ML093280353
| ML093280353 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 11/18/2009 |
| From: | NRC/RGN-II |
| To: | Southern Nuclear Operating Co |
| References | |
| 50-424/09-301, 50-425/09-301 | |
| Download: ML093280353 (218) | |
Text
u.s. Nuclear Regulatory Commission Site-Specific SRO Written Examination Applicant Information Name:
Date: 06-26-2009 Region:
Start Time:
I D II
- III D IV D Reactor Type:
Finish Time:
w.
W.
&2 Facility/Unit: Vogtle 1 &
CEDBWDGED Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with 70.00 percent or better better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80.00 percent to pass. You have 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to complete the combined examination, and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> if you are only taking the SRO portion.
(
Applicant Certification All work done on this examination is my own. I have neither given nor received aid.
Applicant's Signature Results RO/SRO-OnlylTotal Examination Values RO/SRO-Only/Total ~ / 25 / 100 Points Applicant's Scores -- / -- / -- Points Applicant's Grade -- / -- / -- Percent
1.
Given the following:
- A turbine runback occurred and the crew has now stabilzed the plant.
- The UO is making minor turbine adjustments to control RCS Tave.
- Control Rods are currently in AUTO.
- Tave is currently 2.8 degrees higher than Tref.
Which ONE of the following describes the current rod speed and when rod movement would stop?
Current Rod Speed When Rod Motion Stops (In steps per minute) (temperature deviation)
A. 8 steps per minute Tave < 1.0 degrees> Tref B. 8 steps per minute Tave < 1.5 degrees> Tref C. > 8 steps per minute Tave < 1.0 degrees> Tref D. > 8 steps per minute T ave < 1.5 degrees> Tref Tave Page: 1 of 100 6119/2009 6/19/2009
2.
Given the following:
- All RCPs are de-energized.
Based on RVLlS indication, which ONE of the following RCP hand switch positions would lead to a false CORE COOLING CSF status tree condition?
RCP 1 E handswitches RCP non-1 E handswitches A. Open Open B. Closed Open C. Open Closed D. Closed Closed Page: 2oflOO 2 of 100 6119/2009
3.
Given the following:
- Dropped rod recovery in progress.
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- The OATC initiates rod withdrawal for the affected rod.
- The ROD CONTROL URGENT FAILURE annunciator illuminates.
Which ONE of the following completes the following statements?
This is due to a a._ _(a) _ _ failure in the unaffected rod group in the affected bank.
Once rod withdrawal is complete, the OATC should reset the ROD CONTROL URGENT FAILURE alarm using the (b)_ _
(b).
A. (a) regulation (b) ROD CONTROL STARTUP switch B. (a) regulation (b) ROD CONTROL ALARM RESET switch
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C. (a) multiplexing (b) ROD CONTROL STARTUP switch D. (a) multiplexing (b) ROD CONTROL ALARM RESET switch Page: 3 oft 00 of 100 6/19/2009
4.
Initial conditions:
- RCS heatup is in progress.
(
- Filling and venting of the RCS in progress.
- The DATC is ready to start the first RCP.
Which ONE of the following sets of parameters will ALLOW starting of the first RCP in accordance with SDP-13003, Reactor Coolant Pumps?
RCS Pressure (psig) VCT Pressure (psig) Seal Injection Flow (gpm)
A. 360 17 14 B.
- 8. 340 19 10 c.
C. 320 21 7 D. 300 15 11 Page: 4 of oflOO 100 6/19/2009
5.
Given the following:
- AOP-18007-C, "Chemical Volume Control System Malfunction," Section A for "Total
( Loss of Letdown Flow" has been entered.
- Seal injection flows are as follows:
RCP # 1- 8.5 gpm RCP # 2- 9.0 gpm RCP # 3- 8.5 gpm RCP # 4- 9.0 gpm
- Charging flow is currently - 120 gpm
- The OATC is reducing charging flow using FIC-0121, Charging Flow Controller while maintaining the current seal injection flows using HC-0182, "Seal Water Controller."
In accordance with 18007-C, which ONE of the following is the required action the OA OATC TC should take in accordance with 18007-C and the basis for the final charging flow rate?
A. Reduce charging flow to - 35 gpm.
Prevent Regenerative Heat Exchanger damage due to high thermal stresses.
(
B. Reduce charging flow to - 45 gpm Prevent Regenerative Heat Exchanger damage due to high thermal stresses.
C. Reduce charging flow to - 35 gpm.
Prevent an additional high stress thermal cycle on the charging line nozzles.
D. Reduce charging flow to - 45 gpm.
Prevent an additional high stress thermal cycle on the charging line nozzles.
Page: 5 of 100 6119/2009 6/19/2009
6.
Initial conditions:
- The unit is at 100% power.
(
- Control rods are in automatic.
- Boric Acid Potentiometer is required to be set at 3.69
- Boric Acid Potentiometer actual setting is 2.69 Current conditions:
- An automatic makeup to the VCT occurs.
Which ONE of the choices completes the following statements?
To offset the reactivity change from the auto makeup control rods would need to be (a)_ _
At EOL, the effects of the Boric Acid Potentiometer setting error causes rods to move (b) than at BOL.
A. (a) INSERTED
( (b) more B. (a) INSERTED (b) less C. (a) WITHDRAWN (b) more D. (a) WITHDRAWN (b) less.
Page: 6 of 100 6119/2009
7.
The Unit is in Mode 1.
Regarding the following valve:
- RHR PMP-A UPSTREAM SUCTION FROM HOT LEG LOOP 1, 1HV-8701 B.
Which ONE of the following CORRECTLY describes the power supply and the normal valve status for present plant conditions?
C0115, valve is normally energized.
A. 125V DC Inverter CD115, C0115, valve is normally de-energized.
B. 125V DC Inverter CD115, 00116, valve is normally energized.
C. 125V DC Inverter DD116, D. 00116, valve is normally de-energized.
O. 125V DC Inverter DD116, Page: 7 of 100 6/19/2009 6/1912009
8.
The plant is in Mode 5.
The following annunciator is illuminated and cannot be reset;
- B COLD OP ACTU VL VLVV HV-8000B NOT FULL OPEN Which ONE of the conditions below would require entry into the COPS LCO?
(Assume all valve lift setpoints are set correctly)
A. Train A RHR loop suction isolation valves open.
PORV A handswitch in auto, Train A COPS in ARM.
B. RHR Train A loop suction isolation valves open.
PORV B handswitch in auto, Train B COPS in ARM.
C. RHR Train A and Train B loop suction isolation valves open.
PORV A handswitch in auto, Train A COPS in BLOCK
(
D. RHR Train A and Train B loop suction isolation valves open.
PORV B handswitch in auto, Train B COPS in BLOCK.
Page: 8 of 100 6/19/2009 6/1912009
9.
Initial conditions:
- The plant is at 100% power Current conditions:
. During Reactor Trip Breaker "A" testing, the following indications occur.
- Reactor Trip Breaker "A" green light is illuminated.
- Reactor Trip Breaker "B" green light is illuminated.
- Reactor Trip Bypass Breaker "A" red light is illuminated.
- Reactor Trip Bypass Breaker "B" has no light indication.
No operator actions have been taken.
Which ONE of the following is CORRECT regarding plant status and next required operator actions (if any)?
Plant Status Required Action (if any)
A. The Reactor is tripped. Check Turbine Tripped.
/
f B. The Reactor is tripped. Trip the Reactor using both trip handswitches.
\.
C. The Reactor is at power. No action is required.
D. The Reactor is at power. Trip the Reactor using either trip handswitch.
Page: 9 of 100 6/19/2009
10.
Given the following conditions.
- PRZR Safety Valve, PSV-8010A PSV-8010A, is stuck slightly open.
j
- RCS pressure is stable at 1920 psig.
- PZR Vapor Space Temperature is 630 oF.
- PRT pressure is 35 psig and slowly rising.
- Containment pressure is 0 psig.
- PSV-8010A tailpipe temperature is reading - 281 0of. F.
Select the correct choice for:
(a) the tail pipe temperature reading, and (b) the expected response of the tailpipe tem temperature perature if the safety valve does not close.
REFERENCE PROVIDED A. The tail pipe indication is reading correctly.
The temperature should rise to - 338°F, then remain stable.
(
B. The tail pipe indication is reading correctly.
338 0 F, then quickly lower to - 212 oF.
The temperature should rise to - 338°F, C. The tail pipe indication should read - 258°F.
258 0 F.
The temperature should rise to - 328°F, then remain stable.
D. The tail pipe indication should read - 258°F.
The temperature should rise to - 328 00 F, then quickly lower to - 212°F.
(
Page: 10 of 100 6/19/2009
11.
Given the following:
- Maintenance on CCW Train A is to be performed.
During draining of the CCW Train A system, drain flow would normally be routed to (a)_ _
When CCW Train A fill and vent is performed, an operator should be stationed at the CCW Hx vent point and the Control Room should (b)_ _
A. (a) the Radwaste Processing Facility (RPF) for processing and reuse.
(b) start 2 CCW pumps simultaneously, remain in this configuration after fill & vent has been completed.
B. (a) the Clean Water Sump for disposal to the Waste Water Retention Basins.
(b) start 2 CCW pumps simultaneously, remain in this configuration after fill & vent has been completed.
C. (a) the Radwaste Processing Facility (RPF) for processing and reuse.
( (b) start 3 CCW pumps simultaneously, reduce to 2 CCW pumps after fill & vent has been completed.
com pleted.
D. (a) the Clean Water Sump for disposal to the Waste Water Retention Basins.
(b) start 3 CCW pumps simultaneously, reduce to 2 CCW pumps after fill & vent has been completed.
(
Page: 11 of 100 6/1912009 6/19/2009
12.
Given the following plant conditions:
- Unit 1 has a stuck open PRZR Code Safety Valve.
- Appropriate operator response actions have been taken.
- RCS pressure is stable at 1345 psig
- Containment temperature is 160 0 F
- Actual PRZR level is 50%
The effect of RCS pressure at 1345 psig will cause the indicated PRZR level LI-459 to read .J!!L
~ actual level; the effect of containment temperature at 1600 F tends to make the indicated PRZR level read .J.QL
.JQL than actual level. '
A. (a) below (b) lower B. (a) below (b) higher C. (a) above (b) lower D. (a) above (b) higher
(
Page: 12 of 100 6119/2009 6/1912009
13.
Given the following:
- The unit is at full power.
(
- A loss of ACCW occurs at 1432.
- The crew has entered AOP-18022-C, "Loss of Auxiliary Component Cooling Water" and is attempting to restore ACCW flow.
- RCP temperatures are below their immediate trip criteria but are slowly rising rising..
.-;. RCP seal injection flow is - 9 gpm per RCP.
Which one of the following correctly describes:
(a) the MAXIMUM time allowed to trip the RCPs and (b) the MINIMUM temperature setpoint for the motor bearing that requires RCP trip.
A. (a) 1437 (b) 195 0F 195°F B. (a) 1437 (b)3110F (b) 311°F
( C. (a) 1442 195°F (b) 195 0F D. (a) 1442 311°F (b) 311 0F Page: 13 of oflOO 100 6/19/2009
14.
Given the following: .
- Small break LOCA in progress.
(
- A loss of all offsite power has occurred.
DG 1B is tagged out.
- DG1
- The SAT is unavailable.
- The crew transitions to 19012-C, "Post LOCA Cooldown and Depressurization".
- The step for "Isolate SI Accumulators" is being performed.
- The UO has been dispatched to energize the SI Accumulator isolation valves (HV-8808A, B, C, D).
Which ONE of the following is CORRECT regarding isolation of the SI Accumulators?
A. All accumulator isolation valves can be energized.
Close all accumulator isolation valves, leave the valves energized.
( B. All accumulator isolation valves can be energized.
Close all accumulator isolation valves, then de-energize the valves.
C. Two accumulator isolation valves cannot be energized.
Close energized accumulator isolation valves, vent all of the accumulators.
D. Two accumulator isolation valves cannot be energized.
Close energized accumulator isolation valves, vent two of the accumulators.
Page: 14 of oflOO 100 6/19/2009
15.
Actual Pressurizer Code Safety Valve position indication can be read on the (a) which is / are powered from (b) _ _
A. (a) Plasma Displays A and B (b) AY1A and BY1B, 120V AC Vital Instrument Panels B. (a) Plasma Displays A and B (b) NYR and NYS, 120V AC Regulated Instrument Panels C. (a) Main Control Board Panel C (b) AY1A and BY1B, 120V AC Vital Instrument Panels D. (a) Main Control Board Panel C (b) NYR and NYS, 120V AC Regulated Instrument Panels
(
Page: 15 of 100 6/19/2009
16.
Given the following conditions I/ events on Unit 1 :
- Reactor power is 100%.
- PRZR pressure control is selected to 457 I/ 456 position.
- PRZR pressure is at 2235 psig.
- 1PT-456 fails high.
Which ONE of the following is the CORRECT plant I/ system response?
Both spray valves remain closed and all PRZR heaters will energize.
Pressure will stabilize near 2185 psig.
Both spray valves remain closed and all PRZR heaters will energize.
Pressure will stabilize near 2185 psig.
(
Both spray valves open and all PRZR heaters turn off.
Pressure will continue to lower causing a Reactor Trip and Safety Injection.
D. PORV 1 1PV-456 PV-456 opens.
Both spray valves open and all PRZR heaters turn off.
Pressure will continue to lower causing a Reactor Trip and Safety Injection.
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Page: 16 oflOO of 100 6/19/2009
17.
Given the following:
- A DBA LOCA is in progress on Unit 2.
- ECCS has been aligned for "cold leg recirculation".
- RHR pump 2A tripped 5 minutes after completing the recirculation alignment.
- All other components are functioning properly.
Which ONE of the operator actions would be CORRECT regarding the Safety Injection pumps (SIPs) and the Centrifugal Charging pumps (CCPs)?
A. Immediately stop both SIPs.
B. Immediately stop both CCPs.
C. Immediately stop CCP "A" and SIP "A".
D. Allow all CCPs and SIPs to continue running.
(
oflOO Page: 17 of 100 6/19/2009
18.
Given the following:
- A HIGH failure (offscale high) of Pressurizer (PRZR) Pressure Channel PT-457 has occurred.
- Immediate Operator Actions of AOP-18001-C for failure of a Pressurizer Pressure channel have been taken.
- NO other actions have been taken.
Which ONE of the following identifies the MINIMUM additional channels needed to cause a:
a) Reactor Trip b) Safety Injection Reactor Trip Safety Injection A. 1 1 B. 1 2 C. 2 1
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D. 2 2 Page: 18 of 100 oflOO 6/19/2009
19.
A Turbine Driven Auxiliary Feed Water (TDAFW) Actuation has been generated.
- HV-51 06, "TDAFW Pump Steam Supply Valve" has stroked open.
Which ONE of the following conditions is necessary to RESET the TDAFW actuation and close HV-5106?
A. At teast ONE Main Feed Water Pump Turbine RESET.
B. RESET Safety Injection using both OMCB QMCB handswitches.
C. ALL Steam Generator levels above the Low-Low level setpoints.
D. ATWS Mitigation System Actuation Circuit (AMSAC) signal CLEAR.
(
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Page: 19 of 100 6/19/2009 611912009
20.
Given the following:
. - A loss of all AC power has occurred.
- The crew is performing 19102, "Loss of All AC Power Recovery With SI Required".
- Both CCPs are stopped.
(HV-81 03A, B, C, D)
- All RCP Seal Injection Isolations are OPEN (HV-8103A,
- RCP seal temperatures are - 465 00 F Which ONE of the following is the CORRECT mitigation strategy for restoration of charging flow and seal injection to cool the RCP seals?
A. Take HC-0182, Seal Injection Controller to minimum, start a CCP, throttle HC-0182 to establish a slow cooldown rate on the RCP seals.
HV-81 03 valves B. Close all the HV-8103 valves, start a CCP, slowly throttle open the HV-8103 one at a time to establish a slow cooldown rate on the RCP seals.
C. Close the seal injection needle valves, start a CCP, slowly throttle open the seal injection needle valves to establish a slow cooldown rate on the RCP seals.
D. Close all the HV HV-8103
-8103 valves, start a CCP, perform a controlled cooldown of the
( RCS to cool the RCP seals, RCPs should not be started prior to a status evaluation.
(
Page: 20 oflOO of 100 6/19/2009
21.
Given the following:
- The reactor is stable with critical data being taken.
- The crew is preparing to raise power to the POAH.
- 18Y18, 1BY1 B, 120V Vital Instrument bus de-energizes.
Which ONE of the following describes the effects on the Intermediate Range indications and the reason for the indications?
A. IR N-35, remains energized and indication decreases.
IR N-36, remains energized and indication decreases.
- 8. IR N-35, remains energized with indication stable.
B.
IR N-36, remains energized with indication stable.
- c. IR N-35, remains energized and indication decreases.
C.
IR N-36, will de-energize.
(
D. IR N-35, remains energized with indication stable.
IR N-36, will de-energize.
Page: 21 of oflOO 100 6/19/2009
22.
Which of the following choices correctly identifies the expected equipment condition for the initial and current CNMT pressures? Assume no operator actions taken.
( Initial CNMT pressure:
Ch 1- 3.5 psig Ch II - 3.9 psig Ch III - 3.7 psig Ch IV - 3.6 psig Current CNMT pressure:
Ch 1-I - 21.3 psig Ch II - 21.7 psig Ch III - 20.9 psig Ch IV - 22.1 psig A. Initial - Containment Cooler Lo Speed aMCB QMCB MLBs LIT.
Current - Containment Spray Pump aMCB QMCB handswitch RED lights LIT.
B. Initial-Initial - Containment Cooler Lo Speed aMCB QMCB MLBs LIT.
Current - Containment Spray Pump aMCB QMCB handswitch GREEN lights LIT.
QMCB MLBs EXTINGUISHED.
C. Initial - Containment Cooler Lo Speed aMCB Current - Containment Spray Pump aMCBQMCB handswitch RED lights LIT.
(
D. Initial - Containment Cooler Lo Speed aMCB QMCB MLBs EXTINGUISHED.
Current - Containment Spray Pump aMCBQMCB handswitch GREEN lights LIT.
(
Page: 22 of 100 6/19/2009
23.
A failure of a flow transmitter has resulted in the following annunciators:
- REGEN HX LETDN HI TEMP
- LETDN HX OUT HI TEMP
- CHARGING LINE HIILO FLOW All other plant parameters are normal.
Which ONE of the following actions would CORRECTLY mitigate this event?
A. Throttle open HV-0182 (seal flow control).
B. Throttle closed HV-0182 (seal flow control).
8.
C. Throttle open FV-0121 (charging flow control).
D. Throttle closed FV-0121 (charging flow control).
(
Page: 23 oflOO 6/19/2009
24.
Given the following conditions:
- The Reactor was shutdown to Mode 3 conditions 28 days ago ago..
.-:. Core reload has just been completed.
- Unit 1 is in midloop operation.
- All HL and CL nozzle dams are in place.
- A total loss of RHR cooling has occurred.
- RCS temperature was 1OooF GOOF when the loss of cooling event initiated.
Calculate the amount of time it will take from the loss of RHR until saturated conditions are reached in the RCS AND the amount of time until core uncovery occurs.
REFERENCE PROVIDED Time Until Saturated Time Until Core Uncovery A. 38 minutes 145 minutes B. 38 minutes 228 minutes
( C. 58 minutes 145 minutes D. 58 minutes 228 minutes Page: 24 of oflOO100 6119/2009 6/19/2009
25.
Which ONE of the following describes the transition process for the Containment Spray system from the injection flow path to the recirculation flow path after a LOCA?
A. At the RWST LO-LO Level of 29%, the operator manually realigns the flow path.
B. At the RWST LO-LO Level of 29%, the flow path semi-automatically realigns.
C. At the RWST EMPTY level of 8%, the operator manually realigns the flow path.
D. At the RWST EMPTY level of 8%, the flow path semi-automatically realigns.
(
Page: 25 oflOO of 100 6/19/2009
26.
Initial conditions:
- 457 / 456 position selected for PRZR pressure control.
- All channels are reading 2235 psig.
Current conditions:
- PT -455(Ch I) PT-456(Ch II) PT-457(Ch III) PT-458(Ch IV) 2190 psig 2190 psig Offscale High 2285 psig
- Immediate actions to stabilize PRZR pressure have been completed.
- AOP 18001-C, section C for PRZR Pressure Instrumentation, is in effect.
- The step to select controlling channels and recorders is being performed.
Which ONE of the following would be the CORRECT controlling channels to choose and the reason why?
Channel Selection Reason for Channel Selection A. 455/456 Defeat input to the Reactor Protection System.
B. 455/456 Swap to operable control channel inputs.
C. 455/458 Defeat input to the Reactor Protection System.
D. 455/458 Swap to operable control channel inputs.
Page: 26 of oflOO 100 6/1912009 6/19/2009
27.
Given the following:
- CNMT H2 concentration has remained> 4% following a LOCA.
- An SO has been dispatched for local operation of 1-1508-U4-012, "POST LOCA PURGE CTB ISO VALVE".
In accordance with SOP 13130-1, Post-Accident Hydrogen Control", to protect the SO from radiation exposure, which one of the following correctly describes:
The proper sequence of actions and the location where the SO will manipulate the local valve?
A. First action - UO opens the Post LOCA valves on HVAC panel.
Next action - SO opens the valve on the Equipment Building roof.
B. First action - UO opens the Post LOCA valves on HVAC panel.
Next action - SO opens the valve in the Equipment Building ground level.
C. First action - SO opens the valve on the Equipment Building roof.
(
Next action - UO opens the Post LOCA valves on HVAC panel.
D. First action - SO opens the valve in the Equipment Building ground level.
Next action - UO opens the Post LOCA valves on HVAC panel.
Page: 27 of 100 6/19/2009
28.
Given the following conditions:
- Both Reactor Trip Breaker red lights are illuminated.
- Both Reactor Trip Breaker green lights are extinguished.
- The OATC momentarily places the reactor trip handswitch in the TRIP position and Reactor Trip Breaker indications DO NOT CHANGE.
Which of the following choices describes the actions that should have occurred when the OATC placed the handswitch to TRIP?
A. The Undervoltage coils energized.
The Shunt coils energized.
B. The Undervoltage coils de-energized.
The Shunt coils energized.
C. The Undervoltage coils energized.
(' The Shunt coils de-energized.
D. The Undervoltage coils de-energized.
The Shunt coils de-energized.
Page: 28 of 100 6/19/2009
29.
Given the following:
- The unit is at 100% power.
- CNMT pressure is 0.2 psig.
- CNMT Mini-Purge is in service for respirable air quality control.
.-;. A Safety Injection occurs.
Which one of the following is correct regarding the Containment Purge System damper status prior to the SI and the signal that will close the dampers?
A. Only the Mini-Purge exhaust dampers were open.
The dampers would receive a direct auto close signal from CIA.
B. Only the Mini-Purge exhaust dampers were open.
The dampers would receive a direct auto close signal from CVI.
C. The Mini-Purge exhaust and supply dampers were open.
(
The dampers would receive a direct auto close signal from CIA.
D. The Mini-Purge exhaust and supply dampers were open.
The dampers would receive a direct auto close signal from CVI.
Page: 29 of 100 6/19/2009
30.
Given the following conditions:
- Unit 1 core reload is in progress.
- Source Ranges N-31 and N-32 are reading significantly different.
-I& & C determines the power supply for detector N-31 has failed high.
- N-32 is selected for Audio Count Rate in Containment and Control Room.
Due to this power supply failure, N-31 is reading _ _(a)_ _than it should.
The Operator At The Controls is (b )_ _
A. (a) lower (b) required to suspend core alterations B. (a) lower (b) not required to suspend core alterations C. (b) higher (b) required to suspend core alterations D. (b) higher (b) not required to suspend core alterations
(
Page: 30 of 100 6/19/2009
31.
Unit 2 Refueling Outage is in progress:
- Fuel shuffle is in progress in the Spent Fuel Pool (SFP) with the gate to the
( Containment currently closed.
- SPENT FUEL PIT LOW LEVEL annunciator illuminates in Control Room.
- SFP level lowers to 22 feet 10 inches above the fuel due to a leak
- The SS has directed a makeup to the SFP in accordance with SOP-13719-2, "Spent Fuel Pool Cooling and Purification System".
Which ONE of the following describes whether SFP level is above or below the Tech Spec limit, and the source of makeup for this condition, in accordance with SOP.,13719-2.
SFP level is ...
A. BELOW the minimum required by Technical Specifications.
Makeup from the Demin Water Storage Tank (DWST).
B. BELOW the minimum required by Technical Specifications.
(
Makeup from the Refueling Water Storage Tank (RWST).
C. ABOVE the minimum required by Technical Specifications.
Makeup from the Demin Water Storage Tank (DWST).
D. ABOVE the minimum required by Technical Specifications.
Makeup from the Refueling Water Storage Tank (RWST).
Page: 31 oflOO 6/19/2009
32.
Given the following:
- The unit is at 100% power.
- Loop # 3 Inboard MSIV (HV-3016A) goes fully closed.
Given this condition AND before manual or automatic operation of any other system.
- Loop # 2 SG steam FLOW would (a) and
- Loop # 3 SG LEVEL initially would rapidly (b)_ _
A. (a) rise (b) rise B. (a) rise (b) lower C. (a) lower (b) rise D. (a) lower (b) lower
(
Page: 32 of 100 6/19/2009
33.
Which one of the following choices for the SG code safeties:
(a) lists all the correct lift setpoints (psig), and (b) describes the technical specification bases for these valves?
A. (a) 1185,1200,1210,1220,1235.
(b) limits secondary pressure to ~ 110% design pressure during a full power turbine trip without steam dump.
B. (a) 1185,1200,1215,1225,1235.
(b) limits secondary pressure to ~ 110% design pressure during a full power turbine trip without steam dump.
C. (a) 1185,1200,1210,1220,1235.
(b) provide an alternate method for cooling the unit to RHR entry conditions whenever the preferred heat sink via the steam dumps to the condenser is unavailable.
(
D. (a) 1185,1200,1215,1225,1235.
(b) provide an alternate method for cooling the unit to RHR entry conditions whenever the preferred heat sink via the steam dumps to the condenser is unavailable.
Page: 33 of 100 6119/2009 611912009
34.
A rapid power reduction per AOP 18013-C was completed to stabilize condenser vacuum. Rods were inserted below the rod insertion limit (RIL) during the power reduction. An Emergency Boration is in progress and the OATe OATC is withdrawing rods to
(, restore rods above the RIL.
Which ONE of the following is CORRECT regarding:
a) whether SS permission is required for Control Rod insertion during the downpower?
and b) when Emergency Boration may be terminated?
Control Rod Insertion Emergency Boration Termination A. SS permission is NOT required. When Rod Bank LO Limit alarm clears.
B. SS permission is NOT required. When Rod Bank LO-LO Limit alarm clears.
C. SS permission IS15 required. When Rod Bank LO Limit alarm clears.
D. SS permission IS15 required. When Rod Bank LO-LO Limit alarm clears.
Page: 34 of 100 6/19/2009
35.
Given the following:
- At 1330 hours0.0154 days <br />0.369 hours <br />0.0022 weeks <br />5.06065e-4 months <br /> a Loss of All AC Power occurred.
- Restoration of AC power is not anticipated until 1730 hours0.02 days <br />0.481 hours <br />0.00286 weeks <br />6.58265e-4 months <br />.
Which ONE of the following describes the CORRECT action required by 191 OO-C, "Loss of All AC Power" as 1E bus battery voltage decays over time and why?
A. When voltage < 105 VDC, open the battery breaker after its inverters are shutdown.
To prevent damaging the battery cells due to cell reversal.
B. When voltage < 120 VDC, open the battery breaker after its inverters are shutdown.
To prevent damaging the battery cells due to cell reversal.
C. When voltage < 105 VDC, open the battery breaker after its inverters are shutdown.
To prevent a possible explosion due to excessive hydrogen production.
( D. When voltage < 120 VDC, open the battery breaker after its inverters are shutdown.
To prevent a possible explosion due to excessive hydrogen production.
oflOO Page: 35 of 100 6/19/2009
36.
Given the following:
- The unit is at 100% power.
- The loop seal between stages on the in-service SJAE has been lost.
Which ONE of the following is CORRECT regarding the effect of losing the SJAE loop seal on Main Condenser pressure and Main Turbine MW Output?
Main Condenser Pressure (psia) Main Turbine MW Output A. Rises Rises B. Lowers Rises c.
C. Rises Lowers D. Lowers Lowers
(
(
Page: 36 of 100 6/19/2009
37.
Initial Conditions:
- Safety Injection occurs due to an RCS LOCA
- Train "A" SI has been reset
- Train "8" SI is still actuated
- 80th SI pumps are injecting Current conditions:
- A loss of both RATs occurs
- 80th DG's re-energize their respective busses Which ONE of the following would be CORRECT regarding the SI pumps flow and amp responses?
A. SIP "A" - flow and amps lower, then return to previous values automatically.
SIP "8" - flow and amps lower, then return to previous values automatically.
- 8. SIP "A" - flow and amps lower, then have to be manually restored.
SIP "8" - flow and amps lower, then have to be manually restored.
(
C. SIP "A" - flow and amps lower, then return to previous values automatically.
SIP "8" - flow and amps lower, then have to be manually restored.
D. SIP "A" - flow and amps lower, then have to be manually restored.
SIP "8" - flow and amps lower, then return to previous values automatically.
oflOO Page: 37 of 100 6/19/2009
38.
Given the following:
- The Unit is at 70% power.
(
- VCT level is at 46%.
- A loss of a 120VAC Essential bus has occurred.
- VCT auto makeup immediately starts on the bus failure.
Which ONE of the following CORRECTLY describes ...
a) The VCT level channel failure.
b) The effects on VCT auto swapover capability.
A. a) LT-112 T -112 (VCT level) fails LOW.
b) VCT auto swapover will occur on actual VCT 10-10 level.
B. a) L T-185 (VCT level) fails LOW.
LT-185 b) VCT auto swapover will occur on actual VCT 10-10 level.
(
C. a) LT-112 (VCT level) fails HIGH.
b) VCT auto swapover will NOT occur on actual VCT 10-10 level.
D. a) LT-185 LT-185 (VCT Level) fails HIGH.
b) VCT auto swapover will NOT occur on actual VCT 10-10 level.
Page: 38 of 100 oflOO 6/19/2009
39.
The plant is 100% power, all systems in normal alignment.
- ALB04, window D04 for BAT CHARGERS 1CD1 CA 1CD1 CB TROUBLE illuminates.
- 1CD1 B battery amperage reading is 0 amps
- Channel III TSLB status lights are extinguished.
Which ONE of the following is CORRECT regarding the 125V DC bus 1CD1 status?
A. One battery charger tripped the battery is supplying 1CD1 B. Both battery chargers tripped the battery is supplying 1CD1 C. Both battery chargers tripped 1CD1 is completely de-energized D. One battery charger tripped 1CD1 is supplied by the other charger
(
Page: 39 of 100 6119/2009 6/19/2009
40.
Given the following:
- The unit is at 90% power.
- All Main Feedwater (MFW) system controls in AUTO.
- A low failure of PT-508 (Feedwater Header Pressure) occurs.
Both MFPT speeds should (a) due to the (b)
A. (a) raise (b) lower indicated delta P B. (a) raise (b) higher indicated delta P c.
C. (a) lower (b) lower indicated delta P
(
D. (a) lower (b) higher indicated delta P
(
\
Page: 40 oft 00 of 100 6/19/2009
41.
Initial conditions:
- MFPT "A" is operating.
(
- MFPT "8" is tripped.
- All SG levels are stable, AFW flows to all SG's are - 225 gpm
- The following annunciator window is extinguished.
- AL816, window F05 "AFW AUTO START MFPT TRIP RLY CNTL PWR LOSS" Current conditions:
- The UO manually trips MFPT "A".
Which ONE of the following is the CORRECT plant response and actions to take?
A. AFW flows remain as is, SG levels remain stable.
Reset MFPT "A", restart MFPT "A".
(
Reset MFPT "8", startup MFPT "8".
C. AFW flows increase, SG levels begin to rise.
Throttle AFW valves closed before they fully open.
D. AFW flows increase, SG levels begin to rise.
Allow AFW valves to stroke fully open, then throttle.
(
Page: 41 oflOO 6/19/2009
42.
The unit is at 100% power when a Reserve Auxiliary Transformer (RAT) switcher trips open. Reactor power rises above 100%.
Auxiliary Feedwater flow indications are:
- SG # 1 - 650 gpm SG # 2 - 350 gpm SG # 3 - 350 gpm SG # 4 - 650 gpm The sequence of actions the UO should take in AOP-18031, "Loss of Class 1E Electrical Systems" is to ...
A. Reduce Turbine load to reduce Reactor power to ~ 100%.
Then throttle TDAFW speed and MDAFW pump "A" discharge valves.
- 8. Reduce Turbine load to reduce Reactor power to ~ 100%.
Then throttle TDAFW speed and MDAFW pump "8" discharge valves.
C. Throttle TDAFW speed and MDAFW pump "A" discharge valves.
Then reduce Turbine load to reduce Reactor power to ~ 100%.
(
D. Throttle TDAFW speed and MDAFW pump "8" discharge valves.
Then reduce Turbine load to reduce Reactor power to ~ 100%.
Page: 42 of 100 6/1912009 6/19/2009
43.
