2005 NRC SRO Exam Flashcards
The plant is operating at 100% power, with the following:
12: OO on May 1 : Line 5 is declared INOP.
08: OO on May 3: Div 1 DG is declared INOP.
12: OO on May 3: Line 5 is declared OPERABLE
16: OO on May 4: Div 3 DG is declared INOP
14: OO on May 5: Div 1 DG is declared OPERABLE.
06: OO on May 6: Line 5 is declared INOP.
08: OO on May 6: Div 3 DG is declared OPERABLE (delay in obtaining parts).
Which one of the following describes correct implementation of Tech Specs for this sequence of
events?
A. Entry into REQUIRED ACTION requiring a plant shutdown existed at NO TIME during this sequence of events. Restore line 5 to operable status no later than
06:OO on May 9.
B. Entry into REQUIRED ACTION requiring a plant shutdown existed at NO TIME during this sequence of events. Restore line 5 to operable status no later than
12:OO on May 7.
C. At 18:OO on May 4, TS 3.8.1, REQUIRED ACTION F.l was required to be entered but this action was exited as permitted by LCO 3.0.2 before the shutdown was required to be initiated. Restore line 5 to operable status no later than 06:OO on May 9.
D. At 16:OO on May 4, TS 3.8.1 REQUIRED ACTION G.l was required to be entered and the plant was required to be placed into MODE 3 but this action was not completed. Restore line 5 to operable status no later than 12:OO on May 7.
References Provided: TS 3.8.1 (ALL, do not provide
the TS bases).
B is correct. Correctly applying the modified time zero on initial entry into CONDITION A when Line 5 is declared inoperable at 12:OO on May 1 results in Line 5 restoration with completion time 6 days later on May 7 at 12:OO. The completion time of 72 hours which started on 06:OO on May 6 (last Line 5 inop condition results in completion time is 06:OO on May 9) is limited to less time because of the “modified time zero - 6 days from discovery of failure to meet the LCO” which IS DISCUSSED IN THE TS BASES. Per TS BASES 83.8.1, RA A.3 DISCUSSION: the third completion time for required action A.3 established a limit on the maximum time allowed for any combination of required AC sources to be inoperable during any single contiguous occurrence of failing to meet the LCO. The 6 day completion time provides a limit on the time allowed in a specified condition after discovery of failure to meet the LCO. The ‘‘W connector between the 72 hours and 6 day completion times apply simultaneously, and the more restrictive must be met. The completion time of required action A.3 allows for an exception to the normal “time zero” for beginning the allowed outage time clock. This exception results in establishing the “time zero” at the time the LCO was INITIALLY NOT MET, instead of at the time Condition A was re-entered . Therefore, Line 5 must be operable no later than 6 days from declaring the LCO statement not met (which was 12:OO on May 1) which means that Line 5 must be operable by 12:OO on May 7 to avoid entering Condition F which is a Condition/Required Action for a plant shutdown.
The plant is operating at 100% power, with the following:
06: 00:00 - High Pressure Core Spray System is out of service with RED clearance applied to CSH Pump breaker
08: 00:00 - A Control Room Evacuation is directed
08: 00:15 - RO scrams the reactor and all control rods insert
08: 00:25 - A Loss of Offsite Power occurs
08: 01:00 - The MSlVs are closed and the remaining immediate operator actions are completed
08: 04:00 - It is determined that RClC cannot be started and the actions to perform Pseudo-LPCI are directed
08: 06:00: Started RHS’P1A
08: 07:00: Opened 4 ADS valves
08: 08:00: RHS A flow at about 7400 gpm
Which one of the following identifies the station procedure used to lineup injection AND the correct interpretation and action with regards to RPV water level at 08:08:30?
A. N2-SOP-78 is used. Level is above TAF, and will remain above TAF without starting additional RHS pumps.
B. N2-SOP-78 is used. Level is below TAF, and will not be restored above TAF until after RHS B loop injection is established.
C. N2-EOP-6 Attachment 30 is used. Level is above TAF, and will remain above TAF without starting additional RHS pumps.
D. N2-EOP-6 Attachment 30 is used. Level is below TAF, and will not be restored above TAF until after RHS B loop injection is established.
References Provided: None
A. Per N2-SOP-78; step 5.1 1: Control of RCIC, ADS and RHS will be transferred to the RSS immediately upon evacuation of the Control Room. These actions are performed as quickly as possible to allow a determination of the availability of RCIC. If RCIC is not available, a contingency for the use of LPCl as an injection source is provided. This contingency will direct the operator to place RHS pump A or B into service and then open all four remote shutdown ADS valves. To avoid lowering level below the top of active fuel, the ADS valves must be opened within (9) minutes of the Scram and MSlV isolation. RHS will then be used to refill the Reactor and maintain normal Reactor level.
The plant is operating at 100% power, with the following:
- All drywell cooling is lost due to an inadvertent Division 1 Containment isolation signal
- Drywell pressure is 0.70 psig and rising slowly
- Drywell absolute pressure is 15.4 psia
- Drywell temperature is 151F and rising slowly
Which one of the following is the maximum time allowed by Technical Specifications before the plant is required to be in MODE 3?
