Hatch October 2002 Exam 50-321/2002-301 FINAL SRO Writtem ... · Updated correct answer per lesson plan. Resequenced answers. A. Incorrect since the speed control signal was not lost

Final Submittal

E. I. HATCH NUCLEAR PLANT EXAM 2002-301 50-321 & 50-366

OCTOBER 16 - 18, 21 - 25, & OCTOBER 30, 2002,

Senior Operator Written Examination

InTQIN RSB2

rsb2 Final SRO Test.LXRTest 11/21/02 01:42 PM

.......... i ii i ii i ii i ii i ii i ii~ .......... ii iii ii ii

Name: Final SRO Test Form: 0

Version: 0

1. 201002A3.03 001

References: SI-LP-05401 Rev. SI-00, pg. 8 of 26. EO 001.010.a.12

A. Incorrect since the Rod Drift alarm will not actuate due to all control rods still at an even position.

B. Incorrect since turning the Rod Select Power to off will deselect any control rod and this does not guarantee that a Rod Drift alarm will occur.

C. Correct answer.

D. Incorrect since the procedure does not require the operator to wait until the settle light goes out.

Thursday, November 21, 2002 01:37:45 PM

Per 34GO-OPS-O01-2S, Plant Startup, which ONE of the following statements describes how a Rod Drift Alarm test is performed?

A. Turn the Rod Drift Test Switch to test. Verify the ROD DRIFT annunciator alarms. Reset the drift alarm with the test switch.

B. While moving a Group 1 control rod, turn the Rod Select Power Switch to off. Verify the ROD DRIFT annunciator alarms. Turn the Rod Select Power Switch to on and reset the drift alarm with the Test Switch.

C. Select a Group 1 control rod and take the Rod Movement Control Switch to out-notch. During rod travel, place the Rod Drift Alarm Test Switch to test. Verify the ROD DRIFT annunciator alarms. Reset the drift alarm with the Test Switch.

D. Turn the Rod Movement Control Switch to out-notch. When the settle bus light de-energizes turn the Rod Drift Test Switch to test. Verify the ROD DRIFT annunciator alarms. Reset the drift alarm with the Test Switch.

1

Final SRO Test2. 202001A1.07 001

References: SI-LP-00401-01 Rev. SI-01 Pg 9 of 62 FSAR Section 7.7.2, Recirculation Flow Control System

A. Incorrect since increasing recirc pump speed will decrease core void content.

B. Correct answer.

C. Incorrect since increasing Recirc Pump speed causes the steam generation rate to increase which adds negative reactivity.

D. Incorrect since increasing recirc pump speed will decrease core void content.


Which ONE of the following occurs when Recirculation Pump speed is increased?

A. There is a temporary increase in core void content which causes a decrease in moderator density with a resultant decrease in neutron moderation. The steam generation rate decreases which causes a positive reactivity effect.

B' There is a temporary decrease in core void content which causes an increase in moderator density with a resultant increase in neutron moderation. The steam generation rate increases which causes a negative reactivity effect.

C. There is a temporary decrease in core void content which causes an increase in moderator density with a resultant increase in neutron moderation. The steam generation rate decreases which causes a positive reactivity effect.

D. There is a temporary increase in core void content which causes a decrease in moderator density with a resultant decrease in neutron moderation. The steam generation rate increases which causes a negative reactivity effect.

2

Final SRO Test3. 202002K3.05 001

References: SI-LP-00401-01 Rev. SI-01 pg 42 of 62. On test SR 95-01 question 4 Updated correct answer per lesson plan. Resequenced answers.

A. Incorrect since the speed control signal was not lost.

B. Correct answer.

C. Incorrect because of clamped lower limit is 45% speed on the master controller.

D. Incorrect since this happens on a loss of signal from controller (not power loss).


Unit 2 is at 100% RTP with the following conditions:

Both Reactor Recirc Pumps are in Master-Manual Control A fuse for the power supply to the Master Controller fails The Master Controller de-energizes

Which ONE of the following describes the Reactor Recirculation System response?

A. Speed will remain constant because both scoop tubes lock up.

B. BOTH Recirc Pumps will decrease speed to approximately 44% speed.

C. Both Recirc Pumps will decrease to approximately 22% speed.

D. The controller will lock in its previous signal and maintain pump speed constant.

3

Final SRO Test4. 203000A1.01 001

References: LR-LP-20327 Rev. 07, pg 30 - 35 of 53 CP-3 ATWS LEVEL CONTROL EO 201.091.a.15

A. Correct answer.

B. Incorrect since all injection is prevented except Boron, CRD and RCIC until MARPVFP is reached and Table 13 systems are available.

C. Incorrect since Core Spray is not a Table 13 system.

D. Incorrect since all injection is prevented except Boron, CRD and RCIC until MARPVFP is reached and Table 13 systems are available.


During an ATWS on Unit 2, the Shift Supervisor directs injection to be terminated except for Boron, CRD and RCIC. He then directs Emergency Depressurization of the RPV because reactor water level CANNOT be maintained above -185 inches.

Which ONE of the following describes when the Shift Supervisor should recommence injection?

Aý When reactor pressure is less than MARPVFP using RHR or Condensate pumps.

B. As soon as RPV pressure decreases to within the shutoff head of the RHR and Core Spray pumps.

C. When reactor pressure is less than MARPVFP using both Core Spray and RHR pumps.

D. As soon as RPV pressure decreases to within the shutoff head of the RHR and Condensate pumps.

4

Final SRO Test 5. 204000K5.08 001

Unit 1 is at 100% RTP. The I & C Techs have just completed the quarterly functional surveillance for the RWCU Area High Temperature isolation instruments. The foreman is reviewing the paperwork and notes that isolation setpoints for all the areas were set to 155 0F. He immediately notifies the Shift Supervisor.

Which ONE of the following describes the determination the Shift Supervisor should make? (Provide TS Section 3.3.6.1 and Table 3.3.6.1-1)

A. This is not a problem because the setpoint per Tech Specs is < 160 0F.

B." This is a problem and all of the instruments are INOPERABLE. The RWCU system isolation capability must be restored within 1 hour.

C. This is a problem and all of the instruments are INOPERABLE. Each channel must be placed in the tripped condition within 12 hours of INOPERABILITY.

D. This is not a problem if the I & C Techs can recalibrate the instruments within tolerance provided the surveillance frequency hasn't expired.

References: Tech Spec section 3.3.6.1 Tech Spec Table 3.3.6.1-1 Tech Spec Bases B.3.3.6.1

A. Incorrect since the Tech Spec setpoint is < 150 F.

B. Correct answer since isolation capability is not maintained since all channels are inoperable.

C. Incorrect since isolation capability is not maintained per Bases definition.

D. Incorrect since the instruments should be declared INOPERABLE immediately.

Thursday, November 21, 2002 01:37:45 PM 5TT

Final SRO Test6. 205000G2.1.22 001


Unit 2 is shutting down for a maintenance outage due to the failure of the "B" Recirc pump which is out-of-service electrically. At 0100 on 4/12/02 reactor pressure went below 145 psig and reactor temperature went below 3000 F. The following conditions exist at 0400 on 4/12/02:

Reactor pressure 130 psig Reactor temperature 2850F Mode Switch position S/D

At 0415 on 4/12/02 the "A" Recirc pump tripped and cannot be restarted due to bus overcurrent.

Which ONE of the following is required to be taken per Tech Specs? (Provide copy of Tech Spec sections 3.3.6.1, 3.4.1, 3.4.7)

A. No action is required to be taken since Recirc Pumps are only required to be in operation in Modes 1 and 2.

B. Initiate action to place Shutdown Cooling in operation within 1 hour AND monitor reactor coolant temperature and pressure once per hour.

C. Initiate action to restore Shutdown Cooling to OPERABLE status immediately AND verify reactor coolant circulation by an alternate method within 1 hour from discovery of no reactor coolant circulation AND be in Mode 4 in 24 hours.

D!t No action is required for up 2 hours at which time a Recirc Pump must be running or Shutdown Cooling must be in operation.

6

Final SRO Test

References: Tech Spec section 3.3.6.1, Primary Containment Isol Instrument Tech Spec section 3.4.1, Recirc Loops Operating Tech Spec section 3.4.7, RHR Shutdown Cooling

A. Incorrect since Reactor Coolant circulation is required in Mode 3 by Shutdown Cooling or Recirc Pumps.

B. Incorrect since 3.4.7 requires action to place Shutdown Cooling or a Recirc Pump in operation IMMEDIATELY provided NOTE 1 of the LCO is not used or expired.

C. Incorrect since these are the actions to take if Shutdown Cooling is INOPERABLE. At this time Shutdown Cooling is OPERABLE since reactor pressure is below the Shutdown Cooling low pressure permissive.

D. Correct answer.

7. 206000K5.08 001


Unit 1 is operating at 100% RTP. The HPCI isolation valves are being stroked and timed per the Inservice Testing program when MO I E41-Fl 11 HPCI Vacuum Breaker Isolation Valve failed to close. The Shift Supervisor directed HPCI Vacuum Breaker Isolation Valve MO 1E41-F104 to be closed and deactivated.

Which ONE of the following describes the time limit for deactivating MO 1 E41 -F1 04 per Tech Specs and the effect on the HPCI system after the action(s) is/are taken? (Provide Tech Spec section 3.6.1.3)

A. Actions must be taken within 1 hour. HPCI should be declared INOP and a 14 day LCO entered per TS 3.5.1.C.

B./Actions must be taken within 4 hours. HPCI should be declared INOP and a 14 day LCO entered per TS 3.5.1.C.

C. Actions must be taken within 1 hour. HPCI system should still be considered OPERABLE because it can still perform its safety function.

DtActions must be taken within 4 hours. HPCI system should still be considered OPERABLE because it can still perform its safety function.

7

Final SRO Test

References: Tech Spec 3.6.1.3 for PCIVs SI-LP-00501 Rev. 01, LT-00501 Fig. 1 SI-LP-00501 Rev. 01, pg 8 of 46

A. Incorrect since the actions for Tech Spec 3.6.1.3 is 4 hours to isolate the line since there is more than 1 PCIV in the penetration flow path and only 1 valve is INOPERABLE. Also, HPCI can still perform its function and should still be considered OPERABLE.

B. IFn-orr-ct sinCe HPCI can still pcrfor-,- ;ts •, , rdi bivu'd st,'ll bh concidered OPERABL-E. 1,& ,,,w.•,--•,--,--#--'--..,,.-.-'

C. Incorrect since the actions for Tech Spec 3.6.1.3 is 4 hours to isolate the line since there is more than 1 PCIV in the penetration flow path and only I valve is INOPERABLE.

D. Cretaar.As- 1"/x-1 lo- Alc xozacr ce'tý-

8Thursday, November 21, 2002 01:37:46 PM

Final SRO Test 8. 206000K6.08 001

Unit 2 has just experienced a transient with the following conditions present:

Reactor water level -45" increasing Reactor pressure 123 psig Drywell pressure 18 psig HPCI turbine exhaust diaphragm pressure 8 psig

Which ONE of the following indicates the proper HPCI lineup for the present conditions including the cause?

A"' F002 closed F003 closed F001 open HPCI isolated on reactor low pressure with initiation signal still present.

B. F001 closed F002 closed F003 open HPCI isolated on reactor low pressure with initiation signal still present.

C. F002 open F003 open Turbine Stop Valve open HPCI should be injecting at full flow due to high drywell pressure.

D. F002 closed F001 open F006 closed HPCI isolated on high turbine exhaust pressure with initiation signal still present.

References: SI-LP-00501 Rev 01 pg 30 - 33 of 46

A. Correct answer.

B. Incorrect since F001 should be open from initiation signal and F003 should be closed due to auto isolation signal.

C. Incorrect since HPCI should not be injecting due to auto isolation signal from low reactor pressure.

D. Incorrect since HPCI should not have isolated on high exhaust pressure (setpoint is 10 psig).

Thursday, November 21, 2002 01:37:46 PM 9

Final SRO Test9. 209001G2.2.21 001

Reference: Tech Spec Bases SR 3.0.1

A. Incorrect since SR 3.5.1.10 does not need to be performed since no work was done on the Core Spray Logic and SR 3.5.1.2 does need to be performed since valves in the system were out of the normal Operable lineup.

B. Incorrect since SR 3.5.1.10 does not need to be performed since no work was done on the Core Spray Logic.

C. Incorrect since SR 3.5.1.10 and SR 3.5.1.13 do not need to be performed since no work was done on the Core Spray Logic. Also, SR 3.5.1.1 does need to be performed since system was drained.

D. Correct answer.


The "A" Core Spray system on Unit 1 was taken out-of-service to inspect the pump internals due to high vibration. Foreign material was found inside the pump and no additional repairs were necessary. The unit is in Day 5 of a 7 day LCO and the system has been returned to service, filled and vented.

Which ONE of the following indicates the surveillances that are required to be performed prior to declaring "A" Core Spray operable?

A. SR 3.5.1.1 piping filled from pump disch to injection valve, SR 3.5.1.7 flow rate test, SR 3.5.1.10 subsystem actuates on initiation signal.

B. SR 3.5.1.1 piping filled from pump disch to injection valve, SR 3.5.1.2 valve position verification, SR 3.5.1.7 flow rate test, SR 3.5.1.10 subsystem actuates on initiation signal.

C. SR 3.5.1.2 valve position verification, SR 3.5.1.7 flow rate test, SR 3.5.1.10 subsystem actuates on initiation signal, SR 3.5.1.13 ECCS Response time.

Dt SR 3.5.1.1 piping filled from pump disch to injection valve, SR 3.5.1.2 valve position verification, SR 3.5.1.7 flow rate test.

10

Final SRO Test10. 211000A3.04 001

References: SI-LP-01101 Rev. SI-00, pg 21 of 32. EO 011.002.1.03

A. Incorrect since discharge pressure should be higher than reactor pressure.

B. Incorrect since reactor water level increasing is not an indication that SBLC is injecting. Also, neutron level should be decreasing.

C. Incorrect since storage tank level should decrease.

D. Correct answer.


Which ONE of the following contains a correct list of indications that are used to verify the Standby Liquid Control System is operating properly once the system has been initiated? (Not necessarily all the indications)

A. Squib valve loss of continuity alarm annunciated, storage tank level decreasing, discharge pressure slightly lower than reactor pressure.

B. Red light indicating pump is running, reactor water level will increase, neutron level in the reactor will increase.

C. Storage tank level increasing, RWCU suction valve 2G31-F004 closes, squib valve amber light goes off.

D'. System discharge pressure will increase to greater than reactor pressure, neutron level in the reactor will decrease, RWCU suction valve 2G31-F004 closes.

11

Final SRO Test11. 211000A4.01 001

Unit 1 is in an ATWS condition and the Shift Supervisor is directing actions per RCA RPV Control (ATWS). He has determined that Boron injection is required. The CBO initiates Boron injection per 34SO-C41-003-2S, Standby Liquid Control System. The initial tank level indicates 86%. Two minutes later the CBO notices that the SBLC Tank level indication has failed downscale.

Which ONE of the following could be the cause of this and how will the CBO ensure that the Cold Shutdown Boron Weight has been injected?

A. Instrument air has been lost to the level detector and an Operator should be sent to verify level in the tank via the local level indication.

B. Instrument air has been lost to the level detector and an Operator should be sent to verify level in the tank via the top hatch.

C. Too much instrument air is being supplied to the bubbler detector and an Operator should adjust the flow to approximately I scfh.

D. Too much instrument air is being supplied to the bubbler detector and an Operator should adjust the flow to approximately 10 scfh.

Reference: RCA RPV Control (ATWS) flow chart SI-LP-01101-00 Rev. SI-00, Standby Liquid Control EO 011.001.a.03

A. Incorrect due to local level indication is also lost on a loss of instrument air. The level indication fails downscale on a loss of instrument air.

B. Correct answer.

C. Incorrect since too much flow to the bubbler causes level indication to be high.

D. Incorrect since too much flow to the bubbler causes level indication to be high.

Thursday November 21 2002 01:37:46 PM 12I I

Final SRO Test12. 212000K4.06 001

References: 31EO-EOP-103-2S Rev4Ed2 pg 2

A. Incorrect, the scram solenoids are energized, but the next action to be taken is to place the RPS test switches to trip lAW 31-EO-EOP-103-2S.

B. Incorrect, the scram solenoids are not de-energized, however, this action can be taken concurrently lAW 31-EO-EOP-103-2S.

C. Correct, the scram solenoids are energized, and the next action to be taken is to place the RPS test switches to trip lAW 31-EO-EOP-103-2S.

D. Incorrect, the scram solenoids are not de-energized, and the next action to be taken is to place the RPS test switches to trip lAW 31-EO-EOP-103-2S.


After a scram signal is received on Unit 2, control rods fail to insert and a manual scram is inserted. The white RPS scram group lights are ILLUMINATED.

Which ONE of the following describes the state of the scram solenoids and the NEXT

action that should be taken?

A. Energized; Rods should be individually scrammed.

B. De-energized; Rods should be manually driven in.

Ct. Energized; RPS test switches should be taken to trip.

D. De-energized; Links for RPS solenoids should be opened.

13

Final SRO Test13. 214000K4.01 001

References: SI-LP-05401-00 Rev. SI-00, pg 7 & 8 of 26, Reactor Manual Control EO 001.010.a.12, 001.026.a.02

A. Incorrect since this switch does not bypass RPIS.

B. Incorrect since this does not cause an insert and withdraw signal at the same time.

C. Incorrect since being at an even reed switch position ensures a rod drift alarm does not occur.

D. Correct answer since this alarm is actuated when rod is at an odd position and relay buses are not energized.


Which ONE of the following explains why a "ROD DRIFT" alarm is received after

moving a control rod using the "EMERGENCY IN" switch?

A. "EMERGENCY IN" bypasses the Rod Position Indication System.

B. The sequence timer is bypassed causing an insert and withdraw signal at the same time.

C. The rod is at an even reed switch and none of the selected relay busses are energized (insert, withdraw or settle).

D. The rod is at an odd reed switch and none of the selected relay busses are energized (insert, withdraw or settle).

14

Final SRO Test14. 215004K5.03 001

References: Tech Spec 3.3.1.2, Source Range Monitor (SRM) Instrumentation Tech Spec 3.3.1.2 Bases Technical Requirements Manual Table 3.3.2-1 34SV-SUV-019-2S, Surveillance Checks Rev. 32.3 pg 21 of 59 (NOTE) If this question is unacceptable then HATCH99.BNK #96 may be

used in its place.

A. Incorrect since the SRM is currently performing its function and there isn't a requirement that the detector move.

B. Correct answer. The SRM has to be bypassed prior to continuing with the startup since there is a rod block inserted when the SRM reaches the high limit of 7 X 10e4.

C. on

Incorrect since the Rod Block function of the SRM's are required until the IRM's are range 8 or above.

D. Incorrect since the deviation is figured as the Max divided by Min < 20.


A startup is in progress on Unit 1 with all the IRM's on Range 4. The Control Board Operator is in the process of withdrawing SRM's to keep the rod block cleared when it is determined that SRM "A" will not retract. All attempts to free the SRM have failed and Upper Management decides to continue with the startup and to leave the SRM inserted.

Which ONE of the following states IF and WHEN the SRM should be declared INOPERABLE? (Provide copy of TS 3.3.1.2 and TRM 3.3.2 along with appropriate tables.)

A. Declare "A" SRM INOPERABLE immediately, since the SRM cannot be moved.

Bf Declare "A" SRM INOPERABLE when it is bypassed to continue with the startup.

C. You don't have to consider the SRM Inoperable since the SRM's are not required with IRM's on range 3 or above.

D. Declare "A" SRM INOPERABLE when the "A" SRM reading deviates by >200 cps from the other 3 SRM's.

15

Final SRO Test15. 215005A2.03 001

Unit 2 is starting up with the Reactor Mode Switch in the START/HOT STBY position. e following is the present status of each APRM with regard to LPRM inputs and

indi ted power level.

APRM A B C D

Level D LPRM In ts 6 5 6 7 Level C LPRM Input 5 3 8 8 Level B LPRM Inputs 6 6 5 2 Level A LPRM Inputs 5 3 6 6

Indicated Power Level 12% 14% 12% 11%

Which ONE of the following describes the p t response to these conditions and the cause for the response?

A. Half Scram due to High power on APRM "B". -

B. Full Scram due to High power on APRM's "A", "B" and "C

C. APRM UPSC TRIP/INOP SYS B Alarm due to Inputs.

Dt. APRM UPSC TRIP/INOP SYS B Alarm due to Inputs.

APRM "B"

APRM "D"

havin ýo few LPRM

having too few M jvi g

References: SI-LP-01203-00 Rev. SI-00 pg 8-9 of 51 EO 012.003.d.01

A. Incorrect since Full Scram would occur if power START/HOT STBY.

B. Incorrect since power level is too low for scram START/HOT STBY)

reached 13% with Mode Switch in

condition. (13% with Mode Switch in

C. Incorrect since APRM B has the minimum LPRM Inputs required (17).

D. Correct answer.

The alarm identified in the answer does not exist. The answer should have identified that ROD OUT BLOCK is in alarm. The reference material provided to develop the examination was not updated in this area. The answer key was changed to delete this


Final SRO Test16. 216000K2.01 001

Reference: LT-LP-10008 Rev. EO 055.001.a.07

A. Incorrect since a loss of 2B

B. Incorrect since a loss of 2B

C. Incorrect since a loss of 2B

D. Correct answer.

17. 217000K1.01 001

SI-00 pg 7 of 28.

RPS does not affect panels P921 and P923.

RPS does not affect panel P921 or P923.

RPS affects both panels P922 and P924.


Which ONE of the following describes the results of a loss of the 2B RPS bus to the

Analog Transmitter Trip System (ATTS)?

A. A complete loss of power to panels P921 and P923.

B. A loss of one of the two power supplies to panels P921 and P923.

C. A loss of one of the two power supplies to panels P922 and P924.

Dr A complete loss of power to panels P922 and P924.

Unit 2 is operating at 80% RTP. The RCIC system is in standby with a suction from the CST. The quarterly HPCI flow rate test is in progress and is taking longer than expected. Torus level has reached 151" and preparations are being made to pump the torus down to normal level within the 2 hour Tech Spec time limit.

Which ONE of the following describes the effect high Torus level had on RCIC?

A. No effect since the RCIC suction valves do not transfer until 152".

Br" The Torus suction valves (F029 & F031) received an open signal and once both valves were full open then the CST suction valve (F010) received a closed signal.

C. The Torus suction valves (F029 & F031) received an open signal at the same time the CST suction valve (FO10) received a closed signal.

D. The CST suction valve (F010) received a closed signal and when it was full closed then the Torus suction valves (F029 & F031) received an open signal.

17

Final SRO Test

References: SI-LP-03901 Rev. SI-00 pg 10 of 37

A. Incorrect since the U2 RCIC suction swap takes place at a suppression pool level of 150.5" Unit 1 takes place at 152".

B. Correct answer.

C. Incorrect since the CST and Torus suction valves do not get a signal to change position at the same time. The CST gets a closed signal "After" the Torus suction valves are full open.

D. Incorrect since the Torus suction valves come open before the CST suction valve goes closed.

18. 218000G2.2.22 001


On 3/2/02 at 0800 Unit 2 is in Mode 1 when RCIC is declared inoperable and day 1 of a 14 day LCO is entered. On 3/7/02 at 1600 the Instrument Techs start a surveillance on a Drywell Pressure instrument associated with ADS Trip system A by valving the instrument out. At 2200 they report to the Shift Supervisor that the instrument cannot be calibrated and that no other instruments are affected.

Per Tech Specs, which ONE of the following is the latest time the channel shall be placed in the tripped condition? (Provide Tech Spec section 3.3.5.1 and 3.5.3)

Aý 2200 on 3/11/02.

B. 1600 on 3/11/02.

C. 2200 on 3/15/02.

D. 1600 on 3/15/02.

18

Final SRO Test

Reference: Tech Spec section3.3.5.1 and 3.5.3.

A. Correct answer.

B. Incorrect answer. Can delay the actions for Condition F for 6 hours due to note 2 for surveillance requirements even though the instrument is inoperable.

C. Incorrect answer. Completion time in answer is 8 days from required action time. This is wrong since RCIC is inoperable concurrent with this instrument.

D. Completion time in answer is 8 days from instrument being inoperable. This is wrong since you can use the surveillance note of 6 hours and RCIC is also inoperable.

19. 219000A2.08 001

Unit 1 is operating at 100% RTP. The Plant Operator (PO) has placed the 1B RHR pump in Torus Cooling Mode. "Rx Bldg Floor Drains Sump B Leak High" alarm is received followed shortly by "RHR Pump B Trip." A System Operator (SO) in the area notifies the PO that there is excessive leakage in the pump seal area for the 1 B RHR pump.

Which ONE of the following describes the impact on the RHR System and what are the appropriate actions to take per the Alarm Response procedures? (Provide copy of Tech Spec Section 3.5.1)

A. Declare Division 2 of RHR INOPERABLE, send an Operator to the SWGR to look for flags and place the "B" RHR Pump control switch in STOP.

B. Declare the "B" RHR Pump INOPERABLE, verify the suction valve lineup, isolate Division 2 of RHR and reset the sump timers.

C. Declare Division 2 of RHR INOPERABLE, verify the suction valve lineup, isolate the "B" RHR Pump and reset the sump timers.

D. Declare the "B" RHR Pump INOPERABLE, send an Operator to the SWGR to look for flags and isolate Division 2 of RHR.


Final SRO Test

References: 34AR-601-206-IS Rev. 3.1 34AR-602-401-1S Rev. 0.1

A. Incorrect since you need to isolate the pump.

B. Incorrect since you need to call the entire loop INOPERABLE, not just the pump.

C. Correct answer.

D. Incorrect since you do not have to isolate the entire Loop of RHR, just the pump. Also, you should declare the entire loop Inoperable.

20. 223001A2.12 001

Unit 1 has scrammed on vessel low level due to loss of all High Pressure feed. The crew emergency depressurized the RPV before level reached -185". 1A Core Spray pump is operating to restore reactor water level with the following plant conditions:

Reactor Water Level Reactor Pressure Suppression Pool Temperature Suppression Pool Water Level Suppression Chamber Pressure 1A Core Spray Flow

-120 inches and Increasing 45 psig 190OF 145 inches 3 psig 4600 gpm

Which ONE of the following actions should the crew take concerning the continued use of 1A Core Spray pump? (Provide Graph 9, 11A and 11B)

A. REDUCE flow to get back within Vortex limits.

B. CONTINUE flow at the present rate since it is within NPSH and Vortex limits.

C!' REDUCE flow to get back within NPSH limits.

D. CONTINUE flow at the present rate ignoring NPSH and Vortex limits.


Final SRO Test

Reference: Core Spray Pump NPSH Limit Graph 11 B Core Spray Vortex Limit Graph 9 Students will be supplied with both graphs.

A. Incorrect since 1A Core Spray pump is within Vortex limits.

B. Incorrect since EOP's direct you bace to RC RPV CONTROL after vessel level is increasing above top of active fuel and now Vortex and NPSH limits are a concern.

C. Correct answer.

D. Incorrect since CP-1 has been exited to RC RPV CONTROL point B.

21. 223002K1.07 001

U-2 RCIC System is running with the following conditions present 10 minutes after the event:

Reactor Water Level -38" Drywell Pressure +1.5 psig Suppression Chamber Ambient Temp. 170OF RCIC Steam Line pressure 900 psig RCIC Emergency Area Cooler Temp 1 00OF and rising slowly

An operator has been sent to the RCIC room and reports that there is a small steam leak on the line upstream of the Trip and Throttle valve. The Shift Supervisor orders the Reactor Operator to manually isolate RCIC.

Which ONE of the following describes the effect on RCIC when the manual isolation

pushbutton is depressed?

A. Isolation valves F007 and F008 close and the RCIC turbine trips.

B. Inboard Isolation valve F007 closes and the RCIC turbine trips.

C'. Outboard Isolation valve F008 closes and the RCIC turbine trips.

D. No effect on RCIC since the system should already be isolated.

Thursday, November 21. 2002 01:37:48 PM 21• r r

"1"

Final SRO Test

Reference: SI-LP-03901-00 Rev. SI-00 pg 17 of 37. EO 039.012.a.04

A. Incorrect since only the F008 valve is affected by the "Manual Isolation" pushbutton while there is an initiation signal present.

B. Incorrect since the F007 valve is not affected by the "Manual Isolation" pushbutton while there is an initiation signal present.

C. Correct answer.

D. Incorrect since the only isolation signal could be Suppression Chamber Ambient Temp but it has a 30 minute time delay when temp is > 165OF and this is 10 minutes into the event.

22. 226001 K2.02 001

Unit 2 is operating at 100% power with the following equipment out-of-service:

230KV breakers 470 & 480 are open to perform testing on Startup Transformer 2C. EDG 1 B due to oil leak just found (repairs in progress).

While waiting for the repairs to be completed for the 1 B EDG Unit 2 experiences a Reactor Scram on High Drywell pressure due to a failure of Recirc Pump 2B seals. All automatic actions occur as designed with the current plant lineup. The 2B RHR Pump is due to

A. running, power supplied from 2C EDG.

B. not running, power not supplied from I B EDG.

C• running, power supplied from SAT 2D.

D. not running, power not supplied from SAT 2C.

Th, rdav Nnvemhbr 21 2002 01:37:4 PM 223" I I

Final SRO Test

Reference: LT-LP-02701-03 Rev. 03 Electrical Distribution - Switchyard LT-LP-02702-03 Rev. 03 4160 VAC Electrical Distribution Tech Specs and Bases Section 3.8.1

EO 200.017.a.03

A. Incorrect since SAT 2D is still energized.

B. Incorrect due to power still available from SAT 2D.

C. Correct answer.

D. Incorrect since SAT 2D is the power supply.

23. 234000K6.02 001


Unit 2 is in a Refueling Outage with a core reload in progress. All Refueling related surveillances are current. The Unit CBO receives the "600V BUS 2C BREAKER TRIP" alarm and upon investigation determines a loss of 2R24-SO1 1.

Based on this information which ONE of the following describes how Refueling Operations are affected?

A. Core Alterations can continue since a "refuel bridge stopped" alarm has not occured.

Bý' Core Alterations must be suspended immediately OR declare associated AC supported required features inoperable.

C. Core Alterations can continue until the next scheduled AC breaker alignment and voltage checks are due since power is still available to the refueling equipment.

D. Core Alterations can continue for 1 hr provided AC breaker alignment and voltage checks are completed within that hour.

23

Final SRO Test

References: Tech Spec 3.8.8 Distribution - Systems Shutdown Tech Spec Bases 3.8.8 Distribution Systems - Shutdown Bases LT-LP-02705-02 Rev.02 pg 30 of 42

A. Incorrect since the AC distribution system must be determined INOPERABLE due to

the low voltage so Core Alterations must be stopped immediately.

B. Correct answer.

C. Incorrect since equipment must be declared INOPERABLE upon discovery on not being able to meet minimum requirements.

D. Incorrect since 1 hour is not provided to continue Core Alterations in the Tech Specs.

24. 239001K4.04 001


Which ONE of the following describes the purpose of the Main Steam Line Flow Restrictors?

A. To prevent the uncontrolled release of radioactive material to the environs following a steam line rupture outside containment to the extent that the CFR 100 limits are not exceeded at the site boundary.

B. To limit the loading on the steam lines following a steam line rupture outside containment such that the failure of one steam line would not result in a MSIV isolation from high main steam line flow.

C. To limit the pressure reduction following a steam line rupture outside containment such that the safety limit of 785 psig is not reached with reactor power >25% prior to the MSIV's closing.

D. To work in conjunction with the MSIV's to limit flow on a steam line rupture which assures the steam dryer and other internal structures in the vessel remain in place.

24

Final SRO Test

Reference: SI-LP-01401-00 Rev. 00 Pg 9 of 45

A. Incorrect since the isolation of the MSIV's perform this function.

B. Incorrect since the flow restrictors limit the loss of coolant from the vessel on a steam line break.

C. Incorrect since the flow restrictors are concerned with the loss of coolant and not the pressure reduction.

D. Correct answer.


Final SRO Test25. 239002A1.01 001

Unit 1 reactor has just scrammed due to an inadvertent Group I isolation. Supervisor has entered RC RPV CONTROL (NON-ATWS) and ordered a pressure band of 800 - 1080 psig using the relief valves.

The Shift reactor

Which ONE of the following describes the indications you would expect to see when you opened a relief valve?

A. The RED indicator light is lit, annunciator SAFETY BLOWDOWN PRESSURE HIGH is clear, and reactor pressure is decreasing.

B. The GREEN indicator light will extinguish, the YELLOW indicator light is not lit, and tailpipe temperature indication is increasing.

C. The YELLOW indicator light is lit, annunciator SAFETY BLOWDOWN PRESSURE HIGH is alarming, and tailpipe temperature indication is increasing.

D. Reactor pressure is decreasing, the RED indicator light is indicator light is out.

References: 34SO-B21-001-2S, Automatic Depressurization (LLS) System, Rev.13.3 pg 16 of 31

A. Incorrect since the be in alarm.

out, and the GREEN

(ADS) and Low-Low Set

annunciator for SAFETY BLOWDOWN PRESSURE HIGH should

B. Incorrect since the YELLOW light should illuminate.

C. Correct answer.

D. Incorrect since the RED indicating light should be illuminated.


Final SRO Test26. 241000K6.05 001

Reference: SI-LP-02501-00 Rev. SI-00 Pg 7 of 13. EO 200.087.a.01

A. Incorrect since RFP's and Main Turbine trip at the same time at 22.3" Hg Vac.

B. Incorrect since MSIV's close at 10" Hg Vac and Bypass valves close at 7" Hg Vac.

C. Correct answser since RFP's and Main Turbine trip at the same time at 22.3" Hg Vac, MSIV's close at 10" Hg Vac and Bypass valves close at 7" Hg Vac.

D. Incorrect since Main Turbine trips prior to the MSIV's closing.


Unit 2 is holding load at 75% Reactor Power when the operator receives the "Turbine Vacuum Low" alarm.

Which ONE of the following describes the expected sequence of actions as condenser vacuum continues to decrease from 24.7" Hg Vac (alarm setoint) to 0" Hg Vac?

A. 1st - Main Turbine trips. 2nd - RFP turbine trips and Main Turbine Bypass Valves close at the same time. 3rd - MSIV's close.

