 Ability to determine and interpret the following as they apply to PLANT FIRE ON SITE:

Time limit of long-term-breathing air system for control room (CFR: 41.10 /43.5 / 45.13)

RO/SRO Rating: 2.9/3.1 Objective: None

Reference:

None Cog Level: Low Explanation: Smoke in the control room from outside sources is addressed in 001-01.01. If control room evacuation is not required, SCBA5 may be used for long term habitability. There are no installed breathing air systems for the control room. lAW E&RC-0292, SCBA Use and Maintenance, and 001-01.01, the approximate duration of the SCBAs is one hour.

Distractor Analysis:

Choice A: Plausible because E&RC 0292 states expected duration of the Ska-Pak is five to ten minutes.

Choice B: Correct Answer, see explanation Choice C: Plausible because two hours is a period of time that may be reasonable to the candidate.

Choice D: Plausible because four hours is a period of time that may be reasonable to the candidate.

SRO Basis: N/A

V eAfl t ni aJ aadtMtiIM OERC-0292 6.16 The air supply in a Scott Air-Pak 4.5 SCBA is rated at approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 6.17 The air supply in the Ska-Pak with escape bottle is from 5 to 10 minutes.

00 1-0 1 .01

a. General Mitigating Actions (3) In the case of an actual radiological OR chemical event, OR a challenge from smoke where the use of SCBA air packs has been chosen as the mitigating action, Fire Brigade OR FIN team personnel will be directed to deliver additional/replacement SCBA bottles as necessary to CRE occupants. SCBA bottles contain approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> worth of air b Smoke in the Control Room from Outside Sources (1) Place CREV in service.

(2) Review QAOP-32.0.

NOTE: One of the small, two of the large, AND one of the Xlarge SCBA masks are stored with the two SCBA air packs stored in the CAS.

(3) Issue SCBA air packs AND masks from the Control Room emergency inventory to CRE occupants if Control Room evacuation is not necessary.

(8) Attempt to determine estimate of duration of SCBA use for review of continued CRE occupancy plans. For extended SCBA use, the availability of full SCBA bottles AND masks may become a limiting factor for continued CRE occupancy.

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65. 700000 1 Unit Two is operating at rated power when small oscillations due to unstable voltage regulation is observed.

Which one of the following choices completes the statements below lAW 20P-27, Generator and Exciter System Operating Procedure?

Thefirstaction required isto (1)

This action must be reported to the (2)

A. (1) place the generator voltage regulator in Manual (2) System Load Dispatcher B. (1) place the generator voltage regulator in Manual (2) Plant Transmission Activities Coordinator (PTAC)

C. (1) disable the Power System Stabilizer (PSS)

(2) System Load Dispatcher D. (1) disable the Power System Stabilizer (PSS)

(2) Plant Transmission Activities Coordinator (PTAC)

Answer: C K/A: 700000 Generator Voltage and Electric Grid Disturbances G2.04.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator. (CFR:

41.10 /43.5 145.11)

RO/SRO Rating: 2.7/4.1 Objective: LOl-CLS-LP-027-A, Obj. 1 8h Given the plant conditions and one of the following events use plant procedures to determine the actions required to control and/or mitigate the consequences of the event: Unstable Voltage Regulator Operation

Reference:

None Cog Level: High Explanation: The Power System Stabilizer (PSS) provides for dampening of power oscillations resulting from unstable grid perturbations or improper voltage regulator operation. In the event unstable voltage regulation is observed, the first steps required lAW 20P-27 are to disable the PSS to determine if a malfunction of the PSS is the cause of the condition. If the PSS is not the cause and oscillations continue, the next action will be to take the voltage regulator to Manual. The System Load Dispatcher is required to be notified if either of these actions are performed. The PTAC is required to be notified for certain events as specified in 001-01.01, Attachment 30, but not for the actions for the conditions described in the question.

JEltflhILWJasaee g- - .aaaI let aa, CJflCLZt j Distractor Analysis:

Choice A: Plausible because OP-27 provides instructions for placing the voltage regulator in Manual but only after attempts to stabilize voltage oscillations by disabling PSS are attempted. System Load Dispatcher notification is required in either case.