Given the following:
- Loss of All AC power occurs, 19100-C, 191 OO-C, "Loss of All AC Power" is in effect.
- Power is restored to AA02 via DG1A.
Which ONE of the following is the CORRECT action regarding restoration of the NSCW system to service?
A. Verify NSCW discharge valves go closed to prevent water hammer.
Manually start 2 pumps.
B. Verify NSCW discharge valves go closed to to prevent water hammer.
Verify 2 pumps start automatically.
C. Verify NSCW discharge valves are open to prevent high starting currents.
Manually start 2 pumps.
D. Verify NSCW discharge valves are open to prevent high starting currents
(
Verify 2 pumps start automatically.
(
Page: 43 of 100 oflOO 6/19/2009
44.
The "Maintenance I/ Normal" switch located on the QEAB for a 4160V AC bus has been placed in the "Maintenance" position to support bus inspection.
When placed in the "Maintenance" position, this switch ...
A. bypasses the time delay on the overcurrent trips on the bus supply breakers.
B. bypasses the time delay on the overcurrent trips on the bus load breakers.
C. bypasses the instantaneous overcurrent trips on the bus supply breakers.
D. bypasses the instantaneous overcurrent trips on the bus load breakers.
(
Page: 44 oflOO 6119/2009 6/19/2009
45.
The plant is in Mode 3 controlling Tave with the ARVs when 125V DC panel1AD1 de-energizes due to a ground fault. Loop 4 ARV was - 10% open.
( Regarding operation of the ARV, list:
(a) the failure mode on loss of DC power, and (b) where the ARV can be operated.
A. (a) as is (b) locally using the hand pump station B. (a) as is (b) from Shutdown Panel A C. (a) closed (b) locally using the hand pump station
( D. (a) closed (b) from Shutdown Panel A Page: 45 oflOO 6119/2009 6/19/2009
46.
The plant is in Mode 1.
A leak has developed in the DG1A Fuel Oil Storage Tank (FaST). The FaST level decrease is noted by the following times.
- at 1730 level is 90.9% (78,200 gallons)
- at 1745 level is 86.9% (74,800 gallons)
- at 1800 level is 82.9% (71,350 gallons)
- at 1815 level is 78.9% (67,900 gallons)
- at 1830 level is 74.9% (64,500 gallons)
Which ONE of the following is the EARLIEST discovery time that DG1A is inoperable in accordance with Technical Specifications?
A. 1745 B. 1800 C. 1815 D. 1830
(
Page: 46 of 100 6119/2009 6/19/2009
47.
Given the following initial conditions:
- The unit is at 340 oF, RCS cooldown in progress.
(
- RHR Train A isin is in the shutdown cooling mode.
- A loss of instrument air occurs.
- The RHR system AOVs all go to the failure mode for loss of air.
Current conditions:
- Instrument air has been restored.
- No Control Room or local actions have been performed for RHR.
Which ONE of the following is CORRECT regarding:
- RCS cooldown rate response after instrument air is restored.
- RHR system flow rate response after instrument air is restored.
RCS cooldown rate RHR system flow rate
(
A. raises raises B. raises lowers c.
C. lowers raises D. lowers lowers Page: 47 oflOO of 100 6/19/2009
48.
Unit 1 Control Room has been evacuated due to a fire.
- 18038-1, "Operation From Remote Shutdown Panels" is in effect.
- Two ARVs have been placed in the "Fire Emergency" mode of operation.
Which ONE of the following are the fire event qualified ARVs and the method to to control RCS temperature in the Fire Emergency mode?
A. ARVs # 1 and # 4 are fire event qualified.
Using the auto-manual controllers located on the front of the shutdown panels.
B. ARVs # 2 and # 3 are fire event qualified.
Using the auto-manual controllers located on the front of the shutdown panels.
C. ARVs # 1 and # 4 are fire event qualified.
Using a 4 to 20 mA rnA current source connected with banana jacks inside the panels.
( D. ARVs # 2 and # 3 are fire event qualified.
Using a 4 to 20 mA rnA current source connected with banana jacks inside the panels.
Page: 48 of 100 6/19/2009
49.
A gaseous release is in progress in accordance with SOP-13202, "Gaseous Releases".
Considering the following events which occur during the release, determine if the release would be allowed to continue or require termination for each of the following:
a) A-RE-0014 Radiation Monitor fails low b) precipitation (rain, snow, sleet, etc.) is occuring.
A. a) terminate the release.
b) terminate the release.
B. a) terminate the release.
b) continue the release.
C. a) continue the release.
b) terminate the release.
( D. a) continue the release.
b) continue the release.
(
\
Page: 49 of 100 6119/2009 6/19/2009
50.
Given the following:
- A rapid power reduction (> 15% in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) is in progress to comply with the actions of AOP-18009-C, a "Steam Generator Tube Leakage".
- Radiation is indicated on RE-0724 Steam Line monitor.
Which ONE of the following is CORRECT regarding:
- when RE-0724 would indicate a Steam Generator Tube leak
- the source of radiation detected.
A. only when the reactor is at power.
N-16 B. only when the reactor is at power.
1-131 C. any time the SJAE is in service.
(
N-16 D. any time the SJAE is in service.
1-131
(
Page: 50 of 100 6/19/2009 611912009
51.
Given the following:
- The plant is at 100% power.
- NSCW Train B tagged out for repair.
- NSCW Train A pump # 3 trips.
- NSCW Train A pump # 5 cannot be started.
- AOP-18021-C, "Loss "Loss of Nuclear Service Cooling Water" has been entered.
- No other operator actions have been taken.
Which ONE of the following is the CORRECT crew action(s)?
A. Place all Train A NSCW pumps in PTL, Emergency Trip DG1A.
Shutdown to Mode 3 per UOP-12004-C, "Power Operations" within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
B. Place all Train A NSCW pumps in PTL, Emergency Trip DG1A.
Trip the Reactor and go to E-O, align NSCW Train A for single pump operations.
C. Allow NSCW Pump # 1 to continue running, DG1A should be left in AUTO.
Shutdown to Mode 3 per UOP-12004-C, "Power Operations" within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
D. Allow NSCW Pump # 1 to continue running, DG1A should be left in AUTO.
Trip the Reactor and go to E-O, align NSCW Train A for single pump operations.
Page: 51 of 100 6/19/2009
52.
The plant is at 100% power.
- RHR pump A is running for an 1ST surveillance.
(
- CCP "A" is running.
The following annunciator illuminates on the QMCB.
- Various other Train A NSCW low flow annunciators illuminate.
Approximately 10 seconds later the UO notes the following:
- 3 NSCW pumps on Train A red lights are illuminated.
- NSCW Train A return flow is 0 gpm.
- NSCW Train A supply flow is significantly higher than 17,500 gpm.
Which ONE of the following is the CORRECT initiating event and next action to take?
Initiating Event Next action A. Pump sheared shaft Stop one of the running Train A NSCW pumps.
(
B. Catastrophic break Place all NSCW Train A pump handwitches in PTL.
C. Catastrophic break Stop RHR pump A and CCP A, then isolate letdown.
D. Pump sheared shaft Verify the 3rd Train A NSCW pump discharge valve opens.
Page: 52 of oflOO 100 6119/2009 6/19/2009
53.
Following a controlled unit shutdown for a refueling outage, degassification of the RCS is in progress via chemical addition.
Shortly following the chemical addition, a crud burst occurs as expected.
Which ONE of the following radiation monitors would be expected to alarm?
A. RE-0005, Containment - High Range B. RE-0007A, RE-0007 A, Rad Chem Lab C. RE-00011, Seal Table Room D. RE-48000, CVCS Letdown Page: 53 of oflOO 100 6/19/2009
54.
Given the following initial conditions:
- The unit is at 100% power.
-The
- DG1A is paralleled to the grid for surveillance testing at 6500 kW.
Current condtions:
- AOP-18017-C, "Abnormal Grid Disturbances/Loss of Grid" section A has been
-AOP-18017-C, entered.
- Grid Grid frequency has lowered below 60 Hz.
Which ONE of the following would be CORRECT regarding
- the initial conditions of the DG.
- the effects of the abnormal grid disturbance on the DG.
A. DG speed is used to adjust kW.
The governor would throttle open on the fuel racks to attempt to raise DG speed.
( B. DG speed is used'to adjust kW.
8.
~<
The governor would throttle closed on the fuel racks to attempt to lower DG speed.
C. DG speed is used adjust kVARS.
The governor would throttle open on the fuel racks to attempt to raise DG speed.
D. DG speed is used adjust kVARS.
The governor would throttle closed on the fuel racks to attempt to lower DG speed.
Page: 54 of 100 oflOO 6/19/2009
55.
Given the following:
at 100%
- Unit 1 is at1 00% power.
(
- Air compressor # 4 is tagged out.
- An LOSP results in de-energization of 2 of the 3 remaining compressors.
- A slow decrease in air header pressure has occurred.
- Power has now been restored to the 2 previously de-energized compressors.
- Instrument Air Header pressure dropped as low as 82 psig.
-Instrument
- Instrument Air Header pressure is now 97 psig and slowly rising.
Which ONE of the following is CORRECT regarding the current status of the Instrument Air system?
A. PV-9375, Service Air Isolation valve is OPEN.
All 3 available air compressors are running.
( B. PV-9375, Service Air Isolation valve is OPEN.
2 of 3 available air compressors are running.
C. PV-9375, Service Air Isolation valve is CLOSED.
All 3 available air compressors are running.
D. PV-9375, Service Air Isolation valve is CLOSED.
2 of 3 available air compressors are running.
Page: 55 of 100 6/19/2009
56.
Given the following:
- Unit 1 at 25% power, Unit 2 at 100% power.
- The only two available air compressors on each unit are running (4 total).
- Air compressor # 4 is aligned to Unit 2.
Unit 1 has the following indications:
- SERVICE AIR LO PRESS annunciator lit.
- QMCB Service Air Pressure meter - 0 psig
- QMCB Instrument Air Pressure meter - 72 psig and lowering.
Annunciator window INST AIR EQUIP LO PRESS illuminates several minutes later.
Which ONE of the following is the CORRECT action the operators should perform in response to the annunciator?
A. Align the swing air compressor from Unit 2 to Unit 1.
B. Open the cross tie valve to supply Unit 1 air header from Unit 2.
( C. Trip the Unit 1 Turbine and isolate Turbine Building Instrument Air.
D. Trip the Unit 1 Reactor and enter E-O while taking actions to restore air pressure.
oflOO Page: 56 of 100 6/19/2009
- 57. ,
Given the following:
alarm..
- The fire alarm computer goes into alarm
(
..,- The Outside Area Operator reports a fire on RAT "A" You note the following on the QPCP.
~- Pressure gauge ~'Fire Pump CPI - 7918" pressure lowers to 92 psig, then recovers.
Which ONE of the following is CORRECT regarding how many fire pumps are running and WHERE this can be determined?
A. ONLY the Motor Driven Fire pump is running.
QPCP and local observation.
B. ONLY the Motor Driven Fire pump is running.
Local observation ONLY.
Local C. The Motor Driven pump and one Diesel Fire pump are running.
(
QPCP and local observation.
D. The Motor Driven pump and one Diesel Fire pump are running.
Local observation ONLY.
Page: 57 of 100 6/19/2009
58.
Given the following:
- The unit is at full power.
(
- A small air leak inside Containment is causing a slow rise in Containment pressure.
- Containment pressure is currently reading 1.9 psig.
Containment pressure is (a) _ _
To relieve Containment pressure, flow must pass through a restrictive orifice until containment pressure is (b) , then the normal flow path may be aligned.
A. (a) within Technical Specification LCO limits.
(b) less than +0.3 psig B. (a) within Technical Specification LCO limits.
(b) less than -0.3 psig C. (a) in violation of Technical Specification LCO limits.
(
(b) less than +0.3 psig D. (a) in violation of Technical Specification LCO limits.
(b) less than -0.3 psig
(,
Page: 58 of 100 6119/2009
59.
Given the following conditions.
- The OATC has manually actuated Containment Spray using using,_ _(a)_ _on 1 of 2
( QMCB QMCS locations.
- As a result of the manual action, the annunciator window(s) for (b)_ _
should illuminate.
A. (a) 1 of 2 handswitches (b) CNMT SPRAY ACTUATION only B.
S. (a) 2 of 2 handswitches (b) CNMT SPRAY ACTUATION only C. (a) 1 of 2 handswitches (b) both CNMT SPRAY ACTUATION and CNMT VENT ISO ACTUATION D. (a) 2 of 2 handswitches
(
(b) both CNMT SPRAY ACTUATION and CNMT VENT ISO ACTUATION Page: 59 of 100 6/19/2009
10:2~ FAX 706 826 3953 06/22/2009 10:2) VEGP TRAINING CENTER 14! 002 141002 60.
Given the following:
( Cont -01 Room is being evacuated due to a fire.
- Cont'ol DATe (RO) should report to Remote Shutdown Panel _ _ __
The DATC acco-dance with 18038-1/2, In acco'dance 18038-1/2. "Operation From Remote Shutdown Panels" the Q.referrE d method of communications to co-ordinate in plant activities with personnel W'eferre via_ _ _~_
outside the control room is via.---~-
liAR A. "Alit J sound powered phones plugged into dedicated channels (red jacks),
jacks).
B. 118 11 ,
IIsn, 234.
bridge network phone extension # 3145 using codes # 123 or # 234_
"Au ,
C. "A't, bridge network phone extension # 3145 using codes # 123 or # 234.
D. uBn ,
0- u8", sound powered phones plugged into dedicated channels (red jacks).
(
(
Page: 60 60oflOO of 100 6122/2009 6/22/2009
10:28 FAX 706 826 06/22/2009 10:2B B26 3953 VEGP TRAINING CENTER I4i 003 61.
You arEI performing a lineup verification for valves in various positions:
( Which ONE of the following is CORRECT performance of the lineup verification in accord.mce with NMP-OS-002, nVerification accordHnce "Verification Policy"?
A. A vcllve is required to be closed.
Turn valve slightly in the open direction, using reasonab~e force, reclose the valve.
TUrri B. A vcllve is required to be throttled 2 turns open.
Turn valve in closed direction, count turns to ensure correct, re-open valve 2 turns.
C. A valve volve is found not in the required lineup position.
Align valve in the required position, notify the S8 once the valve is properly aligned.
D. A valve position cannot be determined by local observation.
vnlve pOSition Leave valve as is, use remote handswitch light indications to verify valve position.
Page: 61 of 100 6/22/2009
06/22/2009 10:29 FAX 706 826 3953 VEGP TRAINING CENTER 141004 62.
Given tile following:
- Control Room has been evacuated.
- An LClSP has occurred.
- The :rew
-;rew is preparing to close a DG1A output breaker (1AA02-19).
Which I)NE of the following is CORRECT regarding where and how the DG output breaker is closed?
A. At F:emote Shutdown Panel A.
Tak910cal control of the DG output breaker, then close the breaker handswitch.
B. On:he front of DG1A Generator panel.
Take the handswitch located on the front of OG1A DG1A Generator panel to close.
(1M02~OO).
C. At the 4160V 1E Switchgear local control panel (1M02-00).
Take local control of the DG output breaker, then close the breaker handswitch.
(
\
D. On the front of the DG output breaker at 4160V 1E Switchgear 1AA02.
Take the handswitch located on the front of the DG output breaker to close.
Page: 62 of 100 6/2212009 6/22/2009
63.
Which one of the following describes the meaning of an OPERATING PERMIT TAG hanging on a component as described in NMP-AO-003, NMP-AD-003, Equipment Clearance and Tagging?
A.
- The component position can NOT be changed until the tag is cleared.
- The component is under the control of a Tagout Holder.
B.
- The component position can NOT be changed until the tag is cleared.
- The component has been designated as an isolation boundary for personnel safety.
C.
- The component position can ONLY be changed with permission of the Tagout Holder.
- The component is under the control of a Tagout Holder.
O.
- The component position can ONLY be changed with permission of the Tagout D.
Holder.
- The component has been designated as an isolation boundary for personnel safety.
(
Page: 63 of 100 oflOO 6/19/2009
64.
Given the following:
- Reactor power is 30%
- A reactor shutdown is in progress
- A high failure of IR NIS channel N-36 occurs Which ONE of the following is CORRECT regarding the effects of this failure as reactor shutdown progresses?
Power < P-10 Power < P- 6 A. Reactor trip occurs SR High flux trip automatically unblocks B. Reactor trip occurs SR High flux trip must be manually unblocked c.
C. Reactor does NOT trip SR High flux trip automatically unblocks D. Reactor does NOT trip SR High flux trip must be manually unblocked
(
Page: 64 of oflOO 100 6/1912009 6/19/2009
65.
Given the following:
- Waste Monitor Tank (WMT) # 9 release is in progress.
(
- The control room HIGH RADIATION annunciator illuminates.
RE-0018, Liquid Waste Effluent radiation monitor, has failed high.
- WMT 9 discharge flow transmitter, FT-1085A, indicates 35gpm.
35 gpm.
Per ARP 17213-1 for the PLPP annunciator window for WATER DISCH LINE HI RAD, the Auxiliary Building SO should ...
A. fail air to RV-0018.
OR close manual discharge valves to the environment.
B. fail air to RV-0018.
OR using the local manual handwheel, close RV-0018.
(
C. close RV-0018 locally on the PLPP.
OR close manual discharge valves to the environment.
D. close RV-0018 locally on the PLPP.
OR using the local manual handwheel, close RV-0018.
Page: 65 of 100 oflOO 6/19/2009
66.
Given the following:
- The unit is at 100% power.
- A containment entry is to be performed per 00303-C, "Containment Entry" for an RCS leak inspection.
The area inside the bioshield should be posted as a (a),
(a)._ _
Entry into _ _(b)_ _ is prohibited.
A. (a) Very High Radiation Area (Grave Danger)
(b) the Preaccess Filter Unit area (260 ft elevation).
B. (a) High Radiation Area (b) the Preaccess Filter Unit area (260 ft elevation).
C. (a) Very High Radiation Area (Grave Danger)
(b) The elevator.
(
D. (a) High Radiation Area (b) The elevator.
(
Page: 66 of 100 6/19/2009
67.
Given the following:
- IPC computer is unavailable, CSFST monitoring must be performed manually.
( - You are assigned 19200-C, "Critical Safety Function Status Tree" monitoring and are making your first pass through the CSFSTs.
Critical Safety Function Status Trees indicate the following:
- Subcriticality GREEN
- Core Cooling ORANGE
- Heat Sink Not performed yet
- Integrity Not performed yet
- Containment Not performed yet
- Inventory Not performed yet Which ONE of the following describes the:
(a) proper performance of CSFST monitoring on the first pass through.
(b) your responsibilities for CSFST monitoring in accordance with 19200-C and A. (a) Immediately tell the SS to transition to Core Cooling, then continue the first pass through the status trees.
(b) Continuous monitoring is required.
(
B. (a) Complete the first pass through the status trees, then inform the SS to transition to the highest priority FRP.
(b) Continuous monitoring is required.
C. (a) Immediately tell the SS to transition to Core Cooling, then continue the first pass through the status trees.
(b) Monitor every 10 to 15 minutes unless a change in status occurs.
D. (a) Complete the first pass through the status trees, then inform the SS to transition to the highest priority FRP.
(b) Monitor every 10 to 15 minutes unless a change in status occurs.
Page: 67 of oflOO 100 6/19/2009
68.
Given the following:
- Control Room has been evacuated due to a fire.
(
- AOP-18038-1, "Operation From Remote Shutdown Panels" is in effect.
- The operators have just aligned components per Attachment I, "Control Switch Required Positions".
Attachment I is performed at _ _(a)_ _ in ,in order to align _ _(b),
(b)_ _
A. (a) both Remote Shutdown Panels (b) Fire event qualified equipment to prevent spurious actuations.
- 8. (a) both Remote Shutdown Panels (b) switches with maintained contacts to prevent changing status on transfer.
C. (a) Remote Shutdown Panel "8" only (b) Fire event qualified equipment to prevent spurious actuations.
D. (a) Remote Shutdown Panel "8" only (b) switches with maintained contacts to prevent changing status on transfer.
Page: 68 oflOO of 100 6119/2009 6/19/2009
69.
The Emergency Director (ED) has directed you to perform an ENN roll call in accordance with 91002-C, "Emergency Notifications", checklist 4 "Directions for ENN Communicators" Communicators"..
(
- Burke County and State of Georgia have failed to respond to the initial roll call.
Which ONE of the following is the CORRECT actions to perform?
A. Transmit the notification message to the agencies that responded, inform the ED that two agencies couldn't be notified.
Southern Linc would be the next priority to establish communications.
B. Transmit the notification message to the agencies that responded, inform the ED that two agencies couldn't be notified.
The Back-up ENN Bridge would be next priority to establish communications.
C. Promptly notify the ED of the agencies that failed to respond to the initial roll call.
Southern Linc would be the next priority to establish communications.
( D. Promptly notify the ED of the agencies that failed to respond to the initial roll call.
The Back-up ENN Bridge would be next priority to establish communications.
Page: 69 oflOO 611912009 6/19/2009
70.
Given the following plant conditions:
- The Reactor has tripped and Safety Injection has actuated.
(
- 19012-C, "Post LOCA CoolCooldown down and Depressurization" is in progress.
- The crew is performing an RCS cooldown in accordance with the procedure.
The following is a plot of the cooldown:
Time RCS Tcold Time RCS Tcold 1400 549 0 F 1545 427 0 F 1415 532°F 1600 397 0F 397°F 1430 522 0F 522°F 1615 384 0 F 384°F 1445 507°F 1630 366°F 1500 500 0 F 1645 342 0F 342°F 1515 480 0 F 1700 312°F 1530 449 0 F 1715 282°F Which ONE of the following is CORRECT regarding the Tech Spec cooldown limits?
A. The crew is in compliance with Tech Spec cooldown limits.
B. The Tech Spec cooldown limits were first exceeded at 1530.
(
C. The Tech Spec cooldown limits were first exceeded at 1600.
D. The Tech Spec cooldown limits were first exceeded at 1715.
(
Page: 70 of 100 6/19/2009
71.
Which ONE of the following are components I/ parameters for a potential source of RCS leakage that are specifically checked for in 19112-C, "LOCA Outside Containment" and the main concern if this flow path is not isolable?
(
\
A. Safety Injection pump Cold Leg injection valves and RHUT level.
Loss of emergency coolant recirculation due to rupture of low pressure piping.
B. Safety Injection pump Cold Leg injection valves and RHUT level.
Unmonitored radioactive release to environment via the Auxiliary Building HVAC.
C. RHR pump Cold Leg injection valves and Reactor Coolant System pressure.
Loss of emergency coolant recirculation due to rupture of low pressure piping.
D. RHR pump Cold Leg injection valves and Reactor Coolant System pressure.
Unmonitored radioactive release to environment via the Auxiliary Building HVAC.
/
\
oflOO Page: 71 of 100 6119/2009 6/19/2009
72.
RCS Feed and Bleed has been initiated in accordance with 19231-C, "Response to Loss of Secondary Heat Sink".
( - All SG WR levels are 8% and lowering.
- "HI CNMT PRESS SI RX TRIP ADVERSE CNMT" alarm is lit.
- Core Exit TC temperatures are rising rapidly.
- Condensate Pump # 1 is running.
The CORRECT strategy the crew should use to re-establish feed to ...
A. all SG's and feed at 30 -100 gpm until WR level is>is > 31%.
31 %.
B. a selected SG and feed at 30 -100 gpm until NR level is> 32%.
C.
D. a selected SG and feed with no flow restrictions until NR level is> 32%.
(
of 100 Page: 72 oft 00 6/19/2009 611912009
73.
Initial conditions:
- Core Cooling CSFST is ORANGE.
(
\
- 19222-C, "FRP-C.2 Degraded Core Cooling" has been implemented.
- Steam Generator depressurization to 200 psig per 19222-C is in progress.
Current conditions:
- RCS Integrity CSFST RED status has been validated.
- The CSFST status tree path points to 19241-C, "Response to Imminent Pressurized Thermal Shock (PTS).
- The SS continues with the actions of 19222-C.
Which ONE of the following is CORRECT regarding the SS decision during performance of this procedure?
A. The RED condition on Integrity is expected due to the SI accumulators injecting causing rapid lowering of cold leg temperatures, the SS should complete 19222-C.
B. Transitions from FRPs cannot be made until completed or procedurally directed.
( C. Transitions to other procedures is prohibited until ECCS termination criteria is met.
D. The RED condition on Integrity is expected due to the SI accumulators injecting, causing a rapid rise in RCS pressure, the SS should complete 19222-C.
Page: 73 of 100 oflOO 6/19/2009 6/1912009
74.
Given the following:
in effect.
- 19241-C, "Pressurized Thermal Shock (PTS)", is in
(
- An RCS temperature instrument failure occurs.
- The OATC places both trains of COPS to the ARM position without noticing the instrument failure.
- PORV-455 opens.
The temperature instrument which failed is a ...
A. NR Thot B. WR Thot C. NR Tcold D. WR Tcold
(
(
74 of 100 Page: 74oflOO 6/19/2009
75.
Given the following:
- Reactor trip with LOSP in progress.
(
- The SS has directed the crew to perform a natural circulation cooldown using one of the two following procedures.
- 19002-C, "Natural Circulation Cooldown" OR
- 19003-C, "Natural Circulation Cooldown With Void In Vessel (With RVLlS)"
Which ONE of the following is CORRECT regarding cooldown rate, the procedure to utilize, and why?
A. If a cooldown rate of 50 0 F per hour is preferred, use 19002-C.
A lower cooldown rate will help prevent reactor vessel head void formation.
B. If acooldown a cooldown rate of 50 0 F per hour is preferred, use 19003-C.
A lower cooldown rate will help prevent reactor vessel head void formation.
(
C. If a cooldown rate of 10000 F per hour is preferred, use 19002-C.
A higher cooldown rate will help prevent reactor vessel head void formation.
D. If a cooldown rate of 10000 F per hour is preferred, use 19003-C.
A higher cooldown rate will help prevent reactor vessel head void formation.
oflOO Page: 75 of 100 611912009 6/19/2009
76.
Initial conditions:
- Unit is at 100% power
( - All systems are in automatic
- CVCS Letdown is 120 gpm
- The NCP is in service Current conditions:
- PRZR level is lowering
- FI-0121, charging line flow, is fluctuating
- NCP discharge pressure is low and fluctuating
- VCT level is rising
- The following annunciator is illuminated:
CHARGING LINE HI/LO HIILO FLOW Which ONE of the following is the CORRECT procedure to implement and strategy to mitigate the plant conditions?
A. Enter AOP 18004-C Section A for RCS Leakage in Mode 1.
Raise charging flow using FIC-0121, isolate letdown and start a CCP if necessary.
(
B. Enter AOP 18007-C Section B for loss of charging flow.
Isolate letdown, stop the NCP, vent charging pump suctions prior to restarting.
C. Enter AOP 18004-C Section A for RCS Leakage in Mode 1.
Raise charging flow using FIC-0121, reduce letdown flow to 75 gpm as necessary.
D. Enter AOP 18007-C Section B for loss of charging flow.
Isolate letdown, raise charging flow using FIC-0121, swap to a CCP if necessary.
(
Page: 76 of 100 6/19/2009
77.
Given the following conditions:
- The unit is shutdown with Tave at 557 degrees F
( - PV-8000A PV,;.8000A block valve is shut due to excessive seat leakage on PORV- 455
- The breaker for block valve PV-8000B (PORV- 456) has just tripped open
- A risk assessment has NOT been performed for the PV-8000B failure Which one of the following correctly describes allowable technical specification actions and the bases for those actions?
A. Separate condition entry IS allowed for each PORV. Reactor startup may proceed.
Bases require ONE PORV and its associated block valve to be capable of manual operation.
B. Separate condition entry is NOT allowed for each PORV. Reactor startup may NOT proceed. Bases require BOTH PORVs and their associated block valves to be capable of manual operation.
- c. Separate condition entry IS allowed for each PORV. Reactor startup may NOT proceed. Bases require BOTH PORVs and their associated block valves to be capable of manual operation.
D. Separate condition entry is NOT allowed for each PORV. Reactor startup may proceed. Bases require ONE PORV and its associated block valve to be capable of manual operation.
(
(
Page: 77 oft 00 oflOO 6/1912009 6/19/2009
78.
Given the following conditions with the unit at 100% power:
- PRZR pressure channel PT -455 has failed LOW
(
- NR cold leg temperature instrument for RCS loop 2 has failed LOW causing the Loop 2 Delta T indication to go off scale high
- All associated TSLB bistables for each failure are lit
- The unit continues to operate at full power Which one of the following describes the current plant condition and required technical specification entries?
- 3.0.3
- 3.3.1 RPS
- 3.3.2 ESFAS
- 3.3.3 PAMS
- 3.3.4 Remote SID A. An ATWT A TWT is in progress.
LCOs 3.0.3, 3.3.1 and 3.3.2.
(
B. An ATWT is NOT in progress.
LCOs 3.3.1, 3.3.2 and 3.3.4.
C. An ATWT is in progress.
LCOs 3.0.3, 3.3.1, 3.3.2 and 3.3.3.
D. An ATWT is NOT in progress.
LCOs 3.3.1, 3.3.2, 3.3.3 and 3.3.4.
(
Page: 78 of 100 6/19/2009
79.
DG-1A is running with its output breaker open and the unit at full power when the following indications are received:
- Train A MSIV's red & & green lights extinguish
- RTA red & & green lights extinguish
- RCP #1 1E breaker red & & green lights extinguish
- Channell Channel I TSLB bistable lights illuminate Which of the following is the correct...
a) procedure(s) to enter based on the above conditions, and b) corrective actions to take?
A. a) Enter 19000-C, E-O Reactor Trip or Safety Injection and 18034-1, Loss of Class 1 E 125 VDC Power.
b) Stop DG-1A using the Pull-To-Run/Push-To-Stop control, isolate the letdown relief flowpath, energize 1AY1 A and 1AY2A from their regulated transformers, control Tave T ave using SG ARVs 2 & & 3.
B. a) Enter 18034-1, Loss of Class 1E 125 VDC Power, only.
b) Emergency trip DG-1A from the engine control panel, isolate the letdown relief
( 1AY1 A and 1AY2A from their regulated transformers, control flowpath, energize 1AY1A Tave using SG ARVs 2 & & 3.
C. a) Enter 19000-C, E-O Reactor Trip or Safety Injection and 18032-1, Loss of 120V AC Instrument Power.
b) Emergency trip DG-1A from the control room, isolate the letdown relief flowpath, energize 1AY1A and 1AY2A from their regulated transformers, control Tave using the steam dumps.
D. a) Enter 18032-1, Loss of 120V AC Instrument Power, only.
b) Stop DG-1A using the Pull-To-Run/Push-To-Stop control, isolate the letdown relief flowpath, energize 1AY1A reliefflowpath, 1A Y1 A and 1AY2A from their regulated transformers, T ave using SG ARVs 2 &
control Tave & 3.
Page: 79 of 100 6/19/2009
- 80. Given the following conditions at 38% power:
- ACCW pump 2 is in service
& 4 are in service pu m ps 2 &
- CCW pumps
(( - NSCW Pump 5 is danger tagged Train A NSCW indications: Train B NSCW indications:
- Supply header pressure 45 psig - Supply header pressure 58 psig
- Supply header flow 8,000 gpm - Supply header flow 25,000 gpm
- Return header flow 8,000 gpm - Return header flow 10,000 gpm Which of the following choices contains the correct procedural entry and actions?
A. Enter AOP 18021-C, Loss of NSCW, due to loss of both NSCW Trains.
Place all NSCW pumps in PTL, trip the reactor and initiate EOP 19000-C. Trip the RCPs and isolate CVCS letdown.
B. Enter AOP 18021-C, Loss of NSCW, due to leakage on Train A NSCW.
Place all Train A NSCW pumps in PTL, trip the reactor and initiate EOP 19000-C.
Trip the RCPs and isolate CVCS letdown if cooling not restored in 10 minutes.
( C. Enter AOP 18021-C, Loss of NSCW, due to leakage on Train B NSCW.
Place all Train B NSCW pumps in PTL. Shift to Train A CCW pumps. Start ACCW pump #1 and remain in 18021-C.
D. Enter AOP 18021-C, Loss of NSCW, due to loss of both NSCW Trains.
Place NSCW Train B in single pump operation and all Train A NSCW pumps in PTL. Trip RCPs if seal temperatures exceed 230 F and remain in 18021-C.
Page: 80 oft 00 of 100 6/19/2009
- 81. Initial Conditions:
- A large LOCA has occurred
- Neither train of RHR could be aligned for cold leg recirculation
( - EOP 19111-C, Loss of Emergency Coolant Recirculation, has been implemented Current Conditions:
- RWST level is now 7% and lowering
- Integrity CSF Status Tree is Orange Which one of the following describes the correct actions to be taken based on the above conditions?
A. Stop all pumps taking suction from the RWST; Do NOT go to 19241-C, Response to Imminent Pressurized Thermal Shock, because the actions in 19111-C take priority over the Function Restoration Procedures (FRP's).
B. Reduce ECCS flow from the RWST to ONE (1) train running; Do NOT go to 19241-C, Response to Imminent Pressurized Thermal Shock, because the actions in 19111-C take priority over the Function Restoration
( Procedures (FRP's).
