TS

Unit Two reactor pressure instrument B21-PT-N023A has been declared inoperable and has been placed in the tripped condition to comply with LCO 3.3.1.1. Consider the following reasons it may be desired to remove the instrument from the tripped condition:

1. For a PMTR to demonstrate the instrument has been repaired and is operable
2. For additional troubleshooting to determine the cause of the instrument failure
3. To allow surveillance testing of other RPS channel A instruments
4. To allow surveillance testing of RPS channel B instruments

Which of the above is permitted by Technical Specifications?
A. 1 only.
B. 1 and 2 only.
C. 1, 3, and 4 only.
D. All four are permitted.

Unit Two Reactor Water Level trip units B21-LTM-N024A-1-1 (Low Level 2) and B21-LTS-N024A-1-2 (Low Level 3) are being placed in an inoperable status to perform SR 3.3.6.1.2 and SR 3.3.6.1.3. All other instrumentation is Operable.

Which of the following is correct per Technical Specifications?


A. Entry into LCO 3.3.6.1 Condition A may be delayed for up to 2 hours to perform the surveillance testing.
B. Entry into LCO 3.3.6.1 Condition A may be delayed for up to 6 hours to perform the surveillance testing.
C. LCO 3.3.6.1 Condition A must be entered immediately, the channels must be placed in trip within 12 hours.
D. LCO 3.3.6.1 Condition A must be entered immediately, the channels must be placed in trip within 24 hours.

Unit Two is in Mode 1 with SBGT Fan 2A under clearance for the past 12 hours. The
required actions of LCO 3.6.4.3 Condition A have been entered. 480 VAC Substation E8 trips on fault.

What is the status of Secondary Containment Operability in accordance with Technical Specifications? Secondary Containment is:

A. Operable, SBGT does not support Secondary Containment Operability.
B. Inoperable, and actions of LCO 3.6.4.1 are required to be entered per LCO 3.0.6.
C. Inoperable, but actions of LCO 3.6.4.1 are not required to be entered per LCO 3.0.6.
D. Operable, SBGT does support Secondary Containment Operability, but Secondary Containment is still Operable per LCO 3.0.6.

Engineering has written as A/R for Unit 2 stating that the GM detector for 2-D12-N010A, Reactor Building Vent Radiation Monitor, may not function as designed due to age related degradation. A review of other documentation reveals 2-D12-N010B was recently replaced and is not a condern. This information was presented to you at 0800 on 9/04/2003.

Based on this information, the required actions per Tech Specs are:

A. Place channel in the tripped condition no later than 0800 tomorrow (9/05).
B. Place channel in the tripped condition no later than 2000 today (9/04).
C. Restore isolation capability no later than 0900 today (9/04).
D. Immediately isolate the affected penetration.

Unit 2 is conducting a plant startup per 0GP-02. 0PT-11.1.2, “AUTOMATIC DEPRESSURIZATION SYSTEM AND SAFETY RELIEF FALVE OPERABILITY TEST,” is in progress. The following conditions exist:

- Suppression Pool water temperature is 86°F and rising.
- Both loops of Suppression Pool cooling are in service.

While performing the test, SRV “K” control switch fails causing the valve to remain open for an extended period of time. The operator pulled the AOP-30 fuses and the valve was verified closed.

No heat addition to the Suppression Pool is occurring. The Suppression Chamber temperature is now 112°F and lowering.

In accordance with Unit 2 Technical Specifications, which ONE of the following actions is required?


A. Manually Scram the reactor AND depressurize the reactor vessel to less than 200 psig within 12 hour.
B. Manually Scram the reactor AND verify Suppression Pool temperature is < 120°F once per 30 minutes AND be in MODE 4 in 36 hours.
C. Commence reactor shutdown per GP-05 within 1 hour AND reduce thermal power to range 7 on IRM’s within 12 hours AND maximize suppression pool cooling.
D. Suspend all testing that adds heat to the Suppression Pool AND verify supression pool temperature is <120°F once per 30 minutes AND maximize suppression pool cooling.

References: Tech Specs 3.6.2.1, Suppression Pool Average Temperature

A hydrualic snubber was declared inoperable on 06/10/07 @ 1000 IAW LCO 3.0.8 and TRM 3.21 condition A. The snubber was replaced with an OPERABLE snubber on 06/11/07 @ 1600. On 06/12/07 @ 1200, the supported system was determined to be unacceptable for continued operation based on an analysis of the snubber failure mode. The required action is to:


A. Continue unit operation; TRM requirements are met.
B. Declare the supported system inoperable immediately.
C. Declare the supported system inoperable at 1000 on 06/13/07.
D. Initiate astions to repair the snubber and the NRC Regional Administrator shall be notified in writing of the test plan within 7 days.

Unit One (1) is exiting a refueling outage and maintenance has just commenced tensioning the Reactor Head bolts. Operations is venting the North HCU bank. While withdrawing rod 38-15, a rod drift annunciator is received and the CO identifies that rod 38-15 has lost all position indication.
Based on the above conditions what actions are required by Technical Specification?