Choice B: Plausible because OP-27 provides instructions for placing the voltage regulator in Manual but only after attempts to stabilize voltage oscillations by disabling PSS are attempted. PTAC notification is required for numerous conditions specified in 001-01 .01.

Choice C: Correct answer, see explanation.

Choice D: Plausible because disabling of the PSS is the first action specified in OP-27. Notification of System Load Dispatcher is required. PTAC notification is required for numerous conditions specified in 001-01.01.

SRO Basis: N/A SD-27 2.17.7 Power System Stabilizer (PSS)

This circuit is designed to modulate the excitation system to provide supplementary damping for low frequency power oscillations that may exist due to system transmission configurations and plant loading.

Modulation of the generator excitation produces transient changes in the generators electrical output power which counteracts this situation.

A Power System Stabilizer has been added to prevent these oscillations from occurring. Once the unit is above 325MVA load the Power System Stabilizer will be enabled and will automatically dampen possible low frequency oscillations OP-27 should be followed in the event that power oscillations occur while the PSS is ACTIVE or in the event that the PSS is DISABLED.

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J_L a Pi L-- - a 20P-27 8.6 Unstable Voltage Regulator Operation Conir U!

8.6.1 Initial Conditions

1. The Generator and Exciter System is in operation in accordance with Section 5.1.

[ NOTE: Voltage may be oscillating as a result of the Power System Stabilizer modulating exciter field voltage to provide positive dampening of power oscillations.

2. Unstable voltage regulation is indicated (Panel XU-1) by LI oscillations on GEN VOLT PEG 01FF VOLT.

GEN-VM-3495 and oscillations on any of the following:

- GEN-VM-733

- GEN-AM-735

- GEN-MVAR-728

- GEN-AM-729 8.6.2 Procedural Steps

1. NOTIFY System Load Dispatcher prior to disabling PSS LI OR removing voltage regulator from automatic control.

2. PLACE P55 CONTROL, PSSCSI, to DISABLE, at LI Power System Stabilizer control panel, 2-GEN-PSS.

3. PLACE P55 ALARM BYPASS. PSSCS3, in BYPAS& LI fl_

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66. G2.01 .05 1 Following a seven day break in work schedule an operator is assigned to an operating unit with the following work schedule.

Dayl 12 hours Day2 12 hours Day3 12 hours Day4 12 hours Day5 0 hours Day6 12 hours Day7 13 hours Day8 14 hours Day9 0 hours Which one of the following choices will ensure compliance lAW ADM-NGGC-0206, Managing Fatigue and Work Hour Limits, without requiring a waiver?

A. Take day I off.

B. Take day7off.

C. Work only 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> on day 7.

D. Work only 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> on day 8.

Answer: B K/A: G2.01 .05 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc. (CFR: 41.10 /43.5 / 45.12)

RO/SRO Rating: 2.9/3.9 Objective: AOl-CLS-LP-201-D, Obj. 19- Discuss working hour limitations per ADM-NGGC-0206.

Reference:

None Cog Level: High Explanation: ADM-NGGC-0206 calculates work week limits on rolling time periods of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> and 7 days (168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />). These time periods do not reset following a day off but continue to roll. The limitations specified in the procedure are: 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24 hrs, 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> in any 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, and 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any seven days. Turnover time is excluded. A difference exists in the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> calculation if the work is being performed during an outage. In that case, the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> limit may be considered on a calendar day period instead of rolling.

In the example given, work hours are exceeded on day 8 (27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br /> in 48), on the seven days from day 1 through day 7 (73 hours8.449074e-4 days <br />0.0203 hours <br />1.207011e-4 weeks <br />2.77765e-5 months <br /> in one week), and on the seven days from day 2 through day 8 (75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> in one week). Taking day seven off would clear the excessive hours from day 8, and both rolling weeks.

Distractor Analysis:

Choice A: Plausible because taking day one off would clear the excessive hours from the seven day period from day 1 through day 7. However, the second seven day period from day 2 through day 8 and day 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> are still excessive.