C. Stop all pumps taking suction from the RWST; Go to 19241-C, FR-P.1 Response to Imminent Pressurized Thermal Shock.
D. Reduce ECCS flow from the RWST to ONE (1) train running; Go to 19241-C, FR-P.1 Response to Imminent Pressurized Thermal Shock.
(
oflOO Page: 81 of 100 6/19/2009
82.
Initial conditions:
- Unit at 100% power for last 10 weeks
( - All rods out at 228 steps Current conditions:
- Rod at bottom alarm is alarming
- Control Bank 0 rod H-8 rod bottom LED lit
- T ave is lowering Tave
- QPTR & AFD remain within limits Which of the following choices identifies the correct procedure entry and actions to take?
A. Enter AOP 18003-C, Section A, Dropped Rods in Mode 1.
Do not exceed 75% thermal power during rod recovery, rod pulls are limited to 3 step increments. Reset the Bank Overlap Unit to restore the RIL alarm to operable status.
B. Enter AOP 18003-C, Section C, Misaligned Rods in Mode 1 1..
Do not exceed 65% thermal power during rod recovery, rod pulls are limited to 3
( step increments. Reset the Bank Overlap Unit to restore the RIL alarm to operable status.
C. Enter AOP 18003-C, Section A, Dropped Rods in Mode 1.
Do not exceed 75% thermal power during rod recovery, the 3 step rod pull limit may be suspended for this condition. Reset the PIA converter to restore the RIL alarm to operable status.
D. Enter AOP 18003-C, Section C, Misaligned Rods in Mode 1.
Do not exceed 65% thermal power during rod recovery, the 3 step rod pull limit may be suspended for this condition. Reset the PIA converter to restore the RIL alarm to operable status.
Page: 82 of 100 oflOO 6/19/2009
83.
Initial conditions:
- Reactor power is currently 8%
\ - Tave is on program for 8% power
- PRZR LO LEVEL DEVIATION is alarming
- The SS enters AOP 18001-C, section D, Failure of PRZR Level Instrumentation Which of the following contains the correct diagnosis and corrective actions for the indications given?
A. The controlling PRZR level channel has failed high.
Adjust charging to prevent letdown from flashing or isolate letdown.
Apply LCO 3.3.1 RTS Instrumentation actions for the failed channel.
B. The controlling PRZR level channel has failed low.
Adjust charging to prevent letdown from flashing or isolate letdown.
Apply LCO 3.3.1 RTS Instrumentation actions for the failed channel.
C. The controlling PRZR level channel has failed high.
Maintain charging flow approxiametely 10 gpm greater than total seal injection flow.
flowapproxiametely
( Write an INFO LCO 3.3.1 RTS Instrumentation for the failed channel.
D. The controlling PRZR level channel has failed low.
Maintain charging flow approximately 10 gpm greater than total seal injection flow.
Write an INFO LCO 3.3.1 RTS Instrumentation for the failed channel.
Page: 83 of 100 6/19/2009 611912009
84.
Following an RCS LOCA, the following conditions exist:
- CNMT pressure is 29 psig.
- CNMT Spray pump A is tagged out.
- 2 Train A Containment coolers are tagged out.
- CNMT Spray pump B is running.
- An Alert Emergency has been declared due to the Loss of RCS Barrier.
Su~denly:
Suddenly:
- 1BA03 normal incoming breaker trips open.
- DG1 B trips on overspeed.
REFERENCE PROVIDED Which one of the following choices decribes the correct procedures to be implemented and impact on emergency calssification?
"Response to High Containment Pressure" and AOP 18031-C "Loss A. FRP 19251-C "Responseto of Class 1E Electrical Systems" may be performed in parallel.
An EAL classification upgrade is required.
B. FRP 19251-C "Response to High Containment Pressure" and AOP 18031-C "Loss
( of Class 1E Electrical Systems" may be performed in parallel.
An EAL classification upgrade is NOT required.
C. FRP 19251-C "Response to High Containment Pressure" and AOP 18031-C "Loss of Class 1E Electrical Systems" may NOT be performed in parallel.
An EAL classification upgrade is required.
D. FRP 19251-C "Response to High Containment Pressure" and AOP 18031-C "Loss of Class 1E Electrical Systems" may NOT be performed in parallel.
An EAL classification upgrade is NOT required.
(
Page: 84 oflOO 6/19/2009
85.
Given:
- The crew is performing 1901 O-C, "Loss of Reactor or Secondary Coolant" due to a Reactor Coolant System LOCA
- The crew is at the step to determine if a transition to a subsequent recovery procedure is required.
Plant conditions:
- RCS pressure is 410 psig and stable
- RHR flow is reading 0 gpm on both trains
- RWST level is 38% and lowering slowly
- PRZR level is offscale low Which ONE of the following would be the CORRECT procedure(s) to perform based on the given conditions?
A. Go to 19012-C, "ES-1.2 Post-LOCA Cooldown and Depressurization".
A transition to 19111-C, "ECA-1.1 Loss of Emergency Coolant Recirculation" will be required.
B. Go to 19013-C, "ES-1.3 Transfer to Cold Leg Recirculation". When 19013-C is completed, transition back to 19010-C.
1901 O-C.
C. Go to 19012-C. Remain in 19012-C until the RCS cooldown is completed.
D. Go to 19013-C. A transition to 19111-C from 19013-C will be required.
Page: 85 of 100 6/19/2009
86.
Given the following CCW conditions with RCS temperature of 349 F:
Train A:
(
- CCW pumps 1 & 3 are running
- 1AA02 A loss power to 1AA02 occurs
- CCW pump 1 trips during the UV sequence
- CCW pump 5 had to be manually started Train B:
- CCW pump 2 is running with 4 & 6 in PTL
- CCW flow to the SFP HX is isolated
- CCW system pressure is 70 psig Which of the following choices correctly desribes the condition of the CCW system and appropriate technical specification actions to take (if any)?
A. Only CCW train A was inoperable during the time only pump 3 was running.
LCD 3.7.7, Component Cooling Water System, was not met until the standbyCCW LCO standby CCW pump was manually started.
B. Both CCW trains are inoperable.
LCD 3.0.3 until RHR HX pressure is adjusted to > 85 psig on train B. Then Apply LCO apply LCO LCD 3.7.7, Component Cooling Water System, for CCW train A.
C. Both CCW trains are inoperable.
Apply LCO LCD 3.0.3 and restore CCW train B to 2 pump operation. Then apply LCDLCO 3.7.7, Component Cooling Water System, for CCW train A.
D. Both CCW trains are operable.
No technical specification actions are required for the CCW system.
Page: 86 of 100 oflOO 6/19/2009
87.
Given the following:
- RCP's were tripped based on EOP 19000-C foldout page requirements
- Crew has transitioned to EOP 19031-C, ES-3.1 ES-3:1 Post-SGTR Cooldown Using Backfill
- Crew is contemplating starting an RCP Which of the following is the correct action to take and why?
A. Start RCP 4.
This RCP aids in mixing contents of the stagnant loop while minimizing a challenge to core reactivity requirements.
B. Start RCP 1.
This RCP aids in mixing contents of the stagnant loop while minimizing a challenge to reactor vessel integrity requirements.
C. Do not start an RCP.
Inadvertent criticality may occur if an RCP is started following a natural circulation
(
'(
cooldown.
D. Do not start an RCP.
RCP seals may be damaged due to attack from secondary chemicals due to the backfill cooldown.
(
Page: 87 of 100 6/19/2009
88.
Given the following:
- Unit is in mode 1 at 8% power
( - 1BY2B and 1DY1 B de-energize due to bus faults The correct actions to take for this situation are:
A. Enter AOP 18032-1, Loss of 120V AC Instrument Power.
Enter LCO 3.8.9 Distribution Systems - Operating.
BOTH 1BY2B and 1DY1 B must be re-energized within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, otherwise the unit needs to be shutdown.
19000~C, Reactor Trip or Safety Injection, B. Complete EOP 19000-C, then go to AOP 18032-1, Loss of 120V AC Instrument Power.
Enter LCO 3.8.9 Distribution Systems - Operating.
EITHER 1BY2B or 1DY1 B must be re-energized within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> from its associated inverter.
C. Enter EOP 19000-C, Reactor Trip or Safety Injection and AOP 18032-1, Loss of 120V AC Instrument Power, concurrently.
(,
Enter LCO 3.0.3 for this condition due to a loss of safety a function function..
EITHER 1BY2B or 1 DY1 B must be re-energized within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> from its associated inverter.
D. Enter AOP 18032-1, Loss of 120V AC Instrument Power 3~0.3 for this condition due to a loss of a safety function.
Enter LCO 3.0.3 BOTH 1BY2B 1BY2B and 1DY1 B must be re-energized within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, otherwise the unit needs to be shutdown.
Page: 88 oflOO of 100 6/19/2009
89.
Loop 1 SG NR level channels L LT-517 T-517 (red bezel) and LLT-518 T-518 (red bezel) have failed low with the unit in mode 3. All other instrumentation is operable.
Determine the required technical specification actions for these failures.
REFERENCE PROVIDED A. repair one channel within 7 days, repair the second channel within 30 days.
B. repair both channels within 30 days.
- c. repair one channel within 7 days.
D. repair one channel within 30 days.
(
of 100 Page: 89 oflOO 6/19/2009 6/1912009
- 90. Initial condifions:
conditions:
- Unit 1 is at 57% power
- All systems are in their normal alignment for this power level Current conditions:
- 1CD1M CD1 M de-energizes due to a bus fault
- HS-5106 TDAFW Pump Steam Admission Valve indicating lights are dark
- The SS is implementing LCO 3.7.5 for the AFW system.
Which of the following is the correct LCO 3.7.5 condition to enter and the status of the TDAFW Pump governor valve?
A. condition B to declare the TDAFWTRAIN TDAFW TRAIN inoperable.
TDAFW Pump governor valve is shut.
B. condition A to declare one TDAFW Pump STEAM SUPPLY inoperable.
TDAFW Pump governor valve is open.
C. condition B to declare the TDAFW TRAIN inoperable.
( TDAFW Pump governor valve is open.
D. condition A to declare one TDAFW Pump STEAM SUPPLY inoperable.
TDAFW Pump governor valve is shut.
Page: 90 oft 00 of 100 6119/2009 6/19/2009
- 91. Given the following conditions:
- Low main condenser vacuum exists
- AOP 18013-C, "Rapid Power Reduction" Reduction" is in progress
- Reactor power has been reduced from 100% to 70% in 10 minutes
- Boration is in progress
- RCS Tave is 5 F below Tref
- OATC has requested permission to withdraw rods 3 steps to raise Tave The OATe Based on these conditions the SS should ...
A. not approve the rod withdrawal, direct the UO to lower turbine load to match Tave to Tref.
Notify chemistry to sample the Reactor Coolant System to verify DOSE EQUIVALENT 1-131 within technical specification limits.
B. approve the rod withdrawal to restore Tave to Tref.
Notify chemistry to sample the Reactor Coolant System to verify gross activity within technical specification limits.
C. not approve the rod withdrawal, direct the UO to lower turbine load to match match Tave to Tref.
(
Notify chemistry to sample the Reactor Coolant System to verify gross activity within technical specification limits.
D. approve the rod withdrawal to restore Tave to Tref.
Notify chemistry to sample the Reactor Coolant System to verify DOSE EQUIVALENT 1-131 within technical specification limits.
Page: 91 of 100 6119/2009 6/19/2009
92.
Given the following conditions:
- WMT 9 release is in progress
(\. - SS discovers the steps for recirculating WMT 9 in SOP 13216-1, "Liquid Waste Release" were marked N/A What are the impacts and what actions are necessary to correct the consequences of this error?
A. The release permit values are inaccurate. This could result in radioactive nuclides being released to UNRESTRICTED areas greater than license limits.
The release should be stopped, the tank needs to be recirculated for approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> prior to sampling.
B. The activity level of the WMT contents are not at the lowest possible level. This could result in release of radioactive nuclides> ALARA to UNRESTRICTED areas.
The release may continue. The reason for the N/A steps should be annotated on the release permit.
C. The activity level of the WMT contents are not at the lowest possible level. This
( could result in radioactive nuclides being released to UNRESTRICTED areas greater than license limits.
The release may continue. The flowrate must be reduced to the minimum obtainable flow possible.
D. The release permit values are inaccurate. This could result in radioactive nuclides being released to UNRESTRICTED areas greater than license limits.
The release should be stopped. The values used in the release permit need to be recalculated using two independent methods prior to restarting the release.
(
Page: 92 of 100 6/19/2009
93.
A release of Waste Gas Decay Tank # 1 is in progress when the relief valve on Shutdown Decay Tank # 9 fails open causing pressure in Shutdown Decay Tank #9 to rapidly lower.
(
Which ONE of the following is the CORRECT action to take per 001S2-C, 00152-C, "Federal and State Reporting Requirements"?
REFERENCE PROVIDED A. Notify the NRC of the relief valve failure causing an unplanned and monitored release.
B. Notify the NRC of the relief valve failure causing an unplanned unplanned and unmonitored release.
C. Notify the NRC if ARE-0014 exceeds the release permit values, otherwise NRC notification is not required.
D. Notify the NRC if RE-12442C exceeds the release permit values, otherwise NRC notification is not required.
(
r
\
Page: 93 of 100 6/19/2009
94.
Initial conditions:
- The need to generate a new standing order has been identified
- The standing order will provide temporary instructions not covered by a plant procedure.
Which of the following choices is correct concerning review and approval of this new standing order?
A. The duration of the standing order will be limited to 14 days.
The Shift Manager will review and approve. A 10CFR OCFR 50.59 screening is not required.
B. The standing order will not have a termination date.
The Unit Superintendent will review and approve. A 110CFR OCFR 50.59 screening is required.
C. The duration of the standing order will be limited to 14 days.
The Unit Superintendent will review and approve. A 10CFR 10CFR 50.59 screening is not required.
(
D. The standing order will not have a termination date.
The Shift Supervisor will review and approve. A 1 10CFR OCFR 50.59 screening is required.
Page: 94 of 100 6/19/2009
95.
During a review of the surveillance schedules it is discovered that a 31 day surveillance for a particular component was last performed 42 days ago. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of this discovery, the surveillance was performed.
(
- The surveillance failed because it did not meet surveillance criteria.
- Two hours later, a valve was adjusted and retesting provided satisfactory results.
Which one of the following is correct for this condition?
The component was INOPERABLE _ _ _ _ _ _ _ until satisfactory retesting was completed.
A. for the last 42 days B. from the time the grace period expired c.
C. from the time of the failed surveillance D. from the time of identifying the surveillance NOT performed
(
Page: 95 oflOO of 100 6/1912009 6/19/2009
96.
The following equipment is inoperable at 100% power:
- Boric Acid Transfer Pump #1
- CCP- A CCP-A
- RCP Common Thermal Barrier Isolation Valve, HV-2041
- TOAFWP TDAFWP A Loss Of Safety Function (LOSF) evaluation is required for the (a) because it is (b)
A. a. TOAFWP TDAFWP
- b. a Technical Requirement (TR) supported component for the BAST C. a. CCP-A CCP- A
- b. a Technical Specification (TS) required component supported by DG-OG- A
(
D.
O. a. RCP Thermal Barrier Isolation Valve
Page: 96 of 100 6/19/2009
97.
Initial conditions:
- An emergency has been declared
- CNMT pressure is 14 psig and is slowly lowering
- The only available CNMT Spray pump motor has high temperature alarms
- crew' suspects inadequate venting of the motor cooler The crew
- An operator has been selected to vent the motor cooler
- The operator has received 1000 mrem TEDE, but has received NO dose during this event.
Per 91301-C, " Emergency Exposure Guidelines" what is the maximum dose the operator can receive to protect this equipment?
A. 9 Rem B. 10 Rem C. 24 Rem D. 25 Rem
(
Page: 97 of 100 oflOO 6/19/2009
98.
Initial conditions:
- Unit tripped from full power due to an RCS LOCA
( - CNMT radiation monitors RE-005 & & RE-006 indicate 8.1 E+6 mr/hr Determine the appropriate emergency classification based only on the conditions listed above.
REFERENCE PROVIDED A. General B. Site Area C. Alert D. Unusual Event
(
(
Page: 98 of 100 611912009 6/19/2009
- 99. Given the following conditions:
- A loss of All AC power has occurred.
( - Intact SGs are being depressurized to 300 psig per 191 OO-C, Loss of All AC Power.
- PRZR level has rapidly increased from 0% to 55%
Which ONE of the following is the CORRECT action to take in this situation and the reason for the action?
A. SG depressurization should continue.
This is the desired response in order to refill the PRZR and regain control of RCS pressure once the PRZR heaters are covered with water.
B. SG depressurization should be stopped.
This is to prevent loss ofRCS of RCS subcooling and loss of core heat removal.
C. SG depressurization should continue.
This is an expected response to the SG depressurization as a result of upper head voiding in the reactor vessel.
(
D. SG depressurization should be stopped when PRZR level.:::level ~ 75%.
This is to prevent going water solid in the PRZR and ensure that a PRZR steam bubble is available to help control RCS pressure.
Page: 99 of 100 6/19/2009
100.
Initial conditions:
- The unit tripped from 100% power
- Loss of all off-site power occurred
- Both EDG's are powering their respective loads Current conditions:
- RCS pressure 2235 psig and stable
- RCS loop hot leg temperatures 602°F and stable
- RCS loop cold leg temperatures 560°F and stable
- Core exit TCs are 610°F and stable
- SG pressures are 1120 psig and stable Which one of the following choices is correct regarding:
- a. The status of natural circulation?
- b. Procedure used to establish or maintain heat removal?
A. Natural circulation is established.
Verify that natural circulation exists in accordance with Attachment B of 19000-C, Reactor Trip or Safety Injection.
(
B. Natural circulation is not established.
Establish natural circulation in accordance with Attachment B of 19000-C, Reactor Trip or Safety Injection.
C. Natural circulation is established.
Verify that natural circulation exists in accordance with Attachment B of 19001-C, Reactor Trip Response.
D. Natural circulation is not established.
Establish natural circulation in accordance with Attachment B of 19001-C, Reactor Trip Response.
Page: 100 of 100 oflOO 6/19/2009
06/22/2009 10:29 FAX 706 826 3953 VEGP TRAINING CENTER I4i 005 141005
- - - - - - ~--Answers
~-- Answers - ,
1, 001A4.1211 001A4.12 A
( "
2 002G2.1.3 002G2. J.3 1I 1I D 3 003AK2.051 B
.B 4 003K6.141 B 5 004K5.11 1 D 6 004K5.15 1 B 7 005K2.031 B 8 006A4.11 1 B 9 007EA2.0S I B 10 007Kl.01 007K1.01 I B 11 OOSA4.021 008A4.021 B 12 008AK2.C22 D 13 008K3.03 1 C 14 009EG2.4.341 D IS 1S OIOK2.041 010K2.041 A A
16 OlOK6.01 01OK6.01 1 B 17 OllEG2.2371 OIlEG2.2.371 D 1~
l~ 012K3.041 O12K3.041 B 19 013KL071 013K1-071 D 20 015AG2.'1.6 01SAG2.4.611 D 21 015K2.01 OI5K2.01 1 C 22 022Al.021 022Al.011 C 23 022AAl.C21 C 24 02SAK.l.C 025AKl.C 1 1 D
( 25 026K4.01 1I C 26 027AK3J2 2 B 27 028KS.04 1 028KS.041 C 28 029EK2.C66 1 029EK.2.C B 29 029KL051 D 30 032AA2.(*92 032AA2J92 C 31 033AI.OI 033Al.01 1 B 32 035K6.01 1 B 33 039G2.2.::5 2 A A
34 051AAJ.(
051AAU'42 142 B 35 055EK1.C 055EK1.( 1 1 A 36 055K3.01 1 C 37 056AA2.:9 056AA2.:91I D 38 057AAUl4 1 057AAU14 A 39 058AA2.!ll 058AA2.(JJ 1 D 40 059A3.041 A 41 061Al.012 D 42 06lG2.L" 06lG2.1." 1 C 43 062AK3.o3 062ATG.03 1 B 44 062K4.021 062K4.02 I A A
45 063A2.01 1I C 46 064K6.0~ Z 2 C 47 065AAl.!}31 065AAl.o31 D
( 48 068AG2A.62 068AGZ!L62 D I"age:
J."age: 10f3 lof3 612.2/2009 61:2.2/2009
06/22/2009 10:29 FAX 706 826 3953 VEGP TRAINING CENTER I4i 141 006
- -----,---Answers
,--- Answers - - - -
, " '"I'I'I'II'I;rI'il'' ",', " .. '::'1'1'
~
" '::,::;: ::*:i.iW;61fhfi,t*1j ':'f',: .,': :::,,; 'H: :~::!:l~
49 071A2.081 B
( SO 50 073K5.01 I A 51 076A2.012 B 52 076A2.021 B 53 076AA2.032 D 54 077AK2.041 077AK2.C41 A 5S 55 078A3.01 1 A A
56 078G2.4,31 078G2.4.31 1 D 57 086A3,01 086A3.01 1 C 58 103A1.01 1 C 59 103A4.032 J03A4.032 D 60 G2.1.8 02.1.8 1 A 61 G2.1.291 D 62 02.1.301 G2.1.30 1 C 63 G2.2.13 2 C 64 G2.2.441 B 65 G2.3.11 G2.3.1111 C 66 02.3.141 C 67 G2.4.13 02.4.13 1 B 68 G2.4.341 B 69 02.4.431 G2.4.43 1 D 70 WE03EAJ WE03EA.,3 1 C 71 WE04EK;.4 I WE04EK,.41 C 72 WE05EK.3 2 D 0
( 73 WE06EK::.4 3 A 74 WE08EK~.111 WE08EKU B 75 WE09EK ... .2211 A 76 022AA2.C22 B 77 027AG2.1.321 C 78 029EG2.2.22 1 A 79 058AA2.C31 058AA2.C3 1 A 80 062AA2.C21 A 81 WEIIEG:~.4.2 1 WEllEGtA.21 C 82 003A02.4.35 003AG2 ..<1.35 1 C 83 028AA2.C81 028AA2.C8 1 D 84 069AGV.82 A 85 WE03EA!.22 C 86 00SAl.071 008Al.071 C 87 012G2.4.:0 o 12G2.4.:0 1 A 88 013A2.041 013A2.04 1 A 89 05902.4.-'
059G2.4.c 2 D 0
90 061A2.03 1 C 91 055G2A.J 055G2.4.l 12 12 A 92 068A2.021 A 93 071A2.092 A 94 G2.1.151 B 95 9S G2.2.221 C
( 96 G2.2.381 C Page: 20f3 6/22/2009
06/22/2009 10:29 FAX 706 826 3953 VEGP TRAINING CENTER 141 007 141007
-- --~-~--~--*.*~-
- - Answers ----,-~~~---."--
97 G2.3.4 2 8B
( 98 G2.3.5 1 G2.3.51 C
B 99 02.4.203 G2A.20 3 100 G2.4.472 02.4.472 C
(
Page: 3 of3 6/22(2009 6/22{2009
PROCEOURE NO.
PROCEDURE REVISION NO. PAGE NO.
VEGP 18019-C 26.1 41 of 60 Sheet 1 of 1
(
ReS HEAT-UP RATE H 10 E
A 9 \
T U
8 ~\
P 0 7 D
E
\,
G 6 R
, r\\ \
I FULL SPENT CORE I E ~\. "'- ...........
"'i'-....,
E 5
\ '-
S F
P 4
3
~
"-I'-...
~
I'--...
r--
"""- r--
r-
"""- r-- --.
r-- --.
r--
E 2 .~I RELOADED RE LOAD ED CORE CO RE IL R
M 1
( I N
U 0 T 0 50 100 150 200 250 300 350 400 450 500 550 600 650 700 750 800 850 900 9501000 E TIME AFTER SHUTDOWN (HRS)
Assumptions:
- 1) Mid Loop Conditions
- 2) RCS ventedVented To Atmosphere with or without Loop Dams FIGURE 1
PROCEDURE NO. REVISION NO. PAGE NO.
VEGP 18019-C 26.1 42 of 60 Sheet 1 of 1 CORE FLOW TO MAINTAIN 195 Deg F vs. VS.
TIME AFTER REACTOR SHUTDOWN 1440 1360 G1280 P
M1200 \
1120 \
T H1040
\
R R 960 ~
\
0 U 880
\
G 800 \ ~
H \,
720 T 640 H
1\
\.
~
f"".. ,
E 560 C
480 0 400
" "- i'o..
~
r-..... r--- """'-
r-- I FULL H+-L SPENT CORE I
I I RELOADED CORE I --
I R 320
( E 240 -
160 0 40 80 120 160 200 240 280 320 360 400 440 480 520 560 600 640 680 720 760 800 TIME AFTER SHUTDOWN (HRS)
Assumptions:
- 1) Mid Loop Conditions
- 2) RCS Vented To Atmosphere
- 3) Injection Flow Assumed a 100 Degrees F From RWST FIGURE 2
PROCEDURE NO. REVISION NO. PAGE NO.
VEGP 18019-C 26.1 43 of 60 Sheet sheet 1 of 1
(
RCS TIME TO BOILING (FULL SPENT CORE) 90 M 80 I
N U 70 T
E 60 S INITIAL TEMP 75 T 50 lillLll1JJ.lW.H+/-tttt!
II III II 11l.LW+/-t+/-+/-tttTIT1
~IIIIIIIIIIIIIIII
~IIIIIIIIIIIIIIIII 0 INIT~
INITA~
lillill 1 I I I I I l l I I I-1111 II S 40 A ~ITAL TEMP 125 T 30 IIII1111111 IIIII11111I ImrrrTI JJJJI"llffi U INITAL TEMP 150 R
A 20 T
( I 0 10 N
0 0 50 100 150 200 250 300 350 400 450 500 550 600 650 700 750 800 850 900 9501000 TIME AFTER SHUTDOWN (HRS)
Assumptions:
- 1) Full Spent Core Heat Load
- 2) Mid Loop Conditions
- 3) RCS Vented To Atmosphere with or Without Loop Dams FIGURE 3
PROCEDURE NO. REVISION NO. PAGE NO.
VEGP 18019-C 26.1 44 of 60 Sheet 1 of 1
(
RCS TIME TO BOILING (RELOADED CORE) 90 11111111111111111 IIIIIIIIIIIIIIII INITIAL TEMP 75~
M 80 I
N INITAL TEMP 100~
U 70 T
E 60 S
INITAL TEMP TE~ 125.___
IIIIIIIIIIIL~-IIII IIIIIIIIIIII~ IIII T 50 0 u.H+rrH11T111111111 u.H+Htt11T11111111111111IIII 1111111111111111111111111 I IIIIIIIIIIIIIIIIIII IIII INITAL TEMP 150 S 40 A
T 30 U
R A 20 T
( I 0 10 N
0 0 50 100 150 200 250 300 350 400 450 500 550 600 650 700 750 800 850 900 9501000 TIME AFTER SHUTDOWN (HRS)
Assumptions:
- 1) Reloaded Core Heat Load
- 2) Mid Loop Conditions
- 3) RCS Vented To Atmosphere with With or without Without Loop Dams
(
FIGURE 4
PROCEDURE NO. REVISION NO. PAGE NO.
VEGP 18019-C 26.1 45 of 60 Sheet 1 of 1
(
TIME TO CORE UNCOVERY (RCS TEMPERATURE AT SATURATION)
M360 M360 I340 N320 U
T300 T300 E 2S0 E280 S260 T240 T240
, ...- I-I
...- I- -
RELOADED CORE
...- ...- ~
~ l- -
~ -
.... ~-:; ~
~~ ;;;;
0220 ,
1,...000 C200 ...- , -
~
... ~
I FULL SPENT CORE
- r 0 180 01S0
,~
i--" ~ """
...- ~""'"
i.--'
I- .... - -
.... .... t-I-
R160
, .",. ~
~~
I-"" i"""
I-"" f-"
E E140 U120 N100 N100 C
C SO 80 , 1,/
./
V" V'
V
~
~
~
~~
.",. !o"'"
ioo-"
- ~
i--' ~ ~ --
f-" ""'" - f-" ~ f-" -- ~
0 ..... ~ !o"'"
V 60 .. ~
~",
( E 40 R 20 Y
0 0 50 100 150 200 250 300 350 400 450 500 550 600 650 700 750 SOD 800 S50 850 900 9501000 TIME AFTER SHUTDOWN (HRS)
Assumptions:
- 1) Initial RCS Temperature is 212 Degrees F
- 2) Initial RCS Level at Mid-LOOp Mid-Loop
- 3) RCS Ventedvented to Atmosphere with or without Without Loop Dams FIGURE 5
VogtleElectric b~n~rath1g Plant Approved Pr, e Number Rev J.D. Williams , ':,\"",,,,, <,,'. '"",,,,,,,,,, ""'{<',,\.,,"
.\ 91001-C 91001*C 31 Date Approved Page Number EMERGENCY CLASSIFICATION AND IMPLEMENTING INSTRUCTIONS 10/13/2008 10 of 108 REFERENCE USE_u~ USE
- 1. and 3. y OR
- -........1 (p.32) 1----1 1
CORE COOLING CSFST RED N LOSS
...... of CLAD
- 2. Coolant Activity Sample> 300 ).lCi/gm /-lCi/gm Equivalent I1-131
-131 .....;(u;.IP.;....;.3~2"--)_ _ _ _ _ _ _ O_R--II OR BARRIER I
- 5. Containment Radiation Monitors RE-005/006 > 6.0 E+6 mr/hr _\:aJ;I'""...;...-;...;-;.L, (p.33) _ _ _ _ __ ...J OR
~~~----------~
--------------_. OR OR
- 1. and 4. y (p.32)
CORE COOLING ...._......1 I POTENTIAL CSFST ORANGE E N 1----1.......... LOSS of ....
- 7. Judgment: Judgment by the ED that the Fuel Clad Barrier is Lost or Potentially Lost. Consider conditions not addressed and inability to determine the status of the Fuel Clad Barrier (p.33)
FIGURE 1 - FUEL CLADDING INTEGRITY (Modes 1, 2, 3 and 4 only)
Pnnted June 19, 2009 at 11 :41 Printed
- Approve, Approvel. '<;'"'b+" _:. ...:. ,.~~, ".,{'(-/:'?> "~~*".':"{i.t:\;/<,:+(~~<;-i~f~'/ ~H\?:\,ijP"" ..... \ ':. :.- F-. ,re Number Rev
- :/':'i"i, J.D. Williams
'" ~
y
,f'~j Vggtl <<r,. . I¥J~~tr' c, u en e rat in 9 PIa!} ~'j;:~0i,fit;~~0i~ rt~?;;;'? 'i~:~ \' .... .>;?L\~\~:;\ 91001-C 31 Date Approved Page Number EMERGENCY CLASSIFICATION AND IMPLEMENTING INSTRUCTIONS iNSTRUCTIONS 10/13/2008 11 of 108
[I - REFERENCE REFERENCE USE I
[38 ° F ADVERSE] (p.33)
....... of
~
ReS RCS ......
- 4. Containment Radiation Monitors RE-005/006 > 2.0 E+4 mr/hr (p.34) OR BARRIER
-- - --- - -- --- - - - - -------- - - -------- - - - OR
- 1. OR HEAT SINK I: IY (p. 33)
(p. 33) POTENTIAL CSFST RED N
.... lOSS LOSS of .....
(p. 33)
(p. 33) BARRIER
- 2. NON-Isolable RCS leak (including SG tube leakage) GREATER THAN the Capacity of One Charging Pump in the normal charging mode (p.34) OR OR
- 5. Unexplained level rise in any of the following: containment sump, Reactor Coolant Drain Tank (RCDT),
Waste Holdup Tank (WHT) (p.34)
- 6. Judgment: Judgment by the ED that the RCS Barrier is Lost or Potentially Lost. Consider conditions not addressed and the inability to determine the status of the RCS Barrier (p. 34)
FIGURE 2 - REACTOR COOLANT SYSTEM (RCS) INTEGRITY (Modes 1, 2, 3 and 4 only)
Pnnted June 19, 2009 at 11 :41 Printed
- Approve, J.D. Williams ',:t{~< ' :d<
' , M , " ; , , ' , ' ""
Vogtle Electric vdn~l1~ting Plant .\ r ,re Number Rev 91001-C 31 Date Approved Page Number 10/13/2008 EMERGENCY CLASSIFICATION AND IMPLEMENTING INSTRUCTIONS 12 of 108 I REFERENCE USE I
- 2. Rapid Unexplained Containment Pressure Decrease Following Initial Pressure Increase (p.35) OR LOSS
- 2. Intersystem LOCA indicated by Containment Pressure or Sump Level Response Not Consistent OR With a Loss of Primary or SecondaryCoolant Secondary Coolant (p. 35) of
-+ ........
- 4. Primary-to-Secondary Leakage> 10 gpm AND a Non-Isolable Steam Release of Contaminated OR Secondary Coolant is Occurring to the Environment (p. 36)
BARRIER
- 5. Containment Isolation Valve(s) or Damper(s) are NOT Closed Resulting in a Direct Pathway to the OR Environment After Containment Isolation is Required (p. 37)
- 7. Pathway to the environment exists based on VALID RE-2562C, RE-12444C, OR RE-12442C Alarms. (p.37)
- - -------------------- - - - - - - - -- OR y OR
- 1. CONTAINMENT CSFST RED (p.35)
(p. 35)
N POTENTIAL LOSS of
~
CNTMT ...