A. Perform action to enforce a rod withdrawal block within one hour.
B. Immediately fully insert and then electrically disarm rod 38-15.
C. Immediately fully insert all insertable rods or place the reactor mode switch in shutdown within one hour.
D. Disarm all rods within a five by five array centered on rod 38-15.

A plant transient and subsequent safety relief valve malfunction results in reactor steam dome pressure reaching 1300 psig.

Which one of the following choices completes the following statements?

Reactor vessel design pressure __________ been exceeded.

Tech Spec 2.1.2, Reactor Coolant System Pressure Safety Limit ___________ been exceeded.

A. has;
has
B. has;
has not
C. has not;
has
D. has not;
has not

Considering a 31-day surveillance frequency, which one of the following is the last day that the surveillance can be performed and still meet the requirements of Technical Specifications?

The last day the surveillance must be performed by is within of the previous
performance of the surveillance.

A. 32 days
B. 38 days
C. 45 days
D. 62 days

During normal full power operation, I&C has requested removing annunciator card AOG SYSTEM DISCH RAD HIGH from service for trouble shooting of the annunciator. This annunciator is listed on 0OI-01.08, Control of Equipment and System Status, Attachment 11, Technical Specification/TRM/ODCM Identified Annunciators.

The trouble shooting activity will take place early in the shift and last 2 hours.

Which one of the following identifies whether ODCM Radioactive Gaseous Effluent Monitoring Instrumentation 7.3.2 action statement entry is required and also identifies whether 0OI-01.08, Attachment 10, Annunciator Removal From Service Form, is required to be completed for this activity?

A. action statement entry is required;
Attachment 10 must be completed.
B. action statement entry is required;
Attachment 10 is not required.
C. action statement entry is not required;
Attachment 10 must be completed.
D. action statement entry is not required;
Attachment 10 is not required.

During power operation, the 1A Recirc pump has been removed from service due to a failed recirc pump seal.

- Core Flow Recorder B21-R613 indicates 38 Mlb/hr core flow
- Process computer point WTCF indicates 40.5 Mlb/hr core flow

Which one of the following choices completes the following statements correctly?

For the given conditions, the most accurate total core flow indication is ________.

In order for the requirements of LCO 3.4.1, Recirculation Loops Operating to be met, and no shutdown action statements to be entered, the appropriate Single Loop Operating Limits must be applied within ________ hours.

A. 40.5 Mlb/hr
6 hours.
B. 40.5 Mlb/hr
12 hours.
C. 38 Mlb/hr
6 hours.
D. 38 Mlb/hr
12 hours.

Unit Two is operating at rated power when the following alarms are received:

DG-4 CTL POWER SUPPLY LOST
DG-4 LO START AIR PRESS
DG4/E4 ESS LOSS OF NORM POWER
DG-2 CTL POWER SUPPLY LOST

Subsequently, DG4 control power was transferred to its alternate DC source.

Which one of the following identifies the DC panel that was lost and the impact on the operability of DG4 in accordance with LCO 3.8.1, AC Sources Operating and LCO 3.8.7, Distribution Systems - Operating?

A. 125V DC Distribution Panel 1B;
DG4 is operable on its alternate source for up to 7 days.

B. 125V DC Distribution Panel 1B;
DG4 must be declared inoperable the entire time it is on its alternate source.

C. 125V DC Distribution Panel 2B;
DG4 is operable on its alternate source for up to 7 days.

D. 125V DC Distribution Panel 2B;
DG4 must be declared inoperable the entire time it is on its alternate source.

Unit One is operating at 100% power with the following conditions:

- All control rods and control rod scram accumulators are operable
- All control rod scram times are within TS Limits

Subsequently, one control rod scram accumulator has depressurized and cannot be repaired for two days.

Which one of the following identifies the required actions in accordance with Tech
Spec 3.1.5, Control Rod Scram Accumulators?

The affected control rod must be declared:

A. slow only.
B. inoperable only.
C. either slow or inoperable.
D. both slow and inoperable.

Unit One (1) and Unit (2) are operating at 100% power.

At 0945 on 9/10/06, I&C reports that they have measured and verifiec the following battery terminal voltages while on a float charge.

Battery 1A-1: 135.0 V
Battery 1A-2: 135.0 V
Battery 1B-1: 129.0 V
Battery 1B-2: 134.0 V

What action is applicable per Technical Specifications?

A. An Alert.Enter TS 3.0.6 and perform a Safety Function Determination per TS 5.5.1.1.
B. A loss of Safety Function has occurred and entry into TS 3.0.3 is required.
C. Enter Tech Spec 3.0.7, Condition C
D. Enter Tech Spec 3.0.8, Condition A

The following plant conditions exist:

Unit One (1)
- Mode 2
- GP-01 completed
- Reactor Mide Switch STARTUP
- Preparations for startup underway
- NO control rods withdrawn

Unit Two (2)
- Mode 5
- Shutdown for 69 hours
- Reactor Mode Switch REFUEL
- Fuel Shuffle #1 has been competed, no fuel moves in progress.
- Operations with a potential for draining the reactor vessel (OPDRVs) are in progress

Which ONE of the following describes the requirements for the SBGT System?