Choice B: Correct Answer, see explanation Choice C: Plausible because working 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> on day 7 would clear the excessive hours on day 8 and would also clear the excessive hours on the seven day period from day I through day 7.

However, the hours during the seven day period from day 2 through day 8 would still be excessive.

Choice D: Plausible because working 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> on day 8 would clear the excessive hours on day 8 and the seven day period from day 2 through day 8. However, the hours during seven day period from day 1 through day 7 would still be excessive.

SRO Basis: N/A ADM-NGGC-0206 9.2 Covered Worker Work Hour Controls 9.21 On-Line Work Hour Limits

1. A workers work hours shall not exceed the following limits unless a waiver is issued.

a) 16 work hours in any 24-hour period (MWH 16/24) b) 26 work hours in any 48-hour period (MWH26/48)

C) 72 work hours in any 7-day or 168 hour0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> period (MWH72/168) 9.2.2 Outage Work Hour Limits

1. Evaluating work hours to outage rules versus on-line rules is optional.

Should a plant enter an unplanned outage, it is not necessary to reassign personnel to an outage schedule. Nor is it necessary to assign an outage schedule to personnel working on outage activities.

2. A workers work Fiours shall not exceed the following limits unless a waiver is issued.

a) 16 work hours in any 24-hour period (MWHI6/24) b) 26 work hours in any 48-hour period (MWH26148) c) 72 work hours in any 7-day or 168 hour0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> period (MWH721168)

NOTE: EmpCenter evaluates this rule on a raIling 168 hour0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> window. The regulatory limit is aduafly 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7 calendar day period. lr EmpCenter indicates a violation based on 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />, but the worker has not exceeded 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in the 7 calendar day period, the violation may be overridden.

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8.2 Workers shall not exceed the work hour limits defined in the procedure unless authorized by a waiver. If it is determined that a worker has violated the requirements, an NCR must be written and, if applicable, the NCR number added to the comment fields in EmpCenter.

9.1.2 Items To Be Excluded From Work Hour Calculation

4. All turnover time can be excluded from the total hours worked however only one period of turnover time per shift can be excluded for determining if the minimum break requirements are met.

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67. G2.O1.321 Which one of the following identifies the reason IOP-lO, Standby Gas Treatment System Operating Procedure, prohibits venting the drywefl and the suppression pool chamber simultaneously with the reactor at power?

This would cause:

A. the operation of torus to drywell vacuum breaker3 B. the pressure suppression function to be bypassed.

C. the operation of reactor building to torus vacuum breakers.

D. excessive release of radioactivity through the main stack.

Answer: B K/A: G2.01 .32 - Ability to explain and apply system limits and precautions. (CFR: 41.10 /43.2 I 45.12)

ROISRO Rating: 3.8/4.0 Objective:

Reference:

None Bank question that was last used on the 07 NRC exam.

Cog Level: Low Explanation: Per OP-i 0, torus and drywell cannot be vented at the same time in Modes 1, 2 or 3. per the LER reference, this could result in bypassing pressure suppression function.

Distractor Analysis:

Choice A: Plausible because this lineup equalizes pressure between the drywell and the suppression pool free air space since the vacuum breakers operate on a d/p between the spaces this would bypass them, not open them.

Choice B: Correct Answer, see explanation Choice C: Plausible because these vacuum breakers prevent drawing a negative pressure in the suppression pool. Cross connecting the drywell and the suppression pool free air space will not cause a negative pressure in the suppression pool.

Choice D: Plausible because this action would only be done when LOCA conditions do NOT exist SRO Basis: N/A

- s Straw t jtr_ aakasa4aStNSA a.o PRECAUTIONS AND LIMITATIONS 3.4 The Standby Gas Treatment System will NOT automatically start it the control switch is in STAY.

3.5 Venting of the Drywell and Suppression Pool simultaneously shall NOT be performed when the plant is in Mode 1. 2, or 3.

2.0 REFERENCES

2.16 LER 1-97-011, Drywell and Torus Inerting/Deinerting Lineup Results in Unanalyzed Suppression Pool Bypass Path wir prr rr- wr

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68. G2.01.441 During a core reload the initial loading of fuel bundles around each SRM centered 4-bundle cell was completed with all four SRMs fully inserted and reading 50 cps.