AI
"""I spray operable (p.35)
<62% (p.37)
- 2. Containment Pressure> 43 psig (p.35) OR
- 6. Containment Radiation Monitors RE-005/006 > 2.4 E+8 mr/hr (p.37) OR
- 8. Judgment: Judgment by the ED that the CNTMT Barrier is Lost or Potentially Lost. Consider conditions not addressed and inability to determine the status of the CTMT Barrier (p. 37)
FIGURE 3 - CONTAINMENT INTEGRITY (Modes 1, 2, 3 and 4 only)
Printed June 19, 2009 at 11:41 11 :41
PAM Instrumentation 3.3.3 3.3 INSTRUMENTATION 3.3.3 Post Accident Monitoring (PAM) Instrumentation LCO 3.3.3 The PAM instrumentation for each Function in Table 3.3.3-1 shall be OPERABLE.
APPLICABILITY: MODES 1,2, and 3.
ACTIONS
NOT E---------------------------------------------------------
Separate Condition entry is allowed for each Function.
CONDITION REQUIRED ACTION COMPLETION TIME A. One or more functions A.1 Enter the applicable Immediately with one or more Condition referenced in required channels Table 3.3.3-1 for the inoperable. channels.
B. -----------NOT E------------
E------------ B.1 Restore the channel to 30 days For containment OPERABLE status.
isolation valve position indication, separate Condition entry is allowed for each penetration flow path.
One required channel inoperable.
(continued)
Vogtle Units 1 and 2 3.3.3-1 Amendment No. 137 (Unit 1)
Amendment No. 116 (Unit 2)
PAM Instrumentation 3.3.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. One RCS T hot channel That C.1 Restore the channel to 30 days inoperable. OPERABLE status.
AND One channel of core exit temperature per quadrant OPERABLE.
D. Teold channel One RCS Tcold D.1 Restore the channel to 30 days inoperable. OPERABLE status.
AND One channel of steam line pressure OPERABLE in the affected loop.
( E. One SG Water Level E.1 Restore the channel to 30 days (wide range) channel OPERABLE status.
AND One channel of AFW flow to the affected SG OPERABLE.
(continued)
Vogtle Units 1 and 2 3.3.3-2 Amendment No. 96 (Unit 1)
Amendment No.7 4 (Unit 2)
No. 74
PAM Instrumentation 3.3.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME F. One Steam Line F.1 Restore the channel to 30 days Radiation Monitor OPERABLE status.
channel inoperable.
AND One channel of SG Water Level (narrow range) OPERABLE in the affected loop.
G. Required Actions and G.1 Initiate action in accordance Immediately associated Completion with Specification 5.6.8.
Times of Conditions B, C, D, E, or F not met.
H. -----------NOTE
NOT E------------ H.1 Restore at least one 7 days For containment channel to OPERABLE
( isolation valve position status.
\. indication, separate Condition entry is allowed for each penetration flow path.
Two channels inoperable.
OR (continued)
Vogtle Units 1 and 2 3.3.3-3 Amendment No. 96 (Unit 1)
NO.7 Amendment No. 4 (Unit 2) 74
PAM Instrumentation 3.3.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME H. (continued)
NOTE------------
Applicable to those functions with only one required channel per loop, SG, or steam line.
One channel inoperable and no diverse channel OPERABLE.
I. -----------NO T E------------
NOTE------------ 1.1 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Not applicable to Containment Radiation and RVLlS functions.
Required Action and associated Completion
( Time of Condition H not met.
J. -----------NOTE------------
NO T E------------ J.1 Initiate action in accordance Immediately Applicable to with Specification 5.6.8.
Containment Radiation and RVLlS functions only.
Required Action and asociated Completion Time of Condition H not met.
Vogtle Units 1 and 2 3.3.3-4 Amendment No. 134 (Unit 1)
Amendment No. 113 (Unit 2)
PAM Instrumentation 3.3.3 SURVEILLANCE REQUIREMENTS
NOTE------------------------------------------------------------
SR 3.3.3.1 and SR 3.3.3.2 apply to each PAM instrumentation Function in Table 3.3.3-1.
SURVEILLANCE FREQUENCY SR 3.3.3.1 Perform CHANNEL CHECK for each required 31 days instrumentation channel.
SR 3.3.3.2 ----------------------------NOT E-------------------------------
NOTE-------------------------------
Neutron detectors are excluded from CHANNEL CALIBRATION.
Perform CHANNEL CALIBRATION. 18 months
(
Vogtle Units 1 and 2 3.3.3-5 Amendment No. 96 (Unit 1)
Amendment No. 74 (Unit 2)
PAM Instrumentation 3.3.3 Table 3.3.3-1 (page 1 of 1)
Post Accident Monitonng Monitoring Instrumentation FUNCTION REQUIRED CHANNELS CONDITIONS
- 1. Reactor Coolant System (RCS) Pressure (wide range) 2 B,G,H,I
- 2. ThO! (wide range)
RCS Thot 1/1oop 11100p C,G,H,I
- 3. RCS Teold (wide range) 1/1oop 11100p D,G,H,I
- 4. Steam Generator (SG) Water Level (wide range) 1!SG 1/SG E,G,H,I
- 6. Pressurizer Level 2 B,G,H,I
- 7. Containment Pressure 2 B,G,H,I
- 8. Steam line Pressure 2!steam line 2/steam B,G,H,I
- 9. Refueling Water Storage Tank (RWST) Level 2 B,G,H,I
- 10. Containment Normal Sumps Level (narrow range) 2 B,G,H,I
- 11. Containment Water Level (wide range) 2 B,G,H,I
- 12. Condensate Storage Tank Level 2!tank la )
2/tank(a) B,G,H,I
- 13. Auxiliary Feedwater Flow 2!SG 2/SG B,G,H,I
- 14. Containment Radiation Level (high range) 2 B,G,H,J
- 15. Steam line Radiation Monitor 1!steam line 1/steam F,G,H,I
- 16. RCS Subcooling 2 B,G,H,I
- 17. Neutron Flux (extended range) 2 B,G,H,I
- 18. Reactor Vessel Water Level (RVLlS) 2 B,G,H,J
- 19. Deleted
- 20. Containment Pressure (extended range) 2 B,G,H,I 2/penetration flow path 2!penetration lb ) Ie) path(b) (e) B,G,H,I
- 21. Containment Isolation Valve Position
- 22. Core Exit Temperature - Quadrant 1 2(d) 21d) B,G,H,I
- 23. Core Exit Temperature - Quadrant 2 2(d) 21d) B,G,H,I
- 24. Core Exit Temperature - Quadrant 3 2(d) 21d) B,G,H,I
- 25. Core Exit Temperature - Quadrant 4 2(d) 21d) B,G,H,I (a) Only required for the OPERABLE tank.
(b) Not required for isolation valves whose associated penetration is isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.
Applicable for containment isolation valve position indication designated as post-accident monitoring instrumentation (containment isolation valves which receive containment isolation phase A or containment ventilation isolation signals).
(c) Only one position indication channel is required for penetration flow paths with only one installed control room indication channel. -
(d) A channel consists of two core exit thermocouples (CETs).
Vogtle Units 1 and 2 3.3.3-6 Amendment No. 134 (Unit 1 1))
Amendment No. 113 (Unit 2)
PAM Instrumentation B 3.3.3 B 3.3 INSTRUMENTATION B 3.3.3 Post Accident Monitoring (PAM) Instrumentation BASES BACKGROUND The primary purpose of the PAM instrumentation is to display unit variables that provide information required by the control room operators during accident situations. This information provides the necessary support for the operator to take the manual actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for Design Basis Accidents (DBAs).
The OPERABILITY of the accident monitoring instrumentation ensures that there is sufficient information available on selected unit parameters to monitor and to assess unit status and behavior following an accident.
The availability of accident monitoring instrumentation is important so that responses to corrective actions can be observed and the need for, and magnitude of, further actions can be determined. These essential instruments are identified by unit specific documents (Ref. 1) addressing the recommendations of Regulatory Guide 1.97 (Ref. 2)
( as required by Supplement 1 to NUREG-0737 (Ref. 3).
The instrument channels required to be OPERABLE by this LCO include two classes of parameters identified during unit specific implementation of Regulatory Guide 1.97 as Type A and Category I variables.
Type A variables are included in this LCO because they provide the primary information required for the control room operator to take specific manually controlled actions for which no automatic control is provided, and that are required for safety systems to accomplish their safety functions for DBAs.
Category I variables are the key variables deemed risk significant because they are needed to:
- Determine whether other systems important to safety are performing their intended functions;
((continued) continued)
Vogtle Units 1 and 2 B 3.3.3-1 Revision No. 0
PAM Instrumentation B 3.3.3 BASES BACKGROUND
- Provide information to the operators that will enable them to (continued) determine the likelihood of a gross breach of the barriers to radioactivity release; and
- Provide information regarding the release of radioactive materials to allow for early indication of the need to initiate action necessary to protect the public, and to estimate the magnitude of any impending threat.
These key variables are identified by the unit specific Regulatory Guide 1.97 analyses (Ref. 1). These analyses identify the unit specific Type A and Category I variables and provide justification for deviating from the NRC proposed list of Category I variables.
The specific instrument Functions listed in Table 3.3.3-1 are discussed in the LCO section.
APPLICABLE The PAM instrumentation ensures the operability of Regulatory SAFETY ANALYSES Guide 1.97 Type A and Category I variables so that the control room operating staff can:
(
- Perform the diagnosis specified in the emergency operating procedures (these variables are restricted to preplan ned actions for the primary success path of DBAs) e.g., loss of coolant accident (LOCA);
- Take the specified, preplanned, manually controlled actions, for which no automatic control is provided, and that are required for safety systems to accomplish their safety function;
- Determine whether systems important to safety are performing their intended functions;
- Determine the likelihood of a gross breach of the barriers to radioactivity release;
- Determine if a gross breach of a barrier has occurred; and
((continued) continued)
Vogtle Units 1 and 2 B 3.3.3-2 Revision No. 0
PAM Instrumentation B 3.3.3 BASES APPLICABLE
- Initiate action necessary to protect the public and to estimate the SAFETY ANALYSES magnitude of any impending threat.
(continued)
PAM instrumentation that meets the definition of Type A in Regulatory Guide 1.97 satisfies Criterion 3 of 10 CFR 50.36 (c)(2)(ii). Category I, non-Type A, instrumentation must be retained in TS because it is intended to assist operators in minimizing the consequences of accidents. Therefore, Category I, non-Type A, variables are important for reducing public risk.
LCO The PAM instrumentation LCO provides OPERABILITY requirements for Regulatory Guide 1.97 Type A monitors, which provide information required by the control room operators to perform certain manual actions specified in the unit Emergency Operating Procedures. These manual actions ensure that a system can accomplish its safety function, and are credited in the safety analyses. Additionally, this LCO addresses Regulatory Guide 1.97 instruments that have been designated Category I, non-Type A.
The OPERABILITY of the PAM instrumentation ensures there is sufficient information available on selected unit parameters to monitor
( and assess unit status following an accident. This capability is consistent with the recommendations of Reference 1.
LCO 3.3.3 requires two OPERABLE channels for most Functions.
Two OPERABLE channels ensure no single failure prevents operators from getting the information necessary for them to determine the safety status of the unit, and to bring the unit to and maintain it in a safe condition following an accident.
Furthermore, OPERABILITY of two channels allows a CHANNEL CHECK during the post accident phase to confirm the validity of displayed information. More than two channels may be installed if the unit specific Regulatory Guide 1.97 analyses (Ref. 1) determined that failure of one accident monitoring channel results in information ambiguity (that is, the redundant displays disagree) that could lead operators to defeat or fail to accomplish a required safety function.
(continued)
Vogtle Units 1 and 2 B 3.3.3-3 Rev. 1-10/01
PAM Instrumentation B 3.3.3 B
BASES LCO Table 3.3.3-1 lists all Type A and Category I variables identified by (continued) the unit specific Regulatory Guide 1.97 analyses, as amended by the NRC's SER.
Type A and Category I variables are required to meet Regulatory Guide 1.97 Category I (Ref. 2) design and qualification requirements for seismic and environmental qualification, single failure criterion, utilization of emergency standby power, immediately accessible display, continuous readout, and recording of display.
Listed below are discussions of the specified instrument Functions listed in Table 3.3.3-1.
- 1. Reactor Coolant System Pressure (Wide Range) 408, 418, 428, and 438) is a RCS wide range pressure (LOOP 408,418,428, Category I Type A variable provided for verification of core cooling and RCS integrity long term surveillance.
RCS pressure is used to verify delivery of SI flow to RCS from at least one train when the RCS pressure is below the pump shutoff head. RCS pressure is also used to verify closure of manually closed spray line valves and pressurizer power
( operated relief valves (PORVs).
In addition to these verifications, RCS pressure is used for determining RCS subcooling margin. RCS subcooling margin will allow termination of SI, if still in progress, or reinitiation of SI if it has been stopped. RCS pressure can also be used:
- to manually restart low head SI;
- as reactor coolant pump (RCP) trip criteria; and
((continued) continued)
Vogtle Units 1 and 2 B 3.3.3-4 Revision No. 0
PAM Instrumentation B 3.3.3 BASES LCO 1. Reactor Coolant System Pressure (Wide Range)
((continued) continued)
- to make a determination on the nature of the accident in progress and where to go next in the procedure.
RCS subcooling margin is also used for unit stabilization and cooldown control.
RCS pressure is also related to three decisions about depressurization. They are:
- to determine whether to proceed with primary system depressurization;
- to verify termination of depressurization; and
- to determine whether to close accumulator isolation valves during a controlled cooldown/depressurization.
A final use of RCS pressure is to determine whether to operate the pressurizer heaters.
RCS pressure is also a Type A variable because the operator uses this indication to monitor the cooldown of the RCS following a steam generator tube rupture (SGTR) or small break LOCA. Operator actions to maintain a controlled cooldown, such as adjusting steam generator (SG) pressure or level, would use this indication. Furthermore, RCS pressure is one factor that may be used in decisions to terminate RCP operation.
2,3. Reactor Coolant System (RCS) Hot and Cold Leg Temperatures (Wide Range)
(Hot Leg Loops 413A, 423A, 433A, & 443A)
(Cold Leg Loops 413B, 423B, 433B, & 443B)
RCS Hot and Cold Leg Temperatures are Category I, Type A variables provided for verification of core cooling and long term surveillance.
(continued)
Vogtle Units 1 and 2 B 3.3.3-5 Revision No. 0
PAM Instrumentation B 3.3.3 BASES LCO 2,3. Reactor Coolant System (RCS) Hot and Cold Leg Temperatures (Wide Range) (continued)
RCS hot and cold leg temperatures are used to determine RCS subcooling margin. RCS subcooling margin will allow termination of safety injection (SI), if still in progress, or reinitiation of SI if it has been stopped. RCS subcooling margin is also used for unit stabilization and cooldown control.
In addition, RCS cold leg temperature is used in conjunction with RCS hot leg temperature to verify the unit conditions necessary to establish natural circulation in the RCS.
Reactor outlet temperature inputs to the Reactor Protection System are provided by two fast response resistance elements and associated transmitters in each loop. The channels provide indication over a range of 50°F to 700°F.
The core exit thermocouples provide diverse indication for the RCS hot leg temperature.
Steam line pressure provides diverse indication for the RCS cold leg temperature.
(
- 4. Steam Generator Water Level (Wide Range)
Wide range SG water level (Loops 501,502,503, 501, 502, 503, & 504) is a Type A variable used to determine if an adequate heat sink is being maintained through the SGs for decay heat removal, primarily for the response to a loss of secondary heat sink event when the level is below the narrow range. The wide range SG level indication may also be used in conjunction with auxiliary feedwater flow for SI termination. In addition, the wide range level is cold calibrated and provides a complete range for monitoring SG level during a cooldown. Auxiliary feedwater flow provides the diverse indication for wide range SG water level.
((continued) continued)
Vogtle Units 1 and 2 B 3.3.3-6 Revision No. 0
PAM Instrumentation B 3.3.3 B
BASES LCO 5. Steam Generator Water Level (Narrow Range)
(continued)
Narrow range SG water level (Loops 517-519,527-529,537-517-519, 527-529, 537-539, & 547-549) is a Type A variable used to determine if an adequate heat sink is being maintained through the SGs for decay heat removal and to maintain the SG level and prevent overfill. It is also used to determine whether SI should be terminated and may be used to diagnose an SG tube rupture event.
- 6. Pressurizer Level Pressurizer Level (Loops 459, 460, & 461) is used to determine whether to terminate SI, if still in progress, or to reinitiate SI if it has been stopped. Knowledge of pressurizer water level is also used to verify the unit conditions necessary to establish natural circulation in the RCS and to verify that the unit is maintained in a safe shutdown condition.
7,20. Containment Pressure and Containment Pressure (Extended Range)
(
(Containment Pressure Type A Loops 934, 935, 936, & 937; Containment Pressure extended range loops 10942 & 10943)
Containment Pressure is provided for verification of RCS and containment OPERABILITY.
Containment pressure is also used to verify closure of main steam isolation valves (MSIVs) actuation of containment spray, and for accident diagnosis.
- 8. Steam Line Pressure Steam Line Pressure (Loops 514,515,516,524,525,526,534, 514, 515, 516, 524, 525, 526, 534, 535, 536, 544, 545, & 546) is a Type A variable provided for the following:
- Determining if a high energy secondary line rupture occurred and which steam generator is faulted; (continued)
Vogtle Units 1 and 2 B 3.3.3-7 Revision No. 0
PAM Instrumentation B 3.3.3 BASES LCO 8. Steam Line Pressure (continued)
- Maintaining an adequate reactor heat sink;
- Verifying Auxiliary Feedwater flow to the faulted steam generator is isolated;
- Verifying operation of pressure control steam dump system;
- Maintaining the plant in a cold shutdown condition;
- Monitoring the RCS cooldown rate; and
- Providing diverse indication to Cold Leg temperature for natural circulation determination.
Three channels per steam line are installed with sufficient accuracy to determine the faulted steam generator and to verify Cold Leg temperature for natural circulation.
- 9. Refueling Water Storage Tank (RWST) Level
( The RWST level (Loops 990,991,992, 990, 991, 992, & 993) is a Type A variable provided for verifying a water source to the Emergency Core Cooling Systems (ECCS) and Containment Spray, determining the time for initiation of Cold Leg recirculation following a LOCA and event diagnosis.
The RWST level accuracy is established to allow an adequate supply of water to the safety injection and spray pumps during the switch switchover over to Cold Leg recirculation mode. A high degree of accuracy is required to maximize the time available to the switchover operator to complete the switch over to the sump recirculation phase and ensure sufficient water is available to avoid losing pump suction.
(continued)
(continued)
Vogtle Units 1 and 2 B 3.3.3-8 Revision No. 0°
PAM Instrumentation B 3.3.3 BASES LCO 10,11. Containment Sump Water Level (Narrow and Wide Range)
(continued)
(continued)
Containment Sump Water Level (Narrow range Loops 7777 &
7789; wide range Loops 0764 & 0765) is a Type A variable provided for verification and long term surveillance of RCS integrity.
Containment Sump Water Level is used to determine:
- containment sump level accident diagnosis;
- when to begin the recirculation procedure; and
- whether to terminate SI, if still in progress.
- 12. Condensate Storage Tank (CST) Level CST Level (Loops 5101, 5111, 5104, & 5116) is a Type A variable provided to ensure water supply for auxiliary feedwater (AFW). The CST provides the ensured safety grade water supply for the AFW System. The CST consists of two identical tanks with separate piping to each AFW pump suction.
( Inventory is monitored by two 0-.100%
0-,100% level indication channels for each tank. CST Level is displayed on a control room indicator, strip chart recorder, and unit computer. In addition, a control room annunciator alarms on low level.
The CST Level is considered a Type A variable because the control room meter and annunciator are considered the primary indication used by the operator.
The DBAs that require AFW are the loss of.electric of electric power, steam line break (SLB), and small break LOCA.
The CST is the initial source of water for the AFW System.
However, as the CST is depleted, manual operator action is necessary to replenish the CST.
This function is modified by a Note that clarifies only one of the two CSTs must have two channels of level indication. Since only one CST is required OPERABLE, only one set of level channels is required.
((continued) continued)
Vogtle Units 1 and 2 B 3.3.3-9 Revision No. 0
PAM Instrumentation B 3.3.3 BASES LCO 13. Auxiliary Feedwater Flow (continued)
AFW Flow (Loops 5152,15152,5153,15153,5151,15151, 5150, & 15150) is a Type A variable provided to monitor operation of decay heat removal via the SGs. The AFW Flow to each SG is determined from a differential pressure measurement calibrated for a range of 0 gpm to 1000 gpm.
Redundant monitoring capability is provided by two independent trains of instrumentation for each SG. Each differential pressure transmitter provides an input to a control room indicator and the unit computer. Since the primary indication used by the operator during an accident is the control room indicator, the PAM specification deals specifically with this portion of the instrument channel.
AFW flow is used three ways:
- to determine whether to terminate SI if still in progress, in conjunction with SG water level (narrow range); and
(
AFW flow is a Type A variable because operator action is required to throttle flow during an SLB accident to prevent the AFW pumps from operating in runout conditions. AFW flow is also used by the operator to verify that the AFW System is delivering the correct flow to each SG. However, the primary indication used by the operator to ensure an adequate inventory is SG level.
- 14. Containment Radiation (High Range)
Containment Area Radiation (Loops 0005 & 0006) is a Type A variable provided to monitor for the potential of significant radiation releases and to provide release assessment for use by operators in determining the need to invoke site emergency plans.
((continued) continued)
Vogtle Units 1 and 2 B 3.3.3-10 Revision No. 0
PAM Instrumentation B 3.3.3 BASES LCO 14. Containment Radiation (High Range) (continued)
Containment radiation level is used to determine if a high energy line break (HELB) has occurred, and whether the event is inside or outside of containment.
- 15. Steam Line Radiation Monitors The Steam Line Radiation Monitors (Loops 13119, 13120, 13121 & 13122) are a Type A variable provided to allow detection of a gross secondary side radioactive release and to provide a means to identify the faulted steam generator. Steam generator narrow range level serves as diverse indication for the one monitor per loop provided.
- 16. RCS ReS Subcooling RCS Res Subcooling is a Type A variable provided to determine safety injection termination and re-initiation. The RCS Subcooling variable is determined by calculation in the Plant Safety Monitoring System. The ReS RCS Subcooling
( instrumentation provides the information that allows operators to ensure safety injection is terminated at the optimum time and that sufficient subcooling margin exists upon return to normal plant conditions.
- 17. Neutron Flux (Extended Range)
Neutron Flux (Loops 13135A & 13135B) indication is provided to verify reactor shutdown. The extended range is necessary to cover the full range of flux that may occur post accident.
Neutron flux is used for accident diagnosis, verification of subcriticality, and diagnosis of positive reactivity insertion.
(continued)
Vogtle Units 1 and 2 B 3.3.3-11 Revision No. 0
PAM Instrumentation B 3.3.3 BASES LCO 18. Reactor Vessel Water Level
((continued) continued)
Reactor Vessel Water Level (LT1310, LT1311, LT1312, LT1320, LT1321, & L LT1321, LT1322)
T1322) is provided for verification and long term surveillance of core cooling. It is also used for accident diagnosis and to determine reactor coolant inventory adequacy.
A RVLlS channel consists of Full Range, Upper Range, and Dynamic Range transmitters. LT131 T13100 and LT1320 LT1320 are Upper Range, LT1311 and LT1321 are Full Range, and LT1312 and LT1322 are Dynamic Range.
The Reactor Vessel Water Level Monitoring System provides a direct measurement of the collapsed liquid level above the uppercore plate. The collapsed level represents the amount of liquid mass that is in the reactor vessel above the core.
Measurement of the collapsed water level is selected because it is a direct indication of the water inventory.
- 21. Containment Isolation Valve Position CIV Position is provided for verification of Containment OPERABILITY, and Phase A isolation.
(
When used to verify Phase A isolation, the important information is the isolation status of the containment penetrations. The LCO requires one channel of valve position indication in the control room to be OPERABLE for each active containment isolation valve in a containment penetration flow path, i.e., two total channels of containment isolation valve position indication for a penetration flow path with two active valves. This is sufficient to redundantly verify the isolation status of each isolable penetration either via indicated status of the active valve, as applicable, and prior knowledge of a passive valve, or via system boundary status. If a normally active CIV is known to be closed and deactivated, position (continued)
(continued)
Vogtle Units 1 and 2 B 3.3.3-12 Rev. 2-3/05
PAM Instrumentation B 3.3.3 BASES LCO 21. Containment Isolation Valve Position (continued) indication is not needed to determine status. Therefore, the position indication for valves in this state is not required to be OPERABLE. Note (b) to the Required Channels states that the Function is not required for isolation valves whose associated penetration is isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured. The note also identifies the applicable valves (those valves receiving a Phase A or Containment Ventilation Isolation signal). Note C allows an exception to the LCO requirement (2/penetration) to be made for penetration flow paths with only one installed control room indication channel.
22, 23, 24, 25. Core Exit Temperature Core Exit Temperature is provided for verification and long term surveillance of core cooling.
An evaluation was made of the minimum number of valid core exit thermocouples (CET) necessary for measuring core cooling.
( The evaluation determined the reduced complement of CETs necessary to detect initial core recovery and trend the ensuing core heatup. The evaluations account for core nonuniformities, including incore effects of the radial decay power distribution, excore effects of condensate runback in the hot legs, and nonuniform inlet temperatures. Based on these evaluations, adequate core cooling is ensured with two valid Core Exit Temperature channels per quadrant with two CETs per required channel. The CET pair are oriented radially to permit evaluation of core radial decay power distribution. Core Exit Temperature is used to determine whether to terminate SI, if still in progress, or to reinitiate SI if it has been stopped. Core Exit Temperature is also used for unit stabilization and cooldown control.
Two OPERABLE channels of Core Exit Temperature are required in each quadrant to provide indication of (continued)
Vogtle Units 1 and 2 B 3.3.3-13 Revision No. 0
PAM Instrumentation B 3.3.3 BASES LCO 22, 23, 24, 25.
22,23,24,25. Core Exit Temperature (continued) radial distribution of the coolant temperature rise across representative regions of the core. Power distribution symmetry was considered in determining the specific number and locations provided for diagnosis of local core problems. The two thermocouples in each channel must be located such that the pair of Core Exit Temperatures indicate the radial temperature gradient across their core quadrant. A Note specifies that each channel consists of two CETs. Two sets of two thermocouples ensure a single failure will not disable the ability to determine the radial temperature gradient.
APPLICABILITY APPLICABI LlTY The PAM instrumentation LCO is applicable in MODES 1, 2, and 3.
These variables are related to the diagnosis and pre-planned actions required to mitigate DBAs. The applicable DBAs are assumed to occur in MODES 1, 2, and 3. In MODES 4, 5, and 6, unit conditions are such that the likelihood of an event that would require PAM instrumentation is low; therefore, the PAM instrumentation is not required to be OPERABLE in these MODES.
(
ACTIONS A Note has been added in the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed on Table 3.3.3-1. The Completion Time(s) of the inoperable channel(s) of a Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.
(continued)
Vogtle Units 1 and 2 B 3.3.3-14 Rev. 1 - 6/05
PAM Instrumentation B 3.3.3 BASES ACTIONS A.1 (continued)
Condition A applies to all PAM Instrument Functions. Conditions A addresses the situation where one or more required channels for one or more Functions are inoperable. The Required Action is to refer to Table 3.3.3-1 and take the appropriate Required Actions for the PAM instrumentation affected. The Completion Times are those from the referenced Conditions and Required Actions.
B.1, C.1, 0.1.
B.1. C.1. 0.1, E.1.
E.1, and F.1 These Conditions apply to PAM Instrument Functions when a single required channel is inoperable and a redundant or diverse channel for the affected function remains OPERABLE. Conditions C, 0, E, and F each identify the specific acceptable diverse indication channel that is required OPERABLE for the affected PAM function. Condition B applies to those functions with redundant channels except for Containment Isolation Valves. Condition B is modified by a Note that applies to the containment isolation valve position indication function which allows separate Condition entry for that function based on penetration flow paths. This note clarifies that Condition B may be applied to the containment isolation valve position indication function
( separately for each penetration flow path.
Required Actions B.1, C.1, 0.1, E.1, and F.1 require that the affected channel be restored to OPERABLE status within 30 days. The 30 day Completion Time is based on operating experience and takes into account the remaining OPERABLE redundant or diverse channel, the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval.
Condition G applies when the Required Action and associated Completion Time for Conditions B, C, 0, E, or F are not met.
This Required Action specifies initiation of actions in Specification 5.6.8, "Special Reports," which require a written report, approved by the Plant Review Board (PRB), to (continued)
Vogtle Units 1 and 2 B 3.3.3-15 Revision No. 0
PAM Instrumentation B 3.3.3 BASES ACTIONS G.1 (continued) be submitted to the NRC. This report discusses the results of the root cause evaluation of the inoperability and identifies proposed restorative actions. This action is appropriate in lieu of a shutdown requirement since alternative actions are identified before loss of functional capability, and given the likelihood of unit conditions that would require information provided by this instrumentation.
Condition H applies when one or more Functions have two inoperable required channels (i.e., two channels inoperable in the same Function or one channel inoperable and no diverse channel OPERABLE.).
Required Action H.1 requires restoring one channel in the Function(s) to OPERABLE status within 7 days. The Completion Time of 7 days is based on the relatively low probability of an event requiring PAM instrument operation and the availability of alternate means to obtain the required information. Continuous operation with two required channels inoperable in a Function or no diverse indicating channel OPERABLE is not acceptable because the alternate indications may not fully meet all performance qualification requirements applied to the
( PAM instrumentation. Therefore, requiring restoration of one inoperable channel (diverse or primary) of the Function limits the risk that the PAM Function will be in a degraded condition should an accident occur.
Condition H is modified by two Notes. The first Note applies to the containment isolation valve position indication function and allows separate Condition entry for that function based on penetration flow paths. This note clarifies that Condition H may be applied to the containment isolation valve position indication function separately for each penetration flow path. The second Note is applicable to the second part of the Condition and clarifies that this part of the Condition applies only to those Functions that are not redundant and rely on diverse indication channels. Functions that rely on diverse indication are identified in Conditions C, D, E, and F.
(continued)
Vogtle Units 1 and 2 B 3.3.3-16 Revision No. 0
PAM Instrumentation B 3.3.3 BASES ACTIONS 11 (continued)
If the Required Action and associated Completion Time of Conditions H are not met and Table 3.3.3-1 directs entry into Condition I, the unit must be brought to a MODE where the requirements of this LCO do not apply. To achieve this status, the unit must be brought to at least MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The allowed Completion Time is reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
Condition I is modified by a Note that excludes the Containment Radiation and RVLlS Functions. These Functions are addressed by another Condition.
Alternate means of monitoring Reactor Vessel Water Level (RVLlS) and Containment Area Radiation are available. These alternate means may be temporarily installed if the normal PAM channel cannot be restored to OPERABLE status within the allotted time. If these alternate means are used, the Required Action is not to shut down the unit but rather to
( follow the directions of Specification 5.6.8, in the Administrative Controls section of the TS. The report provided to the NRC should discuss the alternate means used, describe the degree to which the alternate means are equivalent to the installed PAM channels, justify the areas (continued)
Vogtle Units 1 and 2 B 3.3.3-17 Rev. 2-3/05
PAM Instrumentation B 3.3.3
( BASES ACTIONS J.J.
.JJ. (continued) in which they are not equivalent, and provide a schedule for restoring the normal PAM channels.
SURVEILLANCE A Note has been added to the SR Table to clarify that SR 3.3.3.1 REQUIREMENTS and SR 3.3.3.2 apply to each PAM instrumentation Function in Table 3.3.3-1.
SR 3.3.3.1 Performance of the CHANNEL CHECK once every 31 days ensures that a gross instrumentation failure has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate
( properly between each CHANNEL CALIBRATION. The high radiation instrumentation should be'compared becompared to similar unit instruments located throughout the unit.
Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including isolation, indication, and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are OPERABLE.
The Frequency of 31 days is based on operating experience that demonstrates that channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the LCO required channels.
( ((continued) continued)
Vogtle Units 1 and 2 B 3.3.3-18 Rev. 1-3/05
PAM Instrumentation B 3.3.3 BASES SURVEILLANCE SR 3.3.3.2 REQUIREMENTS (continued) A CHANNEL CALIBRATION is performed every 18 months, or approximately at every refueling. CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter with the necessary range and accuracy. This SR is modified by a Note that excludes neutron detectors. The calibration method for neutron detectors is specified in the Bases of LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation." The Frequency is based on operating experience and consistency with the typical industry refueling cycle.
REFERENCES 1. Safety Evaluation Report related to the operation of the Vogtle Electric Generating Plant, Units 1 and 2, NUREG-1137, Supplement No.2, Section 7.5, May 1986.
- 2. Regulatory Guide 1.97, Rev. 2.
- 3. NUREG-0737, Supplement 1, "TMI Action Items."
Vogtle Units 1 and 2 B 3.3.3-19 Revision No. 0
Approved By Procedure Number Rev J.D. Williams 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 1 of 77 PRB REVIEW REQUIRED FEDERAL AND STATE REPORTING REQUIREMENTS PROCEDURE USAGE REQUIREMENTS* SECTIONS Continuous Use: Procedure must be open and readily available at the work location. Follow procedure step by step unless otherwise directed.