A. 4 hours
B. 13 hours
C. 16 hours
D. 20 hours
References Provided: Tech Spec 3.6.1.3, 3.6.1.4,
3.6.1.5
D is correct. Per LCO 3.6.1.5, Drywell
temperature is above the LCO limit.
Required Action A.l. Eight (8) hours to
restore, then 12 hour to be in Mode 3
The plant is experiencing a transient, with the following:
- One (1) control rod is at position 48, all other control rods are fully inserted
- RPV pressure is 750 psig and lowering 50 psig per minute
- RPV level is -70 inches (Fuel Zone) and lowering eight (8) inches per minute
- Preferred injection systems cannot be started
- Alignment of alternate injection systems has been directed but none of the these systems are reported as aligned
- RHS Service Water Crosstie will be aligned and available for injection in five (5) minutes
Which one of the following is the correct EOP action to be executed at this time based on the above conditions?
A. EOP-RPV is required.
B. EOP-C4 is required
C. EOP-C2 is required
D. EOP-C3 is required
References Provided: EOP-RPV, EOP-C5, EOP-C4, EOP-C2
D. Under the conditions presented, steam cooling (EOP-C3) is required (EOP-RPV L-12 and L-13). In the steam cooling EOP RPV blowdown will be required in the next 2 minutes (-55 inches) before the alternate injection system is aligned. A. EOP-RPV is exited because no injection sources are aligned with a pump running. EOP-C3 is entered. Plausible because the candidate may determine that it is okay to defer blowdown and steam cooling while the alternate injection system is aligned. Also plausible because EOP-C2 requires re-entry into EOP-RPV after the blowdown is performed. 8. EOP-C4, RPV flooding is not required because there is no reason to believe that RPV water level is not known. The indicated RPV water level is valid and even if in EOP-C5 (ATWS) RPV level is still valid since reactor power level would be less than 4% for the control rod position specified. C. If it is determined that EOP-C5 is appropriate and EOP-RPV is exited based on the control rod positions then RPV blowdown is appropriate. However, EOP-C5 is not entered because the reactor will remain shutdown under all conditions without boron based on the shutdown margin definition.
An ATWS is in progress. Following the actions to terminate and prevent injection the following conditions exist:
- Bypass valves failed to open and cannot be opened
- Reactor pressure is being maintained 800-1 000 psig; 2 SRVs are open
- Suppression Pool average water temperature is 120°F
- Suppression Pool level is 199.8 feet
- Control rod insertion has not been established
- SLS failed to inject and cannot be started
- No alternate boron injection system is injecting
- When indicated level reaches -40 inches (FZ), direction is given to reestablish injection and maintain indicated level -70 to -40 inches (FZ)
- With indicated water level at -40 inches (FZ),RPV injection is re-established
- Ten (10) seconds later indicated water level is -10 inches (FZ);reactor power is 5%
Which one of the following is the correct action in response to this transient?
A. Terminate and prevent injection again.
B. Perform a RPV Blowdown per EOP-C2.
C. Direct a new level control band of -70 to -10 inches (FZ).
D. Reduce the injection rate until in the assigned -70 to -40 inches band.
References Provided: EOP-C5
Correct: A. Level rise will cause reactor power to increase and exceed 4%. Override conditions are met to terminate and prevent injection until reactor power lowers below 4% or level is at TAF(-52” FZ at 800 psig). Distractor: B. There is initially a 20°F margin to HCTL, and rise in suppression pool temperature will not require RPV Blowdown at this time. Reactor pressure will be reduced in a controlled manner to stay within the HCTL if it is being challenged, and then if exceeded and cannot restore within HCTL a blowdown would be performed. Distractor: C. Level rise will cause reactor power to increase and exceed 4%. Override conditions are met to terminate and prevent injection until reactor power lowers below 4% or level is at TAF(-52 FZ at 800 psig). Distractor: D. Restoring to the directed level control band is not appropriate, Override conditions are met to terminate and prevent injection again.
The plant is experiencing a transient, with the following:
- A primary system leak has occurred in the secondary containment
- Attempts to isolate the leak have been successful
- Highest Secondary Containment temperature is 165’ F and temperatures have been stabilized
- Field Survey at the site boundary indicate TEDE dose of 50 mRem
- EDAMS TEDE dose projection is reported as 120 mRem at the site boundary
Which one of the following actions is required based on these radiological conditions?
A. Declare an ALERT. Entry into EOP-C2 is required.
B. Declare an ALERT. Entry into EOP-C2 is NOT required.
C. Declare a SITE AREA EMERGENCY. Entry into EOP-C2 is required.
D. Declare a SITE AREA EMERGENCY. Entry into EOP-C2 is NOT required.
References Provided: EPIP-EPP-02 Attachment 1
EAL, NP-EOP-SCIRR
B is correct. ACTUAL field survey data indicates conditions are above the ALERT value of 10 mRem. EDAMS dose projection is above the SAE level of 100 rnRem, the correct classification is SAE 5.2.4. ACTUAL field survey data is used over the projected. A is incorrect. Dose is significantly below the GE level requiring Blowdown per EOP-RR step RR-1, entry into EOP-C2 is not required. With the temperatures stabilized below 21 2" F RPV Blowdown is also not warranted by EOP-SC. Conditions are not going to reach the GE values because the leak is isolated. C and D are incorrect. ALERT is classified based on ACTUAL plant data and not the projected data when ACTUAL data is available.