After some additional fuel assemblies were loaded (none of which were adjacent to any SRM) the RO observes that all SRMs have increased steadily. Current readings are as follows:

SRMA 120 cps SRMC 26Ocps SRM B 130 cps SRM D 140 cps Which one of the following identifies the required action, if any, at this time lAW OFH-1 1, Refueling?

A. Immediately suspend the movement of fuel.

B. Continue monitoring SRMs during fuel movement as readings are not unusual.

C. Report to the CRS that SRM C is INOPERABLE and recommend bypassing SRM C until l&C can investigate. Fuel movements may continue.

D. Record a new baseline count rate for the SRMs on OFH-1 1, Attachment 6, Documentation for SRM Baseline. Fuel movements may continue.

Answer: A K/A: G2.01 .44 - Knowledge of RO duties in the control room during fuel handling such as responding to alarms from the fuel handling area, communication with the fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation. (CFR: 41.10/43.7 / 45.12)

ROISRO Rating: 3.9/3.8 Objective: LOI-CLS-LP-305, Obj. 18 Given the conditions during a refueling outage state the operator actions for rising SRM count rates and/or inadvertent criticality.

Reference:

None Bank question that was last used on the 04 NRC exam.

Cog Level: High Explanation: FH-1 I requires suspending fuel movement if SRM increase by factor of 5 relative to initial base-line SRM reading (or doubles with any single bundle). Also, malfunctioning of any SRM channel shall be reason to terminate refueling operations until TS compliance can be determined.

r----z a,tss !Lelf)1 Distractor Analysis:

Choice A: Correct Answer, see explanation Choice B: Plausible because factor of five (from 50 to 250 cps) can be confused with five doublings (from 50 to 800 cps).

Choice C: Plausible because fuel movements could continue on a failed SRM if TS actions are completed.

Choice D: Plausible because if a bundle is placed next to an SRM new baseline data is recorded. The question states that all adjacent bundles are loaded.

SRO Basis: N/A 4O PRECAUTIONS AND LIMITATIONS 4.42 Fuel movement shall be suspended and the Reactor Engineer contacted if either of the following occur 4.42.1 An SRM reading increases by a factor of two upon insertion of any single bundle. During a spiral reload, this restriction applies only after the initial loading of fuel bundles around each SRM is complete.

During a Core Shuffle, this restriction does NOT apply to the SRM that is having an adjacent fuel bundle inserted or removed.

4.42.2 An SRM increases by a factor of five relative to the SRM baseline count rate recorded on Attachment 6.

1. OBSERVE AND LOG on Form OENP-24.12-3, the SRM readings as each bundle is lowered in place.

2. IF a fuel bundle is inserted or removed adjacent to an SRM, THEN RECORD new baseline count rate for that SRM on Attachment 6.

3. FOLLOW fuel location on maps by stepping through core component sequence in ShuffleWorics as each fuel move is performed.

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69. G2.02.20 1 Which one of the following time periods are generally part of the MAX/SAFE/GEN operational periods lAW OAl-147, Systematic Approach to Trouble Shooting?

A. January and February ONLY B. June and July ONLY C. July and August ONLY D. January, February, June, July and August Answer: D K/A: G2.02.20 Knowledge of the

- process for managing troubleshooting activities.

(CFR: 41.10 /43.5/45.13)

ROISRO Rating: 2.6/3.8 Objective:

Reference:

None Cog Level: Low Explanation: Al-l47provides specific dates for MAX/SAFE/GEN operational periods. Each of the available choices present options that a student may conclude reasonable, therefore plausible.

Distractor Analysis:

Choice A: Plausible because student may conclude this is reasonable.

Choice B: Correct Answer, see explanation Choice C: Plausible because student may conclude this is reasonable.

Choice D: Plausible because student may conclude this is reasonable.