Reference Use: Procedure or applicable section(s) available at the work location for ready reference by person performing steps.
Information Use: Available on plant site for reference as needed. ALL Pnnted Printed June 19, 2009 at 11.44 11 :44
t~ctric Gen~rating Plant Approved By Procedure Number Rev J.D. Williams t;<:. ;; . '
..a'". 00152-C 40 Date Approved Page Number 12/22/2008 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 2 of 77 1.0 PURPOSE This procedure provides a compilation of federal and state reporting requirements and establishes responsibilities for report preparation, review, approval and submittal for Vogtle Electric Generating Plant. It also identifies report frequency/timing and the appropriate agency or individual to which reports are submitted.
Excluded from this procedure are reporting requirements associated with labor law, health and safety, finance, the reactor oversight process, and the monthly operating report. The reactor oversight process and the monthly operating report are performed in accordance with procedure 00163-C, "NRC Performance Indicator and Monthly Operating Report Preparation and Submittal."
2.0 DEFINITIONS 2.1 CREDIBLE THREAT A threat is considered credible when (1) physical evidence supporting the threat exists, (2) information independent from the actual threat message exists that support the threat, or (3) a specific group or organization claims responsibility for the threat, (4) a message (written or verbal) is received that contains specific information about plant locations, systems or device description an average person would most likely not know. The Shift Manager should make the determination of credibility with input from the On-Shift Captain or their designated representatives.
2.2 FOLLOW-UP REPORT A written report, submitted in addition to a verbal report, to document a verbal notification, corrective actions taken and actions taken to prevent recurrence, or a verbal notification to update status of conditions as related to previous verbal notification.
2.3 IMMEDIATE NOTIFICATION Verbal communication to the NRC or state within the time period (frequency) specified in Table 2 for the conditions specified in Sections 20.1906, 20.2201, 20.2202, 50.72, 70.52, 73.26, 73.27, 73.71 to 10CFR and 14CFR-FAA, 40CFR110, 33CFR153, 40CFR11 GEPD-SPCC, and GEPD-017-0191-05.
0, GEPO-SPCC, GEPO-017-0191-05.
Printed June 19, 2009 at 11 :44
Approved By ~', ~ . Procedure Number Rev J.D. Williams :Vogtle Electric Gel!
, N,,' 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 30f77 3 of 77 NOTE Immediate" and "Immediately," as used in Table 2, imply as soon as practicable after discovery of the associated item (normally within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />).
2.4 MATERIAL ACCESS AREA Any location which contains special nuclear material, within a vault or a building, the roof, walls, and floor of which constitute a physical barrier.
2.5 NON-ROUTINE REPORTS Unscheduled federal and state reports of events at a licensed facility such as Licensee Event Reports as required by 10CFR50. 73 and hazardous waste spills as required by 40 CFR 260-265.
2.6 PROPERLY COMPENSATED Measures as described in the VEGP Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan including backup equipment, additional
( security personnel and specific procedures, taken to ensure that the safeguards effectiveness of the physical security system is not reduced by failure or other contingencies affecting the operation of security-related equipment or structures.
Such measures must be implemented within ten (10) minutes of an event occurrence/discovery and return the physical security effectiveness to the level equivalent to that which existed before the event.
2.7 REPORTABLE EVENTS Any of those conditions specified as such in 1 10CFR50.73 OCFR50. 73 that are documented on License Event Report Form in accordance with Procedure 81030-C, "Preparation and Processing of Draft Licensee Event Reports and Technical Specification Reports."
Printed June 19, 2009 at 11:44 11 :44
Approved By Procedure Number Rev J.D. Williams ? 00152-C 40 Date Approved Approved Page Number 12/22/2008 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 4 of 77 l.8 2.8 REPORTABLE SECURITY EVENTS Reportable security events are those which threaten nuclear activities or lessen the safeguards effectiveness of the the VEGP Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan. These events are categorized into significant and less significant severity levels. Required reports are listed in Table 2. A general description of security events is provided in the Security Reportability Matrix, Table 3.
2.8.1 Significant security events are those that without being properly compensated warrant immediate involvement by the Nuclear Regulatory Commission (NRC) or other government agencies. These events must be telephonically reported to the NRC within one (1) hour of discovery, and a detailed written report must follow within thirty (30) days.
2.8.2 Less significant security events are those which are properly compensated and are required to be logged by Nuclear Security within twenty-four (24) hours and copies of the record maintained by the Nuclear Security Department.
2.9 REPORTING SUSPICIOUS ACTIVITIES 2.9.1 Licensee is requested to notify the following agencies if suspicious activities are
( identified by the on shift Nuclear Security Captain, Shift Manager or other Site Management: (Reference security procedure 90321-C for additional information.)
- Local FAA office for suspicious aircraft activities, @ (904-549-1537/1538)
- Local law enforcement agency, @ (706-554-2133/2136)
- Local FBI field office, @ (706-722-3702)
- The NRC headquarters Operations Center @ (301)816-5100 (301 )816-51 00 2.10 ROUTINE REPORTS Federal and state reports submitted on a frequency pre-established by the applicable regulations, such as Report Of "Changes to Procedures or Facility as described in the FSAR" as required by 10CFR50.59 and "Quarterly* "Quarterly NPDES Report" as required by the State NPDES Permit.
2.11 SECURITY EVENT OCCURRENCE/DISCOVERY The exact date and time an event occurs which marks the start of time constraints for properly compensating, logging, and/or reporting Security Reportable Event/Occurrences. If the exact date and time can not be determined, the date and time when on-site security management or an equivalent level of plant management make the determination that a significant or less significant security event has occurred and is reportable in accordance with 10CFR73.71 (a), (b), or (c).
Printed June 19, 2009 at 11 :44
Approved By Procedure Number Rev J.D. Williams Vogtle Electric Gen'erating Plant 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 5 of 77 l.O 3.0 RESPONSIBILITIES 3.1 VICE-PRESIDENT (Vogtle)
The Vice President (Vogtle) has ultimate responsibility to ensure items identified in this procedure are performed.
3.1.1 The Vice President (Vogtle) may designate qualified site person(s)person( s) to perform functions described in this procedure on his/her behalf with exception to the following:
- a. Approval of the following documents has been designated as requiring a Vice President approval. This list is not conclusive and there may be other documents identified in other plant documents that will require Vice president (Officer of the Company) approval.
(1) Licensing Event Reports (50.73)
(2) Security Event Reports (73.71 (73.71))
(3) Response to "Choice Letters", Regulatory Allegation, Violations or Appeals.
(
(4) ROP Performance Indicator Data (5) Training and Licensed Operator Correspondence.
3.1.2 State and federal reporting requirements for Plant Vogtle are in compliance with the requirements as noted in Table 2.
3.1.3 Performance of special reviews, investigations or analyses and reports, by the PRB, as necessary, which involve state and federal reportability requirements.
3.1.4 Preparation, review, approval and submittal of any reports to the federal or state authorities as determined necessary by the requirements of this procedure.
Printed June 19, 2009 at 11 :44
Approved By Procedure Number Rev J.D. Williams 00152-C 40 Date Approved Page Number 12/22/2008 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 6 of 77 3.2 DEPARTMENT HEADS Department heads will ensure the following:
3.2.1 Appropriate notification is made upon determination that events, conditions, or circumstances are reportable in accordance with this procedure.
3.2.2 Routine reports are prepared or data submitted for follow-up reports to meet the requirements of this procedure.
3.2.3 Follow-up reports are prepared or data submitted for follow-up reports to meet the requirements of this procedure.
3.2.4 Reports are properly reviewed, approved, and submitted in accordance with Table 2 of this procedure.
3.2.5 Federal and state agency reports are prepared in sufficient time for review by the PRB, if required, and be submitted to Corporate Licensing or Corporate Environmental Affairs for final processing.
3.2.6 Submit to the Performance Improvement Commitment Coordinator any Updates to commitments if required.
3.3 PERFORMANCE IMPROVEMENT SUPERVISOR The Performance Improvement Supervisor (PIS) will ensure the following:
3.3.1 Federal and state reports are prepared in sufficient time for review by the PRB and plant management to meet required reporting dates.
3.3.2 Reports are prepared in the proper format for approval and transmittal to the appropriate agency as indicated in Table 2.
3.3.3 Subsequent follow-up reports are prepared in sufficient time to meet required reporting dates.
3.3.4 Ensure adequate and timely periodic reporting to all appropriate agencies (including NRC, federal and state regulatory agencies).
3.3.5 Ensure plant commitment updates provided to the Performance Improvement Commitment Coordinator (PICC) are processed in accordance with appropriate plant procedures.
3.3.6 If requested, will assist other departments in the preparation and approval process of an item.
- 44 Printed June 19, 2009 at 11:44
Approved By Procedure Number Rev J.D. Williams 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 7 of 77 3.4 PLANT REVIEW BOARD (PRB)
The PRB will review all reportable events and other reports as directed by plant management in accordance with Procedure NMP-GM-009, "Plant Review Board."
3.5 PERSONNEL Plant personnel are responsible for immediately reporting any actual or suspected occurrence of events or conditions described in this procedure.
Reports will be made to the Shift Manager or immediate supervisor in accordance with NMP-GM-002, "Corrective Action Program," as appropriate.
3.6 CORPORATE ENVIRONMENTAL AFFAIRS Corporate Environmental Affairs is responsible for tracking, renewing, revIsing revising and the filing of all Non-Radiological Environmental Permits/Reports. If reports will be prepared at the site, EA will notify site personnel of the required reports.
The site will then create and submit the reports to Corporate Environmental Affairs unless otherwise specified in Table 2 of this procedure. (1985307475, 1985305892) 3.7 CORPORATE NUCLEAR LICENSING Corporate Nuclear Licensing is responsible for tracking, renewing, revising and the filing of all licensing related documents unless otherwise stated in this procedure. If reports will be prepared at the site, NL will notify site personnel of the required documentation. The site will then create and submit the reports to Corporate Nuclear Licensing unless otherwise specified in Table 2 of this procedure. (1985307475) 3.8 PERMITS AND LICENSE Upon receipt of a permit/license on site, the responsible manager will review for reporting requirements, expiration date, and any special requirements. The responsible manager will ensure that all conditions of the permit/license are met.
Additionally, the responsible manager should forward the original permit/license to the SNC Corporate responsible department and place a copy into Documentum.
Printed June 19, 2009 at 11 :44
Approved By ,l:li Procedure Number Rev J.D. Williams . 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 8 of 77 4.0 REPORT IDENTIFICATION AND PREPARATION 4.1 REPORT IDENTIFICATION Federal and state regulatory required reports originating at Plant Vogtle are identified in the Federal and State Reporting Matrix, Table 2. The matrix identifies the following:
4.1.1 Source requirement for the reports.
4.1.2 The report title as identified or implied by the source requirement.
4.1.3 Frequency of reporting which is synonymous with timing as used in Regulatory Guide 10.1 Rev 4, Oct., 1981.
4.1.4 Method of submittal of reports, i.e., written, telegram, telephone or other means.
4.1.5 Organizations and/or individuals responsible for the preparation, review, approval, and submittal of reports or conditions.
4.1.6 The federal or state agency or person to whom the report is submitted.
{,
NOTE A listing of the abbreviations used in the matrix is contained in Table 1.
4.2 REPORT PREPARATION Responsible department heads will ensure collection and interpretation of data and preparation of reports identified in the matrix (Table 2) meet the time limitations of the pertinent requirement 4.2.1 Reports not addressed within this section but which are identified in the matrix will be prepared in accordance with the applicable departmental procedures or instructions.
4.2.2 Reports not covered by departmental procedures or instructions will be submitted in accordance with the information supplied in the reporting matrix.
Printed June 19, 2009 at 11 :44 Pnnted
1-~_:D_p~_°':tv_e1_1I~_;m_S J.D. Williams_ _-f-_".;....I~*~~.;....~.;...;g. . .~_le...;.._E_I--"ejbF ~~~er~iirig Plant Approved By Procedure Number Rev
. 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 9 of 77 4.2.3 Licensee Event Reports (LER)
Licensee Event Reports for reportable events, described in 10CFR 50.73 will be prepared, approved by the Vice President-Vogtle, and transmitted in accordance with Procedure 81030-C, "Preparation and Processing of Draft Licensee Event Reports and Technical Specification Reports."
- a. A Condition Report, prepared in accordance with NMP-GM-002 is generally the mechanism for documenting the occurrence of events described in 10CFR 10CFR 50.73.
- b. An individual within the PI Group will be designated as a Licensee Event Report Coordinator (LERC). The LERC is the point of contact for preparation and tracking of LERs.
- c. Vice-President (Vogtle), SRB, and PRB must be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the LER condition, in the event of a safety limit violation 4.2.4 10CFR 50.59 Reports and FSAR Updates These routine reports are described in 10CFR 50.59(d) (2) and 10CFR 50.71 (e) 50.71(e)
(, (4) and will be prepared and submitted to the NRC by Corporate Nuclear Licensing within 6 months after each Unit 2 refueling outage not to exceed 24 months. (1985307344) 4.2.5 Personnel Radiation Exposure and Planned Special Exposure Reports.
Personnel radiation exposure reports and reports of planned special exposures as described by 10CFR parts 20.2105, 20.2203, 20.2206, and 19.13 will be prepared and submitted in accordance with Procedure 45012-C, "Individual Radiation Exposure Records and Reports."
4.2.6 Annual Radioactive Effluent Release Report The Annual Radioactive Effluent Release Report as described in 10CFR 50.36a(a)(2) and Technical Specification 5.6.3 will be prepared and submitted by May 15th of each year.
4.2.7 Asbestos Removal, Use, And Disposal Report Reports of the use, removal and disposal of asbestos as described by 29CFR1926.1101 and 40CFR61 will be prepared and submitted in accordance with Procedure 00265-C, "Asbestos Handling, Removal, And Disposal" and NMP-EN-801, Asbestos Abatement and Demolition Program.
Pnnted June 19, 2009 at 11 :44 Printed
Approved By Procedure Number Rev J.D. Williams 00152-C 40 Date Approved Page Number 12/22/2008 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 10 of 77 4.3 IMMEDIATE NOTIFICATION REPORTING 4.3.1 Reporting of events which require immediate notification in accordance with 10CFR 50.72, as the results of the declaration of the Emergency Classes, will be accomplished in accordance with Procedure 91002-C, "Emergency Notifications.
Notifications.""
4.3.2 Those events which require immediate notification and are not the result of the declaration of one of the emergency classes will be accomplished in accordance with Procedure 10000-C, "Conduct Of Operations." The Event Notification form from Procedure 91002":C 91002-C will be used to record the event information provided in the immediate notification. Acopy A copy of this form is to be forwarded to Performance Improvement for receipt by the next working day. If the notification involves safeguards information a copy of the form (without the safeguards information) will be sent to Performance Improvement with instructions on who to contact to determine the content of the message.
NOTE NRC Form 361 may be used in lieu of the form (Checklist 4) in Procedure 91 002-C for Event Notification.
91002-C 4.3.3 In addition to the conditions as stated in 4.3.1 and 4.3.2, the Vice-President (Vogtle), SRB, and PRB will also be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the LER condition, in the event of a safety limit violation.
Printed June 19, 2009 at 11 :44
Approved By Procedure Number Rev J.D. Williams Vogtle Electric Generating Plant,',' . \ 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 11 of 77 5.0 REVIEW AND APPROVAL NOTE The review and approvals identified in Table 2 are guidelines and allow the review and approval process to deviate to acquire the appropriate reviews, when necessary. However, the responsible department must ensure the required approvals are captured.
5.1 The responsible department head, or designee, will create, review and approve reports as required by the matrix in Table 2.
5.2 Completed reports will be forwarded to the appropriate department for review and further processing as indicated in the matrix. The reports should be processed in a timely manner (50-65% of the allowable reporting time may be preparation)~
used for report preparation).
NOTE Some reports, as noted on the matrix, are submitted directly to the individual or
( agency by the responsible department and are not submitted to Performance Improvement.
5.3 The responsible department head, or designee, review will also include a review of commitment impacts contained in the report thereof. The commitments will be identified and revised as necessary in accordance with Procedure 00409-C, "Commitment And Action Item Tracking." Contact PICC for assistance if required.
5.4 21.21 (c) will be approved and submitted according Reports required by 10CFR 21.21(c) to Table 2.
6.0 RECORDS All reports will be processed as a Quality Assurance record in accordance with Procedure 00100-C, 001 OO-C, "Quality Assurance Records Administration."
Pnnted June 19, 2009 at 11 :44
Approved By Procedure Number Rev J.D. Williams "n9 Plant 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 12 of 77 r.o
7.0 REFERENCES
7.1 Regulatory Guide 10.1 Rev. 4 Oct 1981, "Compilation Of Reporting Requirements For Persons Subject To NRC Regulations."
7.2 10CFR - Energy 7.3 33CFR - Navigation and Navigable Waters (USCG/Corps of Engineers) 7.4 40CFR - Protection of Environment (EPA) 7.5 49CFR - Transportation 7.6 14CFR - Aeronautics and Space (FAA), FAA advisory, AC 70/7460 - IF, Obstruction Marking and Lighting 7.7 29CFR1910 - Safety and Health Standards - General Industry Standards 7.8 Technical Specification, Sections 2,3,4, 2, 3,4, and 5 7.9 State of Georgia Environmental and Water Quality Permits, Georgia Department of Natural Resources i
(
7.10 FSAR Section 13.1 7.11 NUREG-1022, Revision 2 7.12 Management Procedures Manual (GPC), 128-005, Environmental Protection Company Communications With Agencies 7.13 Management Procedures Manual (GPC), 400-004 Power Generation Department General Office Notification of Company Officers per 10CFR 21 7.14 ANSI N18.7 - 1976, Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants 7.15 NPDES Permit 7.16 Georgia Department of Public Health Requirements 7.17 Regulatory Guide 1.21 (June, 1974). "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents From Light-Water-Cooled Nuclear Power Plants" Printed June 1 19, g, 2009 at 11:44
yogtle ElectricG~nerating Pla~r';:~:.\;:~:
Approved By Procedure Number Rev J.D. Williams 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 13 of 77 7.18 r.18 Regulatory Guide 1.33, Revision 2, Feb. 1978, "Quality Assurance Program Requirements (Operation)"
7.19 Regulatory Guide 1.16 (Revision 4, 1975), "Reporting of Operating Information" 7.20 Regulatory Guide 1.35, Revision 2, January 1976, "Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containment Structures" 7.21 Quality Assurance Department Procedure Manual, Procedure No. QA-04-02, 10CFR50.55(e)/1 OCFR21 Significant Deficiency/Defect Reporting, 10CFR50.55(e)/10CFR21 7.22 Code of Georgia Chapter 9, Title 34, State Board of Workers Compensation 7.23 Public Law 91-596 7.24 Georgia Rules For Safe Drinking Water, 391-3-5 7.25 PROCEDURES 7.25.1 00100-C, "Quality Assurance Records Administration" 7.25.2 00163-C, NRC Performance Indicator And Monthly Operating Report Preparation and Submittal 7.25.3 00265-C, "Asbestos Handling, Removal, And Disposal" 7.25.4 00409-C, "Commitment And Action Item Tracking" 7.25.5 10000-C, "Conduct Of Operations" 7.25.6 36025-C "Radioactive Effluent Release reporting" 7.25.7 45012-C, "Individual Radiation Exposure Records and Reports" 7.25.8 81030-C, "Preparation And Processing Of Draft Licensee Event Reports And Technical Specification Reports" 7.25.9 91002-C, "Emergency Notification" 7.25.10 94001-C, "Spill Prevention, Control, Countermeasures (SPCC) And Reportability" 7.25.11 NMP-GM-002, "Corrective Action Program" 7.25.12 NMP-GM-009, "Plant Review Board" Printed June 19,
- 19. 2009 at 11 :44
Vogtl~~:Ele¢tric Generating Pl~~t .. * \
Approved By Procedure Number Rev J.D. Williams ',. . .... .~ v 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 14 of 77 7.25.13 NMP-EN-OO 1, NMP-EN-001, "Management of the Radioactive Effluent Release Reports and the Offsite Dose Calculation Manuals" 7.25.14 NMP-EN-801, "Asbestos Abatement and Demolition Program" 7.25.15 NMP-EN-002, "Actions for Potential Groundwater Contamination Events" 7.26 OTHER DOCUMENTS 7.26.1 Southern Nuclear Operating Company (SNC) Quality Assurance Topical Report (QATR) 7.26.2 ASME NQA-1-1994 Edition, Quality Assurance Requirement for Nuclear Facility Applications 7.27 COMMITMENTS 1984301895 1984301899 1984301952 1984301972 1985304488 1985304490 1985305596 1985305970 1985306628 1985306639 1985307529 1986307668 1986307670 1986310237 1987310348 1987310357 1987310361 1987310368 1987313140 1991321556 1991321693 1992325492 1995329935 1995329943 1995329951 1995330840 1995331607 1995331608 1996332122
( 1996332124 1996334445 1996334446 1997334707 1999339798 1999339799 1999339800 2001342315 2005300009 2005300012 2005300013 2005300015 2005300016 1985307475 1985305982 END OF PROCEDURE TEXT Printed June 19, 2009 at 11 :44
Approved By Procedure Number Rev J.D. Williams 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 15 of 77 ABBREVIATIONS LIST Table 1 CC Corporate Chemistry CEA Corporate Environmental Affairs CFR Code of Federal Regulations CHEM Chemistry Department COE Army Corp of Engineers - Savannah District, P. O. Box 899, Savannah, GA 31402-0889 CM - Chemistry Manager DOE - Department of Energy - Oak Ridge, TN., P.O. Box E, 37830 DOT - Department of Transportation, Washington D.C., -Information Systems Manager, Materials Transportation Bureau
=:0
/
( - Emergency Director ENN - Emergency Notification Network ENS - Emergency Notification System ESC - Emergency Spill Coordinator ESM - Engineering Support Manager EPA - Environmental Protection Agency EPG - Emergency Planning Group ES - Engineering Supervisor FAA - Federal Aviation Administration FEMA - Federal Emergency Management Agency FOS-VP - Fleet Operation Support-Vice President FSS - Financial Services Supervisor Printed June 19, 2009 at 11 :44
Approved By Procedure Number Rev J.D. Williams 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 16 of 77
\ Table 1 (Cont'd)
GDNR - Georgia Department of Natural Resources - Water Supply and Laboratory 47 Trinity Avenue, SW Atlanta, Ga. 30334 GDOPE Ken M. Copeland, Director of the Office of Permits and Enforcement, GA.
DOT, 940 Virginia Ave., Hapeville, GA 30354 GEPD Georgia Environmental Protection Division, Watershed Protection Branch, Water Withdrawal Permitting Program, 4220 International Parkway, Suite 101, Atlanta, Ga. 30354-3902 GEPDA Georgia Environmental Protection Division, Air Protection Branch, 4244 International Parkway, Atlanta, Ga. 30354 HP Health Physics HPM Health Physics Manager HRS Human Resources Supervisor INDIV Individual(s) Involved in Exposure
(
LO Licensed Operator MAINT Maintenance MEA Manager of Environmental Affairs MIS Materials and Inspection Services MM - Maintenance Manager MS Maintenance Support N/A - Not applicable NFD Nuclear Fuel Department NL Nuclear Licensing NLM Nuclear Licensing Manager Printed June 19, 2009 at 11 :44
Approved By Procedure Number Rev J.D. Williams 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 17 of 77 Table 1 (Cont'd)
NOTS Nuclear Operations Training Supervisor NPDES National Pollutant Discharge Elimination System NRC Nuclear Regulatory Commission NRC-DCD - Nuclear Regulatory Commission-Document Control Desk, Washington, D.C.
NRC-DNS Nuclear Regulatory Commission - Director, Division of Nuclear Security, Office of Nuclear Security and Incident Response NRC-EDO - Nuclear Regulatory Commission-Executive Director for Operations NRC-IE Nuclear Regulatory Commission-Office of Inspection and Enforcement (Washington, D.C.), 20333 -Commercial Telephone No. (301) 492-7000 (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)
NRC-IMNS - Nuclear Regulatory Commission-Division of Industrial and Medical Nuclear Safety
(
NRC-MPA - Nuclear Regulatory Commission-Director of Management and Program Analysis, Washington, D.C. 20555.
NRC-NMSS - Nuclear Regulatory Commission-Nuclear Material Safety and Safeguards NRC-NRR - Nuclear Regulatory Commission-Nuclear Reactor Regulation (Washington, D.C.)
NRC-OC Nuclear Regulatory Commission-Operations Center (Bethesda, MD),
Commercial Telephone No.-(301) 816-5100 NRC-ORM - Nuclear Regulatory Commission-Office of Resource Management, Washington, D.C.
NRC-RI Nuclear Regulatory Commission-Resident Inspector NRC-RO Nuclear Regulatory Commission-Regional Office, 61 Forsyth Street, Suite 23T85, Atlanta, Ga. 30323, (Region II, Atlanta, GA), 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Commercial Telephone No. - (404) 562-4400 NRC-RRES - Nuclear Regulatory Commission-Office of Nuclear Regulatory Research Pnnted June 19, 2009 at 11 :44 Printed
Gen~:~ati ng ~lan\~)t . \ :ri~
Approved By ;""" Procedure Number Rev J.D. Williams ". Vogtle Electric ,,"j : :
00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE STATE REPORTING REQUIREMENTS 18 of 77 Table 1 (Cont'd)
ODCM - Offsite Dose Calculation Manual OM - Operations Manager OPS - Operations OSHA - Occupational Safety and Health Administration PI Performance Improvement Group PIS Performance Improvement Supervisor PCP - Process Control Plan PE - Programs Engineering PM - Plant Manager PRB - Plant Review Board
- >TEPM - Plant Training and Emergency Preparedness Manager RE - Reactor Engineer RM - Recovery Manager RWE - Operations Radwaste Engineer SC Supply Chain SCS Supply Chain Superintendent SimC Simulator Coordinator SCBRH - Heyward G. Shealy, Chief, Bureau of Radiological Health & & Environmental Control, 2600 Bull St., Columbia, SC 29201 SE - System Engineer SEC - Security SHIH - Safety and Health Industrial Hygienist Printed June 19, 2009 at 11:44 Pnnted 11 :44
V~;btle Ele?t~ic ~e~~.&~t\ing PI~~~
Approved By ' </
0,' ,
Procedure Number Rev J.D. Williams 1
, 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 19 of 77 Table 1 (Cont'd)
(Cont' d)
SECM - Security Manager Secm - Security Management (see Footnote at end of Table 2)
SSM - Site Support Manager SM Operations Shift Manager SMSO - Senior Member of the Security Organization on-duty/on-call SNSH - Southern Nuclear Safety and Health SNMC - Special Nuclear Material Custodian SNM - Special Nuclear Material SNSE - Southern Nuclear Safety Engineer SRB - Safety Review Board
( rs Training Supervisor VP - Vice President -Vogtle WOG - Westinghouse Owners Group Pnnted June 19, 2009 at 11:44
Approved L, J.D. Williams Vogtle Electric bc{1erating Plant
';;;'\,:i'::::.\~
.\ Pre 00152-C
,Number Rev 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 20 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No, No. Reviewed By By By To 1 10CFR SO Emergency Notification Within 1S minutes of declaring an ENN -ED ED ED State & & local Appendix E Sect emergency 4Commercial N/A gov!.
govt.
IV.D.3 Telephone 2 10CFRSO.72 Immediate Notification 2.1 10CFRSO.72 The declaration of any of the emergency classes specified in the Immediately after notification of state ENS ED ED ED NRC-OC (a)(1 )(i)
(a)(1)(i) licensees approved Emergency Plan or local agencies and not later than 2Commercial
'Commercial N/A NRC-RI one hour after the declaration of the Telephone required NOTE: See 2.3 for follow-up notification reguired Emergency Classes Emergenc},
2.2 10CFRSO.72 For Non-Emergency events that occurred within 3 years of discoverY. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from occurrence of ENS SM SM SM NRC-OC (a)(1)(ii) &
& (b)(1) If not reported as a declaration of an Emergency Class under 2.1 deviation from Technical 2Commercial
'Commercial N/A NRC-RI above, report any deviation from the plant's Technical Specifications Specifications. Telephone 1OC FR SO.S4(x) authorized pursuant to 10CFR follow-up notification required NOTE: See 2.3 for foliow-uE 2.3 10CFRSO.72 For items 2.1, 2.2, 33.1 thru 33.4, 37.1 thru 37.S immediately report Immediately and maintain an open, ENS or SM ED orSM ED or ED or NRC-OC (c) the following: continuous communication channel 2
- 1. Any further degradation in the level of safety of the with NRC Operations Center upon Telephone plant or other worsening plant conditions, including request of the NRC.
those that require the declaration of any of the emergency classes, if such a declaration has not been previously made.
- 2. Any change from one emergency class to another, or a termination of the emergency class.
- 3. The results of ensuing evaluations or assessments of plant conditions.
- 4. The effectiveness of response or protective measures taken.
S. Information related to plant behavior that is not understood.
3 10CFR SO.91 Notice to state of request amendment of operating license At the time of filing amendment Written NLM VPI VP NRCto NRC to (b) request with NRC PRB NLM provide name of State official Footnotes at end of Table 2 Pnnted June 19, 2009 at 11:44 11 :44
Approved L J.D. Williams Date Approved
- . VogtleElectric bc{)erating,~J~ilf'
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. ,'k'*:'!"*** Pre 1:.:':> , .. ';" 00152-C
- Number Rev j
Page Number 40 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 21 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED
. Item Requirement Report Title Or Condition Frequency Method Prepared By Approved By Submitted Submitted No. By Reviewed B;z: By B;z: To 4 Tech Specs 2.2.3 Safety limit violation (NRC approval of restart required). As soon as practical & in all cases ENS SM SM SM (See items 33.1, 40 & 132) within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after determination. ' Commercial 2 N/A (1996334445) Te[ephone Telephone 5 10CFR50.36 Exceeding limiting conditions for operation and associated actions Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ENS SM SM SM NRC-OC (c)(2)&(c)(6) statements as required by the Technical Specifications. (See items 2
' Commercial N/A NRC-R[
NRC-RI 10CFR50.72 2.2,33.1, & 132) Te[ephone Telephone (b) 6 10CFR73.71 Discovery of the loss of any shipment of SNM or spent fuel or Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ENS SM SM SM NRC-OC (a)(1) availab[e after the significant supplemental information that becomes available 2Commercial
'Commercial N/A NRC-RI initial report Te[ephone Telephone (See item 79 for written report)
See footnote #14.
7 10CFR73.71 Recovery of or accounting for lost shipment of SNM or spent fuel or Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ENS SM SM SM NRC-OC (a)(1) significant supplemental information that becomes available after the 2Commercial
'Commercial N/A initial report Te[ephone Telephone (See item 80 for written report)
See footnote #14.
8 10CFR Report of theft or loss of licensed material Immediately Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, ENS HPM VP SM NRC-OC 20.2201 (a)(1)(i)
Reference:
RIS 2005-21 if the material in question is equal to or 2
' Commercial ESM NRC-RO (See Items 102 and 103 greater that 1000 times the quantity Te[ephone, Telephone, NRC-RI specified in Appendix C to 10CFR Part Telegram Or 20, under such circumstances that an Equiv.
exposure could result to person in unrestricted areas.
Pnnted June 19, 2009 at 11:44
\<<<\iogH~~'Electric U,,(lerating Plaril~':~l:<'<<\. < <:}te<"y~
Approved k Pre.. ; Number Rev J.D. Williams ,o<,<j<"<',
00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 22 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved By Submitted Submitted No. Reviewed By By To 8.1 10CFR loss of licensed material Report of theft or 1055 Licensee shall report by telephone, ENS HPM VP SM NRC-OC 20.2201 (a)(1)(ii)
(a)(1 )(ii) Reference; RIS 2005-21 within 30 days after the occurrence of ' Commercial 2 ESM NRC-RO (See Items 102 and 103) any lost, stolen or missing licensed Telephone, NRC-RI material becomes known to the Or Equiv.
licensee. Licensed material that is greater than 10 times the quantity specified in Appendix C to Part 20 that missing at the time.
is still missina 9g 10CFR 1 ng exposure to or release of by-product Notification of incident involving Immediately ENS HP VP SM NRC-OC 20.2202(a) source or special nuclear material that may have caused or threatens ' Commercial 2 HPM to cause an individual present for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to receive 5 times the Telephone annual limit. Or Equivalent 10 10CFR70.52 Accidental criticality Immediately within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after ENS SM SM SM NRC-RO (a) & (b) discovery. ' Commercial 2 N/A NRC-RI Telephone,Or Equivalent Footnotes at end of Table 2.
Printed June 19, 2009 at 11 :44
.\
Approved ..c "" Pre ; Number Rev J.D. Williams ",<<\ Vogtle Electric U~()era!ing;Plant 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 23 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To I 11 10CFR73.67(g) 10CFR73,67(g) Each licensee who receives quantities and types of special nuclear When received Telephone SNMC NA NA Shipper (2)(ii) material of low strategic significance shall notify the shipper of receipt SNMC of the material.
material, 12 Not Used 13 Not Used 14 Not Used 15 Not Used 16 Not Used 17 Not Used ----- ---
Footnotes at end of Table 2.
Pnnted June 19, 2009 at 1111:44
- 44
/ ...