The plant is experiencing an ATWS transient, with the following:
- An Alert was declared 65 minutes ago
- Emergency response facilities are activated
- 849127, FIRE DETECTED PNL128 W WALL / 261, alarms
- Computer Point Fire Detection:
+00-061-1-5, 333XL E261 Div 1 SWGR
+00-061-1-7, 333XL E261 Div 1 SWGR - Discharge Light for Zone 333XL indicates 2FPL-AOV106 is open
Which one of the following is the correct action at this time in response to the above conditions?
A. Per EPIP-EPP-5A, direct an announcement for the fire brigade to report to the fire location.
8. Per EPIP-EPP-SA, direct an announcement for the fire brigade to report to the alarming fire panel.
C. Per EPIP-EPP-28, direct an announcement for the fire brigade to report the fire location.
D. Per EPIP-EPP-28, direct an announcement for the fire brigade to report to the osc.
References Provided: None
D. When credible evidence exists of a fire condition within the protected area, then per EPIP-EPP-28, direct the CSO to implement the CSO fire fighting checklist. The definition of CONFIRMED FIRE is a condition in which credible evidence exists that a fire is actually occurring. A fire may be considered as confirmed given any of the following: fire alarm/annunciator AND suppression system activation accompanied by actual flow or discharge, or Fire BrigadelLeader report, or SSS judgment. Per the fire fighting checklist, EPIPEPP- 28, Attachment 1 : If the OSC has not been activated then the fire brigade shall report to the fire location. If the OSC is activated, the fire brigade shall report to the OSC. If the location of the alarm is a C02 protected area, immediately evacuate all personnel from the fire location and all areas adjacent to and below this area. This area is a C02 area as identified by the "L" in the computer printout.
The plant is operating at 100% power, with the following:
- Annunciator 603103, RPS A REACTOR PRESSURE HIGH TRIP, alarms
- Computer Point, ISCUC05 RPS A1 RX PRESS HI TR, alarms
- I&C reports instrument 2ISCPIS1678 (B22-N678A RPV HIGH PRESS, 2CECPNL609) is failed high
- The plant responded per design
Which one of the following is the correct action in response to the above conditions?
A. Maintain the half scram on RPS Channel A. The ATWS RPT function remains operable.
B. Initiate actions to be in MODE 3 within 12 hours. The ATWS RPT function remains operable.
C. Maintain the half scram on RPS Channel A and trip the associated ATWS RPT channel within 14 days.
D. Initiate actions to be in MODE 2 within 6 hours because of the inoperable ATWS RPT function.
References Provided: P&ID 28A (already provided for
RO) and JUST TS 3.3.1 .I pages 3.3.1.1 -1, 2, 3 and 9
(includes Table 3.3.1 .l-1 page 2 of 3). Must leave
allowable values column intact because this is required
for SRO 17. NO SURV REQUIREMENTS and NO
BASES allowed, TS 3.3.4.2 (NO SURV
REQUIREMENTS and NO BASES). NO SR and NO
BASES for these two specs (3.3.1.1 and 3.3.4.2)
because they could provide the answer other question
intended to be answered from memory.
A. There are four instruments (2 per channel) for the Reactor Vessel Steam Dome Pressure High function of RPS Instrumentation. Refer to TS Table 3.3.1 .l-1, Function 3. With one of the instruments inoperable, place the associated channel in trip within 12 hours or place the associated trip system in trip within 12 hours. For the failed instrument, a half scram automatically occurred satisfying the requirement of TS 3.3.1.1, Condition A, FW A.2. As long as the half scram is inserted on RPS channel A, the plant can operate indefinitely in this condition until the instrument is restored to operable status at which time the half scram can be reset. ATWS RPT function is not provided by this same instrument. This function remains operable
The plant is in MODE 5 with core reload in progress per N2-FHP-13.3 Core Shuffle with the following:
0800 - Annunciator 875111, SPENT FUEL POOL LEVEL HIGH/LOW alarms
0800 - Spent Fuel Pool (SFP) level is 352 ft 9 inches and steady
0802 - 2SFC*AOV33A and B, SFP NORMAL MAKEUP are open and SFP level begins to rise
0815 - An irradiated fuel bundle is being lowered into the core
0816 - A malfunction of the Refuel Bridge causes the grapple and fuel bundle to rapidly lower the last 4 inches into the core
0817 - Annunciator 603209, SRM SHORT PERIOD alarms and clears 5 seconds later
0818 - SRM count rates are rising steadily
Which one of the following actions is required to be taken and what is the reason for that action?