SRO Basis: N/A OAI-1 47 3.10 MaxlSafelGen: The Max/Safe/Gen period is generally January, February, June, July and August. Within this period, troubleshooting activities with potential of loss generation (Medium or High Risk activities) require Plant General Manager approval E i

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70. G2.02.35 1 Which one of the following meets the conditions required to be in MODE 4 lAW Technical Specifications?

The Reactor Mode Switch must be in:

A. either Shutdown or Refuel and the first reactor head bolt fully tensioned.

B. either Shutdown or Refuel and all reactor head bolts fully tensioned.

C. Shutdown and the first reactor head bolt fully tensioned.

D. Shutdown and all reactor head bolts fully tensioned.

Answer: D K/A: G2.02.35 Ability to determine Technical Specification Mode of Operation.

(CFR: 41.7/41.10 /43.2 /45.13)

ROISRO Rating: 3.6/4.5 Objective: CLS-LP-200-B, Obj. 5 Given a set of plant conditions, determine the plant MODE

Reference:

None Bank question that was last used on the 08 NRC exam.

Cog Level: Low Explanation: With the reactor head bolts less than fully tensioned the plant is in Mode 5 provided the Mode switch is in Shutdown or Refuel. As head bolts are being tensioned the plant is still in Mode 5. When the last head bolt is fully tensioned the plant will transition from Mode 5 to Mode 4 provided the Mode switch is in Shutdown. If the Mode switch is left in Refuel this would transition the plant from Mode 5 to Mode 2 (and most likely constitute a violation of LCO 3.0.4)

Distractor Analysis:

Choice A: Plausible since mode transition from Mode 5 to Mode 4 occurs during head bolt tensioning and the Mode switch can be in Shutdown or Refuel for Mode 5 (if in Refuel when the last bolt tensioned, transition is from 5 to 2, not 5 to 4)

Choice B: Plausible since mode transition from Mode 5 to Mode 4 occurs when the last head bolt is tensioned and the Mode switch can be in Shutdown or Refuel for Mode 5 (if in Refuel when the last bolt tensioned, transition is from 5 to 2, not 5 to 4)

Choice C: Plausible since the Mode Switch position is correct and the transition occurs during tensioning of head bolts Choice D: Correct Answer, see explanation SRO Basis: N/A

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Table li-i (page 1 of 1)

MODES REACTOR MODE AVERAGE REACTOR MODE TITLE SWITCH POSITION COOLANT TEMPERATURE (ZF) 1 Power Operation Run NA 2 Startup Refuel or Startup/Hot NA Standby 3 Hot Shutdown 3 Shutdown > 212 4 Cold Shutdown 1 Shutdown 212 S Refueling Shutdown or Refuel NA (a) All reactor vessel head closure bolts fully tensioned.

(b) One or more reactor vessel head closure bolts less than fully tensioned.

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71. G2.02.421 Unit Two is operating at rated power. Drywell leakage calculations are being performed lAW the CO DSR.

The 08:00 drywell leakage calculations were:

Floor Drain Leakage 1.3 gpm Equipment Drain Leakage 3.5 gpm These leakage values have been constant for several days.

At 1200, the difference in integrator readings is:

Floor Drain 816 gallons Equipment Drain 960 gallons Which one of the following choices is correct lAW RCS Operational Leakage Technical Specifications?

A. RCS Operational Leakage is within limits.

B. Unidentified leakage increase is NOT within limits.

C. Average unidentified leakage is NOT within limits.

D. Average total leakage is NOT within limits.

Answer: B K/A: G2.02.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications. (CFR: 41.7/41.10/43.2 /43.3/45.3)

RO/SRO Rating: 3.9/4.6 Objective: LOl-CLS-LP-200-B, Obj. lOd - Explain the following terms as they apply to the Technical Specifications: Condition

Reference:

None Bank question that was last used on the 08 NRC exam.