Approved Lk J.D. Williams Vogtle Electric b~(lerafi'n'gPlant .\ Prl ; Number Rev 001::>2-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 24 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To 18 10CFR73.71 Threatened, attempted or actual: "Individual Items Listed Below" (b)(1) 18.1 10CFR73.71 Theft or unlawful diversion of SNM. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of discovery ENS Secm SMSO SM NRC-OC (b)(1) (see item 81.1) 2, Telephone, N/A
&AppendixG
& Appendix G Telegram or equivalent 18.2 10CFR73.71 Significant physical damage to a power reactor or its equipment or Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of discovery ENS Secm SMSO SM NRC-OC (b)(1) carrier equipment transporting nuclear fuel or spent nuclear fuel, or to 2 Telephone N/A
&AppendixG
& Appendix G the nuclear fuel or spent fuel a facility or carrier possesses. Telegram or (see item 81.2) equivalent 18.3 10CFR73.71 Interruption of normal operations of a power reactor through the Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of discovery ENS Secm SMSO SM NRC-OC (b)(1) unauthorized use of or tampering with its machinery, components, 2 Telephone, N/A
& Appendix G controls including the security system. Telegram or (see item 81.3) equivalent 18.4 10CFR73.71 Actual entry of unauthorized person into a protected area (PA), Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of discovery ENS Secm SMSO SM NRC-OC (b)(1) Material Access Area (MAA) (MAA),, Controlled Access Area (CAA) Vital 2 Telephone, N/A
& Appendix G Area(VA), or transport. (see item 81.4) Telegram or equivalent 18.5 10CFR73.71 Uncompensated failure, degradation, or discovered vulnerability in a Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ENS Secm SMSO SM NRC-OC (b)(1) system that could allow unauthorized 2' Telephone, N/A or undetected access to a PA, MAA, VA or transport Telegram or (see item 81.5) (Also Safeguards Events) equivalent 18.6 10CFR73.71 Actual or attempted introduction of contraband into a PA, MAA, VA, Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ENS Secm SMSO SM NRC-OC (b)(1) & transport 2 Telephone, N/A AppendixG Appendix G (see item 81.6) Telegram or equivalent Discovery of significant supplemental information after initial Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ENS Secm SMSO SM NRC-OC 19 10CFR73.71 (a)(5) response 2 Teleohone Telephone N/A I Footnotes at end of Table 2.
Printed Pnnted June 19, 2009 at 11 :44 11:44
Approved lL J.D. Williams l~i?::;i
- .i1~K*
Vogtle Electric b~.).~ ...ating
. 'I".' . \' ..
Plant .\ Prl 00102-C
,Number Rev 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 25 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. By Reviewed B:£ By By To 20 10CFR26.27 Suspected noncompliance with Fitness-For-Duty requirements by Immediately Telephone SM SM SM NRC-RO (d) NRC employees N/A (NRC-OC if after COB) 21 10CFR19.13 Radiation exposure data to terminating employees Immediately upon termination Written HP N/A HP Terminating (e) e) N/A Employee Emplolee 22 10CFR Report of excessive radioactive contamination on package of Immediately 1 Telephone HP SM SM NRC-RO 20.1906(d)(1 20.1906(d)(1)) radioactive material (Removable surface contamination exceeds the and HPM VP NLM NRC-RI & &
limits of 10CFR71.87.) Telegram Or Final Facsimile Delivering Carrier 23 10CFR Report of excessive radiation levels external to the package of Immediately 1 Telephone HP SM SM NRC-RO 20.1906(d)(2) radioactive material (External radiation level exceeds the limits of and HPM VP NLM NRC-RI & &
10CFR71 .47.)
10CFR71.47.) Telegram Or Final Facsimile Delivering Carrier 24 14CFR-FAA Cooling Tower lights: Any failure or malfunction that lasts for greater Immediately ";; Telephone SM SM SM FAA Flight advisory, AC than 30 minutes and affects as top or flashing light, regardless of N/A Service 70/7460-IK position, should be reported immediately to the nearest flight service Station,
- 1-79-ASO- center.(when problem occurs and when corrected). Also see Item 69 Macon, GA.
GA 1456-0C
- 2-79-ASO- See Footnote #5 for Longitudes and Latitudes and height of cooling 1457-0E towers and phone number. ----------
25 33CFR 153 Discharge of oil and other pollutant materials into navigable Immediately °Telepil()-rie-,
°Telephone, -CHEMCHEM ESC MEA 'GEPD GEPD waterways Telegram or ESC 40CFR110 (see Item 60.1) eguivalent equivalent CM 26 GEPD Hazardous waste spills Immediately 6
° Telephone SM ESC MEA GEPD SPCC (see Item 60 also) ESC CM 27 10CFR Bodily injury or property damage from possession or use of prom ptly as practicable As promptly Written HP VP NLM NRC-NRR 140.6(a) radioactive material resulting in indemnity claim HPM PRB Footnotes at end of Table 2.
Printed June 19, 2009 at 11:44
,~<
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Elect[lc.~.t:(leratlng
,\', . ~
" . ,> , Pro Number Rev J.D. Williams
'1:"';;/ . Vogtle
. ' " " , i "":,,
Plant ..4IIE'a. 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 26 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To 28 10CFR Any material change in proof of financial protection or in any other Promptly Written SCS NLM NLM NRC-NRR 140,15(e) 140.15(e) financial information filed with the NRC under Part 140 Risk Mgt.
Mgt or NRC-NMSS 29 10CFR Activation of the Emergency Response Data System. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after declaring Activation of N/A N/A N/A N/A 50.72(a)(4) emergency class (alert, site area ERDS emergency or general emergency) satisfies this requirement 30 40CFR No Reporting Required N/A N/A N/A N/A 260-265 31 40CFR When releases of radioactivity exceed Tech Spec limits (item 92) and Each release Telephone CHEM VP MEA EPA 302 when subsequent releases, within a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, equal a CM reportable quantity the EPA National Response Center must be notified 32 49CFR Incidents involving hazardous materials Earliest practicable moment Telephone CHEM CM SM DOT 171.15 Note: Responsibility of carrier CM 33 10CFR Events requiring 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> notification 50.72(b)(2) 33.1 10CFR50.72 10CFR50,72 Initiation of any nuclear plant shutdown required by Technical As soon as practical and in all cases ENS SM SM SM NRC-OC (b)(2)(i) Specifications within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 2 Commercial N/A NRC-RI Telephone NOTE: See 2.3 for follow-up notification required 33.2 10CFR50.72 Any event that results or should have resulted in Emergency Core As soon as practical and in all cases ENS SM SM SM NRC-OC (b)(2)(iv)(A) Cooling System (ECCS) discharge into the reactor coolant system as within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 2 Commercial N/A NRC-RI a result of a valid signal except when the actuation results from and Telephone is part of a pre-planned sequence during testing or reactor operation.
NOTE: See 2.3 for follow-up notification required Footnotes at end of Table 2.
Pnnted June 19, 2009 at 11 11:44
- 44
1:._
Approved L.
J.D. Williams '/ .;{ "
Vogtle Electric tn:oerating!:"Plant '
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-:'./,.,
.\ Pre 00152-C Number Rev 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 27 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To 33.3 10CFR50.72(b) Any event or condition that results in actuation of the Reactor As soon as practical and in all cases ENS SM SM SM NRC-OC (2)(iv)(B) Protection System (RPS) when the reactor is critical except when the within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 2
' Commercial N/A NRC-RI actuation results from and is part of a preplanned preplan ned sequence during Telephone, testing or reactor operation.; Other System NOTE; NOTE: See Item 2.3 for follow-up required.
33.4 10CFR Any event or situation, related to the health and safety of the public or As soon as practical and in all cases ENS SM SM SM NRC-OC 50.72 (b)(2)(xi) onsite personnel, or protection of the environment, for which a news within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> '2 Commercial N/A NRC-RI release is planned or notification to other government agencies has Telephone been or will be made. Such an event may include an onsite fatality or inadvertent release of radioactively contaminated materials.
NOTE: See Item 2.3 for follow-up required.
34 Not Used 35 29CFR1904.39 A fatality or hospitalization of 3 or more individuals resulting from a Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of notification of a '"
' Telephone SNSH VP SM OSHA work related incident death or hospitalization.
Death/hospitalization occurring 30 days after an incident is not reportable.
36 NUREG Alert, Site Area Emergency or General Emergency Summary Report Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of close out of Written "Written ED ED ED NRC-RO 0654 Event N/A NRC-RI Appendix 1 State &
& Local Gov!.
37 10CFR Events requiring 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> notification 50.72(b)(3) 50 72(b)(3) 37.1 10CFR50.72(b) 1 OCFR50. 72(b) Any event or condition during operation that results in: Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> ENS SM SM SM NRC-OC (3)(ii) (a) The condition of the nuclear power plant, including 2Commercial
'Commercial N/A NRC-RI its principal safety barriers, being seriously Telephone degraded; or (b) The nuclear power plant being: in an unanalyzed condition that significantly degrades plant safety.
NOTE: See 2.3 for follow-up notification required Footnotes at end of Table 2.
Printed June 19, 2009 at 11:44 Pnnted 11 :44
~"'",;,,:'::
Approved, ',""',;",v ,,;:::' ' , ' , Pre ; Number Rev J.D. Williams .':;": '
VogtleE leotricb"llerati ng Plant .act:,,, :::' '>(::'~;s1flf;~l:r.:~',;'\)\;ii;':);: . 001b2-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 28 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By ~
By ~.
By To 37.2 10CFR50.72(b) Any event or condition that results in valid actuation of any or the Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> ENS SM SM SM NRC-OC (3)(iv) systems listed below except when the actuation results from and is 2 Commercial N/A NRC-RI part of a preplanned sequence during testing or reactor operation.
operation, Telephone, (1) Reactor Protection System (RPS) including: reactor scram and Other-System Other System reactor trip (when reactor not critical).
critical),
(2) General containment isolation signal affecting containment isolation valves in more that one system or multiple main steam isolation valves.
(3) Emergency core cooling system for PWRs including: high-head, intermediate-head and low-head injection systems and the low pressure injection function of residual heat removal systems.
(6) PWR auxiliary or emergency feedwater system. system, (7) Containment heat removal and depressurization systems, including containment spray and fan cooler systems.
systems, (8) Emergency AC electrical power systems, including emergency diesel generators NOTE: See 2.3 2,3 for follow-up notification required Footnotes at end of Table 2.
Printed Pnnted June 19, 2009 at 11 :44 11:44
.\
Approved, Prl ) Number Rev Vogtle Electric l..,,(terating Plant J.D. Williams Date Approved I't .,>>' ",i," 001!:>2-C Page Number 40 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 29 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To 37.3 10CFR Any event or condition that at the time of discovery could have Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> ENS SM SM SM NRC-OC 50.72 prevented the fulfillment fu[fil[ment of the safety function of structures or system 2 CommerCial Commercial N/A NRC-R[
NRC-RI (b)(3) that are needed to: Te[ephone, Telephone, (v)&(vi) Other System (A) Shut down the reactor and maintain it in a safe shutdown condition, (B) Remove residual heat, (C) Control the release of radioactive material, or (0)
(D) Mitigate the consequences of an accident Events covered above may include one or more procedural errors, equipment failures. and/or discovery of design, analysis, fabrication, construction, and/or procedural inadequacies. However, individual component failures need not be reported if redundant equipment in the same system was operable and availab[e available to perform the required safety function NOTE: See Item 2.3 for follow-up required.
37.4 10CFR Any event requiring the transport of a radioactively contaminated Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> ENS SM SM SM NRC-OC 50.72(b)(3) person to an offsite medical facility for treatment 2 Commercial HPM NRC-RI NRC-R[
(xii) Telephone Te[ephone NOTE: See item 2.3 for follow-up required.
37.5 10CFR Any event that results in a major loss of emergency assessment Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> ENS SM SM SM NRC-OC 50.72(b)(3) capability, offsite response capability or communications 2 Commercial HPM NRC-R[
NRC-RI (xiii) capability (e.g., significant portion of control room indication, Te[ephone Telephone Emergency Notification System or offsite notification system as delineated in EPIP EP[P 91204-C and 91706-C) (1991321693)
NOTE: See item 2.3 for follow-up required. _. _ _ .. _ - - - - - - - - - - -
37.6 FSAR Relief and safety valve failures to close wil[
Re[ief will promptly be reported to Promptly LER OM VPP NLM NRC-OC Section 5.4.13 the nuclear regulatory commission. (1989316194) Process PIS NRC-R NUREG-0737 PRB Footnotes at end of Table 2.
Printed June 19, 2009 at 11:44 11 :44
Vogtle Electric lJ~()eratinggJCirii:':!':'\
Approved. \ ': ...'..**.*.Yii'ii1't(?***;: Pre J Number Rev J.D. Williams .... :>;,. ., ,....:',:.<i:',. ,." ,< : .;~;g~ti~l!) 00152-C 40 Date Approved umber Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS of 77 30 .of TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To 38 NUREG Unusual Event Summary Report Within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of close out of Unusual 4 Written "Written ED ED ED NRC-RO 0654 Event N/A NRC-RI Appendix 1 State &
Local Govt.
Gov\.
39 Reg Guide The presence of a loose part is confirmed. Report condition in Promptly ENS SM SM SM NRC-OC 1.133 accordance with prompt notification requirements Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 2 Commercial N/A NRC-RI Section C.6 (1985304490)(1985307529) Telephone report NOTE: See item 66 for written rel20rt 40 Tech Specs Safety limit violation Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Commercial SM SM SM VP 2.2.4 Notify the PM and VP (1995330840) Telephone N/A SRB (See Item 4) PRB PM 41 Operating Violations of the requirements of NPF-68 or NPF-81 Section 2.c, See Items N/A N/A N/A N/A License shall be reported (1987310357) 41.1,41.2, NPF-68 or and,41.3 NPF-81 Section 2.h 41.1 NPF-68 or Exceeding 100 percent power (3625.6 megawatts thermal) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of discovery of ENS SM SM SM NRC-OC NPF-81 (See Item 97.1) violation 2 Commercial N/A NRC-RI Section Telephone, 2.c.(1) Telegram or equivalent 41.2 Not Used.
41.3 NPF-68 or Violation of license conditions as listed in 10CFR 10CFR 50.54 Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Telephone SM SM SM NRC-OC NPF-81 "Conditions of Licenses" N/A NRC-RI Section 2.c Footnotes at end of Table 2.
Printed Pnnted June 19, 2009 at 11 :44
~
Approved ':'
L Pre. ~ Number Rev J.D. Williams )? ' .. ., .: VogtleEle.~tr~p(.,cf1erating Plant , .. ~(( ".:" ..",,:; t;,;ii;;:~ "':';::( . 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 31 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To 42 OL NPF-68 Any Occurrence of an unusual or important event that indicates or Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ENS SM SM SM NRC-OC or NPF-81 could result in significant environmental impact casually related to 2 Commercial CHEM NRC-RI Appendix B plant operations shall be recorded and reported to NRC within 24 Telephone, MEA Section 4.1 hours followed by a written report per item 99. Telegram or LSM (1987310361))
(1987310361 equivalent NOTE: This reporting Requirement is for non-radiological events only. For examples of the types of occurrences to be reported see section 4.1, Appendix B of the operating license. If clarification on reportability is required contact corporate Environmental Affairs (See footnote 8).
43 10CFR Report of incident involving licensed material that may have caused Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ENS HP VP NLM NRC-OC 20.2202(b) or threatens to cause an individual to exceed annual limits in a 24 2 Commercial HPM NRC-RI hour period. Telephone ESM and Telegram PRB NOTE: Report must be submitted in accordance with 50.73 or equivalent 44 10CFR Notify the commission of significant fitness for duty events. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Telephone SM SM SM NRC-OC 26.73(a)(1)& (A) The use, sale, or possession of illegal drugs within the PA N/A (2) (B) Acts by licensed operators or supervisor personnel involving:
(1) The use, sale or possession of controlled substances within the protected area.
(2) Confirmed positive tests.
(3) Use of alcohol in the PA.
PAc (4) Being unfit for work due to alcohol consumption.
45 NPDES Non-Compliance with any daily maximum effluent limitation specified Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> o Telephone U CHEM VP MEA GEPD Permit No. in the permit. CM GA0026786 46 10CFR Notification (by a director or responsible officer) of information Within 2 days of receipt of information Telephone .E.1
.E! VP NLM NRC-OC 21.21 (c)(3)(i) reasonably indicating a failure to comply with the AEA or a defect identifying defect or failure to comply. or PIS affecting construction or operation of a facility or a basic component Facsimile PRB for the facility 47 10CFR Notify the NRC of information identified as having a significant Within 2 working days Telephone, .E.1
.E! VP NLM NRC-RO 50.9(b) implication for public health and safety or common defense and Telegraph or PIS NRC-RI security Written PRB Footnotes at end of Table 2 Printed Pnnted June 19, 2009 at 11:44 11 :44
".' . ,...,""""'::/\': .,.... .F:.' .. ' **.:".'*,::d. '.':"" .,., >:.{k,.
Approved L,
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L . Number Rev J.D. Williams Vogtle Electric b"f1~ra~ing Plant' .-=.. 00102-C 40 Date Approved Page Number umber 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 32 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To 48 Georgia Surface Water Withdrawal Report (Permit violations) Within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> o Telephone 8 SM SM MEA GEPD Surface Water (See Item 59) CM Withdrawal Permit 017-0191-05 49 Georgia Failure to comply with any safe drinking water rule Within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> o Telephone CHEM CM MEA GEPD Rules for safe (See Item 61) (1984301895) N/A drinking water 391-3-5 50 40CFR No Reporting Requirements 260-265 51 NPDES Non-compliance with any daily maximum effluent limitation specified Within 5 days of occurrence Written CHEM VP MEA GEPD No.
Permit No, in the permit. (1985305970) (Waivable by GEPD on a CM GA 0026786 case-by-case basis) 52 29CFR1904.29 29CFR 1904.29 Log of occupational injury and illness Recorded in log within Written SHIH SHIH SHIH SNSE (See Item 157) days of event 7 da)!s N/A 53 10CFR Advance notification of shipment of irradiated reactor fuel and A notification delivered by mail must Written HP VP NLM NRC-RO 71.97 nuclear waste be postmarked at least 7 days before HPM GDOPE the beginning of the 77-day
-day period PRB SCBRH during which departure of the shipment is estimated to occur.
A notification delivered by any other means than mail must reach the office of the governor or of the governor's designee at least 4 days before the beginning of the 7-day period during which departure of the shipment is estimated to occur.
54 Air Quality Permit Deviation from requirements associated with malfunctions that result within 7 days Written OPS VP MEA GEPDA No. 4911-033- in excessive emissions.
0030 -V-02-2 CM 55 10CFR Advance notification of shipment of irradiated reactor fuel 10 days prior to shipment Written SNMC VP NLM NRC-DNS NRC - DNS 73.72(1), (2) and ESM (3) PRB Printed June 19, 2009 at 11 11:44
- 44
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Approved c L
J.D. Williams 'ii, ... ..,\ Vogtle Electric bCilerating'Plant ,',., \
.\ Pre 00152-C J Number Rev 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 33 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To 55.1 10CFR Advance notification of shipment of irradiated reactor fuel 2 days prior to shipment Telephone SNMC N/A N/A NRC-OC NRC - OC 73.72(4))
55.2 10CFR Schedule changes greater than + or - 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for shipment of When discovered Telephone SNMC N/A N/A NRC - OC 73.72(5)) irradiated reactor fuel 56 TR 13.3.2. Seismic instrumentation - With one or more required seismic Within 10 days Written SE VP NLM NRC-DCD Condition B monitoring instruments inoperable more than 30 days ESM NRC-RO (1996332122)(1999339798)(1999339800)
(1996332122) (1999339798) (1999339800) PRB NRC-RI TR 13.3.3 With one of more required meteorological monitoring channels Within 10 days Written SE VP NLM NRC-RO 57 Condition A inoperable for more than 7 days (1999339799) ESM NRC-RI PRB Footnotes at end of Table 2 11:44 Printed June 19, 2009 at 11 :44
Vogtle Electric l>~llerc:tting~PI~u~t A.
Approved ,.*. Pre, .;; Number Rev J.D. Williams ** >>'. . '),'.<.; . 001S2-C 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REPORTiNG REQUIREMENTS 34 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To 58 FSAR 16.3 With all channels of one or more collection regions of the loose parts Within 10 days Written SE VP NLM NRC-RO Requirements 3, detection system inoperable for more than 30 days ESM NRC-RI Reg Guide (1985304488)(1986310237) PRB 1.133 section C.S.b, CSb.
59 Georgia Surface Water Withdrawal Report (Permit violations) Within 10 days Written CEA VP MEA GEPD Surface (See Item 48) (1984301895) CM Water OM Withdrawals Permit 017-0191-05 60 GEPD Hazardous waste spills (See Item 26 also) Within 10 days of occurrence Written CEA VP MEA GEPD SPCC CM PRB 60.1 33CFR 153 Discharge of oil and other pollutant materials into navigable Within 10 days of occurrence Written CEA VP MEA GEPD waterways CM 40CFR 110 (see Item 25) PRB 61 Georgia Failure to comply with any safe drinking water rule Within 10 days Written CHEM VP MEA GEPD Rules for (See Item 49) CM safe drinking water 391-3-5 62 Tech Specs PAM Report Re[1ort Within 14 days Written OPS VP NLM NRC-RO 5.6,8 5.6.8 As required by Condition G or J of LCO 3,3.3 3.3.3 OM NRC-RI (1995331607) (1995331608)
(1995331607)(1995331608) PRB 63 Not Used 64 TR 13,3,2, 13.3.2. One or more required seismic monitoring instruments actuated Within 14 days Written SE VP NLM NRC-DCD Condition C during a seismic event greater than or equal to .01,01g.
- g. ESM NRC-RO Per Condition c,S c.5 The report shall describe the magnitude, PRB NRC-RI frequency spectrum, the results of instrument evaluations, and resultant effect upon facility features important to safety, safety. (Also see Item 56) (1996332124)
Footnotes at end of Table 2 Pnnted June 19, 2009 at 11.44
Approved, *'W*r : ....:., ........ ~ .. Pre* .,~ Number Rev J.D. Williams >:... .:Vogtle.Electric l~~llerating Plant. ~
- .!*.::,.:. .i'.* ..... }); ..**:.:.; .... 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 35 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To 65 Not Used.
66 Reg. Guide Loose Parts Confirmation Follow-up Report (See Item 39) Within 2 weeks Written SE VP NLM NRC-OC 1.133 (1985307529) ESM NRC-RI Section C.6 PRB 67 10CFR 20 App Trace investigation report of Radioactive waste shipment Within 2 weeks of completion of Written HP VP NLM NRC-RO G.III E.2 investigation HPM NRC-RI 68 Not Used 69 14CFR-FAA If Cooling Tower lights are out more than 15 days the flight service 15 days " Telephone SM SM SM FAA Flight advisory, AC center will be called to extend the outage. Extensions will be N/A Service 70/7460-IK requested every 15 days until restored See Item 24 Station, Macon, GA See Footnote #5 for Longitudes and Latitudes and height of cooling towers and phone number.
70 Army Corp of River intake structure maintenance dredging 15 days prior to start of work Written CEA VP MEA COE Engineers Permit (1986307668)(1986307670) MM 200500606 71 49CFR Incidents involving hazardous materials Within 30 days of discovery Written (DOT CHEM VP NLM DOT 171.16 Note: Responsibility of carrier form F5800) CM NRC-NMSS 10CFR71.5(a) PRB 72 Not Used 73 10CFR Notify NRC of permanent reassignment of a licensed operator or Within 30 days Written TS VP VP NRC-RO 50.74a senior operator from a position that requires a license PTEPM 74 10CFR Notify NRC of the termination of any licensed operator or senior Within 30 days Written TS VP NLM NRC-RO 50.74b operator PTEPMS 75 10CFR Notify NRC if a licensed operator or senior operator develops a Within 30 days of diagnosis Written TS VP NLM NRC-DCD 50.74c, permanent physical or mental conditions that causes the operator to PTEPMS NRC-RO 10CFR fail to meet the requirements of the medical examination (temporary 'NRC-RI 55.25 incapacitation is not reportable) (2001200949)
Footnotes at end of Table 2.
Printed Pnnted June 19, 2009 at 11 :44 11:44
~ ";~;;,~;:,t:fP:':::(::" , 00152-C Approved, "
Prl .: Number Rev J.D. Williams ,
"Vogtle,Eleritric ~ ... (Jerating PI~,~t ';,i.:a "" " 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 36 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To 76 10CFR A license operator must notify the NRC of a conviction of a felony Within 30 days Written LO/HRS N/A LO NRC-NRR 55.53g 55,53g NRC-RO 77 10CFR20.2201 10CFR20,2201 license Material> 10 Times 10CFR20 App.
Loss of License App, C Within 30 days Written HP VP NLM NRC-DCD HPM NRC-RI PRB NRC-RO 78 Not Used 79 10CFR73,71 10CFR73.71 Discovery of the loss of any shipment of SNM SN M or spent Within 60 days Written SECM VP NLM NRC-DCD (a)(4)&(5) fuel or significant supplemental information discovered PIS NRC-DNS after initial written report PRB NRC-RO (See Item 6)
See footnote #14.
80 10CFR73.71 Recovery of or accounting for lost shipment of SNM or spent fuel or Within 60 days Written SECM VP NLM NRC-DCD (a)(4)&(5) significant supplemental information discovered after initial written ESM NRC-DNS report. PRB NRC-RO (See Item 7)
See footnote #14.
81 10CFR73.71 Threatened, attempted or actual "Individual Items Listed listed Below" Below" (b)(1) &
Appendix G AppendixG . ._ ---
81,1 81.1 10CFR73.71 Theft or unlawful diversion of SNM SN M Within 60 days Written SECM VP NLM NRC-DCD (b)(1) & (See Item 18.1) PIS NRC-RO AppendixG Appendix G PRB 81.2 10CFR73.71 Significant physical damage to a power reactor or its equipment or Within 60 days Written SECM VP NLM NRC-DCD (b)(1) & carrier equipment transporting nuclear fuel or spent nuclear fuel, or to PIS NRC-RO AppendixG Appendix G the nuclear fuel or spent fuel a facility or carrier possesses (See Item PRB 18.2) 81.3 10CFR73.71 Interruption of normal operations of a power reactor through the Within 60 days Written SECM VP NLM NRC-DCD (b)(1) & unauthorized use of or tampering with its machinery, components, PIS NRC-RO Appendix G controls including the security system. PRB See item18.3)
(See item18,3) 81.4 10CFR73.71 Actual entry of unauthorized person into a protected area Within 60 days Written SECM VP NLM NRC-DCD (b)(1) & (PA), material access area, (MAA) controlled access area (CAA)(CAA),, PIS NRC-RO AppendixG Appendix G vital area, (VA), or transport PRB (See Item 18.4)
Footnotes at end of Table 2 Printed June 19, 2009 at 11.44
Approved L J.D. Williams Vogtle Electric ue()erating"Plarit .\ '1<<<
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00152-C
.; Number Rev 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 37 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To 81.5 10CFR73.71 Uncompensated failure, degradation, or discovered vulnerability in a Within 60 days Written SECM VP NLM NRC-DCD (b)(1) &
& safeguards system that could allow unauthorized or undetected PIS NRC-RO Appendix G access to a PA, MAA, VA or transport (See Item 18.5) PRB 81.6 10CFR73.71 Actual or attempted introduction of contraband into PA, MAA, VA, or Within 60 days Written SECM VP NLM NRC-DCD (b)(1) &
& transport PIS NRC-RO Appendix G (See Item 18.6) PRB 82 10CFR73.71 Follow-up report concerning discovery of safeguards event. Within 60 days Written SECM VP NLM NRC-DCD (b)(2) PIS NRC-RO PRB 83 Not Used 84 Not Used 85 Not Used 86 10CFR71.95(b) Instances in which the conditions of approval in the Certificate of Within 60 days Written HP VP NLM NRC-NMSS Compliance were not observed in making a shipment HPM PRB 87 ODCM 2.1.3.2 Calculated dose release of radioactive materials in liquid effluents Within 30 days Written CHEM VP NLM NRC-RO exceed limits CM CEA PRB 88 ODCM 2.1.4.2 Radioactive liquid waste discharged without treatment and in excess Within 30 days Written CHEM VP NLM NRC-RO of limits and portion of the Liquid Radwaste Treatment Systems not CM in operation. CEA PRB 89 ODCM 3.1.3.2 Calculated air dose from radioactive noble gases in gaseous Within 30 days Written CHEM VP NLM NRC-RO effluents exceed limits CM CEA PRB 90 ODCM 3.1.4.2 Calculated dose from the Release of lodine-131, lodine-133, tritium, Within 30 days Written CHEM VP NLM NRC-RO and all radionuclides in particulate form with half-life of greater than 8 CM days in gaseous effluents above limits CEA PRB Footnotes at end of Table 2 Printed June 19, 2009 at 11:44 11 :44
/'~-~'e Approved l L
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-'.: Number Rev 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 38 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To 91 ODCM 3.1.5.2 Radioactive gaseous waste discharge without treatment & in excess Within 30 days Written CHEM VP NLM NRC-RO of limits CM CEA PRB 92 ODCM 5.1.2 When the calculation of annual dose or dose commitment to any Within 30 days Written (NRC CHEM VP NLM NRC-DCD 10CFR member of the public due to releases of radioactivity and to radiation form 366) CM NRC-RO 20.2203(a)(4) has exceeded limits of specifications CEA 13[ND[V 13INDIV (See Item 31)(1984301952) PRB 93 ODCM 4.1.1.2.2 Confirmed level of radioactivity from plant effluents exceed specified Within 30 days Written ES VP NLM NRC-RO leve[s levels when averaged over any calendar quarter. CM VEGP CEA PRB 94 Not Used 95 Tech Specs 5.6.7 EDG Failure Report Within 30 days Written SE VP NLM NRC-RO
[f an individual emergency diesel generator experiences four or more If ESM NRC-[E NRC-IE valid failures in the last 25 demands, these failures and any non-valid PRB failures experienced by the EDG in that time period. Report per 50.73. (1985305596) (2001342315)(2005300015) 96 Tech Specs 2.2.5 Safety limits violation. 60 day reporting requirement per 10CFR50.73 Within 60 days after discovery Written El E! VP NLM NRC-RO
([tem (Item 132). (A[so (Also see Items 33.1 and 40) PIS NRC-DCD
((1996334446) 1996334446) PRB FOS-VP SRB 97 Operating Violation of the requirements of NPF-68 or NPF-81, See Items 97.1, 97.2, and 97.3 License section 2 c, shall be reported (1987310357)
NPF-68 or NPF-81 Section NOTE:
NOTE; See Item 132 2.h 97.1 NPF-68 Exceeding 100 percent power (3625.6 megawatts thermal) See Item Within 30 days Written El E! VP NLM NRC-DCD or NPF-81 41.1 PIS NRC-RO Section PRB NRC-R[
NRC-RI 2.c.(1) 97.2 Not Used.
Footnotes at end of Table 2 Printed Pnnted June 19, 2009 at 11:44 11.44
Approved l Approved.
J.D. Williams >> "
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- 00152-C
.; Number Rev 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 39 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To 97.3 NPF-68 or Violation of license conditions as listed in 10CFR50.54 "Conditions of Within 30 days Written TS PM NLM NRC-DCD NPF-81 Licenses" PTEPM NRC-RO Section 2c PRB NRC-RI 98 OL NPF-68 Changes to, or renewal of, the NPDES or State Certification is Within 30 days of approval of change Written MEA NLM NLM NRC-DCD or NPF-81 appealed and a stay is granted or renewal or within 30 days of date NLM NRC-RO Appendix B stay is granted NRC-RI Section 3.2 99 OL NPF-68 Non routine reports for occurrence of a non routine event Within 30 days Written MEA NLM NLM NRC-DCD or NPF-81 (1987310368) NLM NRC-RO Appendix B NRC-RI Section 5.4.2 100 Not Used 101 10CFR Changes in Emergency Plan made without prior NRC approval. Within 30 days after change is made Written TS FOS-VP NLM NRC-DCD 50.54(q) (See item 145) PTEPM NRC-RO NRC-RI PRB 102 10CFR Theft or loss of licensed material report Within 30 days of initial report Written HP VP NLM NRC-DCD 20.2201 (b) (See items 8, 8.1 & 103) HPM NRC-RI NRC-RO PRB 103 10CFR Substantive additional information Within 30 days of initial report Written HP VP NLM NRC-OCD NRC-DCD 20.2201 (d) 20.2201(d) (See items 8, 8.1 & 102) HPM NRC-RI PRB NRC-RO 104 Not Used Footnotes at end of Table 2.
Printed June 19, 2009 at 11:44 11 :44
- Approved, Approved",
J.D. Williams
'.':) , ........* ... ':)i:,C:i:;2.' . . . Vogtle Ele~~ri;~ u~l1erating Plant ",\'; <
Pn.