A. Declare an Unusual Event because excessive SFP leakage is occurring.
B. Evacuate the Refuel Floor because an irradiated fuel bundle has been dropped.
C. Initiate Boron injection to the core because an inadvertent criticality event is occurring.
D. Remove the fuel bundle within 1 hour because Shutdown Margin is not within LCO limits.
References Provided: None
C is correct. Inadvertent Criticality is occurring with SRM count rate rising steadily. Per N2-SOP-39, initiate SLS injection. A is incorrect. SFP level is being recovered by the normal makeup valves, 2SFC*AOV33A and B. An Unusual Event (UE 1.5.1) declaration is required if SFP level cannot be restored and maintained above the low water level alarm. B is incorrect. Grapple lowered 4 inches and there is no indication that the bundle is no longer grappled. A dropped fuel bundle event is NOT in progress. D is incorrect. Removing fuel bundle does not restore SDM per the LCO. Mode 5 actions require immediate suspension of core alterations and do not provide a 1 hour completion time.
Which one of the following Radiation Monitoring events requires that you assume the role as Emergency Director (ED)?
A. A coolant leak at one control rod HCU causes the local ARM to indicate yellow on DRMS.
B. When changing a TIP the general area ARM goes offscale high until the TIP is in the transfer cask.
C. During LPRM removal, one local ARM goes upscale before the LPRM is lowered and submerged ten feet.
D. A fuel assembly dropped onto the reactor core causes an automatic containment isolation.
References Provided: None
D. A fuel assembly dropped in the reactor cavity causing containment isolation means that fuel assembly damage occurred. The perforation of fuel cladding caused high radiation levels on the refueling floor due to gaseous release. When radiation monitors 2HVS*RE14A or RE1 48 actuate, a secondary containment isolation results. Per EAL 1.4.2: Valid Rx Bldg above Refueling Floor Radiation Monitor 2HVR'RE14A or 2HVR'RE14B, Gaseous Radiation Monitors (Channel 1 ) isolation due to a refuel floor event, the event is classified as an alert per EPIP-EPP-02 and EAL. Per EPIP-EPP-02, 2.1: Upon initial declaration of an emergency, assumes the role of SSS/Emergency Director (SSS/ED) and functions as the SSS/ED until relieved of those duties by the on-call EDlRM, other SRO, or the emergency is terminated. Declaring an alert requires assuming the role of the Emergency Director. Per N2-SOP-39 discussion 5.4.2: See EPIP-EPP-02, Attachment 1, 9.1, Other for event classification following a dropped fuel bundle or other events which have the potential to have caused damage to fuel bundles.
The plant is in MODE 3, following a planned shutdown. The shutdown was achieved using N2-OP-101C Section G.2.0 Reactor Shutdown by Mode Switch to Shutdown Scram (Soft Scram), with Feedwater Level Control in Manual. Reactor Water Level is being controlled at 150 inches and stable.
THEN …… Reactor water level lowers to 95 inches due to an equipment failure resulting in the following:
- HPCS initiates
- HPCS Diesel Generator fails to start
- HPCS secured per N2-OP-33, Section G.1.0, Shutdown to Standby Following Initiation
- Operators restore level control to the normal band 160 to 200 inches and are controlling Reactor Water Level with Feedwater Level Control in Manual
Which one of the following is the correct reporting requirement for the above conditions?
A. 4 hour report. NO LER is required.
B. 4 hour report. LER is required within 60 days
C. 8 hour report. NO LER is required.
D. 8 hour report. LER is required within 60 days.
References Provided: 10CFR50.72,10CFR50.73
B. 4 hour report. LER is required within 60 days. 1 OCFRS0.72(b)(2)(iv)(A): 4-hour report is required if any event that results or should have resulted in ECCS discharge into the reactor coolant system (HPCS injected into the reactor vessel) as a result of a valid signal except when the actuation results from and is part of a preplanned evolution during testing of reactor operation (not pre-planned). lOCFR50.72(b)(3)(lV)(A): 8-hour report is required if any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B), except when the actuation results from and is part of a pre-planned evolution during testing of reactor operation. EXCEPT when reported under paragraphs (a), (b)(l), or (b)(2), the licensee shall notify the NRC as soon as possible and in all cases within 8 hours of the occurrence. Because a 4-hour report is required the 8-hour report is not required. 4- hour report because there was injection into the reactor vessel. RPS Actuation is part of a preplanned sequence (shutdown by scram reactor mode switch to shutdown, & low level scram signal is expected during this event) and not reportable. 1 OCFR50.73( b)(3)( IV)(A): LER is required if any condition or that resulted in manual or automatic actuation (automatic actuation occurred) of any of the systems listed in paragraph (2)(iv)(B) (HPCS is included) EXCEPT when: The actuation resulted from and was part of a pre-planned sequence during testing or reactor operation (not preplanned), OR . The actuation was invalid (it was invalid) AND occurred while the system was properly removed from service (system was in standby) OR occurred after the safety function had been already completed (system was in standby).
With the plant at 100% power, with the following:
- N2-SOP-97 is entered
- Both RPM-EPAs 2RPMACB1 B and 2RPMACB2B tripped
- 2RPM-MG1 B is not running
- Power source selector switch is swapped from NORM to ALT
- When performing the actions for resetting of RPM-EPAs, the operator is required to manually defeat the overvoltage protective function for 2RPMACB1B and 2RPMACB2B because the voltage on the bus is above the overvoltage reset value
- 2RPMACB1B and 2RPMACB2B are reset and closed
Which one of the following is the correct COMPLETION TIME to restore at least one of the EPA breakers to operable?