Cog Level: High Explanation: Floor drain leakage is 3.4 gpm (816 gal/240 minutes). Equipment drain leakage is 4.0 gpm (960 gallons/240 minutes). Total leakage is 7.4 gpm. Unidentified (floor drain) is <5 gpm and total leakage is <25 gpm, but unidentified leakage increased by 2.1 gpm within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> which does not meet LCO 3.4.4.d (Unidentified leakage increase 2 gpm within previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

9 AIIJ1II Distractor Analysis:

Choice A: Plausible because calculation must be for a four hour period (240 minutes) If one hour (60 minutes) were used, the resulting total leakage would be 29.6 gpm (13.6 floor drains, 16 equipment drains).

Choice B: Plausible because calculation errors could occur resulting in greater than 5 gpm if a four hour period is not used or if candidate associates total floor drain leakage with unidentified leakage.

Choice C: Correct answer, see explanation Choice D: Plausible because calculation errors could occur or candidate could improperly associate identified and unidentified drains.

SRO Basis: N/A RCS Operational LEAKAGE 3.4.4 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Operational LEAKAGE LCO 3.4.4 RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE;

b. S gpm unidentified LEAKAGE averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period;

c. 25 gpm total LEAKAGE averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period; and

d. 2 gpn increase in unidentified LEAKAGE within the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period in MODE 1.

APPLICABILITY: MODES 1, 2. and 3.

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72. G2.03.11 1 Following a large line break in the drywell, H2/02 monitors have been placed in service.

Plant conditions:

Drywell hydrogen 2.5% (ERFIS)

Drywell oxygen 3.5% (ERFIS)

Torus hydrogen 1.4% (ERFIS)

Torus oxygen 3.5% (ERFIS)

Torus level -36 inches Which one of the following choices describes the actions directed by PCCP?

Vent and purge primary containment (1) release rate limits.

Venting from the (2) is preferred.

A. (1) within (2) torus -

B. (1) within (2) drywell C. (1) irrespective of (2) torus D. (1) irrespective of (2) drywell Answer: A K/A: G2.03.11 -Ability to control radiation releases. (CFR: 41.11 /43.4/45.10)

RO/SRO Rating: 3.8/4.3 Objective: LOI-CLS-LP-300-L, Obj. 8c Given the Primary Containment Control Procedure and plant conditions, determine if the following actions are required: Venting the primary containment while staying within radioactivity release rate limits.

LOI-CLS-LP-300-L, Obj. 8d Given the Primary Containment Control Procedure and plant conditions, determine if the following actions are required: Venting the primary containment IRRESPECTIVE of radioactivity release rate limits.

Reference:

None Bank question that was last used on the 04 NRC exam.

Cog Level: High Explanation: PCCP hydrogen leg directs venting and purging primary containment only within ODCM limits when H2 concentration is above 1% provided 02 concentration remains below 4%.

Venting from the torus is preferred due to scrubbing of iodine by the pool.

Distractor Analysis:

Choice A: Correct answer, see explanation Choice B: Plausible because 02 concentration is below 4%, so venting would be within limits. Venting of drywell would be appropriate if torus water level was higher (+6 ft).

Choice C: Plausible because if 02 concentration was 4% or greater PCCP directs CAD initiation and venting lAW OP-24 which vents irrespective of release limits.

Choice D: Plausible because if 02 concentration was 4% or greater PCCP directs CAD initiation and venting lAW 0P-24 which vents irrespective of release limits. Venting of drywell would be appropriate if torus water level was higher (+6 ft).

SRO Basis: N/A EOP-02-PCCP IC B

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16. THROTTLE DWN 2 INLET VLV, CAC-VI63OR LI CAC-V16i AND SUPP POOL N 2 INL VLV, CAC-V162 OR CAC-V16U, as necessary, to maintain drywell AND suppression pool oxygen concentrations less than 5%.

NOTE: Primary containment venting should be initiated to prevent exceeding 30 psig in the drywell and secured when drywell pressure lowers to 25 psig.

The preferred vent path is through the suppression criamber to allow scrubbing of fission products.

18. IF directed by the Site Emergency Coordinator OR Unit CR5 to maintain primary containment pressure less than 30 p519 by venting the suppression chamber, THEN VENT as follows:

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73. G2.03.13 I Access is required to a Unit One plant area for inspection of a suspected steam leak.

E&RC surveys indicate radiation levels in the area of 1100 Mrem/hr at 30 cm and 510 Rads/hr at one meter.