B. 1st - Main Turbine and RFP turbine trip at the same time. 2nd - Main Turbine Bypass Valves close. 3rd - MSIV's close.

C0' 1st - Main Turbine and RFP turbine trip at the same time. 2nd - MSIV's close. 3rd - Main Turbine Bypass Valves close.

D. 1st - RFP turbine trips and Main Turbine Bypass Valves close at the same time.. 2nd - MSIV's close. 3rd - Main Turbine trips.

27

Final SRO Test27. 245000K1.09 001

References: 34AB-R22-001-2S Rev.2.3 Pg 3

A. Incorrect, lAW 34-AB-R22-001-2S the generator output breakers will not open automatically. However, they will open from the control room but the exciter field breaker must be opened locally.

B. Incorrect, lAW 34-AB-R22-001-2S the generator output breakers will not open automatically and the exciter field breaker must be opened locally.

C. Incorrect, lAW 34-AB-R22-001-2S the generator output breakers will fail to open automatically, however, the exciter field breaker must be opened locally.

D. Correct, lAW 34-AB-R22-001-2S the generator output breakers will fail to open automatically and the exciter field breaker must be opened locally.


Which ONE of the following describes the effect that a loss of the "2A" 125/250V DC Switchgear will have on the Unit 2 Main Turbine Generator?

A. Remote trip capability for the generator output breakers AND the exciter field breaker will be lost.

B. Remote trip capability for the generator output breaker will be lost, but the exciter field breaker can still be controlled remotely.

C. The generator output breakers will fail to open automatically, but the exciter field breaker can still be tripped from the control room.

D• The generator output breakers will fail to open automatically AND the excitpr field breaker must be opened locally.

28

Final SRO Test28. 259002K3.02 001

Reference: SI-LP-00201-00 Rev. SI-00 Pg. 27 of 47. Question #LT-LP-002027-0002 EO 002.004.a.08

Minor rewording of question and order of answers.

A. Incorrect since loss of signal results in default level signal being failed controller to the associated RFPT.

B. Incorrect since RFPT with failure does not go to low speed stop RFPT does not go towards the high speed stop.

C. Incorrect since level level.

D. Correct answer.

used (37") for the

and the unaffected

is maintained by the unaffected RFPT at the normal operating


Unit 2 is holding load at 80% Reactor Power. The Feedwater Level Control system is in 3 element control.

Which ONE of the following describes what RPV water level should do if a loss of control signal from RFPT "A" M/A controller occurred?

A. INCREASE due to loss of a feed flow signal resulting in a steam flow / feed flow mismatch.

B. REMAIN THE SAME due to RFPT "A" decreasing speed to it's "low speed stop" and the "B" RFPT increasing speed toward it's "high speed stop".

C. DECREASE initially and STABILIZE approximately 6" lower due to loss of feed flow input.

Dv• REMAIN THE SAME due to RFPT "A" defaulting to the speed setter and RFPT "B" controlling level in auto.

29

Final SRO Test29. 261000K4.01 001

Reference: SI-LP-01302-00 Rev. SI-0 Pg. 16 of 21 Question # LR-LP-200023-0001 EO 013.031.a.05, 013.038.a.08

A. Incorrect since none of these instruments are reading >18 mR/hr so nothing will automatically cause fans to trip and isolate and SBGT to start for both units.

B. Incorrect since both units are affected by any of these instruments reading >18 mR/hr.

C. Incorrect since the Refuel Floor fans and dampers would be affected by these signals if the threshold wee met.

D. Correct answer. None of the instrument readings meet the threshold for actuating the Reactor Building and Refueling Floor vents and starting SBGT (K609 A-D reading >18 mR/hr for Unit 1).

Modified the initial conditions so that the student must realize that the threshold has not been met to cause initiations since the readings are for Unit 1. If the readings were for Unit 2 then the initiations would occur.


The following Unit 1 reactor zone exhaust ventilation radiation levels exist:

1D11-K609A 12 mR/hr 1D11-K609B 14 mR/hr 1D11-K609C 10 mR/hr 1 D11-K609D 11 mR/hr

Which ONE of the following reflects the plant response to the above conditions?

A. Unit 1 and 2 Refueling Floor and Reactor Building supply and exhaust fans trip and isolate and Unit 1 and 2 SBGT starts and aligns to the Reactor Building and Refueling Floor.

B. Unit 1 Refueling Floor and Reactor Building supply and exhaust fans trip and isolate and Unit I SBGT starts and aligns to the Reactor Building and Refueling Floor.

C. Unit 1 and 2 Reactor Building supply and exhaust fans trip and isolate and Unit 1 and 2 SBGT starts and aligns to the Reactor Building only.

D.' Unit I and 2 Reactor Building and Refueling Floor ventilation systems remain in operation and neither SBGT fan starts.

Final SRO Test30. 262001A4.02 001

References: INPO 2001 Exam Bank (question 262001.A4.0 155) 34SO-N40-001-2S, Main Generator Operation Rev. 10.5, pg 13 - 16 of 53.

A. Correct answer.

B. Incorrect answer since you have to use the Auto Voltage Adjust control switch to match voltages.

C. Incorrect since the VAR flow will be from the system to the generator.

D. Incorrect since the VAR flow will be from the generator the Auto Voltage Adjust control switch to match voltages.

Thursday, November 21,2002 01:37:49 PM

to the system and you use

31

Which ONE of the following states how to adjust main generator output voltage and the consequences of improperly setting the main generator output voltage with respect to system voltage during manual synchronization?

A• Adjust the generator output voltage using the Auto Voltage Adjust control switch and if generator output voltage is greater than system voltage then Reactive load will be positive.

B. Adjust the generator output voltage using the Manual Voltage Adjust control switch and if generator output voltage is less than system voltage then Reactive load will be negative.

C. Adjust the generator output voltage using the Auto Voltage Adjust control switch and if generator output voltage is less than system voltage then Reactive load will be positive.

D. Adjust the generator output voltage using the Manual Voltage Adjust control switch and if generator output voltage is greater than system voltage then Reactive load will be negative.

Final SRO Test31. 263000K3.01 001

References: 34AB-R22-001-2S, Loss of DC Buses Rev. 2.3 pg 17,18, and 59 of 66.

A. Correct answer.

B. Incorrect since the 2C Diesel Generator is unaffected by this loss.

C. Incorrect since the 2C Diesel Generator is unaffected by this loss.

D. Incorrect since the 1 B Diesel Generator is unaffected by this loss.


Unit 2 is operating at 50% RTP when the following annunciators alarm:

ARI OUT OF SERVICE LOSS OF OFF SITE POWER 4160V BUS 2E or 600v BUS 2C DC OFF 125/250V BATT VOLTS LOW OR FUSE TROUBLE 125/250V BATT CHGR MALFUNCTION

The Control Board Operator verifies that 125VDC Cabinet 2D has been lost.

Which ONE of the following describes the impact on the Diesel Generators that supply Unit 2 Buses?

A.• 2A Diesel Generator is INOPERABLE due to loss of auto start capability.

B. 2C Diesel Generator is INOPERABLE due to loss of auto start capability.

C. 2A and 2C Diesel Generators are INOPERABLE due to loss of auto start capability.

D. 1 B Diesel Generator is INOPERABLE to Unit 2 due to loss of auto start capability to Unit 2.

32

Final SRO Test32. 264000K5.06 001

Reference: LT-LP-02801 Rev 3 pg 49 and 50 of 87. Copy of Electrical Lineup

A. Incorrect since "B" Core Spray pump is powered from DIG "C" only.

B. Incorrect since "C" PSW pump only starts if the "A" PSW pump fails to start.

C. Incorrect since "B" LPCI pump starts from DIG "C" only.

D. Correct answer. PSW pump "D" starts since the "C" DIG has failed to start which powers the "B" PSW pump.


Unit I is operating at 75% RTP when the following actions occur:

Reactor Scram DIG "A" and "B" start and attain proper speed and voltage D/G "C" fails to start Reactor Water Level -15" increasing Drywell Pressure 4.5 psig Drywell Temperature 200OF Startup Transformers IC and ID are De-Energized

Which ONE of the following lists the major loads on the "1 B" D/G and the sequence that they started?

A. Core Spray "B", LPCI "C", LPCI "D".

B. LPCI "C", LPCI "D", PSW "C".

C. LPCI "B", LPCI "C", LPCI "D".

Dt. LPCI "C", LPCI "D", PSW "D".

33

Final SRO Test33. 268000A1.01 001

Reference: LT-LP-02901-02 Rev. 02 pg 36 of 46. 34AR-601-401-2S Rev. 0.2

A. Incorrect since the high discharge trip setpoint will also auto close the high and low canal discharge lines.

B. Incorrect since the high discharge trip setpoint will auto close the high and low canal discharge lines but do not affect the dilution flow line.

C. Incorrect since the high discharge trip setpoint will also auto close the low canal

discharge line.

D. Correct answer.


The Unit 2 Radwaste Operator is in the process of discharging the Chemical Waste Sample Tank 2B to the discharge canal at 65 gpm, when the discharge radiation monitor exceeds the HIGH trip setpoint.

Which ONE of the following describes ALL of the expected actions for this condition?

A. The Radwaste Effluent High Radiation alarm will annunciate.

B. The Radwaste Effluent High Radiation alarm will annunciate and the dilution flow line will isolate.

C. The Radwaste Effluent High Radiation alarm will annunciate and the High Flow canal discharge line will isolate.

D'. The Radwaste Effluent High Radiation alarm will annunciate and the High and Low Flow canal discharge lines will isolate.

34

Final SRO Test34. 272000K2.05 001

References: LT-LP-10007 Rev. 04 pg 17-27 of 73.

A. Incorrect since this is powered by 24 VDC Cabinet "B", R25-S016.

B. Incorrect since this is powered by 24 VDC Cabinet "A" and "B".

C. Incorrect since this is powered by their NUMAC monitors which receive power from RPS "B".

D. Correct answer.


Unit 1 is operating at 75% RTP when the "A" RPS MG Set inadvertently trips on undervoltage.

Which ONE of the following radiation monitors is INOPERABLE until RPS "A" is restored?

A. RBCCW discharge monitor.

B. Main Stack Radiation monitor.

C. Offgas pretreat rad monitor.

D!t "A" Reactor Building Vent Exhaust monitor.

35

Final SRO Test35. 286000A1.01 001

References: LT-LP-03601Rev.3 pg 19 and 21

A. Incorrect since motor driven pump does not have electrical power and starting pressures are incorrect.

B. Incorrect since pumps start at different pressures..

C. Incorrect since first engine driven fire pump starts at 100 psig.

D. Correct answer.


Unit I is in a Refueling outage with the 4160 VAC I E bus tagged out and de-energized

for maintenance. A fire is detected resulting in main fire header pressure decreasing.

Which ONE of the following is the expected fire protection system response?

A. The motor driven fire pump starts at 110 psig; the first engine driven pump starts at 100 psig and the second engine driven pump starts at 90 psig.

B. The motor driven fire pump does not start; the first engine driven pump starts at 110 psig and the second engine driven pump starts at 100 psig.

C. The motor driven fire pump does not start; the first engine driven fire pump starts at 110 psig and the second engine driven pump starts at 90 psig.

D"' The motor driven fire pump does not start; the first engine driven fire pump starts at 100 psig and the second engine driven pump starts at 90 psig.

36

Final SRO Test36. 288000A2.05 001

References: DI-OPS-36-0989N Rev. 13 pg 1 of 4.

A. Incorrect since D/G does not become INOPERABLE immediately. Also, there isn't direction to install temporary heating.

B. Incorrect since the DIG cooling water can't freeze since it is filled with anti freeze. Also, no direction to install temporary heating and temperature indication.

C. Incorrect since the oil is heated with an immersion heater to maintain proper

temperature. No direction to declare DIG INOPERABLE if temp goes below 600 F.

D. Correct answer.


The weather forcast for the oncoming shift is high winds with the temperature dropping to the low teens. The System Operator (SO) is performing outside rounds and notes that upon entering the 2A DIG room that the temperature is abnormally cold.

Which ONE of the following describes the possible impact on the plant and the correct compensatory actions for this situation?

A. 2A DIG could become INOPERABLE due to cold conditions. Start the DIG to warm up the room. Install temporary heaters to maintain temperature above 600 F.

B. The D/G cooling water system could freeze which would INOP the 2A D/G. Install temporary heating units to supplement the room heaters and install temporary temperature indication in the room.

C. 2A D/G oil could cool down and affect the auto start capability of the DIG. Verify room heaters are operating properly. Declare the DIG INOPERABLE if room temperature is <60OF for 12 hours.

DV 2A D/G could become INOPERABLE due to cold conditions. Verify DIG room and switchgear room louvers are completely closed. Also, verify room heaters are energized and maintaining temperature.

37

Final SRO Test37. 290001A3.01 001

Reference: LT-LP-10007 Rev. 04, pg. 28 EO 200.030.a.10 Revised distractors and stem to make it more plausible.

A. Incorrect since only the inboard isolation valves close.

B. Correct answer.

C. Incorrect since both trains of SBGT and both units Rx Bldg Vents isolate with only the inboard isolation valves going closed.

D. Incorrect since both trains of SBGT and both units Rx Bldg Vents isolate with only the inboard isolation valves going closed.


Units 1 and 2 are operating at 100% power when a Hi-Hi alarm is received on reactor building exhaust ventilation radiation monitor channels 1 D11-K609A and B. Channels 1D11-K609C and D are reading normal.

Which ONE of the following describes the response of the Secondary Containment Systems?

A. Unit I and 2 SBGT systems auto start. Unit 1 and 2 Reactor Building ventilation trips and all isolation valves close.

B.' Unit I and 2 SBGT systems auto start. Unit 1 and 2 Reactor Building ventilation trips and only the inboard isolation valves close.

C. Unit 1 SBGT system auto starts. Unit 1 Reactor Building ventilation trips and all the Unit 1 isolation valves close.

D. Unit 2 SBGT system auto starts. Unit 2 Reactor Building ventilation trips and all the Unit 2 isolation valves close.

38

Final SRO Test38. 290002G2.1.28 001

Which ONE of the following statements CORRECTLY describes a component within the RPV?

A'. The baffle plate provides a mounting surface for the jet pump diffusers and separates the downcomer area from the below core plate area.

B. Flow orifices are mounted in the control rod housing, directly aligned with the fuel support piece.

C. The steam seperator dries the steam/fluid mixture to 99.9% quality.

D. The Standby Liquid Control/Core dp Pipe is a permanently mounted pipe within a pipe with the outer pipe used for SBLC injection.

References: SI-LP-04402 Rev. SI-00 pg 4-10 of 27 LO LT-04402.001

A. Correct answer.

B. Incorrect since the flow orifices are located in the fuel support pieces.

C. Incorrect since the steam separators increase the steam quality from approximately 13% to 90%.

D. Incorrect since the SBLC system uses the inner tube to inject Boron.


Final SRO Test39. 290003K5.02 001

References: Tech Spec section 3.7.4, MCREC System SR 3.7.4.4 Modified answers to reflect only one correct answer

A. Correct answer since dP is > 0.1" wg, subsystem flow rate is < 2750 cfm and outside air flow is < 400 cfm.

B. Incorrect since the dP is adequate.

C. Incorrect since the subsystem flow rate is adequate.

D. Incorrect since the outside air flowrate is adequate.


Both Units are operating at 100% RTP. It is reported that the Control Room HVAC can only maintain a positive pressure of 1/5 inch WG relative to the Turbine Building during the pressurization mode. Outside air flow rate is 399 cfm and subsystem flowrate is 2600 cfm.

Based on these plant conditions, which ONE of the following describes the Control Room HVAC system? (Provide copy of TS 3.7.4 with SR's)

A" would still be OPERABLE because Control Room to Turbine Building dp and ventilation flow rates are adequate.

B. would be INOPERABLE because Control Room to outside Turbine Building dp is too low.

C. would be INOPERABLE because the Control Room ventilation subsystem flow rate

is inadequate.

D. would be INOPERABLE because outside air flow rate is inadequate.

40

Final SRO Test40. 29500 1AK2.07 001

References: 34AB-B31-001-2S Rev 7.2 pg 3 of 7. Hatch 99 Exam question #56 Slight modifications to stem and added LOOP to all flows in answers. Resequenced answers.

A. Correct answer.

B. Incorrect since you cannot subtract the 2A Jet Pump Loop flow from the Total Flow indication and get an accurate reading.

C. Incorrect reading.

since you have to add the two loop flows together to get an accurate

D. Incorrect since the summing circuitry does not account for an idle recirc pump.


Unit 2 startup is in progress with no equipment out of service. Reactor power is 40% and the speed of both recirc pumps was just raised to 30%. A trip of Recirc Pump "2A" occurs and the operator performs the actions of 34AB-B31-001-2S, Reactor Recirculation Pump(s) Trip, or Recirc Loops Flow Mismatch to stabilize the plant.

Which ONE of the following describes how an accurate reading of total core flow is determined under these conditions?

A.• Total core flow must be manually calculated by adding "2A" and "2B" Jet Pump Loop flows to obtain an accurate reading.

B. The Total Core Flow indication must be reduced by the "2KA Jet Pump Loop flow to obtain an accurate reading.

C. Total core flow must be manually calculated by subtracting "2A" Jet Pump Loop flow from the "2B" Jet Pump Loop flow to obtain an accurate reading.

D. The summing circuitry for the Total Core Flow indication automatically accounts for the idle "2A" recirc loop and provides an accurate reading.

41

Final SRO Test41. 295002AA2.02 001

Reference: SI-LP-02501 Rev. SI-00 pg 6 of 13

A. Incorrect since condenser vacuum will decrease with Rx Power <28%. This will cause reactor power to decrease.

B. Correct answer since seal steam is required up to 28% RTP or condenser vacuum will decrease. If condenser vacuum decreases then reactor power will decrease.

C. Incorrect since this amount of steam being drawn off of reactor is negligible..

D. Incorrect since steam seal bypass valve does not automatically open.


Unit 2 is holding load at 25% RTP. The main turbine is on line when the steam seal regulator fails closed.

Which ONE of the following describes the impact this will have on reactor power with no operator action?

A. Reactor power will remain constant since seal steam is not required at this power

level.

B" Reactor power will decrease since condenser vacuum will decrease.

C. Reactor power will decrease since less steam is required from the reactor.

D. Reactor power will remain constant since the steam seal bypass valve automatically opens to maintain seal steam pressure constant.

42

Final SRO Test 42. 295003AK2.01 001

A small LOCA with a LOSS OF OFFSITE POWER occurred on Unit 2 at 0100 on 5/21/02. 600V Bus 2D is being powered from Diesel Generator 2C (Load Shed equipment still de-eneraized) and HPCI is running to maintain reactor vessel water level between -35 and +51.5 inches.

Which ONE of the following predicts how HPCI will respond over the next 3 hours with no operator action?

A. HPCI will continue to operate automatically by maintaining reactor vessel level in a band of -35 inches to +51.5 inches.

B. HPCI will continue to operate automatically by maintaining reactor vessel level between +51.5 inches and the high level reset setpoint.

C. HPCI will operate until it reaches +51.5 inches and will not operate again until the operator depresses the High Level Reset pushbutton.

D. HPCI will continue to operate for approximately 2 hours by cycling between +51.5 inches and -35 inches and then it will fail to operate.

References: SI-LP-00501 Rev. 01 pg 14 - 16 of 46

FSAR Section 8.3.2.1.1,125/250 VDC Station Battery Power System

A. Incorrect since HPCI will not operate after approximately 2 hours.

B. Incorrect since HPCI will not cycle between 51.5 inches and the reset setpoint. It requires manual action to start HPCI at the reset setpoint.

C. Incorrect since HPCI will automatically start again when level drops to -35".

D. Correct answer.


Final SRO Test43. 295004AK3.03 001

References: LT-LP-02704 Rev. 03. pg 37 - 39 of 61 LT-LP-02704 Rev. 03. pg 50 - 55 of 61 EO 200.018.a.01

Changed initial conditions slightly and changed 2 distractors because it is likely that everyone knows that 24/48 VDC systems are mainly for neutron monitoring systems.

A. Incorrect since this Cabinet is powered by Switchgear 2A and it doesn't cause all of these things to happen.

B. Incorrect since this Cabinet affects DC control power and DG loads.

C. Correct answer.

D. Incorrect since this affects HPCI and Core Spray.


Unit 2 is operating at 45% RTP when a loss of DC power causes the following:

Main Turbine trip Reactor Scram and both Recirc pumps trip 125/250V BATTERY VOLTS LOW alarm

Which ONE of the following losses will most likely cause this transient?

A. 125 VDC Cabinet "2B", 2R25-S002.

B. 125 VDC Cabinet '2D", 2R25-S004.

C.• 125/250 VDC Switchgear 2A, 2R22-S016.

D. 125/250 VDC Switchgear 2B, 2R22-S017.

44


Unit 2 was operating at 100% RTP when the reactor scrammed due to a turbine trip

from high vibrations.

Which ONE of the following describes the correct turbine valve response?

A. Turbine Stop Valves Closed Turbine Control Valves ClosedIntercept Valves ClosedIntermediate Stop Valves - Open Bypass Valves - One or more may be open depending on throttle pressure

B. Turbine Stop Valves - Closed Turbine Control Valves Open Intercept Valves - Closed Intermediate Stop Valves - Closed Bypass Valves - All open initially; close to control Rx pressure.

C. Turbine Stop Valves - Closed Turbine Control Valves - Closed Intercept Valves - Open Intermediate Stop Valves - Closed Bypass Valves - One or more may be open depending on throttle pressure

D. Turbine Stop Valves - Closed Turbine Control Valves Closed Intercept Valves - Closed Intermediate Stop Valves - Closed Bypass Valves - All open initially; close to control Rx pressure.

References: SI-LP-01901-00 Rev. 3 pg 44-46 of 79 SI-LP-01701-00 Rev. SI-00 pg 17 of 36 EO 019.010.a.01

A. Incorrect since all valves close except bypass valves on a turbine trip from Hi vibes.

B. Incorrect since all valves close except bypass valves on a turbine trip from Hi vibes.

C. Incorrect since all valves close except bypass valves on a turbine trip from Hi vibes.

D. Correct answer.



Reference: FSAR 15.2.3.6.2.1

A. Correct answer. Voids collapse due to initial pressure transient.

B. Incorrect since level initially decreases from collapsing of voids.

C. Incorrect since level is below the scram setpoint for feedwater from the initial void collapse.

D. Incorrect since level initially increases from swell.


Unit 2 was operating at 100% RTP when a scram occurred from a spurious Group I isolation.

Which ONE of the following describes the expected initial Reactor Water Level response and the reason for that response? (Assume no operator actions occur)

A• Reactor water level will decrease due to collapsing of voids and then will increase due to feedwater injection.

B. Reactor water level will increase due to feedwater injection still at 100% and then will decrease due to level setpoint at scram setpoint.

C. Reactor water level will decrease due to feedwater level control at scram setpoint and then increase due to Startup Level Control Valve (SULCV) leakby.

D. Reactor water level will decrease due to SRV's opening and then will increase due to feedwater injection.

46

Final SRO Test46. 295006G2.2.22 001

Reference: Tech Spec section 3.3.1.1 Tech Spec table 3.3.1.1-1

A. Incorrect since this action would only be required if you cannot meet the Required Action of Condition A. This can be met by bypassing one APRM and placing the other APRM in trip.

B. Incorrect since Condition G would be entered if the APRM were INOPERABLE for the Inop function. They are Inop for the Neutron Flux-High function.

C. Incorrect since you would only go below 28% RTP if the problem was turbine related.

D. Correct answer. Bypass one APRM and then you can meet Condition A by placing the other APRM in trip. This maintains the unit at the current power level as long as you want.


Unit 1 is operating at 100% RTP. The I & C Techs notify you at 0900 that "A" and "B" APRM's will not generate a scram signal until the reactor is at 122% RTP. Adjustments to the APRM's cannot be made for 24 hours.

Which ONE of the following describes the condition of the plant after the 24 hours has expired? (Provide copy of Tech Spec 3.3.1.1 conditions and SR's)

A. The Unit is in Mode 2 as required by Required Action F.1.

B. The Unit is in Mode 3 as required by Required Action G.1.

C. The Unit is <28% RTP as required by Required Action E.I.

Dr. The Unit is at 100% RTP with one APRM bypassed and the other APRM channel in trip.

47


References: 34AR-603-114-1S Rev. 6, Reactor High Pressure RC RPV CONTROL (NON-ATWS) Rev. 7 SI-LP-01401 Rev. 00 Table 4 SI-LP-00501 Rev. 01 Table 6

A. Incorrect since the MSIV's are closed from a valid isolation signal and cannot be reopened.

B. Incorrect since HPCI cannot be used due to high water level isolation signal.

C. Correct answer.

D. Incorrect since the MSIV's are closed from a valid isolation signal and cannot be reopened.


Unit I scrammed due to a Group 1 isolation from high main steam line flow. The following conditions exist:

All rods inserted Reactor Pressure 1090 psig Reactor Water Level +54 inches Drywell Pressure 1.5 psig Torus Level 150 inches

The operator has opened SRV 1 B21 -F01 3G to lower reactor pressure.

Which ONE of the following describes how reactor pressure will be controlled after the operator closes SRV 1 B21-F013G?

A. Reactor pressure will be automatically controlled by the EHC system using the Turbine Bypass Valves.

B. Reactor pressure will be manually controlled by the operator using HPCI with a suction from the condensate storage tank.

C.• Reactor pressure will be automatically controlled by the Low Low Set system between 847 and 1033 psig.

D. Reactor pressure will be manually controlled using the Reactor Feed Pump turbines.

48


Reference: Tech Spec Bases B 3.3.2.2, Feedwater and Main Turbine High Water Level Trip Instrumentation (Background).

A. Correct answer.

B. Incorrect since the turbine trip may cause water hammer in piping due to stop valves going closed.

C. Incorrect since the turbine trip does not perform an anticipatory scram function.

D. Incorrect since the valves closing may be subject to water hammer if the water level continues to increase.

49. 295009AK1.05 002


Which ONE of the following is the reason that the Main Turbine receives a trip signal if

the RPV experiences a high water level condition?

A" Protects the turbine from damage due to water entering the turbine.

B. Protects the main steam piping from damage due to water hammer.

C. Initiates an anticipatory reactor scram due to the simultaneous trip of the reactor feedwater pump.

D. Protects the turbine control and stop valves from damage due to water impingement.

Which ONE of the following is the main reason that RPV level is lowered during an

ATWS condition?

A." Reduce reactor power by reducing the natural circulation driving head.

B. Reduce steam generation rate which reduces the moderator temperature.

C. Prevent thermal stratification which prevents localized power peaks.

D. Reduce reactor pressure which allows more injection from low pressure systems.

49

Final SRO Test

Reference: LR-LP-20327 Rev. 10 Pg 42 of 53 Revised correct answer lAW lesson plan.

A. Correct answer.

B, C or D. Incorrect answer.

50. 29500902.4.1 001

Unit 2 has scrammed with a failure of all the control rods to insert. The following plant conditions exist:

Rx Power 8% Rx Water Level -50"

Drywell Pressure 3.5 psig Torus Pressure 3.0 psig Torus Temperature 141OF LLS is armed ADS is Inhibited

Which ONE of the following is an acceptable order from the Shift Supervisor?

(Distribute copy of BIIT Curve Graph 5)

A. Maintain RWL between -60" and +50" using Condensate and Feedwater.

B. Terminate and prevent all injection except Boron, CRD and RCIC until RWL is below -60".

C. Terminate and prevent all injection except Boron, CRD and RCIC until RWL is below -60" and Rx Power remains below 5% OR RWL is at -155" OR DW press is <1.85 psig and SRV's remain closed and RWL is below -60".

D. Restore and maintain RWL between +3" and +50" using HPCI with a suction from the Condensate Storage Tank.


Final SRO Test

Reference: CP-3 ATWS Level Control RC RPV Control LR-20328 Rev. 06 RPV Control ATWS (RCA)

A. Incorrect since the overide conditions above this step have been met which send you to a terminate and prevent leg.

B. Incorrect since you meet all of the requirements for the overide to send you to another terminate and prevent leg.

C. Correct answer.

D. Incorrect since these actions are in RC RPV Control and not in the CP-3 ATWS Level Control procedure.

51. 295010AA2.06 001


A DBA LOCA has occurred on Unit 2 and the following conditions exist:

Drywell Pressure 51 psig and increasing at 2 psig/min Reactor Water Level -230 inches and increasing at 10"/min with RHR pumps Bulk Drywell Temp 280OF Torus Water Level 218" and increasing slowly

Which ONE of the following should be ordered by the Shift Supervisor?

A. Vent the Drywell IRRESPECTIVE of offsite radioligical release rates.

B. Vent the Torus IRRESPECTIVE of offsite radioligical release rates.

C. Spray the Drywell after verifying within Drywell Spray Initiation Limit.

D. Enter the Severe Accident Guidelines (SAG's).

51

Reference:

Final SRO Test

PC-1 Primary Containment Control Drywell Spray Initiation Limit (Graph 8) (Consider providing PC-1 and Graph 8)

A. Incorrect since Torus water level is below 300".

B. Correct answer.

C. Incorrect since spraying the Drywell is not allowed since RWL is below Top of Active Fuel and RHR pumps are required to maintain adequate core cooling.

D. Incorrect since EOP's have direction to cover this situation.

52. 295012AA1.02 001


Unit I is operating at 50% RTP with all Drywell cooling units in AUTO. The following conditions exist in the Drywell :

CRD cavity area temperature 155OF Recirc pump motor area 135 0 F Vessel upper head area 178 0F Drywell average air temperature 149OF

Which ONE of the following describes which Drywell cooling fans should receive an auto start signal and which parameter provided that signal?

A. Fans B007A, B007B, B009A and B009B due to vessel upper head area high

temperature.

B.' Fans B007A, B007B, B008A and B008B due to CRD cavity area high temperature.

C. Fans B007A, B007B, B008A and B008B due to vessel upper head area high temperature.

D. Fans B008A, B008B, B009A and B009B due to Drywell average air high temperature.

52

Final SRO Test

References: SI-LP-01304 Rev. 01, pg. 16 & 17 of 53

A. Incorrect since drywell cooling fans do not start on vessel upper head area high temperature.

B. Correct answer.

C. Incorrect since drywell cooling fans do not start on vessel upper head area high temperature.

D. Incorrect since drywell cooling fans do not start on Drywell high average air temperature.

53. 295013AK1.03 001


Unit 1 is operating at 75% RTP with Safety Relief Valve (SRV) "G" leaking to the Suppression Pool. All attempts to reseat the valve have failed. One RHR loop is operating in the Suppression Pool Cooling mode with Suppression Pool temperature at 11 1OF and increasing very slowly.

Which ONE of the following actions would the crew be expected to take?

A. Maximize torus cooling by placing the other RHR loop in operation and continue operating.

B. Depressurize the RPV to less than 200 psig within 12 hours.

C. Reduce THERMAL POWER until all OPERABLE IRM channels < 25/40 divisions of full scale on Range 7 within 12 hours.

Dt. Place the reactor Mode Switch in the shutdown position immediately.

53

Final SRO Test

References: LR-LP-20201-13 Tech Spec Section 3.6.2.1 Suppression Pool Average Temperature PC-1 Primary Containment Control

A. Incorrect since this action should have already been performed per PC-1 when Torus temperature went above 1 00OF.

B. Incorrect since this is the Tech Spec action if Torus temperature is > 120 0 F.

C. Incorrect since this is the Tech Spec action if Torus temperature cannot be restored to < 100OF within 24 hours but it is no longer applicable since the Condition has been exited when the temperature went >11 0OF.

D. Correct answer per Tech Spec 3.6.2.1 Condition D.

54. 295013AK3.01 001


Unit 2 is operating at 100% RTP. The quarterly HPCI Flow Rate Test was just suspended with the following plant conditions:

Torus Cooling Both loops in operation Torus Temperature 105 0 F and increasing Torus Level 149" and decreasing

Which ONE of the following describes the reason for suspending adding heat to the Suppression Pool per Tech Spec section 3.6.2.1?

A. Ensures primary containment design limits are not exceeded in the event of a

LOCA.

B• Preserves heat absorption capabilities of the suppression pool.

C. Ensure PCPL is not reached in the event of an emergency depressurization.

D. Ensure HCTL is not reached in the event of a LOCA.

54

Final SRO Test

References: Tech Spec Bases 3.6.2.1

A. Incorrect since this

B. Correct answer.

C. Incorrect since this

D. Incorrect since this

55. 295014AK2.03 001

is the reason for actions at 120 0 F.

is

is

the

not

basis for HCTL.

a basis for suppression cooling action.

References: FSAR section 15.2

A. Correct answer - all occurances increase reactor power and fuel temperature.

B, C and D. Incorrect since at least one occurance decreases reactor power and fuel temperature.


Which ONE of the following describes the Abnormal Operating Occurances (Plant Transients) which increase fuel temperature?

A" Recirc Flow Control Failure-Increasing Flow, Loss of Feedwater Heating, Inadvertent start of HPCI.

B. Loss of Shutdown Cooling, Loss of Condenser Vacuum, Recirc Flow Control Failure-Decreasing Flow.

C. Loss of Feedwater Heating, Trip of one Recirc Pump, Startup of idle Recirc Pump.

D. Recirc Flow Control Failure-Increasing Flow, Loss of Condenser Vacuum, Turbine Trip with Bypass Valves available.

55


References: LR-LP-20305 Rev. 04 pg 17 and 18 of 24

A. Incorrect since Core Spray will not dilute boron flow and Core Spray should not be injecting at 400 psig.

B. Incorrect since level will increase with this valve open.

C. Correct answer.

D. Incorrect since leaving this valve unisolated will not prevent Boron mixing.


During an ATWS on Unit 2 the Operator performing RC-2 fails to close 2N21-F110, SULCV Bypass. RPV pressure subsequently drops to 400 psig with all injection systems still available.

Which ONE of the following describes the potential adverse consequences of this condition?

A. Core Spray injection will occur diluting boron flow.

B. Level will decrease and the MSIV's may close causing a loss of the main condenser and heat sink.

C. A significant power excursion could occur due to uncontrolled injection to the RPV.

D. Boron mixing will be prevented due to less flow in the downcomer.

56


References: HNP-2-FSAR-3 pg 3.1-16 and 17. HNP-2-FSAR-7 pg 7.5-5 thru 7.5-10. Procedure 31RS-OPS-001-1S Rev. 3

A. Incorrect since Technical Specifications do not describe how to shutdown the plant from outside the control room.

B. Correct answer.

C. Incorrect since the Technical Requirements Manual does not describe how to shutdown the plant from outside the control room.