A. 1 hour COMPLETION TIME because a scram logic bus is inoperable
B. 1 hour COMPLETION TIME because a scram solenoid bus is inoperable
C. 72 hour COMPLETION TIME because a scram logic bus is inoperable
D. 72 hour COMPLETION TIME because a scram solenoid bus is inoperable
References Provided: Tech Spec 3.3.8.2 and 3.3.8.3
B. Per SOP-97, 5.6, RPS Scram Solenoids - Alternate Power: When using alternate power to supply the RPS Scram solenoids, bus voltage may be too high to reset the overvoltage trip relay on 2RPM*ACBlA(B) OR 2RPM*ACB2A(B). If any tripped EPA breaker cannot be reset and turned ON due to voltage on the bus exceeding the reset value, shift supervision shall determine which of the following methods is best, considering present plant conditions, to facilitate resetting the affected breaker(s): - Manually defeat the overvoltage protective function at the breaker(s) UNTIL load on the bus is reestablished by resetting the half scram (Preferred method). This will require entry into TS 3.3.8.3, Condition A, with one EPA breaker defeated on one bus (72 hour completion time) OR Condition B, with both EPA breakers defeated on one bus (1 hour completion time). TS 3.3.1 .I also applies with a 1 hour completion time. OR - Lower bus voltage by adjusting the load tap changer for the transformer supplying
Core shuffle in progress, with the following:
- The next step requires movement of the fuel assembly at core location 25-24 to core location 55-34
- Before latching the next step, SRM A and SRM C are declared inoperable
Which one of the following describes which SRM must be restored to OPERABLE status and why?
A. SRM A because it is in the quadrant from which the fuel is removed
B. SRM C because it is in the quadrant into which the fuel is being moved..
C. SRM A because it is in a quadrant into which the fuel is being moved.
D. SRM C because it is in a quadrant adjacent to a fuel move quadrant.
References Provided: Core Map
D. is correct because 3 SRM's must be OPERABLE to complete this particular fuel move. SRM D is in the quadrant form which the bundle will be removed. SRM B is int he quadrant to which the bundle will be moved. SRM A or C are adjacent to either of the above quadrant.
The plant is operating at 100% power, with the following:
- I&C is ready to perform N2-ISP-MSS-Q003, Quarterly Functional Test and Trip Unit Calibration of Main Steam Line Low Pressure Instrument Channels. This test performs channel calibration of the Rosemount Analog Trip System Master Trip Units (MTU) for Main Steam Line Low Pressure instrument Channels 2MSS*PIS1020A-D (B22-N676A-D)
- 2MSS*PIS1020A MTU Trip calibration starts AT 0800
Which one of the following describes the correct implementation and basis for entry into the associated CONDITION and REQUIRED ACTION of Tech Specs?
A. Entry must be logged as of 0800 because associated function does not maintain isolation capability.
B. Entry must be logged as of 0800 because Safety Function Determination Program requirements are not met.
C. Entry may be delayed until 1400 because associated function maintains isolation capability.
D. Entry may be delayed until 1400 because separate entries are allowed for each channel to be tested.
References Provided: None
C. At 0800 when the channel calibration begins, the channel is inoperable. Per TS 3.3.6.1 Surveillance Requirement note, “When a channel is placed in an inoperable status solely for the performance of required surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains isolation capability. "
The plant is operating at 100% power, with the following:
- Power Control provides notification of unstable grid conditions with imminent loss of off site power
Which one of the following actions is required?
A. Per SOP-3, immediately scram the reactor and enter SOP-101C.
B. Per SOP-70, immediately scram the reactor and enter SOP-101C.
C. Per SOP-3, start and run Div 1 and Div 2 Diesel Generators ioaded.
D. Per SOP-70, start and run Div 1 and Div 2 Diesel Generators unloaded.
D. Per SOP-70, start and run Div 1 and Div 2 Diesel Generators unloaded.
Upon completion of a 400-day run and shutdown for a refueling outage, control rod withdrawal using control rod sequence A2UP for has been commenced the subsequent startup. Control rods in RWM Step 1 are being withdrawn.
- The last control rod movement was withdrawing control rod 50-1 5 to position 48
- RO selects control rod 10-1 5 and the following are observed:
> Select Error and WITHDRAW rod block are indicated on the RWM display.
> WITHDRAW BLOCK status light is ON (Rod Select Module).
Which one of the following is the correct action in response to the above conditions?
A. Bypass the RWM and continue control rod withdrawal. Additional staff is not required to be stationed.
B. Bypass the RWM and station an additional RO or SRO to verify control rod movements before continuing control rod withdrawal.
C. Fully insert the withdrawn control rods in reverse order using RMCS. Continue the startup after repairing the RWM hardware failure.
D. Fully insert the withdrawn control rods by a manual reactor scram. Continue the startup after correcting the RWM BPWS error.
References Provided: TS 3.3.2.1 (including Table
3.3.2.1-1), Rod Movement Sheets for the first twelve
rods of sequence A2UP.