Which one of the following choices completes the statements below lAW 0E&RC0040, Administrative Controls for High Radiation Areas, Locked High Radiation Areas, and Very High Radiation Areas?

This area meets the radiological posting criteria for a (1)

Entrance into this area must be approved by E&RC manager or designee, RP Supervisor, and (2)

A. (1) Very High Radiation Area (2) Unit One CRS B. (1) Very High Radiation Area (2) Shift Manager C. (1) Locked High Radiation Area (2) Unit One CRS D. (1) Locked High Radiation Area (2) Shift Manager Answer: B K/A: G2.03.13 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

(CFR: 41.12/43.4/45.9/45.10)

ROISRO Rating: 3.4/3.8 Objective: LOl-CLS-LP-201-F, Obj. 10 Explain the requirement regarding control of High Radiation Areas per E&RC-0040.

Reference:

None Cog Level: High Explanation: Locked high radiation area (LHRA) criteria is> 1000 mrem/hr at 30 cm but < 500 Rads/hr at one meter. Very high radiation area (VHRA) criteria is 1000 mrem/hr at 30 cm and > 500 Rads/hr at one meter. This question provides criteria for a VHRA which requires RP Supervisor, E&RC Manager and Shift Manager approval lAW OE&RC-0040.

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Distractor Analysis:

Choice A: Plausible because changing either mrem or rad would result in the area being a LHRA. Shift Manager not CRS approval is required for VHRA entry.

Choice B: Correct answer, see explanation Choice C: Plausible because changing either mrem or rad would result in the area being a VHRA.

Choice D: Plausible because changing either mrem or rad would result in the area being a VHRA.

SRQ Basis: N/A OERC-0040 LOCKED HIGH RADIATION AREA (LHRA) 9.2 High Radiation Area (HRA) With Dose Rates Exceeding 1,000 MremlHour at 30 Centimeters Prom the Radiation Source or From My Surface Penetrated by the Radiation, But Less Than 500 RadslHour at 1 Meter From the Radiation Source or From My Surface Penetrated by the Radiation 22 LHRAs 2.2.1 Requirements for entering LHRAs with documented stable

( radiological conditions:

1. Appropriate RWP

2. Radiation Protection pre-job briefing (ref. HPS-NGGC-001 9)

3. Knowledge of current work area radiological conditions

4. Electronic dosimeter with pre-set dose/dose rate alarms

5. Consider use of the telemetry system.

6. Qualified HP Technician with dose rate monitoring instrument to provide intermittent coverage during evolutions with potential to change radiological conditions (e.g.. system breaches, draining process piping, start up of systems using main steam, during HWC injection increases, etcj

7. Positive control of LHRA entrance to prevent inadvertent entry VERY HIGH RADIATION AREA 9.3 High Radiation Areas with Dose Rates Exceeding 1,000 MremlHour at 30 Centimeters From the Radiation Source or From Any Surface Penetrated by the Radiation and Greater Than 500 RadslHour At 1 Meter From the Radiation Source or From My Surface Penetrated by the Radiation

iama as sc :na 9.9 Access to Very High Radiation Areas 9.9:1 WHEN a key to a VHRA is needed, COMPLETE applicable portions of Attachment 1, Part 1.

9.9.2 OBTAIN the signature of the RP Supervisor and the Manager E&RC -

or designee.

9.9.3 NOTIFY the affected units Operations Shift Manager tar approval to enter into a VHRA.

9.9.4 WHEN an entry into a VHRA is determined to be completed, ENSURE that the entrance is closed and lacked. AND COMPLETE applicable portions of Attachment 1, Part 2, for First Physical Verification Signature.