D. Incorrect since it is a requirement to be able to perform a prompt hot shutdown from outside the control room.


There is an electrical fire in the Control Room and black smoke has made the Control Room inaccessable. If possible, prior to leaving the Control Room the Reactor Operator inserts a manual Scram per 31RS-OPS-O01-lS, Shutdown from Outside Control Room.

Which ONE of the following describes why the procedure also has steps to Scram the reactor by de-energizing RPS or actuating the Scram Discharge Volume level switches?

A. The Technical Specifications require that Reactor Scram capability from outside the Control Room be maintained.

B1 The FSAR requires the ability for prompt hot shutdown of the reactor from locations outside the Control Room.

C. The Technical Requirements Manual requires the capability to Scram the reactor from outside the Control Room.

D. The capability for prompt hot shutdown of the reactor from outside the Control Room is not required but is a safe operating practice.

57

Final SRO Test58. 295017G2.3.4 001

References: Procedure 73EP-EIP-017-OS, Emergency Exposure Control pg 4 & 6 of 13.

A. Incorrect since the decisions.

Emergency Director has the responsibility to make these

B. Incorrect since the maximum dose allowed without volunteering is 25 Rem.

C. Incorrect since the decisions. ,

D. Correct answer.

Emergency Director has the responsibility to make these


An event has occurred on Unit 1 that resulted in an individual getting injured. The individual is disabled and is in a 100 R/Hr field. An individual is standing by to save the disabled individuals life (He has NOT volunteered). The job will require being in the radiation field for 13 minutes. After conferring with HP Supervision the has determined that is the maximum amount of dose allowed per 73EP-EIP-017-OS, Emergency Exposure Control, for this lifesaving attempt. (CHOOSE the answer that correctly fills in the blanks.)

A. Shift Supervisor, 10 Rem

B. Emergency Director, 10 Rem

C. Shift Supervisor, 25 Rem

D" Emergency Director, 25 Rem

58


References: SI-LP-02301 Rev. SI-00, pg 13 of 21 EO 023.001.c.05

A. Incorrect since a complete loss of stator cooling does not cause an immediate turbine trip.

B. Incorrect since the runback does not need to be complete until 3.5 minutes have elapsed. Initial amps are less than 19080 so the 2 minute timer does not start.

C. Correct answer.

D. Incorrect since 2 minute timer is not in effect.


Unit I is in the process of increasing to full load after a refueling outage when a complete loss of stator cooling occurs. The following conditions existed just prior to the loss of stator cooling:

Generator output 650 MWe Stator current 17000 amps Stator cooling disch press 90 psig Stator cooling conductivity .5 micro mho

Which ONE of the following is the appropriate plant response?

A. The turbine will immediately trip causing a reactor scram.

B. The turbine will trip if it does not complete a runback to less then 24% load in two minutes.

C< The turbine will trip if stator amps are not less than 4500 amps within 3.5 minutes.

D. The turbine will trip if stator amps are not less than 15,000 amps within two minutes and less than 4500 amps within 3.5 minutes. (total time = 3.5 minutes)

59


References: Procedure 34AB-G41-001-IS Rev. 2, Loss of Fuel Pool Cooling pg 2 & 3 of 12.

Procedure 34SO-E11-010-2S Rev. 30.3, Residual Heat Removal System pg 65 of 238.

Procedure 34AB-EI 1-001-1S Rev. 3, Loss of Shutdown Cooling pg 4 of 19.

A. Incorrect since "A" RHR Pump cannot be used for Fuel Pool Cooling Assist.

B. Incorrect since feeding the Unit 1 Fuel Pool is allowed when there are no other means of cooling available.

C. Correct answer.

D. Incorrect since feeding the Unit 2 Fuel Pool is allowed when feeding Unit 1 Fuel Pool cannot be established.


The following conditions exist on Unit 1:

Reactor is in Condition 5 "B" RHR Pump in Fuel Pool Cooling Assist Reactor cavity is flooded Fuel Pool gates are removed Fuel shuffle has just been completed

The supply breaker to I B RHR Pump has tripped and cannot be reclosed. Fuel Pool temperature starts to increase.

Which ONE of the following is the most appropriate action to lower Fuel Pool

temperature?

A. Place the "A" RHR Pump in Fuel Pool Cooling Assist.

B. Feed the Unit 1 Fuel Pool by opening 1G41-F041, Spent Fuel Pool Make-up from CST.

C• Place the "D" RHR Pump in Fuel Pool Cooling Assist.

D. Remove the transfer canal gates and feed the Unit 2 Fuel Pool by opening 2G41-F054, Spent Fuel Pool Make-up from CST.

60

Final SRO Test 61. 295022G2.1.30 001

Unit 2 CRD pump B has been started from the Remote Shutdown Panel. A System Operator (SO) checking the status of the CRD pump reports the following:

CRD Pump B is running. CRD Pump B suction valve is closed. CRD Pump B discharge pressure is about 200 psig. CRD Pump B suction pressure is downscale.

Which ONE of the following describes how the CRD pump should have responded?

A4• SHOULD have started but auto tripped on low suction pressure.

B. SHOULD have started but auto tripped on low discharge pressure.

C. SHOULD NOT have auto tripped because the low suction pressure trip is presently bypassed.

D. SHOULD NOT have auto tripped because the low discharge pressure trip is presently bypassed.

References: SI-LP-05201-00 Rev. 00 Table 8

SI-LP-00101-00 Rev. SI-00 pg 15-17 of 46

A. Correct answer.

B. Incorrect since the CRD pump doesn't have a trip for low discharge pressure.

C. Incorrect since the CRD pump should have tripped on low suction pressure. The LOCA pump trip is bypassed when taking the RSP switch to Emergency and not the low suction pressure trip.

D. Incorrect since the CRD pump should have tripped on low suction pressure. The CRD pump doesn't have a low discharge pressure trip.



References: LT-LP-04502 Rev. 03, pg. 9 of 57.

A. Incorrect since this does not prevent inadvertent criticality but protects site personnel and public from exposure from an accident.

B. Incorrect since SBLC does not prevent inadvertent criticality but it can take care of it once it happens.

C. Correct answer.

D. Incorrect since the Mode Switch in this position will allow a control rod to be withdrawn.


The refueling process (i.e. core offload/reload sequence, moving one component at a time, SRO supervising all core alterations, etc.) is designed to prevent an inadvertent criticality.

Which ONE of the following is used as a backup to the refueling process to prevent an

inadvertent criticality?

A. Secondary Containment is OPERABLE.

B. Standby Liquid Control is OPERABLE.

C." Refueling Interlocks.

D. Mode Switch locked in REFUEL.

62

Final SRO Test63. 295024EA1.04 002

References: PC-1 Primary Containment Control Procedure 34SO-Ell-01 0-2S Rev. 30.3 pg 71 and 72 of 238

A. Incorrect since this is the value for Torus spray flow.

B. Correct answer.

C. Incorrect since this is the flow rate to maintain < if using one pump for drywell spray.

D. Incorrect since this is the maximum flow rate for the loop.


Unit 2 has experienced a LOCA with the following conditions present:

Reactor is Shutdown Reactor water level -100" Reactor pressure 550 psig Drywell pressure 12 psig Drywell temperature 2650F Torus pressure 11.5 psig

The Shift Supervisor has ordered initiation of Drywell Sprays.

In accordance with 34S0-El1-010-2S, Residual Heat Removal System, which ONE of the following is the MINIMUM drywell spray flow rate required to ensure an effective drywell pressure reduction?

A. 700 gpm.

B.3 5000 gpm.

C. 7700 gpm.

D. 17000 gpm.

63


References: SI-LP-00101 Rev. SI-00, pg 21 - 24 of 46. EO 010.024.1.02

A. Incorrect since the low reactor water level setpoint of -35" has not been reached.

B. Incorrect since the high reactor pressure actuation setpoint has already been reached (1170 psig).

C. Correct answer.

D. Incorrect since ARI should have already initiated on high pressure (1170 psig).


Unit 2 is operating at 100% RTP. A turbine trip causes a reactor scram but not all of the turbine Bypass valves open on rising pressure. RPV pressure increases to 1190 psig and reactor water level drops to -32" and is currently increasing.

Which ONE of the following statements correctly describes the response of the Alternate Rod Insertion (ARI) system?

A. ARI WILL automatically initiate due to Rx Low Water Level.

B. ARI WILL NOT automatically initiate until RPV water level decreases further.

Ct. ARI WILL automatically initiate due to high RPV pressure.

D. ARI WILL NOT automatically initiate until RPV pressure increases further.

64

Final SRO Test 65. 295025EA2.04 001

Unit 2 Scrammed from High Drywell pressure due to a small leak in the Recirc piping. The following conditions currently exist:

Drywell pressure 3.5 psig (decreasing) Drywell temperature 2450 F (decreasing) Torus level 185 inches (increasing) Torus temperature 160OF (decreasing) Reactor pressure 600 psig (decreasing) Torus sprays and cooling running

Based upon the above conditions the Shift Supervisor has determined that injection into the RPV from sources external to primary containment must be terminated.

Which ONE of the following identifies the systems that are EXEMPTED from

termination?

A' Systems needed for adequate core cooling, boron injection and CRD.

B. Systems needed to shutdown the reactor, boron injection and CRD.

C. Systems needed for boron injection, CRD and RCIC.

D. Systems needed for adequate core cooling, fire fighting and boron injection.

References: PC-1 Primary Containment Control Suppression Pool Level High SRV Tail Pipe Level Limit (Graph 6)

A. Correct answer due to torus water level CANNOT be maintained below SRV Tail Pipe Level Limit (graph 6).

B. Incorrect since systems needed to shutdown the reactor are not exempted.

C. Incorrect since these systems are excepted when preventing all injection during an ATWS.

D. Incorrect since fire fighting systems are only excepted during SC Secondary Containment Control.


Final SRO Test66. 295026EK2.04 001

Reference: LT-LP-05601 Rev. 03 Safety Parameter Display System EO 056.002.c.03

A. Incorrect because the average temp must be <100OF to be green.

B. Incorrect because there are 2 or more imputs to each group and temp is > 100OF so the box should be red.

C. Incorrect because there are 2 or more imputs to each group and temp is > 100OF so the box should be red.

D. Correct answer since average temp > 100OF and there are 2 or more inputs to each group.


Unit 1 is in Mode 1 with the quarterly HPCI Pump Operability Surveillance in progress. The Suppression Pool average temperature is 102 0 F. The following signals are being sent to SPDS:

Group 1 signals: 4 out of 5 are operable and reading 102 0 F. Group 2 signals: 5 out of 5 are operable and reading 103 0 F. Group 3 signals: 4 out of 5 are operable and reading 1020F.

Which ONE of the following conditions describes the SPDS indication?

A. Green box with the average temp indicated since average temp is <1 050 F.

B. Yellow box with the average temp indicated since all groups have a signal.

C. Yellow box with no temp indicated since all signals are not operable.

D. Red box with the average temp indicated since average temp is >100OF.


Reference: LR-LP-20310-05, p. 59 99 exam question #15 (answers reordered)

A. Incorrect since elevated Drywell temperature has no effect release of Hydrogen to the containment.

B. Incorrect since initiating drywell sprays has no effect on equipment qualification.

C. Correct answer.

D. Incorrect since drywell temperature does not affect the capacity of the suppression chamber to drywell vacuum breakers.


Which ONE of the following is the basis for initiating drywell sprays before the bulk drywell temperature reaches the drywell design temperature limit?

A. To prevent increased degredation of structural concrete and release of hydrogen to the drywell.

B. To maintain the equipment qualification of the drywell valves above the 185' elevation, capable of removing the full decay heat load following a LOCA.

C. To ensure that equipment within the drywell will operate when required.

D. To ensure that the capacity of the suppression chamber - drywell vacuum breakers is not exceeded.

67

Final SRO Test 68. 295029EK1.01 001

In accordance with PC-1 PRIMARY CONTAINMENT CONTROL, drywell sprays cannot be initiated unless torus level is below 215 inches.

Which ONE of the following describes the reason for this restriction?

A. The suppression pool-to-reactor building vacuum breaker connections are submerged preventing their operation if needed.

B.r The drywell-to-suppression pool vacuum breakers are submerged which may cause the containment differential pressure capability to be exceeded.

C. The suppression pool-to-reactor building vacuum breaker connections are submerged and containment integrity would be lost when they open.

D. The drywell-to-suppression pool vacuum breakers are submerged allowing suppression pool water to be siphoned into the drywell.

References: LR-LP-20310 Rev. 05, pg. 31 of 96 EO 201.072.a.27, 201.073.a.06, 201.075.b.15, 201.076.a.14

A. Incorrect since 215" is concerned with covering the drywell-to-suppression pool vacuum breakers.

B. Correct answer.

C. Incorrect since 215" is concerned with covering the drywell-to-suppression pool vacuum breakers.

D. Incorrect since covering the vacuum breakers would not result in siphoning water to the drywell.

Thursday, November 21,2002 01:37:55 PM 68


Unit 2 has developed a leak in the Torus and the Shift Supervisor has entered PC-1 Primary Containment Control. Mechanical Maintenance and Health Physics have been dispatched to investigate and repair the leak. Mechanical Maintenance has reported that the leak should be stopped in approximately 30 minutes. Torus level is currently at 120" and is dropping at a rate of 1" per minute.

Which ONE of the following should be directed by the Shift Supervisor?

A. wait until Torus level drops to 110" and order HPCI tripped.

B. wait until Torus level drops to 98" and order Emergency Depressurization per CP-1.

C. order HPCI tripped and depressurize reactor through the Main Valves irrespective of cooldown rate.

D. order HPCI tripped and depressurize reactor through the Main valves without exceeding a 100OF/hr cooldown rate.

Turbine Bypass

Turbine Bypass

References: LR-LP-20310 Rev. 05 pg. 20 - 23 of 96 RC RPV CONTROL (NON-ATWS) PC-1 PRIMARY CONTAINMENT CONTROL

A. Incorrect since actions should be taken before they are hit if the trend indicates that the limit will be met.

B. Incorrect since actions should be taken before they are hit if the trend indicates that the limit will be met.

C. Correct answer.

D. Incorrect since it is acceptable to exceed the 100 anticipate blowdown per overide in PC RPV Control

F/hr cooldown rate when you (Non-ATWS).



References: LR-LP-20310 Rev. 05, pg. 23 of 96

A. Incorrect since low torus level will not affect RCIC suction trip since it is set at 10" Hg vacuum.

B. Incorrect since Emergency Depressurization is not required before RCIC exhaust line is uncovered.

C. Incorrect since low torus level will not affect RCIC suction trip vacuum and Emergency Depressurization is not required before uncovered.

since it is set at 10" Hg RCIC exhaust line is

D. Correct answer.


Per the EOP's HPCI is required to be tripped on lowering Torus level but RCIC is allowed to continue to operate if necessary.

Which ONE of the following describes why RCIC operation with lowering Torus level is acceptable?

A. The exhaust flow rate of RCIC is approximately equal to decay heat and a low Torus level will cause RCIC to trip on low suction pressure.

B. Elevated Torus pressure will cause RCIC to trip much sooner than HPCI and Emergency Depressurization is required before the exhaust line is uncovered.

C. Low Torus level will cause RCIC to trip on low suction pressure and Emergency Depressurization is required before the exhaust line is uncovered.

Dt. Elevated Torus pressure will cause RCIC to trip much sooner than HPCI and the exhaust flow rate of RCIC is approximately equal to decay heat.

70

Final SRO Test71. 295031G2.4.4 001

References: LR-20308 Entry Conditions LR-20310 Entry Conditions LR-20328 ATWS Conditions

A. Incorrect answer. Does not meet conditions for ATWS. All rods are at position 02 or beyond.

B. Incorrect answer. Does not meet entry conditions yet. Not appropriate to take actions to prevent meeting entry conditions.

C. Correct answer.

D. Incorrect answer. EOP <+3".

entry is required since conditions were previously met. RWL


Unit 1 is operating at 100% power when a leak in the Drywell develops. Reactor water level is trending down and Drywell pressure, temperature and level are trending upward. The SRO orders a reactor SCRAM with the following conditions occuring:

RWL initially reaches +2" and stabilizes at +15" Drywell pressure reaches 1.83 psig and stabilizes Drywell temperature currently at 147 0 F and rising slowly Torus level currently at 148" and rising slowly Reactor pressure at 920 psig and steady 6 control rods stuck at position 02 and all others fully inserted

Which ONE of the following actions are required by the Shift Supervisor under these conditions?

A. Enter RCA RPV Control (ATWS) and take actions to ensure reactor stays shutdown

under all conditions.

B. Enter PC-1 and PC-2 and take actions to prevent reaching entry conditions.

C•. Enter RC RPV Control (Non-ATWS) and take actions to stabilize plant.

D. Entry into EOP's not required since an entry condition does not exist at this time. 34AB-C71-001-2S, Scram Procedure is entered.

71


The SC-SECONDARY CONTAINMENT CONTROL EOP requires Emergency Depressurization if 2 or more areas exceed the Maximum Safe Operating Temperature and a primary system is discharging reactor coolant into secondary containment.

Which ONE of the following statements explain the reason for this action?

A. The rise in secondary containment parameters indicate a wide-spread problem which may pose an indirect but immediate threat to secondary containment integrity or continued safe operation of the plant.

B. The rise in secondary containment parameters indicate substantial degredation of the primary system and may lead to fuel failure if the leaks are not isolated.

C. The rise in secondary containment parameters indicate a wide-spread problem which may pose a direct and immediate threat to secondary containment integrity or equipment located in secondary containment.

D. The rise in secondary containment parameters indicate substantial degredation of the primary system and emergency depressurization places the plant in the safest condition as quickly as possible.

References: LR-LP-20325 Rev. 05, pg 19 and 20 of 40 EO 201.077.a.14, 201.078.a.15, 201.079.a. 19

A. Incorrect since condition pose a DIRECT threat to containment, not an INDIRECT threat.

B. Incorrect since this condition does not indicate substantial primary system degredation.

C. Correct answer.

0. .n..r.. c thia zzn.. to1 doot indicotc . uunttl primary system degredation. ,,. r2,- c-,t,..-,-o-s---c,. A-,--.-r - t


Final SRO Test73. 295033G2.3.10 001

References: 73EP-RAD-001-0S, Radiological Event Rev. 1.1 pg 4 of 7 SC - SECONDARY CONTAINMENT CONTROL

A. Incorrect since a reactor scram is not required by the EOP's.

B. Correct answer.

C. Incorrect, a reactor blowdown is not required since there is not a primary system discharging into secondary containment.

D. Incorrect since the reactor should be shutdown per the EOP's.


Unit 1 is operating at 100% RTP. The Fuel Movement Team is moving fuel bundles in the Unit 1 Fuel Pool to get ready for an upcoming outage. There is a malfunction associated with the mast and a fuel bundle is dropped in the pool. This caused some fuel pins to break and radiation levels are increasing on the Refuel Floor and around the Fuel Pool pumps. The following radiation levels exist on Unit 1:

Refuel Floor 1500 mR/hr Spent Fuel Pool Demin Equip 2000 mR/hr Fuel Pool Demin Panel 75 mR/hr

Which ONE of the following actions should the Shift Supervisor order? (Provide copy of Unit I SC-Secondary Containment Control)

A. Scram the reactor and evacuate the Reactor Building per 73EP-RAD-O01-OS,

Radiological Event.

B.' Commence a Normal Unit Shutdown and evacuate the associated High Rad areas.

C. Scram the reactor and commence a Reactor Blowdown per the EOP's.

D. Announce the High Rad condition over the public address system and evacuate the affected areas. Reactor operation should not be affected.

73


References: SI-LP-01303 Rev. SI-00, Figures 10.

A. Incorrect since the RHRSW side of the HX is not highly radioactive.

B. Incorrect since the drywell will contain the Recirc Pump seal leakage.

C. Incorrect since the feedwater reg valve is in the turbine building.

D. Correct answer since this steam leak is directly off of main steam and has the potential of being highly radioactive.


Which ONE of the following conditions would most likely cause a Secondary Containment Ventilation High Radiation isolation?

A. A leak has developed on the RHRSW side of the RHR Heat Exchanger and the

water level in the room is approximately 1/2" deep.

B. A Recirc Pump seal failure which causes Drywell Pressure to exceed 1.85 psig.

C. The packing is blown on a Startup Level Control Valve (SULCV) and the area is blanketed in steam.

D'. A leak has developed upstream of the HPCI Stop Valve but it is not large enough to cause a high temperature isolation of HPCI.

74

Final SRO Test 75. 295035G2.1.7 001

A tornado was observed moving toward the plant 15 minutes ago. Meteorological instruments have detected wind speeds in excess of 100 mph. The annunciator for "RB INSIDE TO OUTSIDE AIR DIFF PRESS LOW" has just alarmed and the Inside Rounds SO reports air rushing in and then out through a crack in the Reactor Bldg wall on the 158' EL. The following conditions exist for Secondary Containment:

Reactor Power Both Units at 100% RTP Rx Bldg Dp fluctuating between 0" and +.25" Hg Rx Bldg Vent System system isolated Rx Bldg Vent Rad level 1 mRlhr Area water levels normal

Which ONE of the following describes the appropriate actions the Shift Supervisor should take? (Provide copy of 73EP-EIP-001-OS)

A. Declare an ALERT. Initiate actions for Secondary Containment System being Inoperable.

B• Declare a SITE AREA EMERGENCY. Initiate actions for Secondary Containment System being Inoperable.

C. Declare an ALERT. No actions required for Secondary Containment System.

D. Declare a SITE AREA EMERGENCY. No actions required for Secondary Containment System.

References: 73EP-EIP-001-OS Rev. 14.2 pg 16 & 22 of 47 SC - Secondary Containment Control

A. Incorrect since should declare a Site Area Emergency due to tornado damage.

B. Correct answer.

C. Incorrect since should declare a Site Area Emergency due to tornado damage.

D. Incorrect since actions are required by Tech Specs since the Containment is Inoperable.



References: SC - Secondary Containment Control

A. Incorrect since the reactor isn't required to be shutdown until area water level in more than one area is above max safe.

B. Incorrect since the system discharging into the affected area must be isolated with some exceptions. These exceptions are not met.

C. Incorrect since the systems are not required to be declared INOPERABLE just

because the area water level is above max normal.

D. Correct answer since level cannot be restored to normal with the leak in progress.


The following conditions exist on Unit 2:

-Sump alarms in the Control Room indicate a leak in the Southwest Diagonal Area of the Reactor Building.

-The plant operator reports water level is 15" above 87' elevation and increasing. -The source of the leak has been identified as a fire protection pipe in the room. -No increase in area temperature or radiation has been noted. -No other emergency condition exists at this time.

In accordance with SC-SECONDARY CONTAINMENT CONTROL EOP, which ONE of the following actions is appropriate? (Provide copy of SC-SECONDARY CONTAINMENT CONTROL EOP)

A. Shutdown the reactor per the Fast Shutdown procedure.

B. Install a submersible pump to lower level; do not isolate the fire system.

C. Declare the affected systems INOPERABLE and enter appropriate Tech Spec.

Dt. Isolate the fire system header discharging into the area.

76


Unit 2 has scrammed due to low reactor water level. Multiple control rods did not insert and the ATWS procedure is being directed by the Shift Supervisor. The Shift Supervisor has ordered the RO to insert control rods by increasing CRD cooling water differential pressure (dp).

Which ONE of the following describes how this action causes control rods to insert?

A' Increased cooling water dp puts additional pressure on the underside of the CRDM drive pistons.

B. Increased cooling water dp puts additional pressure on the top of the CRDM drive pistons.

C. Increased cooling water dp causes driving flow to increase.

D. Increased cooling water dp causes driving flow to decrease.

Reference: LR-LP-20314 Rev. 03 pg 13

EO 001.034.a.01

A. Correct answer.

B. Incorrect. Additional pressure is placed on the underside of the drive piston.

C. Incorrect answer. Increasing cooling water Dp has no effect on drive flow.

D. Incorrect answer. Increasing cooling water Dp has no effect on drive flow.



An ATWS has occurred on Unit 2 and RCIC is being used to inject boron. The RCIC Minimum Flow Valve F01 9, has been closed and system flow has been verified to be greater than 122.5 gpm. The Minimum Flow Valve's breaker was then opened to prevent operation of the valve.

Which ONE of the following describes the reason that the Minimum Flow Valve was disabled?

A. Ensure that the RCIC pump does not go to run out if the Minimum Flow Valve stuck

in the open position.

B1' Ensure that all boron flow is to the Vessel and not to the Suppression Pool.

C. Ensure that all boron flow is to the Vessel an not to the CST.

D. Ensure that the RCIC pump has the proper amount of cooling flow.

References: LR-LP-20320 Rev. 05, pg. 22 of 28. SI-LP-03901 Rev. 00, Figure 1 EO 039.019.a.11, 039.020.a.07

Changed answer D to relate to pump minimum flow requirements.

A. Incorrect since RCIC minimum flow valve would still operate at the appropriate setpoints.

B. Correct answer.

C. Incorrect since the minimum flow valve goes back to the Torus.

D. Incorrect since the amount of flow is what protects the pump from overheating and not whether the minimum flow valve is deactivated.



Which ONE of the following describes the basis for an offsite radioactivity release rate

of 0.57 mr/hr as an entry condition to Radioactive Release Control (RR)?

A. Represents an immediate threat to the continued health and safety of the public.

B. Corresponds to an entry into a Site Area Emergency in the Emergency Plan.

C. Indicates a primary system break which cannot be isolated.

Dt Represents a release rate that is higher than expected during normal plant operations but does not pose an immediate threat to the public.

Reference: LR-LP-20325 Rev. 05, pg 26,29 EO 201.082.a.09

A. Incorrect answer. This is a bases for the 1000 mr/hr entry to RCA.

B. Incorrect since .57 corresponds to an alert action level.

C. Incorrect answer. This is a bases for the 1000 mr/hr entry to RCA.

D. Correct answer.



As a result of maintenance the control room ventilation system is aligned to the Isolation Mode. A subsequent off-site release resulted in the initiation of a Main Control Room air intake high radiation signal.

Which ONE of the following describes the impact the high radiation signal has on the continued operation of the Control Room HVAC System?

A. Since the control room ventilation system is not in the Normal Mode then the Control Room Operator must manually initiate the Pressurization Mode per 34S0-Z41-001-1S, Control Room Ventilation System.

B. The Control Room Ventilation System will remain in the Isolation Mode until the high radiation signal clears. Then the Control Room Operator must realign the system to the Normal Mode.

CU. 1Z41-F016, Outside Air Intake Damper will automatically open along with the filter inlet valves. This will place the Control Room Ventilation System in the Pressurization Mode to protect the Control Room personnel.

D. The Control Room Ventilation System will automatically shift to the Purge Mode. The Control Room Operator must start the exhaust fan and ensure 1Z41-F018A(B), Suction Damper is open.

References: INPO exam bank (Fermi) 34SO-Z41-001-1S Rev. 17 pg 8 - 12 of 33

A. Incorrect since the Control Room HVAC System will automatically shift to the Pressurization Mode upon high inlet radiation levels.

B. Incorrect since the Control Room HVAC System will automatically shift to the Pressurization Mode upon high inlet radiation levels.

C. Correct answer.

D. Incorrect since the Control Room HVAC System will automatically shift to the Pressurization Mode upon high inlet radiation levels.


Final SRO Test 81. 400000A4.01 001

Unit 1 is operating at 100% RTP with the following lineup for RBCCW pumps:

"A" RBCCW Pump Running "B" RBCCW Pump Running "C" RBCCW Pump Auto not running

Which ONE of the following is the expected plant response from a trip of 4160V Bus I G assuming no operator action is taken?

A. No change is plant status as the "C" RBCCW pump has lost power but was not running.

B. "A" RBCCW pump trips; "C" RBCCW pump cannot start due to a loss of power; Recirc pump, RWCU NRHX outlet, and CRD pump temperatures increase; Recirc pumps trip due to Recirc M-G set high oil temperatures.

C• "B" RBCCW pump trips; "C" RBCCW pump auto starts at 90 psig system pressure; Recirc pump, RWCU NRHX outlet, CRD pump and Recirc M-G set oil temperatures do not measurably change.

D. "A" AND "B" RBCCW pumps trip; "C" RBCCW pump auto starts at 90 psig; Recirc pump, Recirc M-G set, RWCU NRHX outlet and the running CRD pump temperatures increase; Recirc M-G set and CRD pumps eventually trip on high temperatures.

References: LT-LP-02703-03 Rev. 03 pg 48-49 of 63. SI-LP-00901-00 Rev SI-00 pg 13 of 25 EO 200.014.a.05

A. Incorrect since the "C" RBCCW pump auto starts to maintain RBCCW system pressure.

B. Incorrect since the "B" RBCCW pump is the one that trips.

C. Correct answer.

D. Incorrect since the loss of the 2G Bus does not affect the "A" RBCCW pump.



References: 34SO-P33-001

A. Incorrect since the analyzers do not have an auto start signal.

B. Incorrect since the isolation signal is present and it must be bypassed, action would otherwise be correct.

C. Correct answer.

D. Incorrect since the isolation signal is present ain it must be bypassed, otherwise the action would be correct.


After a scram on Unit 2, Reactor Water Level decreases to 0" before being restored to the normal band. When a high Drywell temperature is observed the Shift Supervisor enters the Primary Containment flowcharts. The Shift Supervisor directs H20 2 Analyzers to be placed in service.

Which ONE of the following is required to start the H20 2 Analyzers?

A. No action is required since the H20 2 Analyzers should be running due to an automatic start signal.

B. Depress the H20 2 analyzer reset pushbuttons on the 2H1 1-P700 panel.

Co Place the 2P33-S16/S17 LOCA OVERRIDE switches in Bypass on the 2H11-P700 panel.

D. Place the 2P33-S25(B)A mode switches in ANALYZE on the 2P33-P601 B(A) panels.

82


References: 34AB-X43-001-2S, Fire Procedure Rev. 10 ED 6, pg 2 of 77.

A. Incorrect since the purge mode will draw in more outside air.

B. Incorrect since this lineup still draws in outside air.

C. Correct answer per step 4.6.

D. Incorrect since the control room ventilation system should not be secured under these circumstances.


A fire has been reported outside the Turbine Building which is producing large amounts of smoke. The Fire Brigade has been dispatched and the Brigade Leader reports back that the fire is under control. He also expresses a concern that the Control Room may get smoke drawn into the ventilation.

Per 34AB-X43-O01-2S, Fire Procedure, which ONE of the following describes the action the Control Board Operator (CBO) should take?

A. Place the control room ventilation system in the Purge mode.

B. Maintain the control room ventilation in the normal lineup until the fire is out.

C'. Place the control room ventilation system in the isolation mode.

D. Secure control room ventilation until the fire is out and the smoke clears.

83

Final SRO Test84. G2.1.22 001

References: Tech Spec section 1.1, Table 1.1-1 Modified from question #84 on 1995 SRO exam

A. Correct answer.

B,C,D Incorrect (See table 1.1-1) Unless a Special Operations Tech Spec is invoked then the reactor changes modes when moving the mode switch to refuel.


Unit 2 has been shutdown for a refueling outage. After 16 hours the following conditions exist:

Reactor Mode Switch: Refuel Reactor Temperature: 165OF and steady Reactor Pressure: 0 psig All reactor vessel head closure bolts are fully tensioned. All rods are IN.

Which ONE of the following is the correct Mode of Operation for Unit 2?

A' Mode 2

B. Mode 3

C. Mode 4

D. Mode 5

84


References: LT-LP-30004 Rev. 04 pg 36 of 64 EO 300.022.a.06 Modified the question from a locked open valve to a throttled valve. Question #85 on 1995 SRO exam.

A. Correct answer.

B. Incorrect since there is not a mark on the stem when the valve is initially positioned.

C. Incorrect since the valve should not be operated to verify position.

D. Incorrect since the valve should not be operated to verify position.


During a valve lineup, an operator needs to check a valve in the throttled position. It is

noted that the valve has a locking device installed.

Which ONE of the following describes the actions the operator should perform?

A. Leave the locking device installed and perform a locking device operability check only.

B. Leave the locking device installed, verify mark on stem matches the appropriate position established when the valve was initially positioned.

C. Unlock the valve, turn the handwheel in the closed direction for 1 full turn to verify the valve is throttled, reopen the valve 1 full turn and install the locking device.

D. Unlock the valve, turn the handwheel in the closed direction and count the number of turns until it is seated, reopen the valve to the appropriate position and install the locking device.

85


Reference: 34SO-T46-001-2S Rev. 14 Pg 3

A. Incorrect since the paint fumes do not increase the load on the SBGT system. The fumes affect the operation of the charcoal filters.

B. Incorrect since you do not place the SBGT control switches in OFF when painting is

in progress.

C. Correct answer per 34SO-T46-001-2S Precaution 5.1.4.

D. Incorrect since the paint fumes could not start a fire in the SBGT train and you do not place the SBGT control switches in OFF when painting is in progress.


You have just been notified by the SOS that solvent based painting is scheduled to be performed in the Reactor Building which covers an area of the floor approximately 200 ft2 . The SBGT system takes a suction from this area when the system is in operation.

Which ONE of the following describes the effect of running the SBGT System under these conditions and the actions that should be taken prior to commencing painting?

A. The paint fumes increase the load on the SBGT fan. The CBO for each Unit should log that SBGT cannot be started for 4 hours after painting is complete.

B. The paint fumes will contaminate the SBGT Charcoal filters if the system is started while painting is in progress. Both SBGT switches should be placed in OFF until painting is complete.

C. The paint fumes will contaminate the SBGT Charcoal filters if the system is started within 4 hours of completing the painting. Tags should be hung on the SBGT Fan control switches stating "SOLVENT BASED PAINTING IN PROGRESS".

D. The paint fumes could start a fire in the SBGT Train due to the heater being on when the system initiates. Both SBGT switches should be placed in OFF until painting is complete.