Sequence A2, RSCS Group 1 / RWM Step 1: Control rods in this group are withdrawn in the following sequence; 02-39, 58-39, 58-23, 02-23, 18-55,42- 55, 42-07, 18-07, 10-47, 50-47, 50-15, 10-15. Control rod 10-15 is the 12th control rod.
B. The RWM is inoperable. Although the control rod selected is the next control rod in the control rod sequence and is a control rod within this RWM step (group of control rods), the RWM has incorrectly restricted movement of this control rod. Since the RWM was functioning correctly, the supposed error is that the control rod sequence package is not loaded correctly into the RWM. Therefore, the RWM recognizes the selection and attempted control rod withdrawal as an out-ofsequence rod movement and enforces a Select Error and Rod Block. This requires declaring SR 3.2.2.1.8 not met and the RWM inoperable. With the RWM inoperable, suspend control rod movement except by scram EXCEPT if the following can be satisfied: Determining that more than 12 control rods are withdrawn per TS 3.3.2.1,RAC.2.1.1 (12rodsarenot withdrawn) OR that no startups with the RWM inoperable have occurred in the last year per TS 3.3.2.1, RA C.2.1.2 (met since 400 day run and then shutdown for a refueling outage), control rod withdrawal can be continued by bypassing the RWM after completing TS 3.3.2.1 RA C.2.2. Since N2-OP-95A, Section H.l .O requires additional personnel in the control room for the startup be assigned specifically for verification of control rod movements, TS 3.3.2.1 RA C.2.2 is met. Sequence A2, RSCS Group 1 / RWM Step 1: Control rods in this group are withdrawn in the following sequence; 02-39, 58-39, 58-23, 02-23, 18-55,42- 55, 42-07, 18-07, 10-47, 50-47, 50-15, 10-15. Control rod 10-15 is the 12th control rod.
During the performance of the N2-ISP-ISC-002,CHANNEL CALIBRATION, for the Reactor Vessel Water Level Low Level 3 function of Reactor Protection System, the as-found and as-left values for instrument 21SC*LIS168OC are reported by I&C:
- As-found 157.70 inches
- As-left 159.49 inches
Which one of the following describes the operability of this instrument upon learning the as-found value and after completing the calibration and an instrument drift evaluation?
A. Upon learning the as-found value the channel was inoperable. After calibration the channel became operable.
B. Upon learning the as-found value the channel was inoperable. After calibration the channel remained inoperable.
C. Upon learning the as-found value the channel remained operable. After calibration the channel remained operable.
D. Upon learning the as-found value the channel remained operable. After calibration the channel became inoperable.
References Provided: TS 3.3.1.1 (all, no bases)
A . Per TS Table 3.3.1 .I-1, Function 4, the allowable value is 35l7 .8 inches. Per ARP 603105,21SC*LIS1680C trip set point is 159.3 inches. Because the as-found value (1 57.70 inches) is not within the allowable value, the instrument must be declared inoperable. After the calibration is complete, the as-left value (1 59.49 inches) is within the allowable value and is considered operable. The asleft value is also more conservative than the trip set-point justifying this instrument being restored to operable status. Per the TS bases for RPS Instrumentation, upon learning the allowable value is exceeded, the instrument is inoperable. Provided the instrument as-found value is below the allowable value it is operable even though it may be above the specified trip set point. Provided the as-left value is below the allowable value it is returned to operable status.
The plant is preparing for refueling, with the following:
a The first fuel shuffle is scheduled to start as soon as N2-OSP-RMC-W@002, Reactor Mode Switch Functional Test of Refuel Interlocks, is complete (in about
4 hours).
- Following the first fuel shuffle, the in-vessel servicing window includes the replacement of 10 control rod blades
- Maintenance informs you the monorail hoist load cell on the refueling bridge must be replaced
- The estimated time to replace and calibrate the load cell is 12 hours
Which one of the following is the impact of the monorail hoist load cell problem and the correct action based upon this impact?
A. Starting the first fuel shuffle must be delayed. Although this load cell does not support testing of the other hoist interlocks, this load cell must be operable prior to starting fuel movement.
B. Starting the first fuel shuffle must be delayed. Place the surveillance on hold until this load cell is operable and can be used to support testing of the frame mounted and main hoist interlock checks.
C. The first fuel shuffle can start at the scheduled time by deferring (NA) the monorail hoist checks. The monorail hoist should be tested at a later time to
support fuel support removal for control rod blade change out during the in vessel servicing.
D. The first fuel shuffle can start at the scheduled time by deferring (NA) the monorail hoist checks. The monorail hoist should be tested at a later time to support double blade guide movement for control rod blade change out during the in-vessel servicing.
References Provided: None
C. Refueling interlock checks are only necessary for the equipment used during the actual fuel movements, which include the refueling bridge/trolley and main hoist and the associated support systems such as air. The monorail hoist is not used to move nuclear fuel and does not have to be tested. The monorail hoist and/or frame-mounted hoist are used to remove and install control rod blades, not the main hoist, therefore the monorail hoist should be operable from those activities. Since the cells containing the control rod blades to be changed out will have no fuel, movement of these control rods is not considered a core alteration, however, it is prudent to ensure that equipment being used (i.e., the monorail hoist, frame-mounted) is tested properly prior to its use. If the load cell was not operable, any control rod blade restrictions to its removal (Le. snagged) would not be noticed and interrupted (Hoist Jam) with an inoperable load cell and could result in equipment damage and/or personnel injury. The process for control rod blade change out includes using the monorail hoist to remove the Fuel Support Piece, the Frame-mounted hoist to remove the old control rod blade and then insert the new control rod blade, and then the Fuel Support Piece is reinserted into that location using the monorail hoist.