00152-C
~ Number Rev 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 40 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By Bv By To 105 Item Deleted 106 Permit No. Well or spring water plant report (make-up wells 2 ea) Monthly (By 10th day of the following Written CHEM CM MEA GEPD PG0330017 month) CM to operate potable water system 107 Permit No. Well or spring water plant report (simulator building) Monthly (By 10th day of the following Written CHEM CM MEA GEPD PG0330035 month) CM to operate potable water system s~stem 108 Permit No. Well or spring water plant report (employee recreation area well) Monthly (By the 10th day of following Written CHEM CM MEA GEPD NG0330036 month) CM to operate potable water system 109 10CFR Radiation exposure data to employee (See items 110 & 111) Concurrent with report to NRC Written HP N/A HP Employee 19.13(d) N/A 110 10CFR Overexposure of individuals or excessive levels of radiation or Within 30 days of learning of Written HP VP NLM NRC-DCD 20.2203(a) concentration of radioactive material and/or any incident which occurrence (NRC form HPM NRC-RO 10CFR requires notification under 10CFR20.2202. 366) PRB 131NDIV 13INDIV 19.13(d) 19.13(d (See (1984301952)
See item 109) (1984301952 111 10CFR Radiation exposure data to former employees Within 30 days of request or within 30 Written HP N/A Former 19.13(c) days of determination of exposure, N/A Employee which ever is later.
112 Not Used Footnotes at end of Table 2 11:44 Printed June 19, 2009 at 11 :44
Approved k L
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.;. ,;.j Vogtle Electric ("c:neratlng Plant .&a 001S2-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 41 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Item* Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To 113 10CFR Reduction in effectiveness of authorized package during use Within 30 days Written HP VP NLM NRC-NMSS 71.95(a)(1) (transportation of radioactive material) HPM PRB 114 10CFR Defects in nuclear materials packaging Within 30 days Written HP VP NLM NRC-NMSS 71.95(a)(2) HPM PRB 115 40CFR61 Asbestos removal, use, and disposal notification 10 days prior to event Written CHEM/CEA MEA MEA GEPD 29FCR N/A 1926.1101 116 Not Used 117 10CFR Notification (by a director or responsible officer) of information Within 30 days following receipt of Written El
.El VP NLM NRC-DCD 21.21 (d)(3)(ii) reasonably indication a failure to comply with the AEA or a defect information. PIS NRC-RO affecting construction or operation of a facility or a basic component PRB for the facility 118 10CFR Changes to or errors in an acceptable emergency core cooling Within 30 days if significant Written NL VP FOS-VP NRC-DCD 50.46 system (ECCS) evaluation model, or in the application of such NLM NRC-RO (a)(3)(ii) model, that affect the temperature calculation. PRB NRC-RI (also (See Item 166)
Appendix K)
Printed June 19, 2009 at 11:44 11 :44
Approved, Prl .:j Number Rev J.D. Williams 001 b2-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 42 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To 119 10CFR The licensee shall inform the Director of the Office of Nuclear As stated under "Report Title or Written NL VP FOS-VP NRC-NRR 50.54(w)(4) 50.54{w){4) Reactor Regulation in writing when the reactor is and can be Condition" NLM (ii) maintained in a safe and stable condition so as to prevent any PRB significant risk to the public health and safety. Within 30 days after the licensee informs the Director that the reactor is in this condition, or at such earlier time as the licensee may elect or the Director may for good cause direct, the licensee shall prepare and submit a cleanup plan for the Director's approval. The cleanup plan must identify and contain an estimate of the cost of each cleanup operation that will be required to decontaminate the reactor sUfficiently to permit the licensee either to resume operation of the sufficiently reactor or to apply to the Commission under § 50.82 for authority to decommission the reactor and to surrender the license voluntarily.
Cleanup operations may include one or more of the following, as appropriate: (A) Processing any contaminated water generated by the accident and by decontamination operations to remove radioactive materials: (B) Decontamination of surfaces inside the auxiliary and fuel-handling buildings and the reactor building to levels consistent with the Commission's occupational exposure limits in 10 CFR Part 20, and decontamination or disposal of equipment; (C)
Decontamination or removal and disposal of internal parts. And damaged fuel from the reactor vessel; and (D) Cleanup of the reactor coolant system.
120 10CFR50 Changes to Emergency Plan or implementing procedures. Within 30 days of the change. Written PMTEP VP NLM NRC-DCD Appendix PRB NRC-RO E.V.
EV NRC-RI 121 Emergency Plan Changes to the computer configuration or data protocols (contained At least 30 days prior to installing the Written SE VP NLM NRC-DCD Section H, in the PAL) .(1992325492) change ESM 4.3.H, also PRB 10CFR50 Appendix E.VI.3.b.
EVI.3.b.
Footnotes at end of Table 2 Printed June 19,2009 19, 2009 at 11:44 11 :44
Approved'L Approved'l Pre Pre. J~ Number Rev J,D. Williams J.D. 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 43 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To 122 Emergency Plan Changes to information describing the specific computer data points Within 30 days following the change Written SE VP NLM NRC-DCD Section H, transmitted,(1992325492) ESM 4,3,H, also PRB 10CFR50 Appendix E,V1.3.a.
E.VI.3.a.
123 10CFR Report of Planned Special Exposure Within 30 days following the exposure Written HP VP NLM NRC-RO 20,2204 20.2204 (1984301952) HPM NRC-RI 13 INDIV 13INDIV 124 10CFR Fabrication of package with excess decay heat load or pressure 45 days prior to fabrication Written HP VP NLM NRC-RO 71.93(c) HPM PRB 125 10CFR Changes in Security Plan, that do not reduce safeguards Within 2 months of Written SEC VP NLM NRC-DCD 50 54(p)(2)
SOS4(p)(2) effectiveness, made without prior NRC approval. the change SECM NRC-RO 10CFR (See item 128) 70.32(e) PRB 126 Not Used 127 10CFR Changes made in safeguards contingency plan without prior NRC Within 2 months of the change Written SEC VP NLM NRC-NMSS 70,32(g) approval SECM NRC-RO PRB 128 10CFR Changes made in security plans without prior NRC approval (See Within 2 months of the change Written SEC VP NLM NRC-NMSS 70,32(e) 125) item 12S) SECM NRC-RO PRB 129 10CFR Interim report to notify that evaluation of an identified deviation or Within 60 days from discovery of the Written E! VP NLM NRC-DCD 21 ,21 (a)(2) failure to comply with AEA potentially associated with a substantial deviation or failure to comply, PIS NRC-RO safety hazard cannot be completed within 60 days from date of PRB discovery, Footnotes at end of Table 2 Printed June 19, 2009 at 11:44 11 :44
Approved L L.
J.D. Williams ""',,, ' ;'''
Vogtle Electric u~neratingPlant
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00152-C
- Number Rev 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 44 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To 130 Reg Guide Changes to DMIMS alert level and alert level logic to provide for Within 60 days Written SE VP NLM RC-OC 1.133 background noise as determined from comparison to the two ESM NRC-RI Section previous quarterly measurements. PRB C.3.a(2)(e) If not temporary and the new levels exceed the ability of the DMIMS system to detect a loose part as defined in FSAR sections 1.9.133 1,9,133 and 4.4.6.4, submit as an amendment to the program description.
(See Item 191) 131 Not Used 132 10CFR50.73 License Event Report prepared on Form NRC 366 report of Within 60 days after the discovery of Written .El VP NLM NRC-DCD with NUREG Significant Event as described in 10CFR50.73 (also see Items 4,5, 4, 5, event (Unless otherwise the event. LER Process PIS NRC-RO Rev, 2 1022, Rev. 97, and 133) (2005300013) (1997334707) specified, report an event if it occurred PRB NRC-RI NOTE: All missed surveillances will be reported via LER except as within 3 years of the date of discovery SRB 50,73(a)(2)(B) exempted by 50.73(a)(2)(B) regardless of plant mode or power NOTE: In the case of an valid actuation reported under 50.7350,73 level, and regardless of the (a)(2)(iv), other than actuation of the reactor protection system when significance of the structure, system, the reactor is critical, the licensee may provide notification by event or component that initiated the event.
telephone to the NRC Operations Center with 60 days instead of via an LER 133 50,73 10CFR 50.73 LER supplemental reports to provide information not available for As required or as specified in original Written .El VP NLM NRC-DCD (c) original report or to provide updated information for previous LER. LER PIS NRC-RO (See Items 132)(1997334707) PRB NRC-RI SRB 134 10CFR 50 Reports of effluents released in excess of one half the design Within 30 days from the end of the Written CHEM EAM NLM NRC-DCD Appendix I objective annual exposure based on total cumulative dose from liquid quarter during which the release CM NRC-RO Sec,IV.A Sec.IV.A or gaseous effluents occurred NRC-RI PRB 135 ASME Inservice inspection summary report for Class 1 and 2 components Within 90 days of the completion of Written MIS VP NLM NRC-DCE Section XI See Item 200 & & 201 the inservice inspection conducted PRB NRC-RO IWA-6230 during a refueling outage NRC-RI NRC-NRR 136 Not Used Footnotes at end of Table 2 Printed Pnnted June 1 19, 2009 at 11 :44 11:44
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.,. 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 45 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To 137 Not Used 138 Not Used 139 Not used 140 Not Used 141 Not used 142 NPDES Permit Operation monitoring report (form WQ 1.45) Quarterly: by the 28th day the month Written CHEM VP MEA GEPD No. GA0026786 following the completed period. CM 143 Not Used 144 Tech Specs Steam Generator Tube Inspection Report Within 180 days after the initial entry Written PE VP NLM NRC-RO 5.6.10 into Mode 4 following completion of ESM NRC-RI steam generator eddy current PRB inspection performed in accordance with Tech Specs 5.5.9, SG Program change 145 10CFR70.32 Changes made in emergency plan without prior NRC approval (See Within six months of the change Written EPG VP NLM NRC-RO (i) item 101) PTEPM PRB 146 Georgia Ground Ground water use report Semiannually, Due Jan. and July 15th. Written CHEM/OPS CM MEA GEPD Water Permit No. (1984301899) CM 017-0003 147 10CFR26.71 (d) Fitness-For-Duty program performance data. Biannually within 60 days of end of Written SNSH NLM NLM NRC-DCD June and December. NL VP Footnotes at end of Table 2 Printed June 19, 2009 at 11:44
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- Number Rev 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 46 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted Item Reviewed By By By To No.
148 Air Quality Permit Report copies of diesel fuel supplier certifications for all applicable Semiannually Written SC VP MEA GEPDA No, No. 4911-033- fuel, fuel. SCS 0030-V-02-0 149 Air Quality Permit Deviations from permit conditions not reported in Item 54 Semiannually Written OPS VP MEA GEPDA No, No. 4911-033- CM 0030-V-02-0 CEA 150 OCFR74, 13(a) 10CFR74.13(a) Each licensee, including nuclear reactor licensees as defined in § Annually, submit a report within 60 Electronic NFD NFD NFD NRC-NMSS 5021 and 0,22 50.21 0.22 of this chapter, authorized to possess at anyone time calendar days of the beginning of the and location special nuclear material in a quantity totaling more than physical inventory required by § 350 grams of contained uranium 235, uranium-233, or plutonium, or 74,19(c), Physical inventory is 74.19(c).
any combination thereof, shall complete and submit, in computer- assumed to begin on June 1 of each readable format Material Balance Reports concerning special nuclear year.
year material that the licensee has received, produced, possessed, (2003203266) transferred, consumed, disposed of, or lost lost. The Physical Inventory Listing Report must be submitted with each Material Balance Report.Report, Each licensee shall prepare and submit the reports described in this paragraph in accordance with instructions (NUREG/BR-0007 and NMMSS Report D-24 0-24 "Personal Computer Data Input for NRC Licensees'}
Licensees") .
151 Air Quality Permit Written Certification of compliance with the permit Annually Written CEA VP MEA GEPDA No, No. 4911-033- N/A 0030-V-02-0 152 Not used.
153 10CFRSO.36a(a)(
10CFR50,36a(a)( Annual radioactive effluent release report (includes report of solid Annually by May 15th Written CHEM MEA NLM NRC-DCD 2), Tech Specs waste and irradiated fuel shipments) CM NRC-RO 5,6,3, Reg Guide 5.6.3, (Items 153.1, 153,1, 153.2 153,2 and 153.3 153,3 included with this report) CEA NRC-RI 1.21 ODCM 7.2 1,210DCMn VP 153,1 153.1 PCP 12.0 Changes to "Process Control Program" Same as 153 Written RWE VP NLM NRC-RO OM CM Footnotes at end of Table 2 Printed June 19, 2009 at 11:44
Approved L
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VogtleElectric b,,(1erating Plant ....* {';';:<.:""." 00102-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 47 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To 153.2 Tech Specs 5.5.1 Changes to ""Offsite Dose Calculation Manual" (ODCM) Same as 153 Written CHEM VP CC NRC-RO ODCM 7.2.2.5 CM CEA 153.3 ODCM 7.2.2.7 Major changes to liquid, gaseous, and solid radwaste treatment Same as 153 Written SE VP NLM NRC-RO systems CM 154 OL NPF-68 Annual Environmental Operating Report Annually by May 1st sl Written CEA MEA NLM NRC-DCD or NPF-81 CHEM NRC-RO Appendix B Section 5.4.1 155 Tech Specs Annual Radiological Environmental Operating Report Annually by May 15'" Written CEA MEA NLM NRC-RO 5.6.2 CHEM NRC-IE NLM 156 10CFR Annual personnel exposure and monitoring reports Prior to April 30th Written or HP VP NLM NRC-NRR 20.2206(b)(c) Electronic. HPM NRC-RO (Electronic is 13INDIV preferred method) 157 29CFR 1904.32 29CFR1904.32 Summary of occupational injury and illness (OSHA Form 300A) (See Summary - Posted no later than Feb SHIH VP SHIH SNSE item 52) (H ighest Level Management on site)
(Highest 1 1",t of the year following the year PM covered by the records and keep posted until Apr 30th 158 Not used.
159 10CFR50.54(w)(
10CFR50.54(w)( Current levels and sources of licensee's onsite onsile property propeliy damage Annually by April 1 Written SCS NLM FOS-VP NRC-DCD
- 3) insurance. Risk Mgt NRC-RO NRC-RI 160 10CFR Annual financial report Annually Written NL VP NLM NRC-DCD 50.71(b) NLS NRC-RO NRC-RI 161 Not Used 162 NPDES Annual priority pollutants certification for cooling tower blowdown Annually Written CHEM MEA NLM GEPD Permit No. CM GA 0026786 Printed June 19, 2009 at 11:44
Approved " 'i/;c'; Prl ; Number Rev J.D. Williams .Vogtle Electric l:;."(l~*(~ting Plant '<<;f) 00102-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 48 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed Bl:By By By To 163 Item Deleted 164 10CFR Employee radiation exposure data Annually - at request of employee Written HP N/A HP Employee 19.13(b) N/A 165 NPDES 1 year flow characterization study Annually Written CHEM VP MEA GEPD Permit No. CM GA 0026786 GA0026786 166 10CFR50.46(a) Changes to or errors in an acceptable ECCS evaluation model, or in At least annually if not significant Written NL FOS-VP NLM NRC-DCD (3) the application of such a model, that affect the temperature NLM NRC-RO (ii) (also calculation. NRC-RI Appendix K)
AppendixK) (See Item 118) VP 167 10CFR50.54(a) Changes in licensee's QA OA Program which do not reduce licensee's At least annually in accordance with Written OA QA VP NLM NRC-DCD (3) commitments set forth in the QA OA program previously accepted by the 10CFR50.71 OAS QAS NRC-RO NRC. Changes that do reduce commitments require prior NRC PRB NRC-RI 1OC FRSO.S4(a)(3) approval under 10CFR50.54(a)(3) 168 10CFR140.21 Guarantee of payment of deferred premiums. Annually Written All actions by NRC-NRR GPC or Comptroller Cometroller NRC-NMSS 169 NPDES Inventory of all water treatment chemicals discharged during the Annually Written EA/CHEM VP MEA GEPD Permit No. previous twelve months. MEA/CM GA 0026786 170 Not Used 171 NRC RPV Head Perform visual inspection to identify potential boric acid leaks from Within 60 days after returning the Written PE VP NLM NRC-DCD Inspection Order pressure retaining components above the RPV head. (per procedure plant to operation (if a leak or boron ESM NRC-RO EA-03-009 84008-C. deposit was found either during the PRB NRC-RI Dated 2120/04 2/20104 inspection required by the inspection Paragraph C order or otherwise and regardless of the source)
Printed June 19, 2009 at 11 :44
Approved L "<,;,:",>;",, '., .* ' ;, Prt ; Number Rev J.D. Williams .. ;'.;;, .. . 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 49 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To 172 NRC RPV Head Per Procedure 84008-C the calculation contained in the Order will be Submit appropriate report (172.1, Written PE VP NLM NRC-RO Inspection Order performed prior to each refueling to determine the Effective 172.2, or 172.3) within 60 days after ESM NRC-RI EA-03-009 Degradation Years (EDY) for the end of each operating cycle. The returning the plant to operation PRB Dated 2/20/04 following inspections are required to be performed. The frequency is (2005300012) (2005300016)
(2005300012)(2005300016)
Paragraph C based on the EDY category. (2005300009)
- a. Bare metal visual exam of 100% of the RPV head surface (including 360' 360 0 around each RPV head penetration nozzle.
b(1) Ultrasonic testing of each RPV head penetration (Le.,
(i.e., nozzle base material from 2 inches above the J-Groove weld to the bottom of the nozzle and an assessment to determine if leakage has occurred into the interference fit zone. OR b(2) Eddy current testing or dye penetrant testing of the wetted surface of each J-Groove weld and RPV head penetration nozzle base material at least 2 inches above the J-Groove weld.
172.1 Low category-Plants with a calculated EDY of less than 8 AND no Perform inspection "a" at least every cd previous inspection findings requiring classification as High .. 3rd refueling outage or every 5 years, which ever occurs first (if not performed during the refueling outage immediately preceding the order then it must be completed within the next 2 refueling outages. Additionally either complete inspection "b(1)" or "b(2)"
once within 5 years of issuance of the Order and at least every 4th refueling outage or every 7 years thereafter whichever occurs first 172.2 Moderate category-Plants with a calculated EDY less than or equal Perform either "a" or either of the "b" to 12 and greater than or equal to 8 AND AN D no previous inspection inspections every refueling outage.
findings requiring classification as High. Additionally, perform inspection "a" AND either of the "b" inspections at least once every 2 refueling outages Printed June 19, 2009 at 11:44
Approved LL.
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- Number Rev Page Number 50 of 77 40 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To 172.3 High category-Plants with a calculated value of EDY greater than Each refueling perform inspection "a" 12, OR Plants with an RPV head that has experienced cracking in a and either of the "b" inspections penetration nozzle or J-Groove weld due to PWSCC.
173 10CFR The licensee shall submit, as specified in Section 50.4, 50A, a report Within 6 months of the end of each Written NL VP NLM NRC-DCD
- 50. 59( d) (2) 50.59(d)(2) containing a brief description of any changes, tests, and experiments, Unit 2 refueling outage NLM NRC-RO including a summary of the evaluation of each. A report must be NRC-RI submitted at intervals not to exceed 24 months.
174 10CFR FSAR updating Within 6 months of the end of each Written NL VP NLM NRC-DCD 50.71 (e)(4) Unit 2 refueling outage NLM NRC-NRR NRC-RO NRC-RI 175 40CFR Generator's biennial bien nial hazardous waste report. Applicable only if a Every 2 years, March 1 of even Written CHEM CM MEA GEPD 262A1 262.41 Large Quantity Generator numbered years CM 176 Not used 177 10CFR Status of decommissioning funding Every 2 years, by March 31 of odd Written NL NLM NLM NRC-DCD SO.7S(f)(1) 50.75(1)(1) numbered years NRC-RO NRC-RI 178 10CFR 50 Report of test results of specimens withdrawn from capsules (fracture Within one year of capsule removal Written PE VP NLM NRC-DCD Appendix H toughness tests) ESM NRC-RO Sect IV.A Sect. PRB NRC-RI 179 10CFR140.1S 10CFR140.15 Proof of financial protection. As required by the NRC Written ANI SCS &
Marsh & NRC-NRR (a) N/A Risk Mgt. Mgt McClennon or NRC-NMSS 180 10CFR140.17 Liability insurance policy renewal. 30 days prior to termination of policy Written SCS VP NLM NRC-NRR (b) Risk Mgt. or NRC-NMSS 181 10CFRSO 10CFR50 Vessel fracture toughness level below predicted value At least 3 years prior to the date when Written MIS VP NLM NRC-DCD Appendix G the predicted fracture toughness levels ESM NRC-RO Sect IVA1.c.
Sect. IVA 1.c. will no longer satisfy 10CFR 50 SO App. G PRB NRC-RI Sect IVAi Sect. IVA1 182 10CFR i0CFR Request for registration as user of authorized packages Prior to use of package first time Written HP VP NLM NRC-NMSS 71.12(c)(3) (transportation of radioactive material) HPM PRB 11 :44 Printed June 19, 2009 at 11:44
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Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 51 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To 183 10CFR Each licensee who transfers and each licensee who receives special Upon transfer or receipt of SNM. Electronic SNMC N/A N/A NRC-NMSS 74.15(a) nuclear material shall complete in computer-readable format a Nuclear Material Transaction Report. This should be done in accordance with instructions wherever the licensee transfers or receives a quantity of special nuclear material of 1gram or more of contained uranium-235, uranium -233, or plutonium. Copies of these instructions (NUREG/BR-0006 and NMMSS Report D-24 0-24 "Personal Computer Data Input for NRC Licensees") may be obtained either by writing the U. S. Nuclear Regulatory Commission, Division of Fuel Cycle Safety and Safeguards, Washington, DC 20555-0001, bye- by e-mail to RidsNmssFess@nrc.gov, or by calling (301) 415-7213. This prescribed computer-readable format replaces the DOE/NRC Form 741 which has been previously submitted in paper form.
184 Tech Spec RCS Pressure Temperature Limits Report Upon issuance for each reactor vessel Written SE VP NLM NRC-DCD 5.6.6 fluency period and for any revision or ESM NRC-RO supplement thereto. NRC-RI PRB 185 10CFR Radioactive waste manifest At each shipment Written HP N/A HPM Waste 20.2006(b) HP Collector/
Recipient 186 Georgia Replacement of permitted well must receive prior approval from As needed Written CEA VP MEA GEPD Ground Water GEPD. MEA Permit No.
017-0003 187 NUREG Before engaging in any additional construction or operational Prior to proceeding with activity Written CEA MEA NLM NRC-RO 1087 activities that may result in significant adverse environmental impact N/A Section 6.1 that was not evaluated or that is significantly greater than previously evaluated obtain approval from NRC 188 FSAR 10.3.5 Secondary water chemistry program non-conservative changes Prior to initiation of the change Written CHEM VP NLM NRC-RO Table CM 10.3.5-1 PRB 189 Tech Spec Core Operating Limits Report Upon issuance for each reload cycle Written NFD VP NLM NRC-5.6.5 (1995329951 ) including any mid-cycle revisions or NL NRC-RO supplements thereto. NRC-RI Printed Pnnted June 19, 2009 at 11:44
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Approved L Pr~ ) Number Rev J.D. Williams hi:' *.. . 'y;;.: '.' ..i)" ; > ,. * ~c; >; " , " " '.~
. "'>::/ 00152-C 40 Date Approved Page Number umber 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 52 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. Reviewed By By By To 190 Generic Mishaps involving LLW forms prepared for disposal As needed Written HP vp VP NLM (11)NRC Letter HPM and state 91-02 PRB agency agenc' 191 Reg Guide Changes to the DMIMS alert level for power operation (not required Within 90 days Written SE vp VP NLM NRC-OC 1.133 for temporary changes) or there is a change to the preexisting alert ESM NRC-RI Section levels for power operations and the new levels exceed the ability of PRB C.3.a(2)(a) the DMIMS system to detect a loose part as defined in FSAR sections 1.9.133 and 4.4.6.4. (See item 130) (1984313140) 192 10CFR50.54 Sworn written statements to enable NRC to determine whether or not Upon NRC's request Written NL VP NLM NRC-DCD (f)
(I) license should be modified, suspended or revoked (licensee's NLM NRC-RO agreement to comply is a condition of the license) NRC-RI 193 10CFR55.27 Documentation of Licensed Operator's medical history. Upon request. Written TS PM NLM NRC-NRR PTEPM NRC-RO PRB 194 10CFR71.97 Revised schedule information concerning shipments of nuclear As schedule changes Telephone HP HPM HPM Governor or (e) waste. HPM Governor's designee (retain name of person)
NRC-RO 195 10CFR71.97 Notice of cancellation of shipment of nuclear waste. Upon cancellation of shipment Written HP HPM HPM Governor or (f)
(I) HPM Governor's designee NRC-RO 196 10CFR Uncorrected performance test failures and schedule for correction for Every 4 years on the anniversary of Written SimC VP NLM NRC-RO 55.46(d)(3) licensees using simulation facility consisting solely of a the certification SOP plant-referenced simulator. PMTEP 197 10CFR 50 Submit test methods for supplemental fracture toughness Prior to testing Written PE VP NLM NRC-DCD Appendix G, testing. ESM NRC-RO III.B IILB PRB NRC-RI 198 Not Used Printed Pnnted June 19, 2009 at 11:44 11 :44
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Pre 001S2-C j Number Rev 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 53 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared By Approved Submitted Submitted No. By Reviewed By: By By: By To 199 Note Used 200 10CFR Evaluate acceptability of inaccessible areas in ASME Code Class CC Within 90 days of the completion of Written SE VP NLM NRC-DCD 50,55a(b)(2)(viii)(
50.55a(b)(2)(viii)( components when conditions exist in accessible areas that could the inservice inspection conducted ESM NRC-RO E) indicate the presence of or result in degradation to such inaccessible during a refueling outage NRC-RI areas, For each inaccessible area identified, provide information areas. PRB required by the referenced requirement in report addressed in Item 135, 135.
201 10CFR Evaluate acceptability of inaccessible areas in ASME Code Class Within 90 days of the completion of Written SE VP NLM NRC-DCD 50,55a(b)(2)(ix)(
50.55a(b)(2)(ix)( MC components when conditions exist in accessible areas that could the inservice inspection conducted ESM NRC-RO A)
A} indicate the presence of or result in degradation to such inaccessible during a refueling outage NRC-RI areas For each inaccessible area identified, provide information PRB required by the referenced requirement in report addressed in Item 135, 135.
202 10CFR For flaws or areas of degradation in ASME Code Class MC Within 90 days of the completion of Written SE VP NLM NRC-DCD 50,55a(b)(2)(ix)(
50.55a(b}(2)(ix}( components which exceed acceptance standards, provide the the inservice inspection conducted ESM NRC-RO D)(1 and (2})
(2>> information required by the referenced requirement in report during a refueling outage NRC-RI addressed in Item 135. 135, PRB 203 Georgia A Water Conservation Progress Report must be submitted to the Starting the 6th of August 201 6'" 20100 and Written CHEM VP NLM GEPD th Ground Water GEPD on or before 6th of August 201 2010, thereafter, 0, and every 5 years thereafter. every 5 years thereafter. CM No, Permit No. The progress report should include actions and/or improvements CEA 017-0003 made to conserve water and reduce water losses (e.g., (e,g" leak detection and leak repair, meter installation, meter replacement or etc,), The Water Conservation Progress Report must calibration, etc.).
also include a comparison of unaccounted for water trends as well as comparison of gallons of water used per unit product produced for each of the 5 years presented in the report. report, 204 Georgia A raw groundwater sample shall be analyzed annually for specific Annually Written CHEM CM MEA GEPD Water Ground Water conductance, The results shall be submitted to GEPD on corporate conductance. CEA No, Permit No. letterhead and shall include the date sampled, a map depicting 017-0003 where the sample was collected, temperature of the sample at the testing, and the specific conductance.
time of testin!l, conductance, 205 Georgia EPD Hazardous Waste Management Fee and Record Annually Written CHEM VP MEA GEPD Rule 391-3-19 CEA 206 Hazardous Substance Reporting Fee Annually Written CHEM VP MEA GEPD CEA 207 Landfill D&O Landfill Groundwater Report Semi-Annually Written CHEM MEA MEA MEA. GEPD Plan CEA 208 Landfill D&O Landfill Methane Monitoring Report Quarterly Written CHEM MEA MEA GEPD Plan CEA 209 Army Corps of Completion of intake structure maintenance dredging. dredging, Within 30 days after completion of Written MM MEA MEA MEOA COE Engineers Permit (See item 70) activity CEA GEPD 200500606 Printed Pnnted June 19, 2009 at 11 :44 11.44
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Prl ) Number Rev J.D. Williams : ~ogtl~.,Electric ~""(leratingPlant>,, ", '. . :,.;<::;; <J'",,;",iM' 00102-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 54 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Pregared Pre[1ared Bll B}' Approved By Submitted By Submitted To No. Bll Reviewed B}'
210 40 CFR 262.42 Exception report for non-receipt of copy of manifest from designated For a large quantity generator, 45 days Written CHEM MEA MEA GEPD receiving facility.
facility, after shipment was accepted by CEA transporter, 60 days after shipment transporter.
was accepted by transporter for other status, generator status.
211 General NPDES Annual Report from Appendix B of General Permit GAROOOOOO Initial report must be submitted by the Written CHEM VP MEA GEPD th Permit 26th month after the effective date of CM GAROOOOOO the permit and on an annual basis CEA thereafter, thereafter.
212 Superfund SARA Title III Tier 2 Report Annually Written CHEM VP MEA GEPD Amendment CM Reauthorization CEA Act (SARA) 213 40 CFR 372 Rele.ise Inventory (TRI) Report Toxic Release Annually Written CHEM VP MEA EPA CEA Printed June 19, 2009 at 11:44 11 :44
.\
Approved L'- <"; Prt ) Number Rev J.D. Williams :.):t!' I V Qgtl~.Electric l,,~oeraHn'lrplan! I \. ;"':'\j':~l);:t; 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 55 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Prepared PreJ<ared By Approved By Submitted By Submitted To No. Reviewed BBy 214 40 CFR 141 When an unmonitored and inadvertent contaminated leak or spill Before the end of the next business Telephone
'"Telephone CHEM VP MEA GEPD 40 CFR 190 >100 gallons to the environment occurs, provide the following to the day after the discovery.
discovery, CM 10 CFR 50 state/local agency:
NMP-EN-002
- 1. Location and date of discovery/occurrence 2.
2, Volume (or best estimate) and leak rate if event is a leak
- 3. Radioactivity levels
- 4. Actions taken or being taken to remediate 5, Impact to environment (e.g. groundwater, soil, etc) 5.
- 6. Communication if leak or spill got outside the protected area or is expected to get out of the protected area.
area, See footnote #15.
215 40 CFR 141 When an unmonitored and inadvertent contaminated leak of Before the end of the next business Telephone
'"Telephone CHEM VP MEA GEPD 40 CFR 190 unknown volume to the environment is found, provide the following to day after the discovery.
discovery, CM 10 CFR 50 the state/local agency:
NMP-EN-002 1, Location and date of discovery/occurrence 1.
- 2. Volume (or best estimate) and leak rate if event is a leak
- 3. Radioactivity levels 4.
4, Actions taken or being taken to remediate 5.
5, Impact to environment (e.g.
(e.g, groundwater, soil, etc)
- 6. Communication if leak or spill got outside the protected area or is expected to get out of the protected area.
- 15, See footnote #15.
216 40 CFR 141 When all leaks, regardless of volume or activity from a spent fuel Before the end of the next business Telephone
'"Telephone CHEM VP MEA GEPD 40 CFR 190 pool occurs to the ground outside the buildings, provide the following day after the discovery. CM 10 CFR 50 to the state/local agency:
NMP-EN-002
- 1. Location and date of discovery/occurrence
- 2. Volume (or best estimate) and leak rate if event is a leak
- 3. Radioactivity levels
- 4. Actions taken or being taken to remediate
- 5. Impact to environment (e.g. groundwater, soil, etc) 6.
6, Communication if leak or spill got outside the protected area or is expected to get out of the protected area.
See footnote #15.
- 15, Footnotes at end of Table 2.
11:44 Printed June 19, 2009 at 11 :44
u~{lerati ngFiIiilt"",*,*. A Approved l._ Pro ,Number Rev J.D. Williams
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- r*",'*;*; 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 56 of 77 TABLE 2 FEDERAL AND STATE REPORTING MATRIX PLANT ORIGINATED Item Requirement Report Title Or Condition Frequency Method Pre[1ared By Prel2ared B'i Approved By Submitted By Submitted To No. Reviewed By B'i 217 40 CFR 141 When a confirmed water sample from an offsite or onsite ground Before the end of the next business '"Telephone Telephone CHEM VP MEA GEPD 40 CFR 190 water sample exceeds the reporting criteria for drinking water as day after the discovery. CM 10 CFR 50 listed in Table 4-2 of the ODCM, provide the following to the NMP-EN-002 state/local agency:
(Note: This requirement is for newly identified leaks/spills or suspected new leaks/spills. It does not apply to already existing and reported groundwater conditions.)
- 1. Location and date of discovery/occurrence
- 2. Volume (or best estimate) and leak rate if event is a leak
- 3. Radioactivity levels
- 4. Actions taken or being taken to remediate
- 5. Impact to environment (e.g. groundwater, soil, etc)
- 6. Communication if leak or spill got outside the protected area or is expected to get out of the protected area.
See footnote #15.