86


Fuel movement is on progress on Unit 2 with the following plant conditions:

Mode Switch Position

Run Refuel

Coolant Temperature

545OF

128 0F

Which ONE of the following is the minimum on-site shift staffing required by the Unit 2 Technical Specifications? (Provide Tech Spec section 5.2.2)

I + 1 for Fuel Handling 2 2 1

1 + 1 for Fuel Handling 2 3 1

2 + 1 for Fuel Handling 2 3 0

2+ 3 3 0

I for Fuel Handling


Unit I Unit 2

Reactor Power

80% 0%

A. SRO RO PEO STA

B` SRO RO PEO STA

C. SRO RO PEO STA

D. SRO RO PEO STA

87

Final SRO Test

References: Tech Spec Section 5.2.2 99 exam Question #6 LT-ST-30003-05, p.7 & 8 1 OCFR50.54(m)(2)(i) Modified answer A as follows: PEO from 3 to 2.

A. Incorrect since 3 PEO's are required at all times.

B. Correct answer.

C. Incorrect since Unit 2 does not need an SRO since it is in Mode 5.

D. Incorrect since only 2 RO's are required (one for each unit that has fuel).

88. G2.2.1 001

Unit 2 is in a startup and is currently at approximately 20% RTP. A heat balance has been performed and indicates that Core Thermal Power is 23% RTP. The APRM readings should be adjusted such that

A. each APRM is reading between -2% to +2% of the heat balance value.

B.3 each APRM is reading between 0% to +2% of the heat balance value.

C. each APRM is reading between -2% to 0% of the heat balance value.

D. each APRM is reading between -1% and +1% of the heat balance value.

References: 34GO-OPS-001-2S Rev. 35.1 pg 33 of 61.

Unit 2 Tech Specs Section 3.3.1.1 (SR 3.3.1.1.2)

A.1 Incorrect since the value must be between 0 and 2%.

B. Correct answer.

C. Incorrect since the value must be between 0 and 2%.

D. Incorrect since the value must be between 0 and 2%.



References: 42FH-ERP-014-OS Rev. 15.2, pg 10 of 28 34FH-OPS-001-OS Rev. 21.1, pg 6 and 7 of 42

A. Correct answer.

B. Incorrect since ALL fuel movements must be stopped when an error is found.

C. Incorrect since the SRO cannot approve a movement without proper authorization.

D. Incorrect since the Fuel Movement Change sheet must be approved prior to any further fuel movements.


Unit I is in Mode 5 with a core shuffle in progress. The bridge operator has just inserted a fuel bundle into the core when he notices that the adjacent fuel bundle is mis-oriented.

Which ONE of the following actions are required to be performed by the fuel handling crew with regards to the fuel shuffle?

AX The bridge operator stops fuel movements and informs SRO of condition. The SRO contacts Reactor Engineering to prepare a Fuel Movement Sheet change. The crew reviews the approved change sheet, corrects the orientation error and continues with fuel movements with SRO approval.

B. The SRO allows the crew to continue with fuel movements after correcting the orientation error and notifies Reactor Engineering for documentation on the Core Loading Verification sheet.

C. The SRO stops fuel movements and the crew determines the proper orientation of the adjacent fuel bundle. The SRO approves the actions to re-orient the fuel bundle and the bridge operator notifies the control room when move is complete.

D. The SRO allows fuel movements to continue and contacts Reactor Engineering to prepare a Fuel Movement Sheet change. At the next appropriate opportunity the crew will correct the orientation error per the Movement Sheet as long as it is done on their shift.

89


References: 34AB-G41-002-2S Rev. 2 pg 2 of 5. 1999 Hatch Exam Question 21 Modified answers slightly to make different answer correct.

A. Incorrect since the direction should be to move the fuel bundle to its proper in-core location.

B. Incorrect since direction should be to lower the bundle and you only have to evacuate the refueling floor if there are radiation alarms.

C. Incorrect since direction should be to place the fuel bundle in any fuel pool rack.

D. Correct answer.


Unit 2 is in a refueling outage with a fuel shuffle in progress. A fuel bundle is being transfered from the core to the fuel pool when the Control Board Operator reports that reactor cavity water level is decreasing.

Per 34AB-G41-002-2S, Decreasing Rx Well/Fuel Pool Water Level, which ONE of the following actions should the refueling SRO direct the bridge operator to perform?

A. Return the fuel bundle to the closest in-core location as possible.

B. Stop all movement and evacuate the refueling floor immediately.

C. Continue with movement to the fuel pool and lower it as deep into the pool as possible.

D. Return the fuel bundle to its proper in-core location.

90


Reference: LT-LP-30004-04, Pg. 15-17 99 exam question #19 EO 300.002.a.02

A. Incorrect since this process is used for editorial changes. editorial.

This change is not

B. Correct answer.

C. Incorrect since this process is used for editorial changes. This change is not editorial.

D. Incorrect since this is the normal process that is used if the procedure is not needed now. Since this procedure change is needed prior to the end of the shift then an SRO change is appropriate.


While reviewing a procedure that is required to be completed before the end of the current shift, the SS notices a step that requires the use of a gauge which is broken. Another gauge is available in the system and the SS has confirmed it will operationally function as a substitute.

At a minimum, which ONE of the following actions must be done to perform the

procedure? The SS should:

A. Make a pen and ink change to the procedure.

B" Make a SRO change to the procedure.

C. Make a pen and ink change to the procedure with SOS concurrence.

D. Make a permanent change to the procedure obtaining manager approval prior to use.

91

Final SRO Test 92. G2.3.1 001

Which ONE of the following is the Plant Hatch initial administrative dose limit to the skin or any extremity (SDE)?

A. 2 rem/year

B. 6 rem/year

C. 15 rem/year

DV 20 rem/year

References: LT-LP-30008-02 Rev.2 pg 7 and 8 of 28 SRO exam 95-01 test question #92. Modified the question from LDE limit to SDE limit and changed appropriate answer.

A. Incorrect since this is the annual admin limit for TEDE.

B. Incorrect since this is the annual LDE limit.

C. Incorrect since this is the 10 CFR annual LDE limit.

D. Correct answer.

93. G2.3.2 001

Health Physics technicians have surveyed the main steam chase during an outage and obtained the following results:

Area Dose Rates one foot from the source: 75 mR/hr Airborne Concentration: .23 DAC Smear Results: 650 dpm/100cm2

Bases on these results which ONE of the following states how the area should be posted?

A. Contaminated Area.

B. Airborne Radioactivity Area.

C.€ Radiation Area.

D. High Radiation Area.


Final SRO Test

References: LT-LP-30008 Rev. 2 pg 13-17 of 28 SRO exam 95-01 question #93. Changed numbers in stem but with same results and reordered answers.

A. Incorrect since < 1OOOdpm/100cm2.

B. Incorrect since < .3 DAC.

C. Correct answer.

D. Incorrect since < 100 mR/hr.

94. G2.3.3 001


Unit 1 is at 75% RTP. At 1400 on 8/12/02, after performing scram time testing, the Control Board Operator notes that the Offgas Flow has increased from the steady state level as follows:

Offgas Inlet Flow to Stack prior to scram time testing 100 scfm Offgas Inlet Flow to Stack after scram time testing 175 scfm

Which ONE of the following actions is/are required by Tech Specs for this condition? (Provide copy of TS Section 3.7.6 along with SR's)

A. Notify Chemistry to sample the offgas system by 1900 to verify gross gamma activity is < 240 mCi/second. If greater than 240 mCi/second then isolate SJAE within 12 hours OR BE in Mode 3 within 12 hours.

B1 Notify Chemistry to sample the offgas system by 1800 to verify gross gamma activity is < 240 mCi/second. If greater than 240 mCi/second then enter 72 hour LCO to restore within limits.

C. Notify Chemistry to sample the offgas system by 0200 on 8/13/02 to verify gross gamma activity is < 240 mCi/second. If greater than 240 mCi/second then enter 72 hour LCO to restore within limits.

D. Notify Chemistry to sample the offgas system by 0400 on 8/13/02 to verify gross gamma activity is < 240 mCi/second. If greater than 240 mCi/second then enter 24 hour LCO to restore within limits.

93

Final SRO Test

References: Tech Spec section 3.7.6 (SR 3.7.6.1) LT-LP-03101 Rev. 3 pg 29 of 44

A. Incorrect since the sample time does not allow for 25% grace period and if exceed the LCO limit then have 72 hours to restore.

B. Correct answer.

C. Incorrect since the sample time is based on 8 hours instead of the required 4 hours.

D. Incorrect since the sample time is based on 8 hours plus 25% instead of the required 4 hours. Also, if exceed the LCO limit then have 72 hours to restore.

95. G2.3.4 001

References: 73EP-EIP-017-OS Rev 2.1 pg 6 of 13. SRO exam 95-01 question # 94.

B. Correct answer.

A,C and D. Incorrect. See reference above on page 6.


The Emergency Director decides that is is necessary to send someone into the Reactor Building (with Health Physics) to isolate a leak before the Core Spray and RHR pumps are flooded. (No releases are underway and RPV level is being maintained at 60 inches with the Condensate System)

Which ONE of the following is the maximum allowable dose limit that the Emergency Director may authorize?

A. 5 REM

B1 10 REM

C. 25 REM

D. > 25 REM

94

Final SRO Test96. 02.4.12 001

References: PC-1 Primary Containment Control Rev. 4 30AC-OPS-013-OS Rev. 9.1 pg 5 of 10.

A. Incorrect since Shift Supervisor must re-enter all legs of the flow chart.

B. Incorrect since Shift Supervisor must re-enter all legs of the flow chart.

C. Incorrect since Control Board Operator should immediately announce an entry condition into the EOP's.

D. Correct answer.


Unit 1 was operating at 50% RTP when a steam leak in the Drywell caused a Reactor Scram on High Drywell Pressure. The Shift Supervisor entered all appropriate EOP's for this condition and ordered Torus Sprays to be initiated per PC/P leg. Subsequently, Drywell Temperature increased to 160 0 F.

Which ONE of the following states the appropriate action for the crew to take in this situation?

A. The Control Board Operator announces the new Entry Condition for PC-1 Primary Containment Control. The Shift Supervisor circles this entry condition and continues on in the PC-1 Primary Containment Control flow chart at the current steps.

B. The Control Board Operator announces the new Entry Condition for PC-1 Primary Containment Control. The Shift Supervisor circles the entry condition and orders actions for Drywell Temperature leg only.

C. The Control Board Operator continues to place Torus Sprays in service and updates the crew at the next Brief that Drywell Temperature has increased >1 500 F.

D'. The Control Board Operator announces the new Entry Condition for PC-1 Primary Containment Control. The Shift Supervisor circles the entry condition and re-enters all legs of PC-1 Primary Containment Control.

95


References: RC RPV CONTROL (NON ATWS) Rev. 6 PC-1 PRIMARY CONTAINMENT CONTROL Rev. 4

A. Incorrect since this is improper sequence for getting to the point of placing the Mode Switch in Shutdown.

B. Incorrect since you enter RC RPV CONTROL "before" drywell temperature reaches 2800 F. 340OF is associated with Unit 2.

C. Incorrect since RC RPV CONTROL is entered from PC-1 PRIMARY CONTAINMENT CONTROL.

D. Correct answer.


Unit 1 is operating at 100% RTP. Drywell temperature has been increasing steadily

and is currently at 2750 F and increasing. All other parameters are within their normal operating band.

Which ONE of the following describes the proper directions and sequence of events to the point of placing the Mode Switch in Shutdown? (Provide RO copy of RC RPV CONTROL (NON ATWS) and PC-1 PRIMARY CONTAINMENT CONTROL)

A. Enter RC RPV CONTROL (NON ATWS) at Point A due to Drywell High Temperature. Place the Mode Switch in Shutdown per RC-1 then enter PC-1 to control Drywell temperature.

B. Enter PC-1 when Drywell Temperature exceeds 1500 F. Enter RC RPV CONTROL

at point A before Drywell Temperature reaches 3400F. Perform RC-1 to manually scram the reactor and place the Mode Switch in Shutdown.

C. Enter RC RPV CONTROL (NON ATWS) due to a condition which requires a reactor scram and reactor power is above 5%. Perform RC-1 to manually scram the reactor and place the Mode Switch in Shutdown.

Dt. Enter PC-1 when Drywell Temperature exceeds 1500 F. Enter RC RPV CONTROL

at point A before Drywell Temperature reaches 2800 F. Perform RC-1 to manually scram the reactor and place the Mode Switch in Shutdown.

96


References: LT-LP-10004 Rev. 03 pg 15 of 19. LO LT-10004.008

A. Incorrect since hand held radios are not used in the Control Room.

B. Correct answer.

C. Incorrect answer since the base channel is VHF and uses Channel 2.

D. Incorrect since hand held radios are not used in the Control Room.


"ý-ýee has been reported by the Unit I EHC skid and the Fire Brigade has been dispat ..

Which ONE of the wing describes how the Shift Supervisor maintains constant communications with the Brigade from the Main Control Room?

A. A hand held radio dedicated to U annel 1., a c /

B. VHF base station dedicated to VHF Channel 2.

C. UHF base station dedicated to VHF Channel 1.

0. A hand held radio dedicated to VHF Channel 2.

97


The following conditions exist for Unit 2 at the indicated times:

0200 0300 0400 Reactor Power 100% 85% 60% Reactor Water Level +37" +37" +35"

Drywell Pressure 1.3 psig 1.5 psig 1.7 psig Torus Pressure .9 psig 1.1 psig 1.2 psig Torus Level 150" 137" 123" Torus Temperature 850 F 85 0 F 860F

Which ONE of the following indicates the status of the plant when the operator turns

over to the next shift at 0700 if current trends continue as they are?

A. Reactor in Mode 3 with Drywell Sprays in progress.

B. Reactor in Mode 3 with a cooldown in progress at <100OF/hr.

C. HPCI operation is prevented and Emergency Depressurization has been initiated.

D. Recirc pumps and drywell cooling fans are tripped. Torus cooling and sprays are initiated.

References: PC-1 Primary Containment Control Rev. 4 Att. I

A. Incorrect since drywell sprays are not initiated until you cannot stay below 11 psig

torus pressure or 340OF Drywell temperature. Current trends do not come close to these limits.

B. Incorrect since an Emergency Depressurization is required prior to 0700 due to torus

level <98 inches. This will exceed 100OF cooldown rate.

C. Correct answer due to lowering torus level. At 110 inches then HPCI is prevented from operation and at 98 inches the reactor is depressurized.

D. Incorrect since drywell coolers and recirc pumps are not directed to be tripped unless conditions noted in A above are met.



Unit 1 is in an ATWS condition with Reactor Power oscillating between 15 and 45% RTP. The following indications exist at this time:

SBLC Pump Select Switch in Start Sys A position. SBLC Squib VIv Ready lights are LIT. Rx Water Cleanup VIv, 2G31-F004,Rx Wtr Cleanup Suction VIv, is CLOSED. SBLC Discharge Pressure is greater than reactor pressure.

Which ONE of the following describes the appropriate actions the Shift Supervisor

should order?

A. Inhibit ADS and bypass RWCU filter/demineralizers per 34SO-G31-003-1S.

B. Continue to monitor SBLC and secure when the Cold Shutdown Boron Weight has been added.

C. Inhibit ADS, continue to monitor SBLC, and exit RC/Q when the reactor is

subcritical.

D. Initiate SBLC per 34SO-C41-003-1S using the manual-local initiation method.

References: RCA RPV CONTROL (ATWS) Rev. 6 LR-20328 Rev. 6 pg 44-45 of 58.

A. Incorrect since 2G31-F004 is closed.

B. Incorrect since boron is not injecting.

C. Incorrect since boron is not being injected and RC/A exit requires subcritical with no boron injection.

D. Correct answer.

You have completed the test! PM 99Thursday, November 21, 2002 01:37:59

NRC EXAM 2002 PLANT HATCH

WRITTEN EXAM REFERENCE

BOOK

MASTER SRO

SRO References - NRC Exam 2002

Unit 1 Tech Specs Section 3.3.6.1 and Table 3.3.6.1-1

Unit 2 Tech Specs Section 3.3.6.1 and Table 3.3.6.1-1

Unit 2 Tech Specs Section 3.4.1


Unit 1 Tech Specs Section 3.6.1.3

Unit 1 Tech Specs Section 3.3.1.2

Unit 1 TRM Section 3.3.2 and Tables.

Unit 2 Tech Spec Section 3.3.5.1

Unit 2 Tech Spec Section 3.5.3

Unit 1 Tech Spec Section 3.5.1

Unit 1 EOP Graph 9 (CS VORTEX)

Unit 1 EOP Graph llA (CS NPSH)

Unit 1 EOP Graph 11B(CS NPSH).


Unit I Tech Specs Section 3.3.1.1

Unit 2 EOP Graph 5 (BIIT)

Unit 1 Secondary Containment Control (SCC) EOP Flowchart

73EP-E1P-001-OS, "Emergency Classification Initial Actions"

Unit 2 Secondary Containment Control (SCC) EOP Flowchart



Primary Containment Isolation Instrumentation 3.3.6.1

3.3 INSTRUMENTATION

3.3.6.1 Primary Containment Isolation Instrumentation

LCO 3.3.6.1

APPLICABILITY:

The primary containment isolation instrumentation for each Function in Table 3.3.6.1-1 shall be OPERABLE.

According to Table 3.3.6.1-1.

ACTIONS

--.-.-. ------------------------------------.-. ---.-. ---.. NOTE --..........................................................Separate Condition entry is allowed for each channel.

----------------------------- ----- ---------------------------------------------------

CONDITION REQUIRED ACTION COMPLETION TIME

A. One or more required A.1 Place channel in trip. 12 hours for channels inoperable. Functions 2.a, 2.b,

and 6.b

AND

24 hours for Functions other than Functions 2.a, 2.b, and 6.b

B. - -------------- NOTE ------------- B.1 Restore isolation 1 hour Not applicable for capability. Function 5.c.

One or more automatic Functions with isolation capability not maintained.

C. Required Action and C.1 Enter the Condition Immediately associated Completion referenced in Time of Condition A or B Table 3.3.6.1-1 for the not met. channel.

(continued)

Amendment No. 195HATCH UNIT 1 3.3-47


AflTIfNNS (cnntiniied•

Amendment No. 195HATCH UNIT 1


D. As required by Required D.1 Isolate associated main 12 hours Action C.1 and referenced steam line (MSL). in Table 3.3.6.1-1.

OR

D.2.1 Be in MODE 3. 12 hours

AND


E. As required by Required E.1 Be in MODE 2. 6 hours Action C.1 and referenced in Table 3.3.6.1-1.

F. As required by Required F.1 Isolate the affected 1 hour Action C.1 and referenced penetration flow in Table 3.3.6.1-1. path(s).

G. As required by Required G.A Be in MODE 3. 12 hours Action C.1 and referenced in Table 3.3.6.1-1. AND

OR G.2 Be in MODE 4. 36 hours

Required Action and associated Completion Time of Condition F not met.

H. As required by Required H.1 Declare Standby Liquid 1 hour Action C.1 and referenced Control (SLC) System in Table 3.3.6.1-1. inoperable.

OOR

H.2 Isolate the Reactor 1 hour Water Cleanup (RWCU) System.

(continued)

3.3-48


ACTIONS (continued)


As required by Required 1.1 Initiate action to restore Immediately Action C.1 and referenced channel to OPERABLE in Table 3.3.6.1-1. status.

OR

1.2 Initiate action to isolate Immediately the Residual Heat Removal (RHR) Shutdown Cooling System.

SURVEILLANCE REQUIREMENTS

.. ...............................--------- --------------- NOTES ----------------------------------------------------------a 1. Refer to Table 3.3.6.1-1 to determine which SRs apply for each Primary Containment Isolation Function.

2. When a channel is placed in an inoperable status solely for 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.

- ----------------------------------------------------------------------------------------

SURVEILLANCE FREQUENCY

SR 3.3.6.1.1 Perform CHANNEL CHECK. 12 hours

SR 3.3.6.1.2 Perform CHANNEL FUNCTIONAL TEST. 92 days

SR 3.3.6.1.3 Perform CHANNEL CALIBRATION. 92 days


(continued)

Amendment No. 2323.3-49HATCH UNIT 1


SURVEILLANCE REQUIREMENTS (continued)


SR 3.3.6.1.5 Perform CHANNEL CALIBRATION. 24 months

SR 3.3.6.1.6 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months



Table 3.3.6.1-1 (page 1 of 4) Primary Containment Isolation Instrumentation

APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED

OTHER CHANNELS FROM. SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE

FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE

1. Main Steam Line Isolation

a. Reactor Vessel Water Level Low Low Low, Level 1

b. Main Steam Une Pressure Low

c. Main Steam Line Flow - High

d. Condenser Vacuum - Low

e. Main Steam Tunnel Temperature - High

f. Turbine Building Area Temperature - High

2. Primary Containment Isolation

a. Reactor Vessel Water Level - Low, Level 3

b. Drywall Pressure - High

1,2,3

1

1,2,3

1, 2(a), 3(a)

1,2,3

1,2,3

1,2,3

1,2,3

2

2

2 per MSL

2

6

16(b)

2

2

D SR SR SR SR

E

D

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

SR 3.3.6.1.3 SR 3.3.6.1.6

SR SR SR SR

D

D

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

SR 3.3.6.1.3 SR 3.3.6.1.6

SR SR SR SR

SR SR SR

D

G SR SR SR SR

SR SR SR SR

G

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

-113 inches

a 825 psig

< 138% rated steam flow

>7 inches Hg vacuum

S 194°F

s 200OF

2 0 inches

S 1.92 psig

(continued)

(a) With any turbine stop valve not closed.

(b) With 8 channels per trip string. Each trip string shall have 2 channels per main steam line, with no more than 40 ft separating any two OPERABLE channels.





OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE


2. Primary Containment Isolation (continued)

c. Drywell Radiation High

d. Reactor Building Exhaust Radiation - High

e. Refueling Floor Exhaust Radiation - High

3. High Pressure Coolant Injection (HPCI) System Isolation

a. HPCI Steam Line Flow High

b. HPCI Steam Supply Line Pressure - Low

c. HPCI Turbine Exhaust Diaphragm Pressure - High

d. Drywell Pressure - High

e. HPCI Pipe Penetration Roon Temperature - High

I. Suppression Pool Area Ambient Temperature - High

F1,2,3

1,2,3

1,2,3

SR SR SR SR

SR SR SR

SR SR SR

G

G

F

F

1,2,3

1,2,3

1,2,3

1,2,3

1,2,3

-1,2,3

F

SR SR SR SR

SR SR SR SR

SR SR SR SR

SR SR SR SR

SR SR SR SR

SR SR SR SR

F

F

F

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.3 3.3.6.1.6

3.3.6.1.1 3.3.6.1.3 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

< 138 RPhr

• 80 mR/hr

e 80 mR/hr

< 303% rated steam flow

> 100 psig

S•20 psig

< 1.92 psig

< 169OF

< 169oF

(continued)

Amendment No. 216

I

3.3-52HATCH UNIT 1



APPLICABLE MODES OR

OTHER SPECIFIED

CONDITIONS

REQUIRED CHANNELS PERTRIP SYSTEM

CONDITIONS REFERENCED

FROM REQUIRED ACTION C.1


3. HPCI System Isolation (continued)

g. Suppression Pool Area Temperature - Time Delay Relays

h. Suppression Pool Area Differential Temperature High

i. Emergency Area Cooler Temperature - High

4. Reactor Core Isolation Cooling (RCIC) System Isolation

a. RCIC Steam Line Flow High

b. RCIC Steam Supply Line Pressure - Low

c. RCIC Turbine Exhaust Diaphragm Pressure High


e. RCIC Suppression Pool Ambient Area Temperature - High

f. Suppression Pool Area Temperature - Time Delay Relays

SR 3.3.6.1.4 SR 3.3.6.1.6

1,2,3

1,2,3

1,2,3

SR SR SR SR

SR SR SR SR

F

F

1,2,3

1,2,3

1,2,3

1,2,3

1,2,3

1,2,3

F

SR SR SR SR

SR SR SR SR

SR SR SR SR

SR SR SR SR

SR SR SR SR

F

F

F1

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

SR 3.3.6.1.4 SR 3.3.6.1.6

< 16 minutes 15 seconds

< 420F

< 169OF

S 306% rated steam flow

> 60 psig

S 20 psig

s 1.92 psig

!5 169°F

< 31 minutes 15 seconds

(continued)

Amendment No. 232HATCH UNIT 1

FUNCTIONALLOWABLE

VALUE

3.3-53




OTHER -CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE


4. RCIC System Isolation (continued)

g. RCIC Suppression Pool Area Differential Temperature - High

h. Emergency Area Cooler Temperature - High

5. RWCU System Isolation

a. Area Temperature - High

b. Area Ventilation Differential Temperature High

c. SLC System Initiation

d. Reactor Vessel Water Level - Low Low, Level 2

6. RHR Shutdown Cooling System Isolation

a. Reactor Steam Dome Pressure - High

b. Reactor Vessel Water Level - Low, Level 3

1,2,3

1,2,3

1,2,3

1,2,3

1,2

1,2,3

1,2,3

3,4,5

1

1

1 per area

1 per area

1(c)

2

1

2 (d)

F

F

SR SR SR SR

SR SR SR SR

SR SR SR SR

SR SR SR SR

F

F

H

F

F

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

SR 3.3.6.1.6

SR SR SR SR

SR SR SR SR

SR SR SR SR

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

s42"F

s169F

s 1500F

< 67°F

NA

Ž-47 inches

S145psig

2 0 inches

Amendment No."195

(c) SLC System Initiation only inputs into one of the two trip systems.

(d) Only one trip system required in MODES 4 and 5 when RHR Shutdown Cooling System integrity maintained.

HATCH UNIT 1 3.3-54


3.3 INSTRUMENTATION

3.3.6.1 Primary Containment Isolation Instrumentation

LCO 3.3.6.1

APPLICABILITY:

The primary containment isolation instrumentation for each Function in Table 3.3.6.1-1 shall be OPERABLE.


ACTIONS

-----.............. ............. ............. ............. N O T E --..........................................................Separate Condition entry is allowed for each channel. ----------------------------------------------------------------------------------


A. One or more required A.1 Place channel in trip. 12 hours for channels inoperable. Functions 2.a, 2.b,

and 6.b

AND

24 hours for Functions other than Functions 2.a, 2.b, and 6.b

B. --------- NOTE-------------- B.1 Restore isolation 1 hour Not applicable for capability. Function 5.c.

One or more automatic Functions with isolation capability not maintained.

C. Required Action and C.1 Enter the Condition Immediately associated Completion referenced in Time of Condition A or B Table 3.3.6.1-1 for the not met. channel.

(continued)



ACTIONS (continued)


D. As required by Required D.1 Isolate associated main 12 hours Action C.1 and referenced steam line (MSL). in Table 3.3.6.1-1.

OR


AND


E. As required by Required E.1 Be in MODE 2. 6 hours Action C.1 and referenced in Table 3.3.6.1-1.

F. As required by Required F.1 Isolate the affected 1 hour Action C.1 and referenced penetration flow in Table 3.3.6.1-1. path(s).

G. As required by Required G.1 Be in MODE 3. 12 hours Action C.1 and referenced in Table 3.3.6.1-1. AND

OR G.2 Be in MODE 4. 36 hours

Required Action and associated Completion Time of Condition F not met.

H. As required by Required H.1 Declare Standby Uquid 1 hour Action C.1 and referenced Control (SLC) System in Table 3.3.6.1-1. inoperable.

OR

H.2 Isolate the Reactor 1 hour Water Cleanup (RWCU) System.

(continued)



ACTIONS (continued)


1. As required by Required 1.1 Initiate action to restore Immediately Action C.1 and referenced channel to OPERABLE in Table 3.3.6.1-1. status.

OR

1.2 Initiate action to isolate Immediately the Residual Heat Removal (RHR) Shutdown Cooling System.


*---------------- --------------- NOTES --..........................-----------------------------1. Refer to Table 3.3.6.1-1 to determine which SRs apply for each Primary Containment

Isolation Function.

2. When a channel is placed in an inoperable status solely for 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.

-----------------------------------------------------------------------------------






(continued)







SR 3.3.6.1.7 -------- NOTE----------------Channel sensors are excluded.

Verify the ISOLATION SYSTEM RESPONSE 24 months on a TIME is within limits. STAGGERED

TEST BASIS







1. Main Steam Line Isolation

a. Reactor Vessel Water Level - Low Low Low, Level 1

b. Main Steam Line Pressure - Low

C. Main Steam Line Flow - High

d. Condenser Vacuum - Low

e. Main Steam Tunnel Temperature - High

f. Turbine Building Area Temperature - High

2. Primary Containment Isolation

a. Reactor Vessel Water Level - Low, Level 3

b. Drywell Pressure High

1,2,3 2

2

2 per MSL

2

6

16(b)

2

2

SR SR SR SR SR

D

E

D

D

D

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6 3.3.6.1.7

SR 3.3.6.1.3 SR 3.3.6.1.6

SR SR SR SR SR

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6 3.3.6.1.7

SR 3.3.6.1.3 SR 3.3.6.1.6

SR SR SR SR

SR SR SR

D

G SR SR SR SR

SR SR SR SR

G

3..6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

S-113 inches

a 825 psig

s 138% rated steam flow

S7 inches Hg vacuum

* 1940F

* 1940F

* 0 inches

< 1.92 psig

(continued)

line, with no more than 40 ft

1

1,2,3

1, 2(a), 3(a)

1,2,3

1,2,3

1,2,3

1,2,3

Amendment No. 135

(a) With any turbine stop valve not closed.

(b) With 8 channels per trip string. Each trip string shall have 2 channels per main steam separating any two OPERABLE channels.

HATCH UNIT 2 3.3-51




OTHER CHANNELS FROMSPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE

FUNCTION CONDITIONS SYSTEM, ACTION C.1 REQUIREMENTS VALUE

2. Primary Containment Isolation (continued)

c. Drywall Radiation - High

d. Reactor Building Exhaust Radiation - High

a. Refueling Floor Exhaust Radiation - High

3. High Pressure Coolant Injection (HPCI) System Isolation

a. HPCI Steam Une FlowHigh

b. HPCI Steam Supply Une Pressure - Low

c. HPCI Turbine Exhaust Diaphragm Pressure - High


e. HPCI Pipe Penetration Room Temperature - High

f. Suppression Pool Area Ambient Temperature - High

1,2,3

1,2,3

1,2,3

1,2,3

1,2,3

1,2,3

1,2,3

1,2,3

1,2,3

1

2

2

F SR SR SR SR

SR SR SR

SR SR SR

G

G

F SR SR SR SR

SR SR SR SR

SR SR SR SR

SR SR SR SR

SR SR SR SR

SR SR SR SR

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.3 3.3.6.1.6

3.3.6.1.1 3.3.6.1.3 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.36.1.6

3.&.6.1.1 3.3.6.1.2 3.3.6.1.5 &36.1.6

336.1.1 3.a6.1.2 3.3.6.1.5 3.3.6.1.6

&&6.i.1 3.36.1.2 3.36.1.5 &&6.1.6

336.1.1 3.36.1.2 3.36.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

s 138 Rhr

< 80 mR/hr

< 80 mR/hr

s 303% rated steam flow

z 100 psig

r 20 psig

s 1.92 psig

s 1690F

s 169°F

1

2

2

I

1

1

F

F

F

F

F

(continued)







3. HPCI System Isolation (continued)

g. Suppression Pool Area Temperature - Time Delay Relays

h. Suppression Pool Area Differential Temperature High

i. Emergency Area Cooler Temperature - High

4. Reactor Core Isolation Cooling (RCIC) System Isolation

a. RCIC Steam Line FlowHigh

b. RCIC Steam Supply Line Pressure - Low

c. RCIC Turbine Exhaust Diaphragm Pressure High

d. Drywall Pressure - High

e. RCIC Suppression Pool Ambient Area Temperature - High

f. Suppression Pool Area Temperature - Time Delay Relays

1,2,3

1,2,3

1,2,3

1,2,3

1,2,3

1,2,3

1,2,3

1,2,3

1,2,3 1

F

F

SR 3.3.6.1.4 SR 3.3.6.1.6

SR SR SR SR

SR SR SR SR

F

F

F

F

SR SR SR SR

SR SR SR SR

SR SR SR SR

SR SR SR SR

SR SR SR SR

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

SR 3.3.6.1.4 SR 3.3.6.1.6

• 16 minutes 15 seconds

< 420F

< 169°F

:s 307 % rated steam flow

G60 psig

• 20 psig

S1.92 psig

< 1690F

f31 minutes 15 seconds

(continued)

Amendment No. t74HATCH UNIT 2 3.3-53






4. RCIC System Isolation (continued)

g. RCIC Suppression Pool Area Differential Temperature - High

h. Emergency Area Cooler Temperature - High

5. RWCU System Isolation

a. Area Temperature - High

b. Area Ventilation Differential Temperature High

c. SLC System Initiation

d. Reactor Vessel Water Level - Low Low, Level 2

6. RHR Shutdown Cooling System Isolation

a. Reactor Steam Dome Pressure - High

b. Reactor Vessel Water Level - Low, Level 3

1,2,3

1,2,3

1,2,3

1,2,3

1,2

1,2,3

1,2,3

3,4,5

1

I per area

1 per area

1(c)

2

1

2(d)

F

F

SR SR SR SR

SR SR SR SR

SR SR SR SR

SR SR SR SR

F

F

H

F

F

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

SR 3.3.6.1.6

SR SR SR SR

SR SR SR SR

SR SR SR SR

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

3.3.6.1.1 3.3.6.1.2 3.3.6.1.5 3.3.6.1.6

• 420F

:; 1690F

s 1500F

s 671F

NA

2:- 47 inches

s 145 psig

> 0 inches

Amendment No. 135

(c) SLC System Initiation only inputs into one of the two trip systems.

(d) Only one trip system required in MODES 4 and 5 when RHR Shutdown Cooling System integrity maintained.

3.3-54HATCH UNIT 2

Recirculation Loops Operating 3.4.1

3.4 REACTOR COOLANT SYSTEM (RCS)

3.4.1 Recirculation Loops Operating

Two recirculation loops with matched flows shall be in operation,

OR

One recirculation loop shall be in operation with the following limits applied when the associated LCO is applicable:

a. LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits specified in the COLR;

b. LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," single loop operation limits specified in the COLR; and

c. LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation," Function 2.b (Average Power Range Monitor Simulated Thermal Power - High), Allowable Value of Table 3.3.1.1-1 is reset for single loop operation.