Approvals: Required for Very frtgh Radiation Area)

RP Supervisor (Signaturel Date Manager E&RC or De&gnee (Signature}IDate Operations Shift Manager Notified Shift Managers Name (Print) Date r733wrflr,r

74. G2.04.19 I Which one of the following identifies the EOP flowchart symbol above lAW OEOP-OI-UG, EOP Users Guide?

A. Action Step B. Caution Step C. Critical Step D. Decision Step Answer: C K/A: G2.04.19 - Knowledge of EOP layout, symbols, and icons. (CFR: 41.10/ 4513)

RO/SRO Rating: 3.4/4.1 Objective: LOI-CLS-LP-300-B, Obj. 2d Given the following Emergency Operating Procedure (EOP) flowchart symbols, correctly identify each: Critical Step

Reference:

None Cog Level: Low Explanation: As listed in the Users Guide this is a critical step.

Distractor Analysis:

Choice A: Plausible because an Action step is listed in the Users Guide Choice B: Plausible because a Caution Step is listed in the Users Guide Choice C: Correct Answer, see explanation.

Choice D: Plausible because a Decision Step is listed in the Users Guide SRO Basis: N/A wr . -.,.

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75. G2.04.25 1 Which one of the following choices completes the following statements lAW OFPP-031, Fire Brigade Staffing Roster and Equipment Requirements?

The Fire Brigade Advisor position is filled by (1)

The Fire Brigade Advisor (2) considered available to support activities other than the Fire Brigade during OASSD-02 implementation.

A. (1) anSROONLY (2) is B. (1) any licensed operator (2) is C. (1) anSROONLY (2) is NOT D. (1) any licensed operator (2) is NOT Answer: B K/A: G2.04.25 - Knowledge of fire protection procedures. (CFR: 41.10/43.5/45.13)

ROISRO Rating: 3.3/3.7 Objective: CLS-LP-01 3, Obj. 4d Describe the requirements of the General Fire Plan (PFP-01 3) as they relate to the following: Fire Brigade Advisor FBA.

CLS-LP-01 3, Obj. 9 - Describe the responsibilities of qualified fire brigade members.

Reference:

None Cog Level: Low Explanation: OFPP-031 defines who may fill the Fire Brigade Advisor position and the availability of that position for other tasks. Minimum fire brigade member staffing is specified in the procedure.

The shift incident commander and individual fire brigade members, with the exception of the fire brigade advisor, may not have any other fire duties (i.e. ASSD).

Distractor Analysis:

Choice A: Plausible because there are SRO only positions during emergencies such as Remote Shutdown Panel operator in ASSD procedures.

Choice B: Correct Answer, see explanation Choice C: Plausible because there are SRO only positions during emergencies such as Remote Shutdown Panel operator in ASSD procedures. The shift incident commander and individual fire brigade members, with the exception of the fire brigade advisor, may not have any other fire duties (i.e. ASSD).

Choice D: Plausible because The shift incident commander and individual fire brigade members, with the exception of the fire brigade advisor, may not have any other fire duties (i.e. ASSD).

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1/4eMZW SRO Basis: N/A OFPP-031 3.0 RESPONSIBILITIES 3.5 The Fire Brigade Advisor is responsible for evaluating the operational impact on plant systems and equipment during a fire or other event requiring a response by the Shift Fire Brigade. The Fire Brigade Advisor position is filled by a licensed operator (i.e., RO or SRO). The Fire Brigade Advisor is considered available to support activities other than the Fire Brigade during an ASSD event if required for QASSD-02 or 2ASSD-05, or to ensure control room minimum staffing requirements (1 SRO and 1 RO per unit) are maintained for any unit that is not evacuating its control room.

ATTACHMENT 1 Page 1 of 1 incident Command Fire Brigade Raster Circle one: Dayshift / Nightshift POSITION NAME INITIALS Shift Incident Commander Fire Brigade Member Fire Brigade Member Fire Brigade Member Fire Brigade Member Fire Brigade Advisor Maintenance Contact N/A Security Contact N/A E&RC Contact N/A The Fire Brigade Advisor is considered available to support activities other than the Fire Brigade during an ASSD event if required for OASSD-02 or 2ASSD-05, or to ensure control room minimum staffing requirements (1 SRO and 1 RO per unit) are maintained for any unit that is not evacuating its control room.

OAP-033

c. The Shift Fire Brigade shall not include members of the minimum shift crew necessary for safe shutdown of the unit or any personnel required for other essential functions during a fire emergency.