A. Required to be operable for both units.
B. Not required to be operabe for either unit.
C. Required to be operable for Unit One (1) ONLY.
D. Required to be operable for Unit Two (2) ONLY.

Unit Two (2) Core Reload is in progress with an estimated completion of 4 hours. Spent fuel Pool level is 37 feet 2 inches.

Due to an accident in the Drywell, the 2-E11-F009 vavle was driven closed with no over torque protecton. There is obvious damage to the valve stem (bowed) and the motor wiring caught fire. Initial attempts to reopen this valve have been unsuccessful.

Continued reload activities for the remaining 4-hour duration are?


A. NOT ALLOWED; No decay heat removal system is available.
B. NOT ALLOWED; RHR SDC can only be secured for 2 hours.
C. ALLOWED; RHR is allowed to be removed from service for 2 hours and Spent Fuel Pool level must be maintained >37 feet 1 inch.
D. ALLOWED; Cavity level must be maintained >24 feet 10 inches above RPV flange and alternate decay heat removal section of 2OP-17 has to be placed in service within 2 hours.

Which one of the following conditions is a safety limit violation on Unit Two (2)?

A. Reactor pressure reaches 1300 psig.
B. Core Thermal Power Exceeds 120%.
C. MCPR reaches 1.20 while operating single loop.
D. Reactor pressure lowers to 700 psig with reacotr power at 30%.

Unit One (1) is in refueling outage. Two control rod drive mechanisms were replaced as part of scheduled corrective maintenance. Both control rod drive mechanisms are disarmed and ECs providing administrative control until operability is restored.

Which one of the following is the correct sequence for restoring control rod operability?


A.
(1) Cancel the ECs on both control rods
(2) Conduct strongest rod out testing
(3) Withdraw each control rod one at a time and declare each rod operable
(4) Commence friction testing

B.
(1) Cancel the ECs on both control rods
(2) Test the first control rod when reached in the friction testing sequence
(3) Declare the first rod operable when testing is completed
(4) Test the second rod when reached in the friction testing sequence and declare operable when testing is complete

C.
(1) Cancel the EC for the first inoperable rod when it is reached in the friction testing sequence
(2) Complete testing the control rod and declare operable
(3) Continue friction testing of control rods until the second inoperable rod is reached in the testing sequence
(4) Repeat steps 1 and 2 for second inoperable rod

D.
(1) Cancel the EC for the first inoperable control rod
(2) Begin friction testing
(3) Declare the rod operable when testing is completed for that rod
(4) Cancel EC on second inoperable control rod
(5) Continue friction testing of control rods until the second inoperable rod is reached in the testing sequence
(6) Declare the second inoperable rod operable after testing is complete

The following plant conditions exist during a fuel shuffle on Unit Two (2):

- RHR Loop “A” is in SDC and has been in operation for the past 48 hours
- NO recirculation pumps are running
- Because SDC flow is disturbing the visibility, the Refuel Floor SRO requests SDC be secured to allow completion of several fuel movements
- The request is granted and SDC is secured at 08:00

The Technical Specification implication of this action is:

A. SDC must be started before 10:00 to avoid declaring the LCO statement not met.
B. SDC must be started before 09:00 to avoid declaring the LCO statement not met.
C. At 08:00, TS 3.9.7, Condition C must be entered and the required action taken.
D. At 08:00, TS 3.9.8, Condition C must be entered and the required action taken.

Unit One (1) is in a refueling outage with Alternate Decay Heat Removal utilizing Natural Circulation and the Fuel Pool Cooling System in service for 24 hours. The following data has been recorded on 1OP-17, Attachment 9:

Reactor Cavity Water Surface Temperature At Inlet To Weir
Weir #1 117°F
Weir# 2 114°F
Weir# 3 112°F
Weir# 4 115°F

SFP Water Surface Temperature At Inlet To Weir
Weir #1 104°F
Weir# 2 100°F

LCO 3.9.7, Action C.1 is:

A. Met, all temperature readings are satisfactory.

B. NOT met, SFP water surface temperature at inlet to weir ΔT is excessive.

C. NOT met, reactor cavity water surface temperature at inlet weir ΔT is excessive.

D. NOT met, the difference between reactor cavity and SFP average temperatures is excessive.

Which one of the following identifies whether ODCM Radioactive Gaseous Effluent Monitoring Instrumentation 7.3.2 action statement entry is required and also identifies whether 0OI-01.08, Attachment 10, Annunciator Removal From Service Form, is required to be completed for this activity?


A. action statement entry is required;
Attachment 10 must be completed.

B. action statement entry is required;
Attachment 10 is not required.

C. action statement entry is not required;
Attachment 10 must be completed.

D. action statement entry is not required;
Attachment 10 is not required.