APPLICABILITY: MODES 1 and 2.

Amendment No. 154

LCO 3.4.1

I

I I

HATCH UNIT 2 3.4-1

Recirculation Loops Operating 3.4.1

ACTIONS


A. Requirements of the A.1 Satisfy the 24 hours LCO not met. requirements of the

LCO.

B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not met.

OR

No recirculation loops in operation.



SR 3.4.1.1 .NOTE ------------------Not required to be performed until 24 hours after both recirculation loops are in operation.

Verify recirculation loop jet pump flow mismatch 24 hours with both recirculation loops in operation is:

a. ! 10% of rated core flow when operating at < 70% of rated core flow; and

b. S 5% of rated core flow when operating at • 70% of rated core flow.

SR 3.4.1.2 (Not used.)

Amendment No. 154

I

HATCH UNIT 2 3.4-2

RHR Shutdown Cooling System - Hot Shutdown 3.4.7

3.4 REACTOR COOLANT SYSTEM (RCS)

3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown

Two RHR shutdown cooling subsystems shall be OPERABLE and, with no recirculation pump in operation, at least one RHR shutdown cooling subsystem shall be in operation,

---------------- ---------.------.. .---- NOTES - ----------------- -............ 1. Both RHR shutdown cooling subsystems and recirculation pumps

may be removed from operation for up to 2 hours per 8 hour period.

2. One RHR shutdown cooling subsystem may be inoperable for up to 2 hours for performance of Surveillances.

APPLICABILITY: MODE 3 with reactor steam dome pressure less than the RHR low pressure permissive pressure.

ACTIONS

1. LCO 3.0.4 is not applicable.-NOTES ----...-------------...........--------------------------

2. Separate Condition entry is allowed for each RHR shutdown cooling subsystem.


A. One or two RHR shutdown A.1 Initiate action to restore Immediately cooling subsystems RHR shutdown cooling inoperable, subsystem(s) to

OPERABLE status.

AND

(continued)

Amendment No..135

LCO 3.4.7

HATCH UNIT 2 3.4-114


O ACTIONSCONDITION

A. (continued)

B. No RHR shutdown cooling subsystem in operation.

AND

No recirculation pump in operation.

REQUIRED ACTION COMPLETION TIMEI ___________________________________________

A.2

AND

A.3

B.1

Verify an alternate method of decay heat removal is available for each inoperable RHR shutdown cooling subsystem.

Be in MODE 4.

Initiate action to restore one RHR shutdown cooling subsystem or one recirculation pump to operation.

AND

B.2 Verify reactor coolant circulation by an alternate method.

AND

B.3 Monitor reactor coolant temperature and pressure.

1 hour

24 hours

Immediately

1 hour from discovery of no reactor coolant circulation

AND

Once per 12 hours thereafter

Once per hour

I ____________________________________________________________________ _________________________________________________

Amendment No. .1353.4-15HATCH UNIT 2




SR 3.4.7.1 ----------------------------- NOTE ------...........--------Not required to be met until 2 hours after reactor steam dome pressure is less than the RHR low pressure permissive pressure. ,

Verify one RHR shutdown cooling subsystem or 12 hours recirculation pump is operating.


PCIVs 3.6.1.3

3.6 CONTAINMENT SYSTEMS

3.6.1.3 Primary Containment Isolation Valves (PCIVs)

LCO 3.6.1.3

APPLICABILITY:

Each PCIV, except reactor building-to-suppression chamber vacuum breakers, shall be OPERABLE.

MODES 1, 2, and 3, When associated instrumentation is required to be OPERABLE per

LCO 3.3.6.1, "Primary Containment Isolation Instrumentation."

ACTIONS

------------------- -- NOTES --.-.----------------------------.........-----------1. Penetration flow paths except for 18 inch purge valve penetration flow paths may be

unisolated intermittently under administrative controls.

2. Separate Condition entry is allowed for each penetration flow path.

3. Enter applicable Conditions and Required Actions for systems made inoperable by PCIVs.

4. Enter applicable Conditions and Required Actions of LCO 3.6.1.1, "Primary Containment," when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria.


A. -------- NOTE --------------- A.1 Isolate the affected 4 hours except for Only applicable to penetration flow path by main steam line penetration flow paths with use of at least one two PCIVs. closed and de-activated AND

--------- automatic valve, closed manual valve, blind 8 hours for main

One or more penetration flange, or check valve steam line flow paths with one PCIV with flow through the inoperable except due to valve secured. leakage not within limit.

AND

(continued)


PCIVs 3.6.1.3

ACTIONS


A. (continued) A.2 ---------- NOTE -----------Isolation devices in high radiation areas may be verified by use of administrative means.

Verify the affected Once per 31 days for penetration flow path is isolation devices isolated, outside primary

containment

AND

Prior to entering MODE 2 or 3 from MODE 4 if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days, for isolation devices inside primary containment

B. -------- NOTE -------- B.1 Isolate the affected 1 hour Only applicable to penetration flow path by penetration flow paths with use of at least one two PCIVs. closed and de-activated

automatic valve, closed manual valve, or blind

One or more penetration flange. flow paths with two PCIVs inoperable except due to leakage not within limit.

(continued)


PCIVs 3.6.1.3

ACTIONS (continued)


C. ---------NOTE ------------ C.1 Isolate the affected 4 hours except for Only applicable to penetration flow path by excess flow check penetration flow paths with use of at least one valve (EFCV) line only one PCIV. closed and de-activated

automatic valve, closed AND manual valve, or blind

One or more penetration flange. 12 hours for flow paths with one PCIV EFCV line inoperable except due to leakage not within limits. AND

C.2 ----------.NOTE -----------Valves and blind flanges in high radiation areas may be verified by use of administrative means.

Verify the affected Once per 31 days penetration flow path is isolated.

D. One or more penetration D.1 Restore leakage to 4 hours flow paths with leakage not within limit. within limit.

E. Required Action and E.1 Be in MODE 3. 12 hours associated Completion Time of Condition A, B, C, AND or D not met in MODE 1, 2, or 3. E.2 Be in MODE 4. 36 hours

(continued)


PCIVs 3.6.1.3

ACTIONS (continued)


F. Required Action and F.1 Initiate action to Immediately associated Completion suspend operations. Time of Condition A, B, C, with a potential for or D not met for PCIV(s) draining the reactor required to be OPERABLE vessel. during MODE 4 or 5.

OR

F.2 ----.------ NOTE----Only applicable for inoperable RHR shutdown cooling valves.

Initiate action to restore Immediately valve(s) to OPERABLE status.



SR 3.6.1.3.1 -- ------------- NOTE -----------Not required to be met when the 18 inch primary containment purge valves are open for inerting, de-inerting, pressure control, ALARA, or air quality considerations for personnel entry, or Surveillances that require the valves to be open.

Verify each 18 inch primary containment purge 31 days

valve is closed.

(continued)


PCIVs 3.6.1.3


SURVEILLANCE

SR 3.6.1.3.2

SR 3.6.1.3.3

------------------------------ NOTES --------------------------1. Valves and blind flanges in high radiation

areas may be verified by use of administrative means.

2. Not required to be met for PCIVs that are open under administrative controls.

Verify each primary containment isolation manual valve and blind flange that is located outside primary containment and is required to be closed during accident conditions is closed.

------------------------------ NOTES ---------------------------1. Valves and blind flanges in high radiation

areas may be verified by use of administrative means.

2. Not required to be met for PCIVs that are open under administrative controls.

Verify each primary containment manual isolation valve and blind flange that is located inside primary containment and is required to be closed during accident conditions is closed.

FREQUENCY

31 days

Prior to entering MODE 2 or 3 from MODE 4 if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days

SR 3.6.1.3.4 Verify continuity of the traversing incore probe 31 days (TIP) shear isolation valve explosive charge.

SR 3.6.1.3.5 Verify the isolation time of each power operated In accordance with and each automatic PCIV, except for MSIVs, is the Inservice within limits. Testing Program

(continued)


PCIVs 3.6.1.3



SR 3.6.1.3.6 Verify the isolation time of each MSIV is In accordance with a 3 seconds and < 5 seconds. the Inservice

Testing Program

SR 3.6.1.3.7 Verify each automatic PCIV, excluding EFCVs, 24 months actuates to the isolation position on an actual or simulated isolation signal.

SR 3.6.1.3.8 Verify each reactor instrumentation line EFCV (of 24 months a representative sample) actuates to restrict flow to within limits.

SR 3.6.1.3.9 Remove and test the explosive squib from each 24 months on a shear isolation valve of the TIP system. STAGGERED

TEST BASIS

SR 3.6.1.3.10 Verify leakage rate through each MSIV is In accordance with < 11.5 scfh when tested at a 28.0 psig. the Primary

Containment Leakage Rate Testing Program

SR 3.6.1.3.11 Replace the valve seat of each 18 inch purge 24 months valve having a resilient material seat.

SR 3.6.1.3.12 Cycle each 18 inch excess flow isolation damper 24 months to the fully closed and fully open position.


SRM Instrumentation 3.3.1.2

3.3 INSTRUMENTATION

3.3.1.2 Source Range Monitor (SRM) Instrumentation

LCO 3.3.1.2

APPLICABILITY:

The SRM instrumentation in Table 3.3.1.2-1 shall be OPERABLE.


ACTIONS


A. One or more required A.1 Restore required SRMs 4 hours SRMs inoperable in to OPERABLE status. MODE 2 with intermediate range monitors (IRMs) on Range 2 or below.

B. Three required SRMs B.1 Suspend control rod Immediately inoperable in MODE 2 with withdrawal. IRMs on Range 2 or below.

C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time of Condition A or B not met.

D. One or more required D.1 Fully insert all 1 hour SRMs inoperable in insertable control rods. MODE 3 or 4.

AND

D.2 Place reactor mode 1 hour switch in the shutdown position.

(continued)

Amendment No. 195HATCH UNIT I 3.3-10


ACTIONS (continued)


E. One or more required E.1 Suspend CORE Immediately SRMs inoperable in ALTERATIONS except MODE 5. for control rod insertion.

AND

E.2 Initiate action to fully Immediately insert all insertable control rods in core cells containing one or more fuel assemblies.


----------------------- NOTES -------------------.-..............--------------------1. Refer to Table 3.3.1.2-1 to determine which SRs apply for each applicable MODE or

other specified conditions.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the other required channel(s) is OPERABLE.

------------------------------------------------------------------------------------



(continued)




SURVEILLANCE

SR 3.3.1.2.2 --------------------------- NOTES ---.........------------1. Only required to be met during CORE

ALTERATIONS.

2. One SRM may be used to satisfy more than one of the following.

Verify an OPERABLE SRM detector is located in:

a. The fueled region;

b. The core quadrant where CORE ALTERATIONS are being performed, when the associated SRM is included in the fueled region; and

c. A core quadrant adjacent to where CORE ALTERATIONS are being performed, when the associated SRM is included in the fueled region.

FREQUENCY

12 hours


SR 3.3.1.2.4 - -------------------- NOTES---- --------1. Not required to be met with less than or

equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies in the associated core quadrant.

2. Not required to be met during spiral unloading.

Verify count rate is 2 3.0 cps with a signal to noise 12 hours during ratio a 2:1. CORE

ALTERATIONS

AND

24 hours

.an n.nA~

Amendment No. 195

mILLIIg IU

HATCH UNIT 1 3.3-12




SR 3.3.1.2.5 Perform CHANNEL FUNCTIONAL TEST and 7 days determination of signal to noise ratio.

SR 3.3.1.2.6 --------------------------- NOTE ------------Not required to be performed until 12 hours after IRMs on Range 2 or below.

Perform CHANNEL FUNCTIONAL TEST and 31 days determination of signal to noise ratio.

SR 3.3.1.2.7 ----------------------------- NOTES- ------------1. Neutron detectors are excluded.

2. Not required to be performed until 12 hours after IRMs on Range 2 or below.

Perform CHANNEL CALIBRATION. 24 months



Table 3.3.1.2-1 (page 1 of 1) Source Range Monitor Instrumentation

APPLICABLE MODES OR OTHER

SPECIFIED REQUIRED SURVEILLANCE FUNCTION CONDITIONS CHANNELS REQUIREMENTS

1 Source Range Monitor 2(a) 3 SR 3.3.1.2.1 SR 3.3.1.2.4 SR 3.3.1.2.6 SR 3.3.1.2.7

3, 4 2 SR 3.3.1.2.3 SR 3.3.1.2.4 SR 3.3.1.2.6 SR 3.3.1.2.7

5 2(b)(C) SR 3.3.1.2.1 SR 3.3.1.2.2 SR 3.3.1.2.4 SR 3.3.1.2.5 SR 3.3.1.2.7

(a) With IRMs on Range 2 or below.

(b) Only one SRM channel is required to be OPERABLE during spiral offload or reload when the fueled region includes

only that SRM detector.

(c) Special movable detectors may be used in place of SRMs If connected to normal SRM circuits.

Amendment No. '195HATCH UNIT 1 3.3-14

Control Rod Block Instrumentation T 3.3.2

T 3.3.2 CONTROL ROD BLOCK INSTRUMENTATION

TLCO 3.3.2

APPLICABILITY:

The control rod block instrumentation for each Function in Table T3.3.2-1 shall be OPERABLE.

According to Table T3.3.2-1.

ACTIONS.NOTE

Separate Condition entry is allowed for each channel.


A. One or more Functions A.1 Initiate Reactor Manual 1 hour with one or more required Control System rod channels inoperable, withdrawal block.

HATCH UNIT 1 TRM T 3.3-3


SURVEILLANCE REQUIREMENTS S - --.-.-.------.-.-.--------------- NOTES ------------------------1. Refer to Table T3.3.2-1 to determine which TSRs apply for each control rod block Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours.

SURVEILLANCE

TSR 3.3.2.1 -- NOTES 1. For Function 1, not required to be performed

when entering the MODE 2 IRM range Applicability from a higher IRM range until 12 hours after entering the MODE 2 IRM range Applicability.

2. For Function 2, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.

Perform CHANNEL FUNCTIONAL TEST.

FREQUENCY

7 days

TSR 3.3.2.2 NOTES Not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.

Perform CHANNEL FUNCTIONAL TEST. 184 days

TSR 3.3.2.3 NOTES

1. Neutron detectors are excluded.

2. For Function 4, withdrawal of control rods is not permitted during the CHANNEL CALIBRATION.


HATCH UNIT 1 TRM T 3.3-4 Revision 12


Table T3.3.2-1 (Page 1 of 2) Control Rod Block Instrumentation

APPLICABLE MODES OR REQUIRED

OTHER CHANNELS SPECIFIED PER SURVEILLANCE ALLOWABLE

FUNCTION CONDITIONS FUNCTION REQUIREMENTS VALUE

1. SRM

a. Detector Not Full In 3

5

b. Upscale 2(c)

5

c. Inoperative 2(c)

5

d. Downscale 2(a)

5

2. IRM

a. Detector Not Full in

b. Upscale

c. Inoperative

d. Downscale

2,5

2,5

2,5

2 (d)

3

2 (b)

3

2 (b)

3

2 (b)

6

6

6

6

TSR 3.3.2.1

TSR 3.3.2.1

TSR 3.3.2.1 TSR 3.3.2.3

TSR 3.3.2.1

TSR 3.3.2.3

TSR 3.3.2.1

TSR 3.3.2.1

TSR 3.3.2.1 TSR 3.3.2.3

TSR 3.3.2.1 TSR 3.3.2.3

TSR 3.3.2.1

TSR 3.3.2.1

TSR 3.3.2.3

TSR 3.3.2.1

TSR 3.3.2.1 TSR 3.3.2.3

Not full in

Not full in

• 105 cps

•105 cps

NA

NA

Žt 3 cps

a 3 cps

Not full in

•108/125 of full scale

NA

Ž5/125 of full scale

(continued)

(a) With IRMs on Range 2 or below.-

(b) Only one SRM is required to be OPERABLE during includes only that SRM detector.

spiral offload or reload when the fueled region

(c) With IRMs on Range 7 or below.

(d) With IRMs on Range 2 or above.



Table T3.3.2-1 (Page 2 of 2) Control Rod Block Instrumentation

APPLICABLE MODES OR REQUIRED

OTHER CHANNELS SPECIFIED PER SURVEILLANCE ALLOWABLE

FUNCTION CONDITIONS FUNCTION REQUIREMENTS VALUE

3. APRM

a. Simulated Thermal Power - 1 3 TSR 3.3.2.2 (g) Upscale TSR 3.3.2.3

b. Simulated Thermal Power - 2,5(0 3 TSR 3.3.2.2 (g) Upscale (Setdown) TSR 3.3.2.3

c. Inoperative 1,2,5(0 3 TSR 3.3.2.2 NA

d. Neutron Flux - Downscale 1 3 TSR 3.3.2.2 (g) TSR 3.3.2.3

e. Low LPRM Count 1,2,5(0 3 TSR 3.3.2.2 (g)

f. Reactor Recirculation 1 3 TSR 3.3.2.2 (g) Flow - Upscale TSR 3.3.2.3

4. Scram Discharge Volume 1,2,5(e) 1 TSR 3.3.2.3 •18 gallons Water Level - High

(e) With any control rod withdrawn from a core cell containing one or more fuel assemblies, except control

rods withdrawn under the provisions of Technical Specification LCO 3.10.5 or LCO 3.10.6.

(f) During SDM demonstrations in accordance with Technical Specification LCO 3.10.8.

(g) Allowable value controlled by the Setpoint Index.


ECCS Instrumentation 3.3.5.1

3.3 INSTRUMENTATION

3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation

LCO 3.3.5.1

APPLICABILITY:

The ECCS instrumentation for each Function in Table 3.3.5.1 -1 shall be OPERABLE.


ACTIONS

---. ---------. ---------.. --. -------. ---. --------- --------- N O T EO--------------------------------------------- ---------------Separate Condition entry is allowed for each channel.


A. One or more channels A.1 Enter the Condition Immediately inoperable, referenced in

Table 3.3.5.1-1 for the channel.

B. As required by Required B.1 - --------- NOTES----Action A.1 and referenced 1. Only applicable in in Table 3.3.5.1-1. MODES 1,2,

and 3.

2. Only applicable for Functions 1 .a, 1.b, 2.a, and 2.b.

Declare supported 1 hour from discovery feature(s) inoperable, of loss of initiation

capability for feature(s) in both divisions

AND

(continued)



K

Amendment No. 135

ACTIONS


B. (continued) B.2 ----------- NOTE -----------Only applicable for Functions 3.a and 3.b. --------------- -------- I --------------

Declare High Pressure 1 hour from discovery Coolant Injection (HPCI) of loss of HPCI System inoperable, initiation capability

AND

B.3 Place channel in trip. 24 hours

C. As required by Required C.1 ---------- NOTES ----------Action A.1 and referenced 1. Only applicable in in Table 3.3.5.1-1. MODES 1, 2,

and 3.

2. Only applicable for Functions 1.c, 2.c, 2.d, and 2.f.

--------------- -------------------------


capability for feature(s) in both divisions

AND

C.2 Restore channel to 24 hours OPERABLE status.

(continued)

HATCH UNIT 2 3.3-34


ACTIONS (continued)


D. As required by Required D.1 --------NOTE ----Action A.1 and referenced Only applicable if in Table 3.3.5.1-1. HPCI pump suction is

not aligned to the suppression pool.

Declare HPCI System 1 hour from discovery inoperable, of loss of HPCI

initiation capability

AND

D.2.1 Place channel in trip. 24 hours

OR

D.2.2 Align the HPCI pump 24 hours suction to the suppression pool.

E. As required by Required E.1 ---------- NOTES ----------Action A.1 and referenced 1. Only applicable in in Table 3.3.5.1-1. MODES 1,2,

and 3.

2. Only applicable for Functions 1.d and 2.g.


capability for subsystems in both divisions

AND

E.2 Restore channel to 7 days OPERABLE status.

(continued)



ACTIONS (continued)


F. As required by Required F.1 Declare Automatic 1 hour from discovery Action A.1 and referenced Depressurization of loss of ADS in Table 3.3.5.1-1. System (ADS) valves initiation capability in

inoperable, both trip systems

AND

F.2 Place channel in trip. 96 hours from discovery of inoperable channel concurrent with HPCI or reactor core isolation cooling (RCIC) inoperable

AND

8 days

G. As required by Required G.1 Declare ADS valves 1 hour from discovery Action A. 1 and referenced inoperable, of loss of ADS in Table 3.3.5.1-1. initiation capability in

both trip systems

AND

G.2 Restore channel to 96 hours from OPERABLE status. discovery of

inoperable channel concurrent with HPCI or RCIC inoperable

AND

8 days

H. Required Action and - H.1 Declare associated Immediately associated Completion supported feature(s) Time of Condition B, C, D, inoperable. E, F, or G not met.




-.--------------.----..........---------------------...... NOTES ----------------------------------------------------------1. Refer to Table 3.3.5.1-1 to determine which SRs apply for each ECCS Function.

2. When a channel is placed in an inoperable status solely for, performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours for Functions 3.c and 3.f; and (b) for up to 6 hours for Functions other than 3.c and 3.f provided the associated Function or the redundant Function maintains initiation capability.









Table 3.3.5.1-1 (page 1 of 5) Emergency Core Cooling System Instrumentation

APPLICABLE CONDITIONS MODES REQUIRED REFERENCED

OR OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE

FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

1. Core Spray System


b. Drywell Pressure High

c. Reactor Steam Dome Pressure Low (Injection Permissive)

d. Core Spray Pump Discharge Flow Low (Bypass)

2. Low Pressure Coolant Injection (LPCI) System


b. Drywell Pressure - High

1,2,3, 4(a), 5(a)

1,2,3

1,2,3

4(a), 5(a)

4(b)

4(b)

4

4

1,2,3, 1 per 4(a), 5(a) subsystem

1,2,3, 4(a), 5(a)

1,2,3

4(b)

4(b)

Amendment No. 137

B

B

C

SR SR SR SR

SR SR SR SR

SR SR SR SR

SR SR SR SR

SR SR SR SR

B

3.3.5.1.1 3.3.5.1.2 3.3.5.1.4 3.3.5.1.5

3.3.5.1.1 3.3.5.1.2 3.3.5.1.4 3.3.5.1.5

3.3.5.1.1 3.3.5.1.2 3.3.5.1.4 3.3.5.1.5

3.3.5.1.1 3.3.5.1.2 3.3.5.1.4 3.3.5.1.5

3.3.5.1.1 3.3.5.1.2 3.3.5.1.4 3.3.5.1.5

3.3.5.1.1 3.3.5.1.2 3.3.5.1.4 3.3.5.1.5

3.3.5.1.1 3.3.5.1.2 3.3.5.1.4 3.3.5.1.5

a -113 inches

< 1.92 psig

> 390 psig and

S476 psig

> 390 psig and

< 476 psig

t 570 gpm and

< 745 gpm

2 -113 inches

< 1.92 psig

E

B

B

SR SR SR SR

SR SR SR SR

(a) When associated subsystem(s) are required to be OPERABLE.

(b) Also required to initiate the associated diesel generator (DG) and isolate the associated plant service water (PSW) turbine building (T/B) isolation valves.

(continued)

I

I

HATCH UNIT 2 3.3-38






2. LPCI System (continued)

c. Reactor Steam Dome Pressure - Low (Injection Permissive)

d. Reactor Steam Dome Pressure - Low (Recirculation Discharge Valve Permissive)

e. Reactor Vessel Shroud Level - Level 0

f. Low Pressure Coolant Injection Pump Start Time Delay Relay

1,2,3

4(a), 5(a)

1(c), 2(0), 3(c)

1,2,3

1,2,3, 4(W), 5(a)

4 C

4 B

4

2

1 per pump

SR SR SR SR

SR SR SR SR

SR SR SR SR

SR SR SR SR

C

B

C

3.3.5.1.1 3.3.5.1.2 3.3.5.1.4 3.3.5.1.5

3.3.5.1.1 3.3.5.1.2 3.3.5.1.4 3.3.5.1.5

3.3.5.1.1 3.3.5.1.2 3.3.5.1.4 3.3.5.1.5

3.3.5.1.1 3.3.5.1.2 3.3.5.1.4 3.3.5.1.5

390 psig and

S 476 psig

> 390 psig and

S 476 psig

> 335 psig

> -202 inches

SR 3.3.5.1.4 SR 3.3.5.1.5

Pumps A, B, D

Pump C

g. Low Pressure Coolant Injection Pump Discharge Flow Low (Bypass)

1,2,3, 1per 4(a), 5(a) subsystem

E SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.5

> 9 seconds and

• 15 seconds

< 1 second

a 1675 gpm and

<2215 gpm

(continued)

(a) When associated subsystem(s) are required to be OPERABLE.

(c) With associated recirculation pump discharge valve open.

Amendment No..1743.3-39HATCH UNIT 2



,APPLICABLE CONDITIONS MODES REQUIRED REFERENCED



3. High Pressure Coolant Injection (HPCI) System

a. Reactor Vessel Water 1, 4 B SR 3.3.5.1.1 Ž -47 inches Level - Low Low, 2(d), 3 (d) SR 3.3.5.1.2 Level 2 SR 3.3.5.1.4

SR 3.3.5.1.5

b. Drywell Pressure - High 1, 4 B SR 3.3.5.1.1 < 1.92 psig 2(d), 3(d) SR 3.3.5.1.2

SR 3.3.5.1.4 SR 3.3.5.1.5

c. Reactor Vessel Water 1, 2 C SR 3.3.5.1.1 S 56.5 inches Level - High, Level 8 2(d), 3(d) SR 3.3.5.1.2

SR 3.3.5.1.4 SR 3.3.5.1.5

d. Condensate Storage 1, 2 D SR 3.3.5.1.3 ? 2.61 ft Tank Level - Low 2(d), 3(d) SR 3.3.5.1.5

e. Suppression Pool 1, 2 D SR 3.3.5.1.1 5154 inches Water Level - High 2 (d), 3(d) SR 3.3.5.1.2

SR 3.3.5.1.4 SR 3.3.5.1.5

f. High Pressure Coolant 1, 1 E SR 3.3.5.1.1 >590 gpm Injection Pump 2 (d), 3(d) SR 3.3.5.1.2 and Discharge Flow - Low SR 3.3.5.1.4 •845gpm (Bypass) SR 3.3.5.1.5

(continued)

(d) With reactor steam dome pressure> 150 psig..






FUNCTION CONDITIONS FUNCTION , ACTION A.1 REQUIREMENTS VALUE

4. Automatic Depressurization System (ADS) Trip System A


b. Drywell Pressure - High

c. Automatic Depressurization System Initiation Timer

d. Reactor Vessel Water Level - Low, Level 3 (Confirmatory)

e. Core Spray Pump Discharge Pressure High

f. Low Pressure Coolant Injection Pump Discharge Pressure High

g. Automatic Depressurization System Low Water Level Actuation Timer

1, 2(d), 3(d)

1, 2(d), 3(d)

1, 2(d), 3(d)

1, 2(d), 3(d)

1, 2(d), 3(d)

1, 2(d), 3 (d)

1, 2(d), 3(d)

2 F

2

SR SR SR SR

SR SR SR SR

F

I G

F

2 G

4

3.3.5.1.1 3.3.5.1.2 3.3.5.1.4 3.3.5.1.5

3.3.5.1.1 3.3.5.1.2 3.3.5.1.4 3.3.5.1.5

SR 3.3.5.1.4 SR 3.3.5.1.5

SR SR SR SR

SR SR SR SR

SR SR SR SR

G

2 G

3.3.5.1.1 a3.5.1.2 3.3.5.1.4 3.3.5.1.5

3.3.5.1.1 3.3.5.1.2 3.3.5.1.4 3.3.5.1.5

3.3.5.1.1 3.3.5.1.2 3.3.5.1.4 3.3.5.1.5

SR 3.3.5.1.4 SR 3.3.5.1.5

a -113 inches

• 1.92 psig

S 114 seconds

2 0 inches

a 137 psig and

< 180 psig

> 112 psig and

< 180 psig

s 12 minutes 18 seconds

(continued)

(d) With reactor steam dome pressure > 150 psig.







5. ADS Trip System B

a. Reactor Vessel Water 1, 2 F SR 3.3.5.1.1 >-113 inches Level - Low Low Low, 2(d), 3(d) SR 3.3.5.1.2 Level 1 SR 3.3.5.1.4

SR 3.3.5.1.5

b. Drywell Pressure - High 1, 2 F SR 3.3.5.1.1 S 1.92 psig 2(d), 3(d) SR 3.3.5.1.2

SR 3.3.5.1.4 SR 3.3.5.1.5

c. Automatic 1, 1 G SR 3.3.5.1.4 < 114 seconds Depressurization 2(d), 3(d) SR 3.3.5.1.5 System Initiation Timer

d. Reactor Vessel Water 1, 1 F SR 3.3.5.1.1 > 0 inches Level - Low, Level 3 2(d), 3(d) SR 3.3.5.1.2

'(Confirmatory) SR 3.3.5.1.4 SR 3.3.5.1.5

e. Core Spray 1, 2 G SR 3.3.5.1.1 a 137 psig Pump Discharge 2(d), 3(d) SR 3.3.5.1.2 and Pressure - High SR 3.3.5.1.4 < 180 psig

SR 3.3.5.1.5

f. Low Pressure Coolant 1, 4 G SR 3.3.5.1.1 > 112 psig Injection Pump 2(d), 3(d) SR 3.3.5.1.2 and Discharge Pressure - SR 3.3.5.1.4 180 psig High SR 3.3.5.1.5

g. Automatic 1, 2 G SR 3.3.5.1.4 < 12 minutes Depressurization 2(d), 3(d) SR 3.3.5.1.5 18 seconds System Low Water Level Actuation Timer

(d) With reactor steam dome pressure > 150 psig.


.rnwrr�r-

RCIC System 3.5.3

3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE COOLING (RCIC) SYSTEM

3.5.3 RCIC System

LCO 3.5.3

APPLICABILITY:

ISOLATION

The RCIC System shall be OPERABLE.

MODE 1, MODES 2 and 3 with reactor steam dome pressure > 150 psig.

ACTIONS


A. RCIC System inoperable. -------------------- NOTE ---------------LCO 3.0.4 is not applicable.

A.1 Verify by administrative 1 hour means high pressure coolant injection (HPCI) System is OPERABLE.

AND

A.2 Restore RCIC System 14 days to OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. AND

B.2 Reduce reactor steam 36 hours dome pressure to S5 150 psig.


RCIC System 3.5.3



SR 3.5.3.1 Verify the RCIC System piping is filled with water 31 days from the pump discharge valve to the injection valve.

SR 3.5.3.2 Verify each RCIC System manual, power 31 days operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.5.3.3 ---------------------------- NOTE -----------------------------Not required to be performed until 12 hours after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure < 1058 psig and 92 days 2:920 psig, the RCIC pump can develop a flow rate Ž 400 gpm against a system head corresponding to reactor pressure.

SR 3.5.3.4 A ------------------- NOTE ----------------Not required to be performed until 12 hours after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure s 165 psig, the RCIC 24 months pump can develop a flow rate ? 400 gpm against a system head corresponding to reactor pressure.

SR 3.5.3.5 --- ------------------------. NOTE--------------Vessel injection may be excluded.

Verify the RCIC System actuates on an actual or 24 months simulated automatic initiation signal.


ECCS - Operating 3.5.1

3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION

COOLING (RCIC) SYSTEM

3.5.1 ECCS - Operating

LCO 3.5.1

APPLICABILITY:

Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS),function of six of seven safety/relief valves shall be OPERABLE.

MODE 1, MODES 2 and 3, except high pressure coolant injection (HPCI) and ADS

valves are not required to be OPERABLE with reactor steam dome pressure s 150 psig.

ACTIONS


A. One low pressure ECCS A.1 Restore low pressure 7 days injection/spray subsystem ECCS injection/spray inoperable, subsystem to

OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not AND met.

B.2 Be in MODE 4. 36 hours

C. HPCI System inoperable. CA1 Verify by administrative 1 hour means RCIC System is OPERABLE.

AND

C.2 Restore HPCI System 14 days to OPERABLE status.

(continued)

Amendment No. 204

'I

HATCH UNIT 1 3.5-1

- - � 'N: . - -


ACTIONS (continued)


D. HPCI System inoperable. D.1 Restore HPCI System 72 hours to OPERABLE status.

AND OR

One low pressure ECCS injection/spray subsystem D.2 Restore low pressure 72 hours is inoperable. ECCS injection/spray

subsystem to OPERABLE status.

E. Two or more ADS valves E.1 Be in MODE 3. 12 hours inoperable.

AND OR

E.2 Reduce reactor steam 36 hours Required Action and dome pressure to associated Completion < 150 psig. Time of Condition C or D not met.

F. Two or more low pressure F.1 Enter LCO 3.0.3. Immediately ECCS injection/spray subsystems inoperable.

OR

HPCI System and two or more ADS valves inoperable.

I,

Amendment No. 204

I

HATCH UNIT 1 3.5-2




SR 3.5.1.1 Verify, for each ECCS injection/spray subsystem, 31 days the piping is filled with water from the pump discharge valve to the injection valve.

SR 3.5.1.2 ---------------------------- NOTE----------------Low pressure coolant injection (LPCI) subsystems may be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the Residual Heat Removal (RHR) low pressure permissive pressure in MODE 3, if capable of being manually realigned and not otherwise inoperable. -- ---------------------------------------------------

Verify each ECCS injection/spray subsystem 31 days manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.5.1.3 Verify ADS air supply header pressure is 31 days ý 90 psig..

SR 3.5.1.4 Verify the RHR System cross tie valve is closed 31 days

and power is removed from the valve operator.

SR 3.5.1.5 (Not used.)

SR 3.5.1.6 -------------------- NOTE- -.........----------------Only required to be performed prior to entering MODE 2 from MODE 3 or 4, when in MODE 4 > 48 hours.

Verify each recirculation pump discharge valve 31 days cycles through one complete cycle of full travel or is de-energized in the closed position.

(continued)


*.�-w� -�---W:--��w - I




SR 3.5.1.7 Verify the following ECCS pumps develop the In accordance with specified flow rate against a system head . the Inservice corresponding to the specified reactor pressure. Testing Program

SYSTEM HEAD CORRESPONDING

NO. OF TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF

CS a4250 gpm 1 a 113 psig LPCI k 17,000 gpm 2 a 20 psig

SR 3.5.1.8 -------------------- NOTE --------- Not required to be performed until 12 hours after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure s 1058 psig and 92 days a 920 psig, the HPCI pump can develop a flow rate a 4250 gpm against a system head corresponding to reactor pressure.