The plant is operating at 100% power, with the following:
00: 00 on May 1 : SGT A and SGT B are declared inoperable
00: 55 on May 1 : Recirc flow is lowered for a plant shutdown
08: 00 on May 1 : Reactor mode switch placed to shutdown
09: 00 on May 2: Average RCS temperature is lowered to 199°F
Which one of the following is the implication of the above actions?
A. The power reduction was NOT initiated within the specified time.
B. The plant was NOT placed in Mode 2 within the specified time.
C. The plant was NOT placed in Mode 3 within the specified time.
D. The plant was NOT placed in Mode 4 within the specified time.
References Provided: T.S. 3.6.4.3
B. The plant was not placed in Mode 2
within the specified time.
With both SGT trains inoperable, TS
LCO 3.0.3 applies.
The plant is operating at 100% with the following:
09: 00 on May 1 : Per TRSR 3.4.1.3 the last satisfactory channel check for Reactor water continuous conductivity recorders and conductivity was 0.8 umho/cm
09: 00 on May 4: Per TRSR 3.4.1.3 RCS A Loop & WCS filter continuous conductivity recorders are declared inoperable
12: 00 on May 4: Conductivity measurement reported at 1.02 umhos/cm at 25°C
- SRO determines entry into TRM 3.4.1 CONDITION B is required prior to 12:00 on May 5 because Required Action A.2 cannot be completed
Which one of the following describes the SRO determination regarding the TLCO?
A. SRO correctly applied the TLCO. CONDITION B applies to REQUIRED ACTIONS A.1. A.2 and A.3. COMPLETION TIME for ACTION A.2 is not met.
B. SRO incorrectly applied the TLCO. Application of TRSR 3.0.2 allows an extension of CONDITION A.2 COMPLETION TIME to 30 hours prior to entering
CONDITION B.
C. SRO incorrectly applied the TLCO. CONDITION B only applies to REQUIRED ACTION A.3. REQUIRED ACTION A.2 is addressed by TRSR 3.4.1.1.
D. SRO incorrectly applied the TLCO. CONDITION B should be entered at 12:00 on May 4 because TRSR 3.4.1.3 cannot be performed.
References Provided: TRM: T3.4.1 and TRM Table
T3.4.1-1
C. is right. Condition B of TLCO 3.4.1 applies only if the required actions and associated completion times of A.3 are not met. If there is a problem with the continuous conductivity recorder TRSR 3.4.1 .I takes affect. A and D are wrong because CONDITION B only applies to A.3. B. is wrong because CONDITION B does not apply if A.2 is not done and TRSR 3.0.2 is irrelevant in this situation.
The plant is operating at 100% power, with the following:
- A single Division 1 Drywell Pressure ECCS initiation instrument fails downscale requiring corrective maintenance
- The SM declares the Division 1 Drywell Pressure instrument inoperable
Which one of the following describes the ACTIONS required by Technical Specifications?
A. Declare CSL and A RHS subsystems inoperable immediately and trip the High Drywell pressure channel with a COMPLETION TIME of 24 hours.
B. Declare CSL and A RHS subsystems inoperable within 1 hour and trip the High Drywell pressure channel with a COMPLETION TIME of 24 hours.
C. CSL and A RHS subsystems remain OPERABLE and trip the High Drywell pressure channel with a COMPLETION TIME of 12 hours.
D. Declare CSL and A RHS subsystems inoperable immediately and trip the High Drywell pressure channel with a COMPLETION TIME of 12 hours.
References Provided: Tech Spec 3.3.5.1 and 3.5.1
A. is correct because the definition of OPERABILITY requires all support instrumentation to be OPERABLE. Therefore CSL and A RHR are both declared inoperable. The completion time of 24 hour applies to 1 .c. in Tech Spec 3.3.5.1. B: is incorrect because the trip function C: is incorrect because CSL and A RHR do not meet the definition of OPERABLE. Also the completion time for channel trip is 24 hours. D: is incorrect because the completion time is 24 hours.
A core shuffle is in progress. Which one of the following conditions meets the Fuel Handling Procedure (FHP) criteria for stopping fuel movement?
A. Last performance of the Refueling Platform Interlocks Test was completed fortyeight (48) hours ago.
B. A fuel assembly is lowered four (4) feet into a core location when it is recognized that this is the incorrect location.
C. The fuel assembly nose piece is lowered to two (2) feet above the core top guide before establishing the correct orientation.
D. A fuel assembly is moved to the spent fuel pool and the rod block interlock light clears when the bridge is clear of the reactor core.