SR 3.5.1.9 --- ------------------ -NOTE--------------------Not required to be performed until 12 hours after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure s 165 psig, the HPCI 24 months pump can develop a flow rate > 4250 gpm against a system head corresponding to reactor system pressure.

SR 3.5.1.10 ------------------- NOTE- -.......------------------Vessel injection/spray may be excluded. .............- .--7 ------------ ------.-

Verify each ECCS injection/spray subsystem 24 months actuates on an actual or simulated automatic initiation signal.

(continued)

Amendment No. '232

I

3.5-4HATCH UNIT 1




SR 3.5.1.11 -------------------- NOTE--------Valve actuation may be excluded.

Verify the ADS actuates on an actual or simulated 24 months automatic initiation signal.

SR 3.5.1.12 Verify each ADS valve relief mode actuator 24 months strokes when manually actuated.


GRAPH 9 UNIT 1

CORE SPRAY VORTEX LIMIT

1000

73

68

C

F-4

/ / /2

2000 3000 4000 5000

CORE SPRAY FLOW (gpm)

NOTE: Ma-" "-e SPDS Emergency Displays in place of this Graph.

SSA F E A/I z

f . . f l f I I" J f f f f I" /" f7 / • p

63

UNSAFE UNSAFE /

58

53 /

/48-I

3

'eui einssued elqoO!lddD eqjG Molq si UIObJ Oulm.Jedo ejS,, eJnsseJd IeqwoL4o uolsselddnS

l4doj0 sit4 4o ecoid ul sAoldsia AoueJeuw3 SCdS esn A•N :e9oN

(wd6) MOl-

0008 OOOz 0009 1 1 1 I

0009 00017 000C 0006 1

000L 0 I I

([,9v7 [ eAoqv JO V jEAG-1 'JOJOM lOOld uo!sseJddnS) I!W!1 HSdN dwnd AoJdS enoo

09 L

bl6sd 01 LMOlG9

is

16d0 A~ JOI

3±VSNn E- 017;5 0-9O6

0 m

C') 06 L

(D)

00

0£0 -(D

066, 3 (D

096I 5+

VII HdVUO

[--

'Gull ejrseja eaqcoiaaoo eL4 mcleq a, uopjce awjaco jsin. emnsseid ieqwDL]Q uop~seiddnS

ytdoiE) siy jo ecold ui sAnldsrc] AoueDJewB SCdS esn ADA~ eoN

(ujdb) MvOld

0

LIINf 9

(,,9t7 M01OleO leA JGeIM lOOd uo!sseJddns)

lW!-I HSdN dwnd ADJdS ejoC SLti HdVHdO 99

0009 O0001 0009 0009 000t 000C 000O 0001

...... ........ -- -------- ------ -I-------- ------ ------- --- -- --

... ... ...f.l.-- ---

017

09L

09L

OZ L

06 L CD)

061 _cn,

0

0 --I OLZ (

3

C:

nl177 cuVV •

MCREC System 3.7.4

3.7 PLANT SYSTEMS

3.7.4 Main Control Room Environmental Control (MCREC) System

LCO 3.7.4 Two MCREC subsystems shall be OPERABLE.

----------------------------.----------------------------- NOTE r -- --- ------------------------------------------------The main control room boundary may be opened intermittently under administrative control.

APPLICABILITY: MODES 1,2, and 3, During movement of irradiated fuel assemblies in the secondary

containment, During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel

(OPDRVs).

ACTIONS


A. One MCREC subsystem A.1 Restore MCREC 7 days inoperable, subsystem to

OPERABLE status.

B. Two MCREC subsystems B.1 Restore control room 24 hours inoperable due to boundary to inoperable control room OPERABLE status. boundary in MODE 1,2, or 3.

C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time of Condition A or B AND not met in MODE 1, 2, or 3.

C.2 Be in MODE 4. 36 hours

(continued)

Amendment No. -2253.7-8HATCH UNIT 1

MCREC System 3.7.4

ACTIONS (continued)


D. Required Action and --- --------------- NOTE --------associated Completion Time LCO 3.0.3 is not applicable. of Condition A not m et -------------------------------------------during movement of irradiated fuel assemblies in D.1 Place OPERABLE Immediately the secondary containment, MCREC subsystem in during CORE pressurization mode. ALTERATIONS, or during OPDRVs. OR

D.2.1 Suspend movement of Immediately irradiated fuel assemblies in the secondary containment.

AND

D.2.2 Suspend CORE Immediately ALTERATIONS.

AND

D.2.3 Initiate action to Immediately suspend OPDRVs.

E. Two MCREC subsystems E.1 Enter LCO 3.0.3. Immediately inoperable in MODE 1, 2, or 3 for reasons other than Condition B.

(continued)


MCREC System 3.7.4

ACTIONS (continued)


F. Two MCREC subsystems ----------------- NOTE ------------------inoperable during LCO 3.0.3 is not applicable. movement of irradiated fuel ------------------- ......----------------assemblies in the secondary containment, F.1 Suspend movement of Immediately during CORE irradiated fuel ALTERATIONS, or during assemblies in the OPDRVs. secondary containment.

AND

F.2 Suspend CORE Immediately

ALTERATIONS.

AND

F.3 Initiate action to Immediately suspend OPDRVs.



SR 3.7.4.1 Operate each MCREC subsystem a 15 minutes. 31 days

SR 3.7.4.2 Perform required MCREC filter testing in In accordance with accordance with the Ventilation Filter Testing the VFTP Program (VFTP).

SR 3.7.4.3 Verify each MCREC subsystem actuates on an 24 months

actual or simulated initiation signal.

(continued)


MCREC System 3.7.4



SR 3.7.4.4 Verify each MCREC subsystem can maintain a 24 months on a positive pressure of Ž 0.1 inches water gauge STAGGERED relative to the turbine building during the TEST BASIS pressurization mode of operation at a subsystem flow rate of s 2750 cfm and an outside air flow rate s 400 cfm.


RPS Instrumentation 3.3.1.1

3.3 INSTRUMENTATION

3.3.1.1 Reactor Protection System (RPS) Instrumentation

LCO 3.3.1.1

APPLICABILITY:

The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE.


ACTIONS -------- .........----------------------------------- -----.NOTE ------...................................................... Separate Condition entry is allowed for each channel.


A. One or more required A.1 Place channel in trip. 12 hours channels inoperable.

OR

A.2 ----------- NOTE -----------Not applicable for Functions 2.a, 2.b, 2.c, 2.d, and 2J.

Place associated trip 12 hours system intrip.

B. --------- NOTE ------------- B.1 Place channel in one 6 hours Not applicable for trip system in trip. Functions 2.a, 2.b, 2.c, 2.d, and 2.f. OR

B.2 Place one trip system in 6 hours One or more Functions with trip. one or more required channels inoperable in bothl trip systems.

(continued)

Amendment No. 213

½�>

I

I

HATCH UNIT 1 3.3-1


ACTIONS (continued)


C. One or more Functions with C.1 Restore RPS trip 1 hour RPS trip capability not capability. maintained.

D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition A, B, Table 3.3.1.1-1 for the or C not met. channel.

E. As required by Required E.1 Reduce THERMAL 4 hours Action D.1 and referenced POWER to < 28% RTP. in Table 3.3.1.1-1.

F. As required by Required F.1 Be in MODE 2. 6 hours Action D.1 and referenced in Table 3.3.1.1-1.

G. As required by Required G.1 Be in MODE 3. 12 hours Action D.1 and referenced in Table 3.3.1.1-1.

H. As required by Required H.1 Initiate action to fully Immediately Action D.1 and referenced insert all insertable in Table 3.3.1.1-1. control rods in core

cells containing one or more fuel assemblies.

(continued)

Amendment No. 214

I

HATCH UNIT 1 3.3-2


ACTIONS (continued)


1. As required by Required 1.1 Initiate alternate method 12 hours Action D.1 and referenced to detect and suppress in Table 3.3.1.1-1. thermal-hydraulic

instability oscillations.

AND

1.2 Restore required. 120 days channels to OPERABLE.

J. Required Action and J.1 Be in MODE 2. 4 hours associated Completion Time of Condition I not met.


------.--------.............------------------------- NOTES ............. ----------------------------- 1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains RPS trip capability.

-...-..- .......................................... .......... =........------------------------------------------------.-.. ..



SR 3.3.1.1.2 ------------------------------ NOTE ---------------------------Not required to be performed until 12 hours after THERMAL POWER a 25% RTP.

Verify the absolute difference between the 7 days average power range monitor (APRM) channels and the calculated power is S 2% RTP while operating at Ž 25% RTP.

(continued)



N

Amendment No. 213



SR 3.3.1.1.3 (Not used.)

SR 3.3.1.1.4 A--------------------------- NOTE - - ----Not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2. ---------------- y------------------------------



SR 3.3.1.1.6 Verify the source range monitor (SRM) and Prior to intermediate range monitor (IRM) channels withdrawing SRMs overlap. from the fully

inserted position

SR 3.3.1.1.7 ------------------------------ NOTE -......-------------.-Only required to be met during entry into MODE 2 from MODE 1. ---------------------------------------------------------

Verify the IRM and APRM channels overlap. 7 days

SR 3.3.1.1.8 Calibrate the local power range monitors. 1000 effective full power hours


SR 3.3.1.1.10 - --------------------....... NOTE ---- ...-------------------For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.


(continued)

HATCH UNIT 1 3.3-4




SR 3.3.1.1.11 Verify Turbine Stop Valve.- Closure and 24 months Turbine Control Valve Fast Closure, Trip Oil. Pressure - Low Functions are not bypassed when THERMAL POWER is ? 28% RTP.

SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST. 24 months

SR 3.3.1.1.13 ---------------------------- NOTES ----------- ......---------1. Neutron detectors are excluded.

2. For Function 1, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.


SR 3.3.1.1.14 (Not used.)


SR 3.3.1.1.16 ------------------------------ NOTE-------------Neutron detectors are excluded. -- ------------------------------------------------

Verify the RPS RESPONSE TIME is within limits. 24 months on a STAGGERED TEST BASIS

(continued)

Amendment No. 232

I

HATCH UNIT 1 3.3-5




SR 3.3.1.1.17 Verify OPRM is not bypassed when APRM 24 months Simulated Thermal Power is a 25% and recirculation drive flow is < 60% of rated recirculation drive flow.


CGRAPH 5 UNIT' 2

BORON INJECTION INITIATION TEMPERATURE

/

/9. 9. 9.

i//

4

N

///

/ /

/III

/

/

//A

//

// //

Wfrffi I �CC Y

/ UNSAFE 4W747771

/

//

'9

/

//

'9

.7< I

/ /

///

//

0

N //•

__ I' 1,- - .---

6

9,

/

81 I I I I I 10:23 12 14 16 18 20

REACTOR POWER (%)

NOTE: May use SPDS Emergency Displays in place of this Graph.

//'9

/

/

/

/ // •/•

7'

'9

170

i/

/

160

150

140

130-

120-

110

1 flA1 00

0 21

-•SAFE

____1 2..... 4 .. 5 7 1 8 1 9

SC - SECONDARY CONTAINMENT CONTROLT rao miien.l xrdirreranliol ieplupe aso bove

( A.y odr foordratin upterasrolevelohmi Table 5 Meximem Nu.s.mal Operellyg Water LeveAmfex or HVAS .0ioato rdlollou leelabv Table 0 Muuimem Normal Operating Raalost Orn 11

DiOtroeonlt ossure .at aove 0 In. of wator .0

10 11PLANT E.A HATCH JUNIT

... - 0SOCONY ... eIrwtNT to~ra APA. FunloAxraxYaEmes, CpRgo.

31EO-mOP.014-IS REV. 7 ATTACHMENT I

IF PRIMARYCONTAINMENT FLOODING MEN. etllleOPsaendIS OR HAS BEEN REQUIRED I enter th Severi Acidrt Guidelines

'COperate available area cootran in Olfeaoed areas,

)E a moundary ona d condilio does NOT 017l

THEN operatele Interloitg: o ReIelp Foor HVAC

per 34SO-T40-1 O S o Recolor Buidisg HVAC

par 40-SOT41.005iIS

WA IT UNTIL

area ambient or ditferential temperature

Maximum Normal Operating Temperature (Table 4)

Isolate ALL systerms diSoparglfg p1o area EXCEPT systems reoutred to: osareodeqaae core COOilig o sliertdown reactor o suppress pre o mairla, primorycontorlsnt

iteegrity

rPERFORM CONCURRENTLy

WAITUNTIL

to dioirrgiug racerorcou loaspecondary conlenmers

(table?7)

ANY~aree ambietrdiyaroviu .amour rmonies Meardisu Site Operating

Temperature (Tables4)

PERFORM CONCURRENTLY FOA~ODSemmA

WAIT UNTIL

.. s ard1.dfpeabove me

emourasm rat Opeatiu Temperature Is noi then onearea

EMERGENCY OEPREyS5IS REQUIRED

WATUNI

araabeto d ileW= fereIGo P&lla eprtr

G Table 4 SECONDARYCONTAINMENT OPERATING TEMPERATURES

AMBIENT TEMP (I021B.S1IS7) Opering rperuieg

on IH11-P814, 1G31-R604 V lue V elUe •F .

shae Sop Room (1031-No01A) 2 Hy Room (IG31-No16B) 3 APumprpruom (North) (IG31-N16C) 4 pimp roon (Snelh) (G31.-N 16D) 5 Ha Roo (No31-N0b6l 6 Phase Sep ROOm (0IG)1-N16F)

SOUTHE•AST DIAGONAl AREA

7 RHRJCSA(lEIl-NM9OA) NORPTH EAST DIAGONAL A'EA

8 RHR/CS B (IE11-No09B)

9 Pump Room (1E41-N024) 10 EmerAreo Cir (1E41-NO30A) 11 Emer Area COr (1E41-H3xOI)

RM C ROORAREA 12 Praup No•on•, 1 .-P1r) 13 drorr r CJt0151-N023A) 14 Ere-AerafeCl(lf5i~H2E

TOUIRROOMAREA - 15 West Wall (IE.1-NO2MA)

16 Soulhwesi Wall (IESI-N0251) 17 Southeast Wall (IE51-oz25c) Is Nxthimest~al LQE1N0ý2ý

CIAiOITIAMIRTINNFI AGFA s9 0B20-N0l4

HPCI PIPE PENIETRATION ROOM I IE41-N046A 2 1641400468

150 160 ISO 150 ISO 15o

15O

I5

2125 212.5 212.5 212.5 212,5

212.6 -1.5

2125

1675 245 107.5 245

167.5 245

.ar 310 1.1•:. 241_

167.1 310 167.5 210

167.5 212,5 1675 212.5 167.5 212,5 167.s_ _ 1 . _

192.5 N00

150 212.5 IN 212.5

lJax NvrrnejMaxale

I FF TEMP (1 B21B-S2) Orera g on i 41l-P6t4. 1 G31R60B 1 ale VaR60

1 Ha Room (1G31.N022XVN0)3A) 2 A pump room (G31-N022EBN023B) $ Opump ream (1G31-N2SC/NO23C) 4 H1 Room (1G31.N022D/N023D) SPase rap Room( )IG31"N022E1N023E

o Phase Sep Room 1G31-NO2FINO23F)

SOU TH EASTD•IAGONAL ARFA

7 RNR/CS A (IE11-N02ONNOSSA)

NORTHFA5T DIAGONAL AREA 8 RHRJCS B (IE1-N0290/No03)

TRCROOM ARF 13 1051-NO2EN0i27A 14 1E51-NO26W/N0278 15 1E51•N0261CN027C

MAIN STEAM I Np TINNFI ARFA

17 1E21.N016NN016B

67 07 67 67 67 67

40

40

42 42 42 42

60

90 9a 98$

98

980

102 102

102

SOIL

WAIT UNTIL

ONE or the fol•loing

Iswe abovanlied eo aiv

MaNirau NOrmng Opratin , War., Leve (Table 5):

o Afloorbdein sreup water loyel

O MNarea water lae

SiOplate AvaLaLe surtes discegpg toretoer

enld rmrp or oal bEXlo sMaximsm Normal 0eaing Water Lyerd able 5)

IF ONE ofgassiure g CANNOTbor res=Wtoe oralm rulnale beamMa-imu

o MY Beot dron trump e war ve o ANYe•awaer 'lemel

WEN r•nlate ALL systnes dornaslung m ter Into sump or are EXCEPT lytems reuair to: Oaoarepuadradqale cporecding

o moderasatoe o manein promary conlatnmori

inteRFOty

SPERFORM CONCURRENTLy

WAIT UNTIL

prtpat, sytemur Is diuscharging madrnclt coolant

nr eotimunarcoaraemast

BEFORE

ANY' are wlertnt araac..es Maximum mete Operate, Waler Levl

PERFORMACONCURRENTLY

WA IT UNTIL

Maximum Sale Opeetog, Water Level Inpropomthan oneerso

(Table 5)

EMERGENCY DEPRESS IS REQUIRED

Table 5 SECONDARYCONTAINMENT Tbe5OPERATING WATER LEVELS

AREA WATER LEVELANNUNCIATORS Opereliug rairg DO t11,l-P657 Vatue Vale

NORTHWEST DIAGONAL AREs

I RD N-WDIAGONAL FLOOR DRN SUMP LEVEL HIGH (657-233) (IT45-N007)

2 CRO N-W DIAG INSTR SUMP LVL HIGH-HIGH (6574681) (1T45-NSM5)

NORTHEAST DIAGONAl ARFA

O NO N-E DIAGONAL FLOOR DRN SUMP LEVEL HIGH-HIGH (657,034) (1T45-NOo)

4 RHR-CS N-E DMI0 INSTR SUMP LVA HIGH-HIGH (65R-0I5) (IT4•.N003)

High

High High

High

High High

14 in above

19 Is, chose ht" el,

SOU TH5•J;T DIAGONAl APRs.1 32 Ia, sboe

RCIC SW DIAG INSTR SUMP LI8 High Sl.T HIGH-HIGH (657-,06) (1T45-NO04) I High

ORJTHFA•T DIAGONAL AQEA R

6RIIR-S S-IEDIAG INSTR SUMP LVL High HIGH-HIGH (t57-07G) (IT4S-EI00A) High

7 HPCI ROOM INSTR SUMP LVL HIGH-HIGH (657s071)(1T41"NO0I)

High High

14 In. above Oa el.

37 BIn above OT al.

Sgln. 8 TORUS -WMEAIHSTR SUMPLVL High above

HIGH-HIGH (657-104) (1T45-N(02C) High a7 al.

O TORUS N-WAREA INSTR SUMP LVL HIGH-"IIGH (657-105) (1T45-NM02D)

10 TORUS N-EAREA INSTR SUMP LVL HIGH-HIGH (657-106) (1T45-N(028)

11 TORUS S-EAREAINSTR SUMP LVL HIGH-HIGO (657-107) (1T45-E002A)

High High

High High

High Hig h

WA IT UNTIL

area wae Iee Isabe

Mairliurn Ieee qhele Watar Leve In Orre lonOeae

bhutd n itactorfpori)AGO-OPSi3I .rs400-OpS014-1S

T SCONDARYCONTAINMENT OPERATING RADIATION LEVELSaim NmrrIa Ma•x Sate

HVACEXHAUSTRADIATIONAIINUNCIATORS Comraktr coantig

on ItIll-601 VelUe Val e. _____________________ mRhr mERty

HI-MI Raditoed WA

- RBLDOGPOTCONTAMAREA 18 NtA (1D11-KMi AD)

- REFUELING FLOOR VENT EXHAUST 18 NIA (l0ll-Roil A-D)

ax. NMax a AREA RADIATION MONITORS operadna raden oan IHII-P600,1D21-P600 Value Value

_ _ _ _ _ _ _ _ oat jmR

Table 6 eF

O Reaorea lFyd a (1021- 1A) 2 Refaeno P'lum evabmay (I D21-KSO1O) 2 Refuceoc F•s (1D21-l<60100 4 Deyree Sl~eld Flug (I D21-l<EO1E) 5 Spent Fuel por & Nar Fuel Semgs (tD21-K6GIM

203'tFVATIONrARFA 6 R8 203 WosedonAa (1D21.K0 IX)

"seT EUPalnAE

7 Sent Fuel Pom enln. Eip (ID21-Kl60C) SFuelPool Denm. Panel (1D21,-t17)

15, (IEVARTION AREA 9 RB 1OSI Ue)aD21-t(0GI)

to Roamapl Ral 0o121.KIIL)

H3IP IELEVATION NORTH AR6 11 TIp icea (1021-ecO01F)

12 North CRD HCU (1D21-KMOIP) 13 TIPPrsbaD•vesANra(1D21-K6I1U)

13" ELEVATION EASTASEA

14 RB 130 N-E WucitegAea (1D21-KIGG 15 EqulpmeotArcesxAbleco (1D21-K601S)

lIS ELEVATION SOUTHAREA 16 RB 130 N-E WosllrinAnea (1021-KMOUH) 17 Soud CR0 HCU(tD21-K MIN)

SOUTHWSVST DIAGONAI.ARJ5E

18 RCIC Equip S-W Diagonal (1D21-.KS1V)

NORTOAWPSTORIAGONALAREA

19 CR0 Pump N-W Dagonal (1D2140601h)

NORTH EAST DIAGONAL AREA 2 CS&RHRMN-Ecdaaoal(1021-K601y)

SOUTH EAST DIAGONAL AREA 21 CS & RHR SE DIagoal (1D2t-KtOIR)

130 ELEVATION AREA OSOUTHWEST) SHPCI Turlona Area (1O21-K60iT)

SO

SO

loo

100 old0

50 1(

150 SO

So

SO lSO

60

SO 5O ED

so

&0

1(00

11000

1000

100(f

1 ON

1(ow

ON0

I ON

1 OOD

S0 l

10 No(

SIC'

WAIT UNTIL.

area radiation loya

Maximumh Noemal OuPeralirg Falroniron Level (Tible 6)

Isolate ALL sysnems dIschargipg into ores EXCEPT sysEems rriud to oassureadequste core coding o slstjsdown aaeucor "o suppress fire O orann pdm.• containment

lelegrity

PERFORM CONCURRENTLY

WAIT UNTIL

pdmaprhgtrimrsysteom

Intro seonerdary corvuppeut (Table?7)

B EFO RE

ANY ardea roadtlrlasleverreches Moelmum Saefo Operfullg Radroto Levol

PERFORM CONCURRENTLY RC(A) polet A

WAIT UNTIL

Itaboven e

Maximmsi Sate Operallng Rdlialit Level In .. re rn one area

(Table 6)

EMERGENCY DEPRESS IS RECUIRE

W A ITd U N T IL O ` P ~ H ý S

RR- RADIOACTIVITY RELEASE CONTROLOfftst radloictiviy releoae ruse

(L WILE PERFORMING THE FOLLOWING

I PRIMARYCONTAINMENTIFLOODINGt IbDEN erssheaEOPaadaenotr he

S ,OR HAS BEEN REOQIREO I S.e•ire.ccdepro Guire•ines

MAIILE PERFORMING THE FOLLOWING

LI If Turri HYWAS to shudivofd MEN retulaulngHVAC as ItI • •uRhed perA34SOU41401-IS

1 Isolate ALL pdfary ,ystem disholrartn rfauor. oolant klo areas outside pFrmity and seondary ronlah-eomto (Table 7) EXCEPT systers requited to: 0 assure adequate core cooteig o shutdowr reotar O mpalle gopdmryocnalmunt

Wmooily

WAIT UNTIL

ps roumary tysem sTabxle?

containmensie

BEFORE mowlierdnefossvit reles rae reaches


EMERGENCY OEPRESS IS RE•• IRE

Table 7 PRimarY SYSTEMSPrimary system dischargral Into area is derned as REACTOR COOLANT leak from:

o CRD o RCIC o Com Spry o Recrprsompleg O Feedatar 0 RWICU oHPCI o RHR o Main steam o SaLC O MaI st•ram draios

A system is consIdered a "p6r.y sytauer Ir lowerg maclot frew•pre eppdf reduce the teak ralt Forom he urodated ystelm,

SECONDARYCONTAINMENTHVACE Table 14 EXHAUSTRAIDIATION)SOLATIONSETPOJNTSIVAC EXHAUST RADIATION ANNUNCIATORS e(Wter

on 1(2)HIIPSOI m

"-UNITI XELDGPOTCONTAMAREA 18 (10114<69/bD

-UNIT 1 REFUELING FLOOR VENT EXHAUST 18 (ID11-K61OiAD)

- UNIT 2 RX BLDG POT CONTAM AREA (2D11-s6OoA-D)

- UNIT 2 REFUELING FLOOR VENT EXHAUST (21DI1K611A-D)

- UNIT 2 REFUELING FLOOR VENT EXHAUST (2D11-K(634A-O)

- UNIT 2 REFUELING FLOOR VENT EXHAUST (2DI 1l<$6A-D)

I.... l.,I-J - - - I I

1 I 2 3 4 5 6 7 8S 9

15,0

6.9

5.7

1U

J

A


A

Oo SAGs

IE AMY Ut oit I Unit 2 aeco•ary enortaiement THEN comfirm hIAC Axhaust hadlotit.enet exceeds the I . Unt I end Unit 2 Roetcor Building iuoletiou seupot (Table 14) t HVAC isslarion

I oUnlx I lod Unit 2 Refueal Flor HVAC Isslaslon

I o Unit I and Unit 2 NSOT initiaion I per34AE-T22-0n-1S

1E Rmtuel Floor HVAC liomlats IEN reeted ReMuel Floor HVAC per DI 3450-T41-O(06-1S

a Unit i or Unit)2 secoay .onlalnren I Ir nee ary defeat high dirywll pressure radiation conditiondoe. NOTealsA

I loRe L Aoolalion rntarleoks I pal 31EO-E0P-100-1S

JE Reactor Buis.HVAC Isloates M3Lj resNai Reactorai BdIlng HVAC per I 34SO-T41-005-1S

a Unit I or Unit 2 seco, IdarVcoalprxfrnt I If reesar y defeat high drpywall pressure modioon cmetdilnt dson NOT exist

I or Islaio Intortobs I per 31fO.EOP.100-1S

D

E

B

C

D

E

Fl

J

K

G

H

J

K

11I I

i

1

9.6

SOUTHERN NUCLEAR DOCUMENT TYPE: PAGE PLANT E.I. HATCH EMERGENCY PREPAREDNESS PROCEDURE 1 OF 46 DOCUMENT TITLE: DOCUMENT NUMBER: REVISION/VERSION

EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NO14.3

EXPIRATION APPROVALS: EFFECTIVE DATE: DEPARTMENT MANAGER JCL DATE 10/28/99 DATE:

7/9/2002 N/A NPGM/POAGM/PSAGM CTM DATE 10/28/99

1.0 OBJECTIVE

This procedure establishes the methodology for emergency classification. Specific Emergency Action Levels (EALs) and minimum initial actions to respond to a given emergency are established in this procedure.

2.0 APPLICABILITY

This procedure applies to emergency classification determinations and associated initial responses. This procedure is performed as required.

3.0 REFERENCES

3.1 10AC-MGR-006-0, Hatch Emergency Plan

3.2 73EP-EIP-004-0, Duties of Emergency Director

3.3 73EP-EIP-005-0, On-Shift Operations Personnel Emergency Duties

3.4 73EP-EIP-015-0, Offsite Dose Assessment

3.5 73EP-EIP-018-0, Prompt Dose Assessment

3.6 73EP-EIP-073-0, Offsite Emergency Notifications

3.7 Hatch Unit 1 Technical Specifications (TS), Sections 2.0, 3.2 through 3.9, 3.11

3.8 Hatch Unit 2 Technical Specifications (TS), Sections 2.0, 3.2 through 3.9, 3.11

3.9 Edwin I. Hatch Nuclear Plant Unit 1 and Unit 2 System Evaluation Document

MGR-0002 Rev. 8

SOUTHERN NUCLEAR PAGE 2 OF 46 PLANT E.I. HATCH DOCUMENT TITLE: DOCUMENT NUMBER: REVISION/VERSION

EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NO-14.3

4.0 REQUIREMENTS

4.1 PERSONNEL REQUIREMENTS

4.1.1 Any personnel trained and qualified as an Emergency Director (ED) may use this procedure.

4.1.2 The Emergency Director may modify emergency plan implementing procedures and staffing to meet the needs of emergency response.

4.1.3 Personnel who have received instruction in applicable emergency procedures are required to perform this procedure.

4.1.4 Initially, the Emergency Director position is filled by the Superintendent of Shift (SOS). If the SOS is unavailable, then the affected unit's Shift Supervisor (SS) will become the Emergency Director. IF the SOS is unavailable and the event involves both units, the Unit 1 Shift Supervisor (SS) will become the Emergency Director. Any of these persons will assume the position of Emergency Director in the Control Room until a qualified relief, as specified in step 4.1.5, can arrive on site and receive an adequate turnover.

4.1.5 Any one of the following persons may assume the Emergency Director (ED) duties after he is given proper turnover from the off going ED.

"* Nuclear Plant - General Manager "* Plant Operations - Assistant General Manager (POAGM) "* Plant Support - Assistant General Manager (PSAGM) "* Vice President - Plant Hatch "* Other qualified Emergency Director

4.2 MATERIAL AND EQUIPMENT

N/A - Not applicable to this procedure

4.3 SPECIAL REQUIREMENTS

4.3.1 Portions of this procedure require the results from calculations of projected doses at or beyond the site boundary to determine the appropriate emergency classification. Refer to procedures 73EP-EIP-01 5-0 and 73EP-EIP-01 8-0.

4.3.2 Portions of this procedure will require actual dose measurements (onsite OR off-site) to determine the appropriate emergency classification. Refer to procedures 73EP-EIP-015-0 and 73EP-EIP-018-0.

MGR-0001 Rev. 3

SOUTHERN NUCLEAR PLANT E.I. HATCH

PAE3 OF 46

DOCUMENT TITLE: DOCUMENT NUMBER: REVISION/VERSION EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NO:

14.3

5.0 PRECAUTIONS/LIMITATIONS

5.1 PRECAUTIONS

The value of any emergency actions, which may require movement of plant personnel, must be judged against the danger to personnel or nuclear safety.

5.2 LIMITATIONS

5.2.1 The Operating Facility is defined to be areas within the Protected Area and the 230 Kv and 500 Kv switchyards.

5.2.2 Onsite is defined to be anywhere within the Owner Controlled Area.

7.0 PREREQUISITES

This procedure will be utilized for drills, exercises and actual emergencies.

I REFERENCE

7.0 PROCEDURE

7.1 EMERGENCY CLASSIFICATION AND INITIAL ACTIONS

7.1.1 Upon notification of an abnormal condition OR observation of abnormal instrument readings, notify the Unit Shift Supervisor immediately.

7.1.2 Confirm abnormal conditions by comparing redundant instrument channels OR other related parameters, observation AND field reports, as applicable.

7.1.3 Assess the abnormal condition and classify the emergency by referring to subsection 7.2, Emergency Classification Chart.

MGR-0001 Rev. 3


PAGE 4 OF 46

DOCUMENT TITLE: DOCUMENT NUMBER: IREVISIONNERSION EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NoO

14.3U>

7.1.3.1 The Emergency Classification Chart details abnormal plant conditions that meet specific emergency class entrance requirements. These emergency classes are defined, in theory, in steps 7.1.3.1.1 through 7.1.3.1.4.

7.1.3.1.1 NOTIFICATION OF UNUSUAL EVENT (NUE)

Unusual events are in progress OR have occurred which indicate a potential degradation of the level of safety of the plant. No releases of radioactive material requiring offsite response OR monitoring are expected UNLESS further degradation of safety systems occurs.

7.1.3.1.2 ALERT EMERGENCY

Events are in progress OR have occurred which involve an actual OR potential substantial degradation of the level of safety of the plant. Any releases are expected to be limited to small fractions of the Environmental Protection Agency (EPA) Protective Action Guideline (PAG) exposure levels.

7.1.3.1.3 SITE AREA EMERGENCY

Events are in progress OR have occurred which involve actual OR likely major failures of plant functions needed for protection of the public. Any releases are NOT expected to exceed PAG exposure levels, except near the site boundary.

MGR-0001 Rev. 3

CAUTION

THE REVIEW OF ALL EMERGENCY CLASSES ASSOCIATED WITH A GIVEN CONDITION IS ESSENTIAL. FAILURE TO DO SO COULD RESULT IN A LOWER CLASSIFICATION THAN WARRANTED.

CAUTION

IN THE UNLIKELY EVENT AN ABNORMAL CONDITION MEETS THE DEFINITIONS STATED IN 7.1.3.1.1 THROUGH 7.1.3.1.4 BUT ARE NOT COVERED IN THE EMERGENCY CLASSIFICATION CHART, OR THE INITIATING CONDITION IS MET BUT EQUIPMENT STATUS PARAMETERS VALUES ARE NOT, THE SOS/ED WILL USE HIS JUDGMENT, BASED ON THE AVAILABLE INFORMATION, TO DECLARE THE APPROPRIATE LEVEL OF EMERGENCY.


EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NO.14.3

7.1.3.1.4 GENERAL EMERGENCY

Events are in progress OR have occurred which involve actual OR imminent substantial core degradation OR melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed PAG exposure levels offsite for more than the immediate site area.

7.1.4 IF a potentially declarable emergency exists, inform the SOS immediately. The SOS will evaluate the abnormal condition and operator actions.

7.1.5 IF a declarable emergency exists, the SOS shall assume the duties of the Emergency Director in accordance with 73EP-EIP-004-0, Duties of Emergency Director AND declare the appropriate emergency classification within 15 minutes of the condition requiring the classification.

7.2 EMERGENCY CLASSIFICATION CHART

Refer to the applicable section of the emergency classification chart to assess an abnormal condition and classify the emergency. An index of each emergency action level in the chart is listed on the next page for reference. The key words of an initiating condition are indicated in BOLD print. The supporting data / parameters are listed below each emergency action level. The logical connectors (AND and 0R) used in the supporting data / parameters are to be used as described in Technical Specification section 1.0 "Use and Application", part 1.2 "Logical Connectors."

MGR-0001 Rev. 3

SOUTHERN NUCLEAR PLANT E.I. HATCH PAGE 6 OF 46DOCUMENT TITLE: DOCUMENT NUMBER: REVISION/VERSION

EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NOI 14.3

Index of Emer ency Action Levels Page

1.0 AUTOMATIC INITIATION OF ECCS ------------------------------------------------------------------------ 7 2.0 RADIOLOGICAL EFFLUENTS ----------------------------------------------- 7 ------------------------------ 8 3.0 CORE DAMAGE --------------------------------------------------------------------------------------------- 11 4.0 PRIMARY SYSTEM (STEAM) LINE BREAK OR SAFETY'RELIEF VALVE FAILURE -------------------- 12 5.0 LOSS OF AC POWER --------------------------------------------------------------------------------------- 14 6.0 LOSS OF ONSITE DC POWER ----------------------------------------------------------------------------- 15 7.0 LOSS OF CONTAINMENT ----------------------------------------------------------------------------------- 16 8.0 FIRE IN PLANT ----------------------------------------------------------------------------------------------- 17 9.0 SECURITY EVENT ------------------------------------------------------------------------------------------ 18 10.0 NATURAL PHENOMENON ---------------------------------------------------------------------------------- 19

EARTHQUAKE -------------------------------------------------------------------------------------------- 19 HIGH WINDS ---------------------------------------------------------------------------------------------- 21 HIGH RIVER WATER LEVEL ----------------------------------------------------------------------------- 22 LOW RIVER WATER LEVEL ------------------------------------------------------------------------------ 22

11.0 HAZARDS TO PLANT OPERATIONS ----------------------------------------------------------------------- 23 AIRCRAFT ACTIVITY ------------------------------------------------------------------------------------- 23

EXPLOSIONS ----------------------------------------------------------.---------------------------------- 24 TOXIC GAS ------------------------------------------------------------------------------------------------ 25 FLAMMABLE GAS ----------------------------------------------------------------------------------------- 26 TURBINE FAILURE/MISSILE IMPACT ------------------------------------------------------------------- 27

12.0 THIS SECTION INTENTIONALLY LEFT BLANK 29 13.0 CONTROL ROOM EVACUATION --------------------------------------------------------------------------- 29 14.0 CONTROL ROD DROP -------------------------------------------------------------------------------------- 30 15.0 FAILURE OF REACTOR PROTECTION SYSTEM ------------------------------------------ 31

16.0 LOSS OF CONTROL ROOM INDICATION/ ALARMS/ ANNUNCIATORS --------------------------------- 32 17.0 LOSS OF SHUTDOWN FUNCTIONS ----------------------------------------------------------------------- 33 18.0 FUEL DAMAGE BY FUEL HANDLING ACCIDENT- ----------------------------------------- 34 19.0 HIGH RADIATION OR AIRBORNE CONTAMINATION --------------------------------------- 35 20.0 Loss OF COOLANT ---------------------------------------------------------------------------------------- 36 21.0 LOSs OF ENGINEERED SAFETY FEATURES ------------------------------------------------------------ 37 22.0 MULTIPLE SYMPTOMS AND OTHER CONDITIONS ------------------------------------------------------ 38

TECHNICAL SPECIFICATION SAFETY LIMITS ---------------------------------------------------------- 38 PRECAUTIONARY ACTIVATION OF TSC -------------------------------------------------------------- - 39 PRECAUTIONARY ACTIVATION OF MONITORING TEAMS -------------------------------------------- 40 POTENTIAL LARGE RELEASES OF RADIOACTIVITY -------------------------------------------------- 41 FIRE IN PLANT ----------------- -------------------------------------------------------------------------- 43 CLADDING/CONTAINMENT/COOLANT BOUNDARY PARAMETER ASSESSMENT TABLE -------- 44

23.0 ISFSI OPERATIONS -------------------------------------------------------------------------------------- 45 LOSS OF CASK CONFINEMENT BOUNDARY FOR ANY LOADED SPENT FUEL CASK ------------- 45

DEGRADATION OF SPENT FUEL CASK - AN OPERATIONAL EVENT----------------------- 45 DEGRADATION OF SPENT FUEL CASK - ENVIRONMENTAL PHENOMENA EXTERNAL EVENTS --------------------------------------------------------------------------------------------------- 46

MGR-0001 Rev. 3

K->

SOUTHERN NUCLEAR PLANT E.I. HATCH -I PAGE 7 OF 46

MGR-0001 Rev. 3

DOCUMENT TITLE: DOCUMENT NUMBER: REVISION/VERSION EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NO.

14.3

1.0 - AUTOMATIC INITIATION OF ECCS

Emergency conditions exist WHEN: N A S G UILIA E

AUTOMATIC INITIATION, OR DEMAND FOR ECCS, TO RECOVER WATER LEVEL as indicated by: E E N R T

HPCI, Core Spray, or LPCI Automatic Initiation has occurred. AND

HPCI, Core Spray, or LPCI is discharging to the vessel. AND

Reactor Water Level <- 113 inches OR

Drywell Pressure > 1.92 PSIG (TS)

See Section 20.0, Loss of Coolant, for determination of Site Area Emergency Classification.

See Section 22.0, Multiple Symptoms and Other Conditions, for determination of the General Emergency Classification.

END AUTOMATIC INITIATION OF ECCS

SOUTHERN NUCLEAR PLANT E.I. HATCH PAGE 8 OF 46DOCUMENT TITLE: DOCUMENT NUMBER:

EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0REVISIONNERSION

NO-

2.0 - RADIOLOGICAL EFFLUENTS

Emergency conditions exist WHEN: NASG

R T

LIMITS FOR GASEOUS EFFLUENT RELEASES BEYOND THE SITE BOUNDARY HAVE EXCEEDED TS as indicated by either actual field measurements OR effluent monitor readings corresponding to:

> 0.057 mR (TEDE) in an hour*

(*TS yearly limit divided by the number of hours in a year)

OR

> 500 mR (TEDE) in a year (TS)

LIMITS FOR LIQUID EFFLUENTS HAVE BEEN EXCEEDED [as given in the Offsite Dose Calculation Manual (ODCM)] as indicated by Chemistry analysis as follows:

> 1.5 mR to the total body in a quarter OR

> 3.0 mR to the total body in a year

A GASEOUS EFFLUENT RELEASE IS UNDERWAY WITH OFFSITE DOSE RATES BEYOND THE SITE BOUNDARY, as indicated by either field measurements OR effluent monitor readings corresponding to:

> 0.57 mR (TEDE) in an hour** (** 10 times the TS yearly limit divided by the number of hours in a year.) OR

> 5000 mR (TEDE) in a year (10 X T.S.)

-> -> [CONTINUE TO THE NEXT PAGE] '4'4

MGR-0001 Rev. 3


EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0

2.0 - RADIOLOGICAL EFFLUENTS (continued)

Emergency conditions exist WHEN:

NOTE Adverse meteorological conditions is defined as Stability Class F AND 1 in/sec (= 2 mph) wind speed, OR inclement weather.

A GASEOUS EFFLUENT RELEASE IS UNDERWAY WITH OFFSITE DOSE AT THE SITE BOUNDARY, as indicated by either field measurements OR effluent monitor readings (using adverse meteorological conditions) corresponding to:

> 50 mR (TEDE)in an hour for > 1/2 hr but< 1000 mR (TEDE) in an hour OR

> 500 mR (TEDE)in an hour for 2 min. but < 1000 mR (TEDE) in an hour OR

> 250 mR (CDE thyroid) in an hour for 1/2 hr but < 5 REM (CDE thyroid) in an hour OR

> 2500 mR (CDE thyroid) in an hour for 2 min. but < 5 REM (CDE thyroid) in an hour

DOSE BEYOND THE SITE BOUNDARY IS PROJECTED TO BE > EPA PAGs based on dose projections from plant parameters as follows:

> 1 REM (TEDE)

OR

> 5 REM (CDE thyroid)

REVISION/VERSION NO114.3

-> -> [CONTINUE TO THE NEXT PAGE]---->

MGR-0001 Rev. 3

SOUTHERN NUCLEAR PAGE 10 OF 46 PLANT E.I. HATCH

DOCUMENT TITLE: DOCUMENT NUMBER: REVISION/VERSION EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NO

1 14.3

2.0 - RADIOLOGICAL EFFLUENTS (continued)


R T

A GASEOUS EFFLUENT RELEASE IS UNDERWAY WITH OFFSITE DOSE BEYOND THE SITE BOUNDARY, as indicated by either field measurements OR effluent monitor readings (using actual meteorological conditions) corresponding to:

> 1 REM (TEDE) in an hour OR

> 5 REM(CDE thyroid) in an hour

DOSE BEYOND THE SITE BOUNDARY IS PROJECTED TO BE> EPA PAGs based on dose projections from plant parameters as follows:

A gaseous release is ongoing or imminent AND

> 1 REM (TEDE) OR

> 5 REM (CDE thyroid)

END RADIOLOGICAL EFFLUENTS

MGR-0001 Rev. 3


EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NO.-14.3

3.0 - CORE DAMAGE

Emergency conditions exist WHEN: N A S G

R T

CORE DAMAGE IS INDICATED BY HIGH OFF-GAS ACTIVITY WITH PRETREAT MONITOR (D1l1K601) AT Hi ALARM PLUS Pretreat Monitor reading exceeding eithe of following as indicated on pretreat graph located in Unit 1 OR Unit 2 OFF-GAS Release Curve book.

> 500,000 gCi/sec OR

> 100,000 gCi/sec increase WITHIN a 30 minute period

CORE DAMAGE IS INDICATED BY HIGH OFF-GAS ACTIVITY WITH PRETREAT MONITOR (D 1K601) AT HI-HI ALARM PLUS >_ 5 CI/SEC as indicated on pretreat graph located in Unit 1 OR Unit 2 Off-Gas Release Curve book

CORE DAMAGE IS INDICATED BY HIGH COOLANT ACTIVITY LAB SAMPLE WITH 1-131 DOSE EQUIVALENT COOLANT ACTIVITY > 100 gCi/gm _ !

CORE DAMAGE IS INDICATED BY HIGH COOLANT ACTIVITY LAB SAMPLE WITH 1-131 DOSE EQUIVALENT COOLANT ACTIVITY > 300 pCi/gm

CORE DAMAGE IS INDICATED BY DEGRADED CORE WITH POSSIBLE LOSS OF CORE GEOMETRY as indicated by the following:

Containment Post LOCA Hi Rad Alarm > 138 REM/hr (TS) AND

Reactor Low, Low, Low, Level Alarm < -113 inches OR

Noble Gas Fission Product Monitor (D11 -K630) upscale (7.0 x 105 cpm) OR

Noble Gas Fission Product Monitor (D11 -K630) (variable setpoint) Hi-Hi Radiation Alarm

See Section 22.0, Multiple Symptoms and Other Conditions for determination of General Emergency Classification.

END "CORE DAMAGE

MGR-0001 Rev. 3

SOUTHERN NUCLEAR PLANT E.I. HATCH PAGE 12 OF 46

DOCUMENT TITLE: DOCUMENT NUMBER: REVISION/VERSION EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NO.

1 14.3

4.0 - STEAM LINE BREAK OR SAFETY RELIEF VALVE (SRV) FAILURE


R T

A MAIN STEAM LINE RELIEF VALVE FAILED TO CLOSE WHEN system pressure is reduced below setpoint of safety relief valve (S/RV) and fuses pulled as indicated by:

S/RV tailpipe temperature remaining > 2300 F AND

S/RV tailpipe pressure switch remaining > 80 psig AND

Temperature continuing to increase on any suppression pool local water temperature indicator

A PRIMARY SYSTEM (AS DEFINED BY EOPs) STEAM LINE BREAK OCCURS OUTSIDE CONTAINMENT WITH significant isolation valve leakage as indicated by the following:

Any valid Reactor or Turbine Bldg. leak detection indication

OR

Hi MSL Tunnel Temperature > 1940 F (TS) AND

Any Reactor Bldg. ARM above maximum Normal Operating Values AND increasing

OR

Any Turbine Bldg. ARM above alarm setpoint AND increasing

-' -> [CONTINUE TO THE NEXT PAGE]-- ->

MGR-0001 Rev. 3


IPAGE 13 OF 46

DOCUMENT TITLE: DOCUMENT NUMBER: REVISION/VERSION EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NO4

4.0 STEM LNE BEAKOR SFET RELEF ALVE(SR) FALUR

4.0 - STEAM LINE BREAK OR SAFETY RELIEF VALVE (SRV) FAILURE (continued)

Emergency conditions exist WHEN: [NFA1

AN UNISOLABLE PRIMARY SYSTEM (AS DEFINED BY THE EOPS) BREAK OUTSIDE CONTAINMENT as indicated by:

A primary containment isolation failure (cannot be isolated automatically OR manually) has occurred on the affected primary system.

AND

Entry conditions into Secondary Containment Control Emergency Operating Procedures

OR

Any indications of significant leakage into the Turbine Bldg. from the Main Steam system WITH Turbine Bldg. ARMs above alarm setpoint AND increasing.

OR

SOS/ED judgment

See Section 22.0, Multiple Symptoms and Other Conditions, for determination of General Emergency Classification.

ENDSTEAM LINE BREAK OR SAFETY RELIEF VALVE (SRV) FAILURE

MGR-O001 Rev. 3

SOUTHERN NUCLEAR PAGE 14 OF 46 PLANT E.I. HATCH

DOCUMENT TITLE: DOCUMENT NUMBER: REVISION/VERSION EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NO

1 14.3

5.0 - LOSS OF AC POWER

Emergency conditions exist WHEN: N U

A LOSS OF OFFSITE POWER OR LOSS OF ONSITE AC POWER CAPABILITY HAS OCCURRED and E is indicated as follows:

Loss OF OFFSITE POWER is indicated by: Zero voltage on all 500 kV incoming lines

AND

Zero voltage on all 230 kV incoming lines OR

Loss of startup transformers (SUTs) 1C AND 1 D OR

Loss of startup transformers (SUTs) 2C AND 2D

LOSS OF ONSITE AC POWER CAPABILITY is indicated by: Loss of all emergency diesel generators on Unit One OR Unit Two for any reason

Loss OF OFFSITE POWER WITH LOSS OF ALL ONSITE AC POWER •15 MINUTES (on Unit One OR Unit Two) is indicated by:

All 4.16 kV buses (Unit One OR Unit Two) reading zero volts AC AND

The inability to energize at least one Unit One AND one Unit Two 4.16 kV bus WITH diesel generators

Loss OF OFFSITE POWER WITH LOSS OF ALL ONSITE AC POWER >15 MINUTES (on Unit One OR Unit Two) is indicated by:

All 4.16 KV buses (Unit One OR Unit Two) reading zero volts AC AND

The inability to energize at least one Unit One AND one Unit Two 4.16 kV bus WITH diesel generators

See Section 22.0, Multiple Symptoms and Other Conditions, for Determination of General

Emergency Classification.

END LOSS OF AC POWER

MGR-0001 Rev. 3

SOUTHERN NUCLEAR PLANT E.l. HATCH

IPAGE 15 OF 46

DOCUMENT TITLE: DOCUMENT NUMBER: REVISION/VERSION EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 -No0

14.3

6.0 - LOSS OF ONSITE DC POWER

Emergency conditions exist WHEN: N AS SG

A LOSS OF ALL VITAL ONSITE DC POWER OCCURS and is indicated as follows: E E E N R T

A LOSS OF ALL VITAL ONSITE DC POWER OCCURS FOR < 15 MINUTES as indicated by:

Low voltage AND/OR fuse trouble on ALL the affected unit's 125v/250v station batteries AND Low voltage AND/OR fuse trouble on the affected unit's 125v D/G batteries (including the swing D/G)

A LOSS OF ALL VITAL ONSITE DC POWER OCCURS FOR > 15 MINUTES as indicated by:

Low voltage AND/OR fuse trouble on ALL the affected unit's 125v/250v station batteries AND Low voltage AND/OR fuse trouble on the affected unit's 125v D/G batteries (including the swing D/G)

END LOSS OF ONSITE DC POWER

MGR-0001 Rev. 3


PAGE 16 OF 46"

DOCUMENT TITLE: DOCUMENT NUMBER: IREVISIONNERSION EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NO4

14.3

7.0 - LOSS OF CONTAINMENT


T

NOTE NUE is to be declared upon commencing Load Reduction.

A LOSS OF PRIMARY OR SECONDARY CONTAINMENT INTEGRITY OCCURS as indicated by the inability to meet any one of the requirements WITHIN the time limit established by the applicable unit's TS.

See Section 11.0, Hazards to Plant Operation, for determination of Alert Classification.

See Section 11.0, Hazards to Plant Operation for determination of Site Area Emergency Classification.

See Section 22.0, Multiple Symptoms and Other Conditions, for determination of General Emergency Classification.

END LOSS OF CONTAINMENT

MGR-0001 Rev. 3


EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NO03 114.3

8.0 - FIRE IN PLANT

Emergency conditions exist WHEN: NA S G U LA E

EEE N R

A FIRE CONTINUING > 10 MINUTES (AFTER DISCOVERY) EXISTS WITHIN THE PROTECTED AREA. INCLUDING 230 KV AND 500 KV SWITCHYARDS, as indicated by:

Fire Alarm WITH visual confirmation OR SOS/ED judgment

NOTE Refer to the System Evaluation Document (SED) for a listing of safety systems.

A FIRE CONTINUING > 10 MINUTES (AFTER DISCOVERY) EXISTS POTENTIALLY AFFECTING SAFETY SYSTEMS, required for the present mode of operation, as indicated by:

Fire Alarm AND Location, observation AND judgment of SOS/ED

A FIRE CONTINUING > 10 MINUTES (AFTER DISCOVERY) COMPROMISING THE FUNCTIONS OF SAFE SHUTDOWN SYSTEMS as indicated by:

Fire defeating redundant safety system trains required for the current mode of operation OR Loss of safety system due to fire that affects shutdown capability by the inability to perform ONE of the following functions:

* Prevent excessive reactor pressurization "* Provide adequate makeup inventory "* Depressurize the reactor "* Remove decay heat from the reactor

OR Location, observation AND judgment of SOS/ED

See Section 22.0, Multiple Symptoms and Other Conditions for determination of General Emergency Classification.

END FIRE IN PLANT

MGR-0001 Rev. 3

SOUTHERN NUCLEAR PAGE 18OF46 PLANT E.I. HATCH

DOCUMENT TITLE: DOCUMENT NUMBER: REVISION/VERSION EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NO:,

14.3

9.0 - SECURITY EVENT


R T

"A SECURITY ALERT OCCURS as indicated by Nuclear Security Shift Supervisor advises SOS/ED of Security Alert condition AND SOS/ED judgment

"A SECURITY EMERGENCY OCCURS as indicated by: Nuclear Security Shift Supervisor advises the SOS/ED of a Security Emergency condition AND SOS/ED judgment

"A LOSS OF PHYSICAL CONTROL OF THE PLANT IS IMMINENT as indicated by: Loss of physical barrier capability or control of the protected area

OR Attempted unauthorized entry into the protected area by force or covert action AND SOS/ED judgment based (n Nuclear Security Shift Supervisor advice

"A LOSS OF PHYSICAL CONTROL OF THE PLANT IS IMMINENT as indicated by: Loss of physical barrier capabilities of any vital building

OR Loss of control of any vital area including:

"* Intake Structure "* Main Control Room "* Diesel Generator Bldg. "* CAS/SAS "* Power Block

AND SOS/ED judgment based on Nuclear Security Shift Supervisor advice

END SECURITY EVENT

MGR-0001 Rev. 3

PLANT E.I. HATCH I I DOCUMENT TITLE: DOCUMENT NUMBER: REVISION/VERSION

EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NO4s 114.3

10.0- NATURAL PHENOMENON


EARTHQUAKE DETECTED:

ANY EARTHQUAKE IS DETECTED WITHIN THE PLANT as indicated by:

Felt by Personnel OR Confirmed "Seismic Instrumentation Triggered" (Unit 1) alarm indicating horizontal acceleration > 0.005 g

ANY EARTHQUAKE IS DETECTED WITHIN THE PLANT nsR inrlinn•tnr h'&

"Seismic Instrumentation Triggered" (Unit 2) alarm indicating horizontal acceleration > 0.08g Operating Basis Earthquake (OBE Level)

OR Any horizontal (N-S, E-W) peak shock annunciator 12.7 hz AMBER light illuminated indicates 100% OBE actuated on Panel I H1 1 -P701

AND

"Seismic Instrumentation Triggered" (Unit 1) alarm indicating horizontal acceleration > 0.005g

OR Unit 1 AND/OR Unit 2 Seismic Peak Shock Recorder High "G" Alarm

OR Unit 1 AND Unit 2 Time-History Recorders start

1.-

-4 -> [CONTINUE TO THE NEXT PAGE]->

MGR-0001 Rev. 3

SOUTHERN NUCLEAR [PAGE 19 OF 46


EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0REVISION/VERSION

NOt-

10.0 - NATURAL PHENOMENON, (continued)


EARTHQUAKE DETECTED: (continued) E E E N R T

NOTE The actual maximum g acceleration may be determined by having I & C play back the Time-History Recorder's tapes per the Earthquake Response Manual, SX18271 (located in Document Control) and the applicable I & C procedure(s).

ANY EARTHQUAKE IS DETECTED WITHIN THE PLANT as indicated by:

Same parameters as in the Alert classification AND Any horizontal (N-S, E-W) peak shock annunciator, 12.7 hz RED light illuminated on Panel 1 H11-P701 indicating maximum g level measured by Time-History Recorders as Ž 0.15g Design Basis Earthquake (DBE) AND EITHER unit NOT in Cold Shutdown

AN EARTHQUAKE THAT COULD CAUSE MASSIVE DAMAGE TO ANY PLANT SYSTEM WHICH COULD LEAD TO CORE DEGRADATION OR CORE MELT as indicated by:

Loss of systems needed to maintain integrity of all three fission product barriers:

* Fuel Integrity * RCS Integrity * Containment Integrity

OR Observation and judgment of SOS/ED.

END - EARTHQUAKE

-- [NATURAL PHENOMENON - CONTINUED TO NEXT PAGE]-->

MGR-0001 Rev. 3

I

SOUTHERN NUCLEAR PLANT E.I. HATCH IPAGE 21 OF 46

DOCUMENT TITLE: DOCUMENT NUMBER: EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0

10.0 - NATURAL PHENOMENON, (continued)


HIGH WINDS EXIST:

HIGH WINDS are indicated by:

Any tornado observed onsite OR Any hurricane force winds projected onsite with windspeed > 75 mph

Any tornado observed striking the operating facility (areas within the protected area and the 230 Kv and 500 Kv switchyards) OR Any hurricane observed onsite with sustained windspeeds at design level (> 94.5 mph) OR SOS/ED judgment

REVISION/VERSION NOt 14.3

The observation of damage from an onsite tornado with windspeed in excess of meteorological instruments range (>100 mph)

OR Sustained windspeeds in excess of meteorological instruments range (>100 mph)

AND Either unit NOT in Cold Shutdown

END - HIGH WINDS

-> [NATURAL PHENOMENON - CONTINUED TO NEXT PAGE]"-MGR-0001 Rev. 3

CAUTION The wind speed instrumentation will not reflect the actual wind speeds

of a tornado. Consideration should be given to the distance of a reported tornado from the met tower and the extent of the reported

damage when attempting to determine if the wind speed "exceeds the range of the instrumentation (> 100 mph)".

I

I

SOUTHERN NUCLEARPLANT E.I. HATCHDOCUMENT TITLE: DOCUMENT NUMBER: REVISION/VERSION

EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NO114.3

10.0- NATURAL PHENOMENON, (continued)

Emergency conditions exist WHEN: NAS G ULA E

HIGH / LOW RIVER WATER LEVEL INDICATED: E N R T

HIGH RIVER WATER LEVEL is indicated by:

Plant Service Water Intake Pump well level indication > 88.6 ft Mean Sea Level (MSL)

Plant Service Water Intake Pump well level indication > 100 ft MSL

Plant Service Water Intake Pump well level indication > 120 ft MSL OR

Actual OQ projected hurricane surge ORF flood levels > 120 ft MSL AND Either unit NOT in Cold Shutdown

LOW RIVER WATER LEVEL is indicated by:

Plant Service Water Intake Pump well level indication < 60.7 ft Mean Sea Level (MSL)

Plant Service Water Intake Pump well level indication < 59.9 ft MSL Plant Service Water Intake Pump well level indication < 57.2 ft MSL


END - HIGH / LOW RIVER WATER LEVEL

END NATURAL PHENOMENON

MGR-0001 Rev. 3

PAGE 2 OF 6


PAGE 23 OF 46


14.3

11.0 - HAZARDS TO PLANT OPERATION


AIRCRAFT ACTIVITY

UNUSUAL AIRCRAFT ACTIVITY IS OBSERVED over the operating facility (areas within the protected area and the 230 Kv and 500 Kv switchyards)

OR AIRCRAFT CRASH OCCURS within the owner controlled area AND SOS/ED judgment

AIRCRAFT CRASH OCCURS WITHIN THE OPERATING FACILITY (areas within the protected area and the 230 Kv and 500 Kv switchyards)

AIRCRAFT CRASH OCCURS AFFECTING VITAL OPERATING PLANT STRUCTURES by impact OR fire including:

* Intake Structure"* Main Control Room "* Diesel Generator Bldg. "* CAS/SAS * Power Block


OR SOS/ED judgment

END - AIRCRAFT ACTIVITY

MGR-0001 Rev. 3

--> [HAZARDS TO PLANT OPERATION - CONTINUED TO NEXT PAGE]--->

SOUTHERN NUCLEAR PLANT E.I. HATCH IPAGE 24 OF 46DOCUMENT TITLE: DOCUMENT NUMBER: REVISION/VERSION

EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NO1 1 14.3

11.0 - HAZARDS TO PLANT OPERATION, (continued)


EXPLOSIONS

ANY EXPLOSION OBSERVED WITHIN THE OPERATING FACILITY (areas within the protected area and the 230 Kv and 500 Kv switchyards)

KNOWN EXPLOSION DAMAGE TO FACILITY (ONSITE) AFFECTING PLANT OPERATION

SEVERE DAMAGE TO SAFE SHUTDOWN EQUIPMENT FROM MISSILES OR EXPLOSION THAT AFFECTS SHUTDOWN CAPABILITY by the inability to perform ONE of the following functions:

Prevent excessive reactor pressurization OR

Provide adequate makeup inventory OR

Depressurize the reactor OR

Remove decay heat from the reactor AND Either unit NOT in Cold Shutdown

END - EXPLOSIONSI-�.��1�

-> [HAZARDS TO PLANT OPERATION - CONTINUED TO NEXT PAGE]J-

K--->

MGR-0001 Rev. 3

SOUTHERN NUCLEAR PLANT E.. HATCHDOCUMENT TITLE: DOCUMENT NUMBER:


PAGE 25 OF 46REVISION/VERSION

NO:14.3



TOXIC GAS RELEASED: E E N R T

OBSERVATION OF SIGNIFICANT TOXIC GAS RELEASE WITHIN the operating facility (areas within the protected area and the 230 Kv and 500 Kv switchyards) AND SOS/ED judgment

UNCONTROLLED TOXIC GAS ENTRY INTO PROTECTED AREA FACILITY ENVIRONS

UNCONTROLLED TOXIC GAS ENTRY INTO A VITAL AREA restricting access and constituting a safety problem:

"* Intake Structure "* Main Control Room "* Diesel Generator Bldg. "* CAS/SAS * Power Block


END - TOXIC GAS

MGR-0001 Rev. 3

NOTE Toxic gas releases may hamper the ability of personnel to perform activities related to plant safety. Releases within the protected area of the plant may jeopardize the operation of equipment or safety functions necessary to establish or maintain cold shutdown. Releases which may fall into this category include, but are NOT limited to Carbon Dioxide, Nitrogen and Chlorine.

CAUTION DO NOT LIMIT EVALUATION OF THE CONDITION BASED ON THE CHEMICAL DEFINITION OF THE MATERIAL IN QUESTION. THE WORD 'TOXIC" IN THESE EALS IS A BROAD CATEGORY OF MATERIALS WHICH HAVE THE POTENTIAL FOR LIMITING THE ABILITY OF PERSONNEL TO PERFORM WORK ACTIVITES ASSOCIATED WITH PLANT SAFETY.

-- [HAZARDS TO PLANT OPERATION - CONTINUED TO NEXT PAGE]-->

SOUTHERN NUCLEAR PLANT E.I. HATCH PAGE 26 OF 46DOCUMENT TITLE:

EMERGENCY CLASSIFICATION AND INITIAL ACTIONSDOCUMENT NUMBER: REVISION/VERSION 73EP-EIP-O01-0 N-0o

1 14.3


Emergency conditions exist WHEN: N A S G U L A E

FLAMMABLE GAS RELEASED: E E E N R T

NOTE Flammable gas releases may jeopardize the operation of equipment or safety functions necessary to establish or maintain cold shutdown.

OBSERVATION OF SIGNIFICANT FLAMMABLE GAS RELEASE WITHIN the operating facility (areas within the protected area and the 230 Kv and 500 Kv switchyards) OR PIPING RUPTURE IN ANY FLAMMABLE GAS SYSTEM (i.e., hydrogen, propane, etc.) OR SOS/ED judgment

UNCONTROLLED FLAMMABLE GAS ENTRY into any Protected Area facility environs

UNCONTROLLED FLAMMABLE GAS ENTRY INTO VITAL AREAS INCLUDING:

"* Intake Structure "* Main Control Room "* Diesel Gen. Bldg. "* CAS/SAS "* Power Block

AND Either unit not in cold shutdown

END - FLAMMABLE GAS

-> [HAZARDS TO PLANT OPERATION - CONTINUED TO NEXT PAGE]->

MGR-0001 Rev. 3


PAGE 27 OF 46


14.3



TURBINE FAILURE/MISSILE IMPACT EEEN R T

A TURBINE FAILURE GENERATING PROJECTILES is indicated by: Main Turbine Trip AND Confirmation of rotating component failure

OR SOS/ED judgment

A TURBINE FAILURE GENERATING PROJECTILES is indicated by: Main turbine trip

AND Turbine casing penetration by internal components OR Projectile from any source, affects plant operation OR SOS/ED judgment

END - TURBINE FAILURE/MISSILE IMPACT

END HAZARDS TO PLANT OPERATION

MGR-0001 Rev. 3


PAGE 28 OF 46

DOCUMENT TITLE: DOCUMENT NUMBER: REVISION/VERSION EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NO4t

1 1 14.3

12.0 - THIS SECTION INTENTIONALLY LEFT BLANK

MGR-0001 Rev. 3

SOUTHERN NUCLEAR 7 IPAGE 29 OF 46PLANT E.I. HATCH DOCUMENT TITLE: DOCUMENT NUMBER: REVISION/VERSION

EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NO:14.3

13.0 - CONTROL ROOM EVACUATION

Emergency conditions exist WHEN: N A S G U LIA E

AN EVACUATION OF THE MAIN CONTROL ROOM IS IMMINENT as indicated by: E E E N R T

Entry into the Remote Shutdown procedures used to shutdown the plant from outside the Control Room.

An evacuation of the Main Control Room is ordered AND Control of shutdown systems from local stations is NOT established within 15 minutes after Main Control Room evacuation.

END CONTROL ROOM EVACUATION

MGR-0001 Rev. 3


EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0NO"• 14.3

14.0 - CONTROL ROD DROP

Emergency conditions exist WHEN: NAS G U LAE

A CONTROL ROD DROP ACCIDENT OCCURS as indicated by: EEE N R T

Local power range monitors (LPRM) indicate abnormal neutron flux in the vicinity of the suspected dropped rod AND MSL high rad monitors > 3X normal background

OR Average power range monitor (APRM) upscale trip of RPS channels "A" and/or "B"

"* Unit 1 > 120% RTP "* Unit 2 > 120% RTP

OR Intermediate range monitor (IRM) upscale trip of RPS channels "A" and/or "B" Either unit Ž 120/125 divisions of full scale

END CONTROL ROD DROP

MGR-0001 Rev. 3

SOUTHERN NUCLEAR PLANT E.I. HATCH DOCUMENT TITLE: DOCUMENT NUMBER:


15.0 - FAILURE OF REACTOR PROTECTION SYSTEM


A FAILURE OF THE REACTOR PROTECTION SYSTEM (RPS) TO INITIATE A SCRAM as indicated by:

Valid automatic scram signal AND Reactor NOT subcritical OR subcriticality cannot be maintained

A FAILURE OF THE REACTOR PROTECTION SYSTEM (RPS) TO INITIATE AND COMPLETE A SCRAM which brings the reactor subcritical, is indicated by:

Valid automatic AND manual scram signal AND Reactor NOT subcritical OR subcriticality cannot be maintained

A TRANSIENT REQUIRING OPERATION OF SHUTDOWN SYSTEMS WITH FAILURE TO SCRAM (continued power generation but no core damage immediately evident) is indicated by

Valid automatic AND manual scram signal AND < 3% power generation cannot be achieved OR maintained AND Standby Liquid Control initiation required

See section 22.0, Multiple Systems and Other Conditions, for determination of the General Emergency Classification

END FAILURE OF REACTOR PROTECTION SYSTEM

PAGE 31 OF 46

REVISION/VERSION NO:14.3

MGR-0001 Rev. 3

SOUTHERN NUCLEAR PLANT E.I. HATCHDOCUMENT TITLE: DOCUMENT NUMBER:



NO>14.3

16.0 - LOSS OF CONTROL ROOM INDICATION/ALARMS/ANNUNCIATORS

Emergency conditions exist WHEN: FNTAT

ANY SIGNIFICANT LOSS OF ANY ONE OF THE FOLLOWING MAIN CONTROL ROOM INDICATION OR ALARMS, THAT REDUCE ASSESSMENT CAPABILITY TO THE EXTENT REQUIRING PLANT SHUTDOWN BY TS:

"* Plant Process Computer "* Safety Parameter Display System "* Radioactive Effluent Instrumentation

AND The plant NOT shut down WITHIN the time limit specified by TS

MOST OR ALL MAIN CONTROL ROOM ALARMS (ANNUNCIATORS) LOST as indicated by: Observation OR failure in alarm check OR SOS/ED judgment

MOST OR ALL MAIN CONTROL ROOM ALARM (ANNUNCIATORS) LOST WITH PLANT TRANSIENT INITIATED OR IN PROGRESS as indicated by:

Observation of plant transient (i.e., reactor trip, turbine trip, loss of feedwater,etc.) OR SOS/ED judgment

END LOSS OF CONTROL ROOM INDICATION/ALARM/ANNUNCIATORS

K>

MGR-0001 Rev. 3




14.3

17.0 - LOSS OF SHUTDOWN FUNCTIONS


E E E N

R T

A COMPLETE LOSS OF ANY FUNCTION NEEDED FOR PLANT COLD SHUTDOWN is indicated by:

Both trains of RHR shutdown cooling mode unavailable for any reason AND Loss of alternate shutdown cooling modes AND Inability to maintain reactor coolant temperature < 2120 F, WHEN required.