References Provided: None
B. FUEL LOADING ERROR: The placement of a fuel assembly in the core in a location other than that specified by the fuel movement instructions. This includes partial insertion of a fuel assembly into the reactor core. Per FHP 13.3 Attachments 5, criteria for stopping fuel movement, a fuel loading error requires stopping fuel movement. A. This interval for this surveillance is 7 days C. It is acceptable to lower the fuel assembly before establishing the correct orientation; however, the correct orientation is to be established before lowering the fuel assembly into the assigned core location. D. This is correct operation of the refueling interlocks. The rod block interlock light is lit when the refueling bridge is over the reactor core and the main hoist is loaded. When clear of the reactor core (proximity switch) the rod block clears.
A plant startup is in progress with the following:
- Reactor power is 14%
- Containment Purge is required to reduce oxygen concentration to within Tech Spec limits
- Standby Gas Train A has just been declared inoperable
Which one of the following identifies the required containment purge outlet flowpath?
A. Simultaneously from the Suppression Chamber and Drywell through 2GTSSOV102, the 2 inch purge valve.
B. Simultaneously from the Suppression Chamber and Drywell through 2GTSAOV101 the 20 inch purge valve.
C. From the Suppression Chamber first through 2GTSSOV102, the 2 inch purge valve.
D. From the Suppression Chamber first through 2GTSAOV101, the 20 inch purge valve.
References Provided: None
C is correct. N2-OP61A Precaution and Limitation D.12.0 prohibits simultaneous purging from the DW and SC to prevent establishing a Suppression Pool bypass leakage pathway. SR 3.1.6.3.1 and N2-OP- 61A D.3.0 allow purging with only one GTS subsystem operable provided that 2GTS*AOV101 (20 inch) valve is closed. Suppression Chamber purge should be performed before Drywell purge to minimize chance of DW SC vacuum breaker opening, per N2-OP- 61A E.2.18 Note 1.
The plant is experiencing an event, with the following:
- Plant conditions justify the declaration of an ALERT
- Before an ALERT is actually declared, plant conditions improve and now plant conditions only justify the declaration of an UNUSUAL EVENT.
- Plant conditions justifying the declaration of an ALERT are no longer present.
Which one of the following describes correct actions to properly classify the event and notification in response to the above conditions?
A. Declare and report an UNUSUAL EVENT; circle “Unusual Event’‘ in block 4 of the notification fact sheet. In block 8 of the same notification fact sheet, state that conditions for an ALERT were momentary and the time/date of its termination.
B. Declare and report an ALERT by circling “Alert” in block 4 of the notification fact sheet. Submit a separate notification to indicate the change in classification to an UNUSUAL EVENT by circling “Unusual Event“ in block 4 of the notification fact sheet.
C. Declare and report an UNUSUAL EVENT; circle “Unusual Event” in block 4 of the notification fact sheet. No mention of the momentary ALERT and its termination is required in this notification, nor in any preceding or subsequent notifications to this one.
D. Declare and report an ALERT by circling “Alert” in block 4 of the notification fact sheet. In block 8 of the same notification fact sheet, state that conditions for an ALERT were momentary and the timeldate of its termination, and current conditions only justify an UNUSUAL EVENT and the time/date of its declaration.
References Provided: NOTIFICATION FACT SHEET
- PART 1 (ensure the procedure number and revision in the footer of the fact sheet is removed - blacked out)
A. IF: An EAL has been met or exceeded, but the EAL threshold or emergency condition no longer exists prior to making the emergency declaration (Transitory event), THEN: Classify current conditions (UNUSUAL EVENT) and declare the emergency, if necessary. Make notifications required for the declared emergency in accordance with EPIP-EPP-20. Notify State, County and NRC of transitory event (even if no emergency is declared).
Given the following conditions:
- An ATWS concurrent with a LOCA has been in progress for twenty (20) minutes
- Reactor power is 15% and steady
- Reactor pressure being maintained 800-1 000 psig
- Suppression Pool water level is within the TS limit
- Suppression Pool Temperature is 130°F and rising 5°F per minute
- RPV water level is -75 inches (Fuel Zone) and lowering 5 inches per minute
- EOP-C2, RPV Blowdown, has iust been entered
Which one of the following is the correct emergency classification won entrv into EOP-C2, RPV Blowdown, assuming the correct classification is initially made based on the stated conditions?
A. Continue at an ALERT
B. Continue at a SITE AREA EMERGENCY.
C. Reclassify from an ALERT to a SITE AREA EMERGENCY
D. Reclassify from a ALERT to a GENERAL EMERGENCY
References Provided: Unit 2 EAL Matrix
B. Before the low RPV water level occurred, a Site Area Emergency should be declared because of the Failure of RPS, including RRCS, to bring the reactor subcritical. Upon entry into EOP-C2, RPV Blowdown, the classification should remain at a Site Area Emergency until after an evaluation it is determined if low pressure systems are capable of restoring and maintaining RPV water level. Escalation to a General Emergency is based on actual or imminent substantial core damage or melting with potential loss of primary containment. These conditions are not present at this time because adequate core cooling is still assured by presence of the mechanisms for adequate core cooling. Also, General emergency is based on not being able to maintain within the HCTL. The HCTL is not exceeded and is not the reason for the blowdown, the margin to the HCTL will increase substantially as rector pressure is rapidly lowered from the blowdown.