See section 22.0, Multiple Symptoms and Other Conditions, for determination of the General Emergency Classification

END LOSS OF SHUTDOWN FUNCTIONS

MGR-0001 Rev. 3

SOUTHERN NUCLEAR PLANT E.1. HATCH PAGE 34 OF 46


14.3

18.0 - FUEL DAMAGE BY FUEL HANDLING ACCIDENT


R T

A FUEL HANDLING ACCIDENT WITH RELEASE OF RADIOACTIVITY TO REACTOR BUILDING is indicated by:

Valid Refueling Floor ARM Hi Alarm > 50 mR/hr

OR

Valid "REFUELING FLOOR VENT EXHAUST RADIATION HI-HI" Alarm (601-403) AND

Any of the following process radiation monitors indicating > 20 mR/hr

"* 1D11-K611A-D "* 2D11-K611A-D "* 2D11-K634A-D "* 2D11-K635A-D

OR

Valid "REFUELING FLOOR VENT FLTR DISCH RADIATION HIGH" Alarm (601-42 AND

Any of the following process radiation monitors indicating > 20 mR/hr

"* 1D11-K616A,IB "* 2D11-K616A,B

MAJOR DAMAGE TO SPENT FUEL IN REACTOR BUILDING as indicated by: Spent Fuel Storage Pool Low Level Alarm

AND More than one Refuel Floor ARM exceeding Max Safe Operating Value OR Large object damages spent fuel in pool

AND SOS/ED judgment (based on refueling floor radiation levels)

END FUEL DAMAGE BY FUEL HANDLING ACCIDENT

MGR-0001 Rev. 3

SOUTHERN NUCLEAR [PAGE 35 OF 46 PLANT E.I. HATCH I DOCUMENT TITLE: DOCUMENT NUMBER: REVISION/VERSION

EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NOr14.3

19.0 - HIGH RADIATION OR AIRBORNE CONTAMINATION

Emergency conditions exist WHEN: NASG ULAE EEEN

R T

HIGH RADIATION LEVELS OR HIGH AIRBORNE CONTAMINATION WHICH INDICATE A SEVERE DEGRADATION IN CONTROL OF RADIOACTIVE MATERIAL is indicated by:

ARMs are offscale high (readings confirmed)

OR An increase by factor of 1,000 in direct radiation readings

END HIGH RADIATION OR AIRBORNE CONTAMINATION

MGR-0001 Rev. 3



PAGE 36 OF 46REVISIONNERSION

NOt-

20.0 - LOSS OF COOLANT


R T

NOTE NUE is to be declared based upon commencing Load Reduction.

ANY CONFIRMED REACTOR COOLANT SYSTEM (RCS) OPERATIONAL LEAKAGE AS DEFINED BY TS is indicated by:

Any RCS pressure boundary leakage ANY CONFIRMED REACTOR COOLANT SYSTEM (RCS) LEAK OR UNISOLABLE SYSTEM LEAK CAUSING THE DIRECT LOSS OF VESSEL INVENTORY GREATER THAN 50 GPM as indicated by:

Calculation of RCS leak rate greater than 50 gpm using Drywell Equip AND/OR Floor Drain Sump level integrators on Panel H11-P613 OR SOS/ED judgment that an unisolable RCS leak greater than 50 GPM into the Reactor Building has occurred and may be indicated by one OR more of the following indications:

"* Reactor Building Equip AND/OR Floor Drain Sump level high alarms "* Valid leak detection alarms "* Any confirmed ARM in the Reactor Building above Max Normal Operating

Values. OR SOS/ED judgment

ANY CONFIRMED REACTOR COOLANT SYSTEM (RCS) LEAK is indicated by: RCS leak greater than all available ECCS pump capacities AND Reactor low, low, low level alarm < -113 inches AND level decreasing with available makeup pumps running and discharging to vessel AND Drywell High Temp Alarms AND Drywell temperature increasing

OR Drywell high pressure initiation alarm > 1.92 psig AND increasing

See section 22.0, Multiple Symptoms and Other Conditions for determination of the General Emergency Classification

END LOSS OF COOLANT

MGR-0001 Rev. 3

SOUTHERN NUCLEAR [PAGE 37 OF 46 PLANT E.I. HATCH DOCUMENT TITLE: DOCUMENT NUMBER: REVISION/VERSION

EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NO:14.3

21.0 - LOSS OF ENGINEERED SAFETY FEATURES

Emergency conditions exist WHEN: NAS G UAlE

EEEIN

T THE LOSS OF ENGINEERED SAFETY FEATURES (ESF) WITH CONTINUED OPERATION OF EITHER UNIT BEYOND THE TIMEFRAME SPECIFIED IN THE APPLICABLE TS REQUIRED ACTION STATEMENT (RAS):

The following are engineered safety features (ESFs): "* Automatic Depressurization System "* Containment Heat Removal System "* Containment Isolation System "* Control Rod Velocity Limiters "* Core Spray "* CRD Housing Supports * Diesel Generators "* High Pressure Coolant Injection System "* Low Low Set Relief Logic System "* Low Pressure Coolant Injection System "* Main Control Room Environmental Control System "* Main Steam Line Flow Restrictor "* Main Steam Line Isolation Valves "* Post LOCA Hydrogen Recombiner System (i.e., Combustible Gas Control

System) * Reactor Protection System * Standby Gas Treatment System

END LOSS OF ENGINEERED SAFETY FEATURES

MGR-0001 Rev. 3


IPAGE 38 OF 46


14.3

22.0 - MULTIPLE SYMPTOMS AND OTHER CONDITIONS


TECHNICAL SPECIFICATION SAFETY LIMITS ARE EXCEEDED: R T

PLANT CONDITIONS THAT EXCEED ANY SAFETY LIMIT AS REQUIRED IN TS are indicated by the following categories:

Thermal Power OR

Minimum Critical Power Ratio (MCPR) OR

Low reactor water level with irradiated fuel in the reactor vessel < -139" in Unit 1 OR < -158" in Unit 2 OR

Reactor vessel steam dome pressure > 1325 psig with irradiated fuel in the reactor vessel OR

Other condition that in the SOS/ED judgement warrant increased awareness of the plant operating staff or State and/or local authorities.

END - TECHNICAL SPECIFICATION SAFETY LIMITS

-> [MULTIPLE SYMPTOMS AND OTHER CONDITIONS

CONTINUED TO NEXT PAGE]-> ->

MGR-0001 Rev. 3


EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-0010 -No3 14.3

22.0 - MULTIPLE SYMPTOMS AND OTHER CONDITIONS, (continued)

Emergency conditions exist WHEN: N A S G U L A E,

PRECAUTIONARY ACTIVATION OF TSC IS WARRANTED: EREN

T Plant conditions exist that warrant precautionary activation of the TSC and placing the EOF AND other key emergency responders on standby, as indicated by the following:

Observation

AND

SOS/ED judgment

END - PRECAUTIONARY ACTIVATION OF TSC

-- [MULTIPLE SYMPTOMS AND OTHER CONDITIONS

CONTINUED TO NEXT PAGE]+-->

MGR-O001 Rev. 3


PAGE 40 OF 46

DOCUMENT TITLE: DOCUMENT NUMBER: REVISION/VERSION EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NOt

1 14.3

22.0 - MULTIPLE SYMPTOMS AND OTHER CONDITIONS, (continued)


PRECAUTIONARY ACTIVATION OF MONITORING TEAMS IS WARRANTED:

Plant conditions exist that warrant activation of emergency centers and monitoring teams, OR a precautionary notification to the public near the site, as indicated by the following:

Observation

AND

SOS/ED judgment

END - PRECAUTIONARY ACTIVATION OF MONITORING TEAMS


CONTINUED TO NEXT PAGE]'• --->

MGR-0001 Rev. 3


IPAGE 41 OF 46

DOCUMENT TITLE: DOCUMENT NUMBER: REVISION/VERSION EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-O01-O N403

1 1 14.3

22.0 - MULTIPLE SYMPTOMS AND OTHER CONDITIONS (continued)

Emergency conditions exist WHEN: N A SG U L AE

POTENTIAL LARGE RELEASE OF RADIOACTIVITY EXISTS: E E E N R T

PLANT CONDITIONS EXIST WHERE THE POTENTIAL RELEASE OF LARGE AMOUNTS OF RADIOACTIVITY IN A SHORT TIME PERIOD ARE POSSIBLE (e.g., any core melt situation) is indicated by the following conditions:

Transient (e.g., scram, loss of offsite power, etc.) AND

Failure of required core shutdown system (could lead to core melt in several hours) [e.g., CRD system, SLC system, RPS, ECCS, DG'S, RHRSWJ

AND

Containment failure likely OR

Small or large LOCA AND

Failure of ECCS to perform (leading to core degradation or melt in minutes to hours)

AND

Loss of containment imminent OR

Small or large LOCA AND

Containment performance is unsuccessful (affecting longer term success of ECCS. Could lead to core degradation OR melt in hours)

---- [CONTINUE TO THE NEXT PAGE]---

MGR-0001 Rev. 3

SOUTHERN NUCLEAR PLANT E.I. HATCH PAGE 42 OF 46

DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NO

114.3


Emergency conditions exist WHEN: N ASG U LAE

POTENTIAL LARGE RELEASE OF RADIOACTIVITY EXISTS: (continued) E EEN R T

OR

Shutdown occurs AND

Required decay heat removal systems (e.g., RHR) are rendered unavailable or non-safety systems heat removal capabilities are rendered unavailable

AND

Core degradation OR melt could occur in about ten hours WITH subsequent containment failure OR

Any major internal OR external event which could cause massive damage to plant systems resulting in any of the conditions listed in multiple symptoms of potential larger releases of radioactivity OR

SOS/ED judgment

END - POTENTIAL LARGE RELEASE OF RADIOACTIVITY


CONTINUED TO NEXT PAGE]-> --

MGR-0001 Rev. 3

DOCUMENT TITLE: DOCUMENT NUMBER: REVISION/VERSION EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 N0.

14.3


Emergency conditions exist WHEN: N ASG U LAE

FIRE IN PLANT OCCURS: E EEN R T

PLANT E.I. HATCH

A FIRE IN THE PLANT THAT COULD CAUSE MASSIVE DAMAGE TO ANY PLANT SYSTEM WHICH COULD LEAD TO CORE DEGRADATION OR CORE MELT as indicated by the following:

Loss of systems due to fire, needed to maintain integrity of all three fission product barriers.

"* Fuel Integrity "* RCS Integrity "* Containment Integrity

OR Location, observation AND judgment of SOS/ED (Based upon Fire Brigade Leader's report.)

END - FIRE IN PLANT

->[MULTIPLE SYMPTOMS AND OTHER CONDITIONS

CONTINUED TO NEXT PAGE]--"

MGR-0001 Rev. 3

SOUTHERN NUCLEARPAGE 3 OF 6

SOUTHERN NUCLEAR PLANT E.I. HATCH PAGE 44 OF 46DOCUMENT TITLE: DOCUMENT NUMBER: REVISION/VERSION

EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NO4

22.0 - MULTIPLE SYMPTOMS AND OTHER CONDITIONS (continued

Emergency conditions exist WHEN: ULAE

Any of the following are indicated, using the Parameter Assessment Table below: R T

A Failure of the Fuel Cladding ANDPrimary Containment with a potential lass of the Primary Coolant Boundary

OR A Failure of the Fuel Cladding AND Primary Coolant Boundary with a potential loss of Primary Containment

OR A Failure of the Primary Coolant Boundary AND Primary Containment with a potential loss of the Fuel Cladding

A General Emergency should be declared when TWO boundaries (cladding, coolant, or containment) have an ACTUAL failure AND a THIRD boundary has an ACTUAL or POTENTIAL failure. IF a parameter is approaching emergency action level criteria and mitigation systems are unavailable, assume the barrier will be lost. Exceeding ONE of the parameters below is an indication of an actual or potential loss of the associated boundary.

PARAMETER ASSESSMENT TABLE

LCLADDING COOLANT CONTAINMENT

Actual Actual Actual O 1-131 > 100gCi/cc 01 Unisolable primary system break CI Integrity breached

outside containment E DWRRM > 500 R/hr 0 Significant leakage in TB With TB I: Drywell OR Torus Ž 6%

ARMs above alarm setpoints and hydrogen with Ž> 5% oxygen increasing.

O1 DW Pressure >Ž25 psig 0 SOS judgement that containment is lost OR loss is imminent

O1 DW Temperature > 300OF OI Gap activity in DW

Potential Potential Potential "O Failure of ECCS to 0] Failure of SRVs to open with pressure E Containment pressure

maintain RWL high off-scale approaching 56 psig "O RWL < -158" for 3.5 min 11 AIl4160/600 V buses undervoltage E4 Drywell OR Torus Ž6%

AND MCUTL hydrogen with Ž 5% oxygen

All 4160/600 V buses undervoltage

El Failure of ECCS to maintain RWL r-

-d is-i ii-�

SOS/ED judgement that containment loss is imminent

END MULTIPLE SYMPTOMS AND OTHER CONDITIONS

MGR-0001 Rev. 3

SOUTHERN NUCLEAR PLANT E.I. HATCH IPAGE 45 OF 46

DOCUMENT TITLE: DOCUMENT NUMBER: REVISION/VERSION EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NOO

14.3

23.0 - ISFSI OPERATIONS


T A Loss OF CASK CONFINEMENT BOUNDARY FOR ANY LOADED SPENT FUEL CASK OCCURS as indicated by:

Direct Radiation levels outside the ISFSI protected area boundary exceed 2 mrem in an hour

AND Contamination levels outside the ISFSI protected area boundary exceed the technical specification limits for spent fuel storage cask surface contamination

OR

Direct Radiation Readings for a Loaded Spent Fuel Cask exceed the technical specification limit for overpack average surface dose rates.

DEGRADATION OF ANY SPENT FUEL CASK DUE TO AN OPERATIONAL EVENT as indicated by:

Direct observation of a loaded spent fuel cask indicates cask confinement boundary or shielding damage due to an operational event

"* Cask handling "* Cask drop "* Cask tip-over

AND

SOS/ED judgment

"--' -> [ISFSI OPERATIONS - CONTINUED TO NEXT PAGE]-> ->

MGR-0001 Rev. 3


PAGE 46 OF 46


1 14.3

23.0 - ISFSI OPERATIONS (continued)

Emergency conditions exist WHEN: NA S G ULIA E E]EE N

R T

A Loss of cask confinement boundary for any loaded spent fuel cask occurs as indicated by: Degradation of any Spent Fuel Cask due to environmental phenomena or external events

Direct observation of a loaded spent fuel cask indicates cask confinement boundary or shielding damage due to environmental phenomena or external events

* Tornado * Explosion * Lightning * Flooding * Earthquake * Extreme environmental temperatures * Burial under debris * Fire * Explosion * Aircraft Crash * Missile or projectile impact * Security Event

AND

SOS/ED judgment

END ISFSI OPERATIONS

MGR-0001 Rev. 3

3 4 5 7 8

SC - SECONDARY CONTAINMENT CONTROLPLANT E. I. HATCH

"' an, 70ECON .~ cn N TAsn MumE NTICOG RAP,

310O-EOP-014-2S REV. 6 ATTACHMENT 1

BWROG ERG/SAG REV. I

Area a Iblenlor diflfrealial temperatur e aboiabOe Ame. or IWAC exhaust radtla Ievel a bove Ofe ilal Pessam atOrove

A Table4 Mxm umo raOaainTeprlueabe5ximiu Normals Ororaling WalerLeve Table 6 Maximoum NoG alOerm a tn g r aFaliot oLevel 0In. of ivaterA

L E PRIMARY CONTAINMENT FLOODING IS OR HAS BEEN REQUIREC

NG EFOLLOWING enter the Severs Accident Guieldates


8C r SOI SCLSc

Epsrana availaWl are coolers WNAIT TLl WAIT UNTIL

in affboveelae rado$iEOfon lelma

[- a necordasy vmntasvNv.r radiation MaMaluvm Normal Operainng WredevalO Labol mcndir w~ut etion dos O)tts (TableS) 6)lvnNrsa pveigRdsg ee

0rdean T t------- A A or is b waler lveable)

Sopaae le Re l0owrt A am atr "o Refuel Fsoo -SAC pe r 34 SC,.T4 1.00 0.2f

"

per B40 idT4 1 HVA Operale rvailable bump pampe to restore Isolater ALL syslems discharging Into area rOs maintain waterlevel beloo Maximum EXCEPT aystemo meLied Is: Normal 0 arab V•Wsal Level (Table5)I

o assureadaviaua bm ose•Ting

WAIT UTnha d.n reactor WATUTI supportss fin,

JE NE f te flisyCANOT he hOmainti primaty counlenmest

area ambiento dourerential temperature Ir OetofE, aC TdoI ned oCnNOrbo MXm Inr Noarmal Opemsng Waler Level (Table 5): t

Maximum Normal Oparaling Temperature A My to drain bump water level (Table 4) o -- aY area waler level PERFORM CONCURRENTLY

WA sump orIT UNTIL WAITEUNTIL

I solate ALL systems d i na rnto dOe•

Nol level

EXCEPTiSysmary system I a•o ve

. nur wrmw uig liso .. w dSlcatarging ra.y, roola Maximum Sale Oparalir Redalin Lvel o ssecl ue°o coigo mahundOnptmaryorlllfln tl actaybiariataetata ~eae shutrawt reapeor 0 3ppre- I1no1

o rupma mre 0 maolnprlmarryromlrmment Mm ar lir Raao Le I maintaids pharrg•ingtramernt iclan (al

i i ai r ts my a RpERFORMLCONCURRENTLY

SPERFORM CONCURRENTLY Shu wer r=ea4cPu 4oPr3"S

LRC(A) poirtA

hto re°0€P-ee WAT UNTILOpS 1GE2

ANY TUNTL ..... UNTIL

MY

di~eon u•area

rdiatio1n level

reahes Mau te Ipe tgMdmum Safe Operaling Radiation Level

T e m p r a l le ( a b le 4 ) II s m o r e mh a n s ale a r ea

isO e o b n r w unB r p a n n n w us( - ) n m r . a PERFORM CONCURRENTLY

(Tabl(abl 7)(T).a

BEFORE shuu w n .R R " R A D IO A C T IV IT Y R E LE A S E C O N T R C

Erandom muR Safe rEaUiREg sG PR

s m R I(.

Opsite rad releoyase rate

Thomson~bov I 57l 4)mnm.hio

PEFR IOCRRNL PRMAYCNAINbTLOOI 6) xshsf~adelrvi

PRWOILE PERFORMINCNTHCURRENTIY

R

I IMAy CONTAINMENTFLOODING - SeveAcdentGuidelines IS GEit°

vihilLE~ ~ ~~G PmFOJI4 THEFALSSN

essb son TUEN nrestar Turbine ODinuie HVAO as r l V h raqimd Per34SOU41-WI"2S

solate ALL pemery systerm dads

SECONDARY CONTAINMENT acsexr lolmolantinomeans ol•a

Table 6 OPERATINGRADIATION LEVELS areid•coryc°antigy alreameowtdTeN Ma~x Normal Max Safe EXCEPT systemso required to:

HVAC EXHAUSTRADLAT[ONANNUNCIATORS onA tlng 0 assure adequalesa coteoling on 21111-1`601 Vassu V&ler 0 °sUt door, mosm,~

Table 4 SECONDARY CONTAINMENT r Inte imay

OPERATING TEMPERATURES HIeH0

RadIl-N NIA as. Norm ax an uSremdeoat v

AMBIENT TEMP (2B21BS-1S7) Operwang )Pran RACTORRUII INRn u WAIT UNTIL n ZH -ti-P614, 2G 31 .-R 604 Value V alue

f/•llI ae°

on 2F - RE BLOG POT CONTAM AREA RAOIATION 9S NoA prmory system In Table?7 A(211-KOOSA-O) In .rlet oxolart 1- EL EVATION AREAIRWORU SECONDARY CONTAINMENT REFUFLINO Fl 0OR

A A mnptoomwSoukln(2331.NGI6A) ISO 215 Table 5 OPERATINGWATERLEVELS oti0sWd patmiseal

Opump roam (Noilh) (2G31.NGI6B) I5O 215 1, /

2 Hx p mpro w (o nn(2 4331-N 0160 n 1 SO 215 -o~ n M xa e -REFUEL ING FLOOR VENT EXHAUST RADIAT ION

4 HxRoo(2G31.NolO0) a50 215 AREA WATER LEVEL ANNUNCIATORS Opemong Orreeg I2OI1-KS11A-f)

5 Pp.Gap T!,bRowm(23314006E) % 2H217 Melee Malue REFUELING FLOOR VENTE EXHAUST RADIATION 6.9 WVA

TF.FLENVATiONARPAIRWCiUS 5.7 GoBEFORE

MT-r 8 ~ojesEPOrE 6 RWCU ValNet 4(2331-i0160F 150 215 Man.THW sTOiAGONAS ARFAabo le REFUEULNG FLOOR VENT EXHAUST RADIATION 0.7 OVA .d)

.

"NR-A- A S1 RS-WIAGONAL FLOOR ORN SUMP LEVE Nig H al 87' W.

IT T Nno HIGH (657-033) (2T45N007) Macormea Max Sure

7 R.R..CSA. ....SOGA) -.. 190 2 CR0 S.W DIAG INSTR SUMP LVL High AREARADIATIONPERFORM CONCURRENTMSr

- OLTHEAST DIAGNA AR --- HIGH.HIGO-HIGH (RI7-32) (2T45-NHo5) High on 2HI1.P1,00 2D21-P600 Vatee Valne

8 RHACS 150_ 190 Hih nalmir mlr IL RC(A) pot

8 R HPJ C iPODMARFA B-1"-..._.RFFI.Fi.OONA• i S... .. ..-.. )LTHAy !15i, 1 Rh dorhsilomn e (OND21-K601A) 50 ION0

Pump Room (2E40-NG24) 167.5 245 SOUTHET DIAGONalyAREA I , Eeorh l (20)1-EA 1000 I

10 ESoertsea Cur (2E41.No3OA) 107,5 245 3 RE S-E OIAGONAL FLOOR DRN SUMP High ao 2ert FIpator&eoo u202121- 50E N E 11 EmarAxsaClr(2E41-N0S)5 167.5 245 High 67 a ent F.elPel&etifita(2021-K1NAtM 50 1003 I•S. NUý1T 25 HG-IOH-hl(65?-034)(2T4)-tNON) Pah ReWaalser V gml nfsoori fO21K1) MI o nS 1000 C MRECO5RSiRG

5 RaaddXrVieeflaIitxiaah Romr(2D21-t03llL) DO VOK * S ........ •_+• ....... . 31 4 FJRH-ýS-EDAB GNSTRSUMPUV1 High •l or•Z-~iL 0 10

12 Pump ROom (2E61-N01 i675 310 JIR.CS saM H TRsAel High

13 EnstAeaCJr(2S1 T3A) 167.5 310 HtGH-HIGH.NtON(657103)(2T46N0o3E) High 3OAiFICVATOONARAImM ASý I

54High 6 Caflrapalr anr(2021-iKatIT) 50 1t11

I __otra 22 • -KoI 50 o Table 7 pRIMARySYSTEMS

in West Wall (2E51-N025A) 1a7,5 2125 NORTHWFST DIAGONAL AREA In.T 16 tC Nhwy OtAG1 HE14 SU5MP1 LW High aove A •biIEIbSE Pitmeny system daichooging intoew Iam i dettnad CE

17 Nortreast Wall (2E51-No2SC) 117.5 2125 hr atISUPPHhW10AORlEA

85 SoutheaslWahi(2E51-NtO256LC 107.5 2126 HIGH.HIGH-HIGH (657-014) (2T45-N004) High a .e. Seetlfea roo passgeway (2D21 1OP)00R OWED ....... ...... . -... High P R 185iepera .(tD21 o2O21-K0 R) 5A 1003

TMON STEAM I INS TUNNELAREA RB 18n sepatl pael area (2021-K601S) 55 10500e CR0 0 ROiC

1 -- I -92. 11 RBIT5RWCU crarcra1xneIO22-KatlU) 15t 1000 o CoxeSplay o Reaclorsaamplin

.... . RRWCU 112.5 310 NOITHEASTDIAGOvNALARrA 141.. o FeedwCter 0 RPR

H. . . . . .. .. . .P.N.TR ..TION.. . . .... .abo ve 165' EL SMAT ION A tEA iN O RTH ) H C

5ION ROOM 6 RHRCSEOBIAS INSTR SUMP LVL High she" 12 RE 15a, NE(2021.K601C) 50 1000 o edanshenn o SBLC

I 2H4-N S HIOhtGH-HIGH (657-086)(2T45-NO3A) High V.I. 12 RE 158 area N-W(2021-l(01O) 50 1000 0 Malnsteals drns

2E41.N046B 167.5 21S. H-Jh 1_, EI.FVATION AREAIOUTHt Asystem in considered a"Pmary tysem-f I satng •rny"Ohn, V•I• • [2D '1--- -16)I50 1000 m~easure mon, mosesc the Ile at ie mare fte unhares:

DI"? TEMP (2821g0$2) Operaotng IraEn 14 In. 14 RhlSetepan.E(202I-ntatlE) 15r m

on 2H11 .P614, 2 03 i-R608 Va lue V ia.e lag er 8b w 2 2 .~ 0L0 1000

T T oF37 HPCI COMPT INSTR SUMP LEVEL ao, l, S p 156

HIGHlHIGH (657-6) (2f4"5OOl) High b SECONOARy CNTAINM ENT HVAC 1 N M FI VAT iO N A R EA , (RW CU 1

!T EL E WAT O N SFAR SA( O RI AMHW ST)

I Apumproom(2331-N025ArN023A) 67 Op 1 Table 14 EXHAUST. .A..0TIONISOATIO

SE ______ ils. _____________ -HAC EHAUS RIAON ANUNIATORUS

pump room (2G31.N02204N023B) 67 D9n 16 Tnp axee (20214o601F 5VACE10AUTRADATONANNU 3 Nc ROOm (2G31 .NO22C3INOC) 67 99 5 on 1(2)Hl6 -P0t

4 H. Room (2G31.N0,22O'NO23D) 07 99 8 TORUS N-EARFEAINSTRSUMPLVL High 8b.1 1 13 EyryVATIOArVA(2N3THbEAT)

5 Phas.SepTkROsm(2$G1-14)22Et023E9 67 en HIGH.HIGt.HIGHH(6t7013)(2ZT4t002A) High UNIT 17 RB135NEwod eX2D21 1) St 1003DT pOTC AM

PIE Ft FVATION ARFANIRWCUI Hg13 ELEVATIONhARFAISOUTHEAST)

6~1 RWCUCRHU(121Kl Valve NetfO1502FN2P 7 62 2FN2 - 9 TORUSS-EAREAiNSTRSUMpLW Nigh 16 SeuthCROHCU(2D21-Kt01N) 50 v000 -UNIT1 REFUELINGFLOORVENTEX AUST

NORTHEAST DIAONSAL HIGH-HIGH.-IGH (657.$1)( 2T45-NOO2S) High 130 ELEVAIONREAISOUTK'vWS7)

7 RHK/CSA(2EII.N02SAJN0NA) 40 74 High g RB 13 S.W eorsking area (2021.K6011) 50 100N SO. ....... IA.O.l.ARA.. UNIT 2 M BLDG POT CONTAMAREA

. .. . AV S n" AGES~ i ,3N•L A . . . . . . NOR RTHWEST IAGON &LAREA (2D011-I6n A'D

8 RIDIRIS B (2ElI-NG29BSE MSOB) 40 74 10 TORUS N-WAREA INSTR SUMP LVL High 20 RCIO equimeatt N.W diagoINa (2021-K60t) St 1000

.. .. . . . HIGS ROIGOI- N65704) (2T45-N2C Hih .-.-.... .. .- - UNIT 2 REFUELING FLOORVENT EMXHAUST High N oTH ES O ISOALD• A 2 14 611A -D )

13 2ES1-NO26ANNO27A 42 98 Nigh 21 SW iagona (2021LWE 50 1009

14 2E51-NO26Bt/ii27B 42 98T..A. . RA UNIT 2 REFUELING FLOOR VENT EXHAUST

10 2E51-NO2600tD27O 42 By 11 TORUS S.WAREMINSTR SUMPLVL High T A R&R1RN-Ed-g034I(2021-K6oIX) 50 1l0

16 .............. .... ....... 4 ....... HIGN.HIGH.HIGH (t5707) (2T454NO0 2

) High . C UNIT 2 REFUELING FLOOR VENT EXHAUST

MAIN STE•M LIN TUNNE L ARPA High SUTHEAST DIAGONAL/'REA -N 2D1R.F HOFU)

17 2621.NO16N016B 70 I0N 23 CS&RHRS-EdLn I2D2t-K001bI I5 1000w

1 2 1 3 4 5 6 7 8 9

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SANY Unit I Or Unit 2 secondary cnal amen I coesfirmo r HVAC diat e r o adioin le exeNO s axi a Unit I aRm UnIt 2 Reactor Suld g

plerr selp3nt 100bO 14) 1WAS ioa U i~t 1 and Unit 2 Rfouel Fluor IIIHVAC Iveso~ain

I OUnlit I a nd Unift 2 SBGT mi ntamin

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AND 3480-T4 1O06-2S

A U Ast 1 Or U oft 2 reoonmery cossloenr,•t I • n SS ra a ,deouat high dr/wall Presuare

meniticon hrrfione does NOT exust I lmw RARE "I ,olto Iner 's I per31EC-EOP-O02

• Rueator ull,1l• 1VAC isolstes, Ml( h[iF start Reactor, Buiding WVAG per

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Organization 5.2

5.0 ADMINISTRATIVE CONTROLS

5.2 Organization

5.2.1 Onsite and Offsite Organizations

Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. 'the onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant.

a. Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements, including plant specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications, shall be documented in the Plant Hatch Unit 1 FSAR;

b. An assistant plant manager shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe operatidn and maintenance of the plant;

c. The corporate executive responsible for Plant Hatch shall take any measures needed to ensure acceptable performance of the staff in operating,.maintaining, and providing technical support to the plant to ensure nuclear safety; and

d. The individuals who train the operating staff, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.

5.2.2 Unit Staff

The unit staff organization shall include the following:

a. A total of three plant equipment operators (PEOs) for the two units is required in all conditions. At least one of the required PEOs shall be assigned to bach reactor containing fuel.

(continued)


Organization 5.2

5.2 Organization

5.2.2 Unit Staff (continued)

b. At least one licensed Reactor Operator (RO) shall be present in the control room for each unit that contains fuel in the reactor. In addition, while the unit is in MODE 1, 2, or 3, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room.

c. The minimum shift crew composition shall be in accordance with 10 CFR 50.54(m)(2)(i). Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and 5.2.2.a for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

d. An individual qualified to implement radiation protection procedures shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

e. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety related functions (e.g., licensed and non-licensed operations personnel, health physics technicians, key maintenance personnel, etc.).

Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work a nominal 40 hour week while the unit is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance, or major plant modification, on a temporary basis the following guidelines shall be followed:

1. An individual should not be permitted to work more than 16 hours straight, excluding shift turnover time;

2. An individual should not be permitted to work more than 16 hours in any 24 hour period, nor more than 24 hours in any 48 hour period, nor more than 72 hours in any 7 day period, all excluding shift turnover time;

3. A break of at least 8 hours should be allowed between work periods, including shift turnover time;

4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.

(continued)

HATCH UNIT 1 5.0-3 Amendment no I•znv.• v

Organization 5.2

5.2 Organization

5.2.2 Unit Staff

e. (continued)

Any deviation from the above guidelines shall be authorized by an assistant plant manager or by hibher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation.

Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by an assistant plant manager or designee to ensure that excessive hours have not been assigned. Routine deviation from the above guidelines is not authorized.

f. The operations manager shall hold an active or inactive SRO license.

g. The Shift Technical Advisor (STA) shall provide advisory technical support to the shift supervisor in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. In addition, the STA shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.


Main Condenser

3.7 PLANT SYSTEMS

3.7.6 Main Condenser Offgas

LCO 3.7.6

APPLICABILITY:

The gross gamma activity rate of the noble gases measured at the main condenser evacuation system pretreatment monitor station shall be 5 240 mCi/second.

MODE 1, MODES 2 and 3 with any main steam line not isolated and steam jet air

ejector (SJAE) in operation.

ACTIONS


A. Gross gamma activity rate A.1 Restore gross gamma 72 hours of the noble gases not activity rate of the noble within limit, gases to within limit.

B. Required Action and B.1 Isolate all main steam 12 hours associated Completion lines. Time not met.

OR

B.2 Isolate SJAE. 12 hours

OR

B.3.1 Be in MODE 3. 12 hours

AND

B.3.2 Be in MODE 4. 36 hours

Amendment No. 195

Offgas 3.7.6

HATCH UNIT 1 3.7-16

Main Condenser Offgas 3.7.6



SR 3.7.6.1 ----------------------------- NOTE -..........----..--------Not required to be performed until 31 days after any main steam line not isolated and SJAE in operation.

Verify the gross gamma activity rate of the noble 31 days gases is S 240 mCi/second.

AND

Once within 4 hours after a > 50% increase in the nominal steady state fission gas release after factoring out increases due to changes in THERMAL POWER level


NRC EXAM 2002 PLANT HATCH

WRITTEN EXAM REFERENCE

BOOK

MASTER SRO

Documents

Hatch October 2002 Exam 50-321/2002-301 FINAL SRO Writtem ... · Updated correct answer per lesson plan. Resequenced answers. A. Incorrect since the speed control signal was not lost