218 40 CFR 141 When a surface water sample exceeds the reporting criteria for water Before the end of the next business '"Telephone Telephone CHEM VP MEA GEPD 40 CFR 190 as listed in Table 4-2 of the ODCM, provide the following to the day after the discovery. CM CFR 50 10 CFR50 state/local agency:
NMP-EN-002
- 1. Location and date of discovery/occurrence
- 2. Volume (or best estimate) and leak rate if event is a leak
- 3. Radioactivity levels
- 4. Actions taken or being taken to remediate
- 5. Impact to environment (e.g. groundwater, soil, etc)
- 6. Communication if leak or spill got outside the protected area or is expected to get out of the protected area.
See footnote #15.
219 NMP-EN-002 Notify the NRC in accordance with site or company applicable When state/local agency is provided Telephone CHEM VP MEA NRC-OC 10 CFR 50.72 procedures using the guidance and contents of Figure 1 of with information related to potential CM NRC-RI NMP-EN-002. groundwater contamination event.
Footnotes at end of Table 2.
Printed June 19, 2009 at 11 :44
'I ~~r,~vog~le Ele,ctric Generating Plant Approved By Procedure Number Rev J.D. Williams .\ 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 57 of 77 Footnotes for Table 2
- 1. NRC-RO Commercial Phone No.(24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) - (301) 415-7197
Maintain an open, continuous communications channel with the NRC Operations Center upon request of the NRC. Backup number (301) 951-0550
- 3. NRC-IE Commercial Phone No. (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) - (301) (301)992-7000.
992-7000.
- 4. Notifications and reports to the state and local government Agencies as identified by Procedure 91002-C, 91 002-C, "Emergency Notifications."
- 5. FAA Flight Service NOTAM# line is 877-487-6867 ask for Georgia information. Provide Longitude and Latitude and height for Cooling Towers as follows:
Antenna registration numbers (ARN's) are not required for these towers.
Unit 1: 33 degrees 08 minutes 31 seconds Unit 2: 33 degrees 08 minutes 38 seconds north 81 degrees 45 minutes 24 seconds north 81 degrees 45 minutes 24 seconds
~~
west ~~
west Tower Height 543 feet AGL (Above Ground Level) 758 feet AMSL (Above Mean Sea Level)
( FAA study numbers Unit 1 Cooling Tower -- 1979-ASO-1456-0E Unit 2 Cooling Tower -- 1979-ASO-1457-0E
- 6. Refer to Procedure 94001-C, "Spill Prevention, Control, Countermeasures (SPCC) and Reportability" for phone numbers and proper reporting chain. If SNC Environmental Affairs (EA) can not be contacted, the ESC or designee will notify the Georgia EPD directly.
- 7. DOT - Toll free number (800) 424-8802 or commercial no. (202) 426-2675. For notices involving etiologic agents also call Director for Disease Control, Public Health Health Service, Atlanta, GA, (404) 633-5313.
- 8. Environmental Affairs 1-800-522-2246/0444071 (Environmental - on call).
- 9. Reportable events which require reports to other federal, state and local agencies will be submitted per those requirements. A copy of such reports shall be sent to the NRC in lieu of reports required by this section.
- 10. Security Management (Secm) may be anyone of the following:
-Security Manager
-Nuclear Security Captain Printed June 19, 2009 at 11 Pnnted 11:44
- 44
Electtj~ Gener~ftn~ Plant ~'----'-------1__00152-C ~_r~c_1e_~~_r~_~_um_b_e~_;_ev-l Approved By Procedure Number Rev J.D. Williams ie,'" Vogtle 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 58 of 58 of 7777 Footnotes for Table 2 (Cont'd)
- 11. NRC - Division of LLW Management and Decommissioning and the State Disposal Site Regulatory Agency.
12 Occupational Safety and Health Administration 1-800-321-6742
- 13. INDIV: The report(s) must be sent to the individual(s) no later than the time the report(s) are transmitted to the Commission. (Reports to individuals will be by written method).
- 14. VEGP does not currently arrange for the physical protection of special nuclear material of low strategic significance while in-transit as described in 10 CFR 73.67(g)(3). If at such time as VEGP arranges for such protection, the reporting requirements of 10 CFR 73.71 as referenced by this footnote are applicable.
- 15. Inform the State Radiological Control or equivalent department(s) (and any local official if requested), in accordance with NMP-EN-002 with as much information as known. Georgia Environmental Protection Division - Department of Natural Resources 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> emergency contact number: 1-800-241-4113. Verbal notification may be followed by written notification if requested by state/local agency.
(
Printed Pnnted June 19, 2009 at 11 :44
Vogtl~Electric u~()eratingf>lant Approved ..'- PrL ) Number Rev J.D. Williams Date Approved
.' .. '. .. * \ ",,,,. ,;,,, 00152-C Page Number 40 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 59 of 77 TABLE 3 SECURITY REPORTABILITY MATRIX The time constraints to implement compensatory measures and initiate NRC reports typically start at the time the event occurs. If the exact time cannot be determined, the time of event discovery by on-site security management or an equivalent level of plant management will be used for initiating compensatory measures and event reporting.
Compensatory Measures Taken Within the Timeframe Described Event Description System/Component Status in Procedure 9010S-C, 5.62, 90106-C, RG 5.S2, Reportability NUREG-1304, and Generic Letter 91-03 A. THREATS OR ACTS
- 1. Any event in which a Theft of unlawful diversion of N/A One Hour person has committed or any special nuclear material caused, or attempted to atVEGP at VEGP commit or cause a credible threat to commit or cause:
- 2. Any event in which a Significant physical damage N/A One Hour person has committed or to either VEGP unit, spent caused, or attempted to fuel pool, or nuclear fuel commit or cause a credible transport vehicle to the threat to commit or cause: extent that it cannot perform its normal function Printed June 19, 2009 at 11 :44
c~,_.
Vogtle Electric \:;,~(teratinggl~nt Approved, Prl ; Number Rev J.D. Williams " .," ;':."".,." ','
..\ 00102-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 60 of 77 TABLE 3 (CONT'D)
SECURITY REPORTABILITY MATRIX Compensatory Measures Taken Within the Timeframe Described Event Description System/Component Status 90106-C, RG 5.62, in Procedure 9010S-C, 5.S2, Reportability NUREG-1304, and Generic Letter 91-03 B. LOSS OF VA BARRIER SYSTEM/COMPONENT
- 1. VA Barrier(s) (1) Physically/Forcibly Breached N/A One Hour and or Doors(s)
Design flaw or vulnerability Yes Log in VA barrier No One Hour
- 2. Keycard and Alarmed-Not Locked Yes Log Alarmed Doors(s) No One Hour Locked-Not Alarmed Yes Log No One Hour Not Alarmed-Not Locked Yes(2) Log No One Hour
- 3. Card Reader Failure allows Yes Log Failure unauthorized access No One Hour through door Printed Pnnted June 19, 2009 at 11:44 11 :44
Approved, Pre ) Number Rev J.D. Williams "",',c, ' "
00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 61 of 77 (CONTO)
TABLE 3 (CONT'O)
SECURITY REPORTABILITY MATRIX Compensatory Measures Taken Within the Timeframe Described Event Description System/Component Status 5.S2, 9010S-C, RG 5.62, in Procedure 90106-C, Reportability NUREG-1304, and Generic Letter 91-03 C. LOSS OF VA BARRIER SYSTEM/COMPONENT
- 1. PA Barrier(s) (1) Physically/Forcibly Breached N/A One Hour or Turnstiles(s)
Design flaw or vulnerability Yes Log in PA barrier No One Hour
- 2. Turnstile(s)/PA Alarmed-Not Locked Yes Log Entry Doors(s) No One Hour Locked-Not Alarmed Yes Log No One Hour Not Alarmed-Not Locked Yes Log No One Hour D. LOSS OF ALARM ASSESSMENT CAPABILITIES
- 1. Loss of E-field when in Inoperable/ Out of Yes Log service (one or more Service due to No One Hour zones) Maintenance/Repai rlT esting Maintenance/RepairlTesting
- 2. Loss of Microwave Inoperable/ Out of Service due Yes Log (one or more zones) Maintenance/Repair/Testing to Maintenance/RepairlTesting No One Hour Printed Pnnted June 19, 2009 at 11 :44 11:44
Vogtle Electric ~~f1erciting;'Plant Approved t..
- l. Pro
- Number Rev J.D. Williams .\. 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 62 of 77 TABLE 3 (CONT'D)
REPORTABILITY MATRIX SECURITY REPORTABIUTY Compensatory Measures Taken Within the Timeframe Described Event Description System/Component Status in Procedure 9010S-C, 90106-C, RG 5.S2, 5.62, Reportability NUREG-1304, and Generic Letter 91-03 D. LOSS OF ALARM ASSESSMENT CAPABILITIES
- 3. Loss of CCTV Surveillance Inoperable/ Out of Service due to Yes Log (one or more Maintenance/Repairs/Testing No One Hour cameras/monitors)
- 4. Loss of MUX Keycard/Alarm Doors Yes Log (Total) Inoperable/Out of Service due to No One Hour Maintenance/Repair/Testing
- 5. Loss of MUX Keycard/Alarm Doors Yes Log (Partial) Inoperable/Out of Service due to No One Hour Maintenance/Repair/Testing
- 6. Loss of both Inoperable/Out of Service due to Yes Log Computers Maintenance/Repair/Testing No One Hour
- 7. Loss of one Computer Fail over to Backup Yes(2) Log No log Log
- 8. Loss of CAS/SAS Alternate Alarm Station N/A Log Operable log Log Both Stations inoperable Yes One Hour No Printed June 19, 2009 at 11 :44 11:44
Vogtl~ Electric (:n:nerating'<i~J\~;~t':'" . \
Approved Pre, ;"Number Number Rev L
J.D. Williams
,i;\i:f"":,:;, . ,;: .
..<;,. 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REPORTiNG REQUIREMENTS 63 of 77 TABLE 3 (CONT'D)
SECURITY REPORTABILITY MATRIX Compensatory Measures Taken Within the Timeframe Described Event Description System/Component Status in Procedure 90106-C, 9010S-C, RG 5.62, 5.S2, Reportability NUREG-1304, and Generic Letter 91-03 D. LOSS OF ALARM ASSESSMENT CAPABILITIES
- 9. Loss of PA lighting Inoperable/Out of Service Yes Log (Total (T otal or Partial) due to No One Hour Maintenance/RepairlTesting Maintenance/Repair/Testing Momentary Loss of PA N/A Log lighting E. LOSS OF POWER
- 1. Loss of AC Power Standby Power Yes Log (Total (T otal or Partial) Maintains Integrity No One Hour
- 2. Loss of Standby during Total Power Loss N/A One Hour loss of Power F. LOSS OF LLEA COMMUNICATIONS
- 1. Partial Loss At least one method of N/A Log communicating operable
- 2. Total Loss All on-site communications N/A One Hour capabilities Inoperable/Out inoperable/Out of Service due to Maintenance/Repair/Testing Printed June 19, 2009 at 11:44 11 :44
Electric b<:neratingP,I~~l\";'*'<<\.
Approved,L Approved "., Pre J Number Rev J.D. Williams Vogtle
'. t ", 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 64 of 77 TABLE 3 (CONT'O)
SECURITY REPORTABILITY MATRIX Compensatory Measures Taken Within the Timeframe Described Event Description System/Component Status in Procedure 9010S-C, RG 5.S2, Reportability NUREG-1304, and Generic Letter 91-03 G. INTERRUPTION OF NORMAL OPERATIONS
- 1. Fire or Explosion of N/A N/A One Hour suspicious or unknown origin within the PA, including the Isolation Zone
- 2. Suspect explosive device N/A N/A One Hour discovered
- 3. Unauthorized use of, N/A N/A One Hour vandalism, or tampering, with machinery, components or controls, including the security system/components
- 4. Credible Bomb or Extortion N/A N/A One Hour Threat
- 5. Results of Bomb Search N/A N/A One Hour
- 6. Unsubstantiated Bomb or N/A N/A Log Extortion Threat Pnnted June 19, 2009 at 11 :44 11:44
Approved, , "
' , ' " , PrL .:l
.l Number Rev J.D. Williams . .};t 5?: 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 65 of 77 (CONT'D)
TABLE 3 (CO NT' D)
SECURITY REPORTABILITY MATRIX Compensatory Measures Taken Within the Timeframe Described Event Description System/Component Status in Procedure 901 9010S-C, OS-C, RG 5.S2, Reportability NUREG-1304, and Generic Letter 91-03 G. . INTERRUPTION OF NORMAL OPERATIONS
- 7. Mass Demonstrations or N/A N/A One Hour Civil Disturbance near the site which may pose a threat to the facility
- 8. An assault on the reactor, N/A N/A One Hour regardless of whether perimeter penetration is achieved.
- 9. Attempted or confirmed N/A N/A One Hour intrusion, or actual entry of an unauthorized individual into PANA
- 10. Discovery of/or willful N/A N/A One Hour attempted introduction of weapons, explosives, or incendiary devices in/into the PA.
- 11. Discovery of contraband N/A N/A Log inside the PA that is not a significant threat.
Printed Pnnted June 19, 2009 at 11:44
Approved J.D. Williams
- x
.&a Prl
. 00152-C
.;J Number Rev 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 66 of 77 (CONT'O)
TABLE 3 (CONT'D)
SECURITY REPORTABILITY MATRIX Compensatory Measures Taken Within the Timeframe Described Event Description System/Component Status 5.S2, in Procedure 9010S-C, RG 5.62, Reportability NUREG-1304, and Generic Letter 91-03 G. INTERRUPTION OF NORMAL OPERATIONS
- 12. Reasonable suspicion of N/A N/A One Hour illegal use, sale, possession, NOTE: The requirements of or introduction of 10CFR 26.73 are applicable controlled substances(s) to item 12 regarding on-site or other criminal controlled substances. See acts involving personnel Table 2, item 44.
with unescorted access which may directly affect operations
- 13. Suspension of Safeguards N/A Yes Log controls by the Shift No One Hour Manager during Emergencies including Security.
H. SECURITY FORCE PERSONNEL RELATED
- 1. Unavailability of Minimum N/A Yes Log Number actual, or No One Hour impending strike of Security Personnel.
Pnnted June 19, 2009 at 11:44 Printed 11 :44
/-,
Approved l J.D. Williams r '; " , ,", ',;" '; "";,;' ',:<' eY'
,~,: <;'Mogt',,~~I:I,~~tl1;Q:~,.~i1~rCltil'1g ,
," Pre 00102-C
- Number Rev 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 67 of 77 TABLE 3 (CONT'D) (CONT'O)
SECURITY REPORTABILITY MATRIX Compensatory Measures Taken Within the Timeframe Described Event Description System/Component Status 9010S-C, RG 5.62, in Procedure 90106-C, 5.S2, Reportability NUREG-1304, and Generic Letter 91-03 H. SECURITY FORCE PERSONNEL RELATED
- 2. Loss of Security N/A N/A One Hour Weapon On-site
- 3. Compensatory N/A Yes Log Officer asleep on Post No One Hour I. PROCEDURAL RELATED FAilURE FAILURE
- 1. Loss or Theft of N/A N/A One Hour Safeguards Information which could SIGNIFICANTLY degrade Physical Security Effectiveness (1991321556)
- 2. Inadvertent/accidental N/A Yes(4) Log/Non-Event(5) removal of active No keys/badge from PA
- 3. Discovery of intentionally N/A N/A One Hour falsified identification or Key Cards Printed June 19, 2009 at 11:44
Approved l ,,;;" Pre ,; Number Rev J.D. Williams 00102-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 68 of 77 (CONT'O)
TABLE 3 (CONT'D)
REPORTABILITY MATRIX SECURITY REPORTABllITY Compensatory Measures Taken Within the Timeframe Described Event Description System/Component Status in Procedure 9010S-C, 90106-C, RG 5.S2, 5.62, Reportability NUREG-1304, and Generic Letter 91-03 I. PROCEDURAL RELATED FAILURE
- 4. Unauthorized use of, N/A Yes (4) Log unaccounted for, lost, or No One Hour stolen badge on-site
- 5. Visitor without Escort N/A Yes Log/Non-Event(5)
No Log (5)
- 6. Individual authorized only N/A Yes Non-Event PA access is incorrectly No Log issued a badge granting VA access (initial badge issue), but does not enter any VAs or does not enter any VAs with malevolent intent.
J. MISCELLANEOUS
- 1. Discovery of off-site N/A Yes Log criminal acts by personnel No One Hour with unescorted access which could cause concern about their trustworthiness and reliability or a program weakness.
Printed June 19, 2009 at 11 :44
Approved L J.D. Williams " . ...... ........ ...*ii *....* ..... ': .": Vogtl~. Electri~ u"O.~.(~ti:f1;§'*Plant . \ .., Pre
.: 001b2-C
- Number Rev 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 69 of 77 TABLE 3 (CONT'D)
(CONTD)
REPORTABILITY MATRIX SECURITY REPORTABIUTY Compensatory Measures Taken Within the Timeframe Described Event Description System/Component Status in Procedure 9010S-C, RG 5.S2, Reportability NUREG-1304, and Generic Letter 91-03 letter J. MISCELLANEOUS
- 2. Any criminal act by a N/A Yes Log licensee employee No One Hour(3) committed offsite, which receives media attention, that has the potential for affecting the radiological safety of licensee activities needs to be reported.
- 3. Any other threatened N/A Yes Log attempted, or committed No One Hour act not previously defined having potential for reducing Physical Security Effectiveness
- 4. Vehicle Barrier System N/A Yes Log Degradation No One hour For longer time period (not to Log exceed 30 days)
Printed June 19, 2009 at 11:44
'--'" ~.
- Approved, Approved L ". ,.* Pre Prl ); Number Rev J.D. Williams . V<?gtle'Electtic~~Ogr~~i;; I~~ 001 b2-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 70 of 77 (CONTO)
TABLE 3 (CONT'D)
SECURITY REPORTABILITY MATRIX Compensatory Measures Taken Within the Timeframe Described Event Description System/Component Status in Procedure 9010S-C, RG 5.S2, Reportability NUREG-1304, and Generic Letter 91-03 letter K. PARTIAL FAILURE OF ACCESS CONTROL PROGRAM
- 1. A badged person N/A Yes Log inadvertently enters the PA No One Hour through a vehicle gate before being searched.
- 2. A badged person walks N/A N/A N/A Log into a VA behind another person without using a badge, if the improper entry was inadvertent or without malevolent intent.
- 3. A person enters a VA to N/A N/A log Log which he/she is authorized unescorted access by inadvertently using a badge intended for another person who is authorized access to the area.
Printed June 19, 2009 at 11 :44
Approved" .. ' ..
PrL ; Number Rev J.D. Williams .... , /, ~,';~ 001 o2'-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 71 of 77 TABLE 3 (CONT'O)
(CONT'D)
SECURITY REPORTABILITY MATRIX Compensatory Measures Taken Within the Timeframe Described Event Description System/Component Status 9010S-C, RG 5.62, in Procedure 90106-C, 5.S2, Reportability NUREG-1304, and Generic Letter 91-03 L. INCOMPLETE PRE-EMPLOYMENT SCREENING RECORDS (to include falsification and willful omission from the screening application)
- 1. Unescorted access of the individual may need to be cancelled or suspended until the identified anomaly is resolved. If the licensee determines that unescorted access would have been denied based on developed information that is viewed as significant as outlined in Nuclear Fleet Security (NFS)-003 Adjudication of Derogatory Information and Appeals Process, note the following:
- Discovery of significant information, access is denied and the event will be reported to the NRC within 1-hour.
- Discovery of less significant information, access is denied and the event is recorded in the security safeguards log within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
NOTES (1) Examples include floor or wall plugs, bisco seals, barriers in HVAC ducts, etc.
(2) To be properly compensated, measures must be in place within the timeframe described in procedure 901 06-C.
(3) Events reportable under 50.72 or 50.73 do not require duplicate reports under 73.71 (4) To be properly compensated, both invalidating the badge and initiating actions to ensure no unauthorized use of the badge must occur within the timeframe described in procedure 90106-C.
901 06-C.
(5) May be a log-able event depending on circumstances.
Printed June 19, 2009 at 11 :44
Approved By ".. , Procedure Number Rev J.D. Williams y 00152-C 40 Date Approved
" Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 72 of 77
( TABLE 4 CROSS REFERENCE INDEX SOURCE REQUIREMENT ADDRESSED BY ITEM(s) 10 CFR 19.13 (b) 164 10 CFR 19.13 (c) 111 10CFR19.13(d) 10 CFR 19.13 (d) 109,110 10 CFR 19.13 (e) 21 10 CFR 20.1906 (d) (1) 22 10 CFR 20.1906 (d) (2) 23 10 CFR 20.2006 (b) 185 10 CFR 20 App G, III, (E)(2) 67 10 CFR 20.2201 77 10 CFR 20.2201 (a) 8 10 CFR 20.2201 (b) 102 10 CFR 20.2201 (d) 103 10 CFR 20.2202 (a) 9 10 CFR 20.2202 20 .2202 (b) 43 10 CFR 20.2203 (a) 110 10 CFR 20.2203 (a)(4) 92
( 10 CFR 20.2204 123 10 CFR 20.2206 (b) & & (c) 156 10CFR21.21 10 CFR 21.21 (a) (2) 129 10 CFR 21.21 (c) (3) (i) 46 10 CFR 21.21 (d) (3) (ii) 117 10 CFR 26.27 (d) 20 10 CFR 26.71 (d) 147 10 CFR 26.73 (a) (1) & (2) 44 10 CFR 50.9 (b) 47 10 CFR 50.36a (a) (2) 153 10 CFR 50.36 (c) (2 & & 6) 5 10 CFR 50.46 (a) (3) (ii) 118, 166 10 CFR 50.54 (a) (3) 167 10 CFR 50.54 (f) 192 10 CFR 50.54 (p) (2) 125 10 CFR 50.54 (q) 101 10 CFR 50.54 (w) (3) 159 10 CFR 50.54 (w) (4) (ii) 119 Pnnted Printed June 19, 2009 at 11:44 11 :44
Approved By Procedure Number Rev J.D. Williams 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 73 of 73 77 of 77
( TABLE 4 (CONT'D)
CROSS REFERENCE INDEX SOURCE REQUIREMENT ADDRESSED BY ITEM(s) 10 CFR 50.55a (b) (2) (VIII) (C) 198 10 CFR 50.55a (b) (2) (VIII) (D) 199 10 CFR 50.55a (b) (2) (VIII) (E) 200 10 CFR 50.55a (b) (2) (IX) (A) 201 10 CFR 50.55a (b) (2) (IX) (D) (1) & & (2) 202 10 CFR 50.71 (b) 160 10 CFR 50.95(d)(2) 173 10 CFR 50.71 (e)(1) 174 10 CFR 50.72 2, 33 2,33 10 CFR 50.72 (a) (1) (i) 2.1 10 CFR 50.72 (a) (4) 29 10 CFR 50.72(b) 5 10 CFR 50.72 (a)(1)(ii) & & (b) (1) 2.2 10 CFR 50.72 (b) (2) 33 10 CFR 50.72 (b) (2) (i) 33.1 10 CFR 50.72 (b) (2) (iv) (A) 33.2
- 50. 72 (b) (2) (iv)(8) 33.3 10 CFR 50.72 (b) (2) (xi) 33.4 10 CFR 50.72 (b) (3) 37 10 CFR 50.72 (b) (3) (ii) 37.1 10 CFR 50.72 (b) (3) (iv) 37.2 10 CFR 50.72 (b) (3) (v) 37.3 10 CFR 50.72 (b) (3) (vi) 37.3 10 CFR 50.72 (b) (3) (xii) 37.4 10 CFR 50.72 (b) (3) (xiii) 37.5 10 CFR 50.72 (c) 2.3 10 CFR 50.73 132 10 CFR 50.73 (c) 133 10 CFR 50 Appendix E, Section IV 0.3 1 10 CFR 50 Appendix E, Section V1.3.a 122 10 CFR 50 Appendix E, Section V1.3.b 121 10 CFR 50 Appendix E, Section V 120 10 CFR 50 Appendix G, Section III.B 111.8 197 10 CFR 50 Appendix G, Section IV.A.1.c 181 10 CFR 50 Appendix H, IV, A 178 Printed June 19, 2009 at 11:44
Approved By .,}y Procedure Number Rev J.D. Williams ,,00152-C 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 74 of 77 TABLE 4 (CaNT' (CONT'D)
D)
CROSS REFERENCE INDEX SOURCE REQUIREMENT ADDRESSED BY ITEM(s) 10 CFR 50 Appendix I, Section IV. A 134 10 CFR 50.74a 73 10 CFR 50.74b 74 10 CFR 50.74c 75 10 CFR 50.91 (b) 3 10 CFR 55.25 75 10 CFR 55.27 193 10 CFR 55.46 (d) (3) 196 10 CFR 55.53 (g) 76 10 CFR 70.32 (c) (2) (i) 131 Techs Specs 5.6.10 144 10 CFR 70.32 (d) 126 10 CFR 70.32 (e) 125,128 10 CFR 70.32(g) 127 10 CFR 70.32(i) 145 10 CFR 70.52(a) 10
( 10 CFR 70.52(b) 10 10 CFR 104 10 CFR 71.5(a) 71 10 CFR 71.7(c)(3) 182 10 CFR 71.93(c) 124 10 CFR 71.95(a)(1) 113 10 CFR 71.95(a)(2) 114 10 CFR 71.95(c) 86 10 CFR 71.97 53 10 CFR 71.97(e) 194 10 CFR 71.97(f) 195 Not Used 12 Not Used 13 Not Used 14 Not Used 15,16,17 Not Used 11 10 CFR 73.71(a)(1) 6,7 Printed June 19, 2009 at 11:44
Approved By Procedure Number Rev J.D. Williams 00152-C 40 Date Jo\DDI'DveD Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 75 of 77
( TABLE 4 (CONT'D)
CROSS REFERENCE INDEX SOURCE REQUIREMENT ADDRESSED BY ITEM(s) 10 CFR 73.71 (a)(4) & (5) 79, 80 10 CFR 73.71(b)(1) 18,18.1,18.2,18.3, 18, 18.1, 18.2, 18.3, 18.4, 18.5, 18.6, 81,81.1,81.2,81.3, 81.4,81.5, 81.6 81.4,81.5,81.6 10 CFR 73.71 (b)(2) 82 10 CFR 73.72 55 74.11 (b) 10 CFR 74.11(b) 11 10CFR74.13 10 CFR 74.13 150 10.CFR.74.15 183, 150 10 CFR 50.75 (f)(1) 177 10 CFR 140.6(a) 27 10 CFR 140.15(a) 179 10 CFR 140.15(e) 28 10 CFR 140.17(b) 180 10 CFR 140.21 168 14 CFR FAA advisory AC 70/7460-IK 24
- 1 ASO-1456-0C
- 2 ASO-1457-0E 29 CFR 1904.29 52 29 CFR 1904.32 157 29 CFR 1904.39 35 29 CFR 1926.1101 115 33 CFR 153 25 40 CFR 61 115 40 CFR 110 25 40 CFR 260-265 30 40 CFR 262.41 175 40 CFR 262.42 210 40 CFR 302 31 40 CFR-372 213 49CFR171.15 32 49CFR171.16 49 CFR 171.16 71 TR 13.3.2 Condition B 56 TR 13.3.3 Condition A 57 Tech. Spec. 5.6.8 62 Tech. Spec. 5.6.9 84 TR 13.3.2 Condition C 64 Tech. Spec. 5.6.7 95 Printed June 19, 2009 at 11:44 Pnnted 11 :44
Approved By Procedure Number Rev J.D. Williams 00152-C 40 Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 76 of 77
( TABLE 4 (CONT'D)
CROSS REFERENCE INDEX SOURCE REQUIREMENT ADDRESSED BY ITEM(s)
Tech. Spec. 2.2.3 4 Tech. Spec. 2.2.4 40 Tech. Spec. 5.6.2 155 Tech. Spec. 5.6.3 153 Tech. Spec. 5.6.5 189 Tech. Spec. 5.5.1 153.2 Tech. Spec. 5.6.6 184 Reg. Guide 1.21 153 Reg. Guide 1.35 Section C.8 85 Reg. Guide 1.108 95 Reg. Guide 1.133 Section C.3.a(2)(a) 191 Reg. Guide 1.133 Section C.3.a(2)(e) 130 Reg. Guide 1.133 Section C.5.b 58 Reg. Guide 1.133 Section C.6 39,66 FSAR 10.3.5, Table 10.3.5-1 188 FSAR 16.3 Requirement 3 58
( NUREG 0654 Appendix 1 36,38 36, 38 NUREG 0737 37.6 NUREG 1022 132 NUREG 1087 Section 6.1 187 Operating License NPF-68 or NPF-81 Section 2.h 41,97 Operating License NPF-68 or NPF-81 Section 2.c(1) 41.1, 97.1 Operating License NPF-68/NPF-81 Section 2.c 41.3,97.3 Operating License NPF-68/NPF-81 Appendix B, Section 3.2 98 Operating License NPF-68/NPF-81 Appendix B, Section 4.1 42 Operating License NPF-68/NPF-81 N PF -68/N PF -81 Appendix B, Section 5.4.1 154 Operating License NPF-68/NPF-81 Appendix B, Section 5.4.2 99 Printed June 19, 2009 at 11 :44
Approved By J.D. Williams Date Approved Page Number 12/22/2008 FEDERAL AND STATE REPORTING REQUIREMENTS 77 of 77 of 77 77
( TABLE 4 (CONT'D)
CROSS REFERENCE INDEX SOURCE REQUIREMENT ADDRESSED BY ITEM(s)
ASME Section XI 135 COE Permit 200500606 70,209 FEA, Act 1974 93-275 161 Generic Letter 91-02 190 Air Quality Permit No. 4911-033-0030 -V-02-0 54,148,149,.151 54,148,149,151 GA Permit No. PG0330017 106, GA Permit No. PG0330035 107, GA Permit No. NG0330036 108, GA Permit No. 017-0003 146, 186, 203 204 203204 GA Permit No. 017-0191-05 48,59, Georgia Rules 391-3-5 49,61,205 NPDES Permit No. GA0026786 45,51, 162, 165, 169 162,165,169 NUREG 1022 132 GEPD-SPCC 26,60 ODCM 2.1.3.2 87 ODCM 2.1.4.2 88 ODCM 3.1.3.2 89 ODCM 3.1.4.2 90 ODCM 3.1.5.2 91 ODCM 5.1.2 92 ODCM 4.1.1.2.2 93 ODCM 7.2 153 ODCM 7.2.2.5 153.2 ODCM 7.2.2.7 153.3 PCP 12.0 153.1 Emergency Plan Section H, 4.3.H 121, 122 121,122 EA-03-009 (2/20104) 171, 172 171,172 Landfill 070 Plan 206,207 Pnnted June 19, 2009 at 11:44
- 44
Answers 001A4.121 001A4.12 1 A 002G2.1.31 1 D 4
S 5
003AK2.0511 003AK2.0S 003K6.141 004K5.11 1 004KS.11
- B B
D 6 004K5.15 1 004KS.lS B 7 005K2.031 00SK2.031 B 8 006A4.11 1 B 9 007EA2.061 B 10 007Kl.Ol 1 B 11 008A4.021 B 12 008AK2.022 D 13 008K3.031 C 14 009EG2.4.34 1 D IS 15 010K2.041 A 16 01OK6.01 1 B 17 01IEG2.2.37 1 *D D
18 012K3.041 B 19 013Kl.071 D 20 015AG2.4.6 0ISAG2.4.6 1 D 21 015K2.01 1 01SK2.01 C 22 022Al.021 C 23 022AAl.021 C 24 025AKl.Ol 02SAKl.Ol 1 D 026K4.01 1 C
( 027AK3.022 B 27 028K5.041 028KS.041 C 28 029EK2.061 B 29 029Kl.051 029Kl.0S 1 D 30 032AA2.092 C 31 033Al.Ol 1 B 32 035K6.01 1 03SK6.01 B 33 039G2.2.25 2 039G2.2.2S A 34 051AAl.042 OS1AAl.042 B 35 3S 055EKl.Ol 1 OSSEKl.Ol A 36 055K3.01 OSSK3.01 1 C 37 056AA2.391 OS6AA2.391 D 38 057AAl.041 OS7AAl.041 A 39 058AA2.01 1 OS8AA2.01 D 40 059A3.041 OS9A3.041 A 41 061Al.Ol 2 D 42 061G2.1.71 06IG2.1.71 C 43 062AK3.03 1 B 44 062K4.021 A 4S 45 063A2.01 1 _C 46 064K6.082 C 065AAl.031 06SAAl.031 D
'-' 068AG2.4.6 2 D Page: lof3 6/19/2009
Answers filii-49 071A2.081 B t/ 073K5.01 1 A
\
)2 53 54 076A2.012 076A2.021 076AA2.032 077AK2.041 B
B D
oA 55 078A3.01 078A3.0111 A 56 07802.4.31 1 078G2.4.31 'D "D
57 086A3.01 1 C 58 103A1.01 103Al.01 1 C 59 103A4.032 "D 60 02.1.29 1 G2.1.29 D 61 02.1.301 G2.1.30 1 C 62 02.1.81 G2.1.8 1 -A
'A 63 02.2.13 G2.2.13 2 C 64 G2.2.4411 02.2.44 B 65 02.3.11 G2.3.11 1 -c
- C 66 02.3.141 G2.3.141 C 67 02.4.13 G2.4.13 1 B 68 02.4.341 G2.4.341 ,~ B 69 02.4.43 1 G2.4.43 ,~ D 70 WE03EA1.3 1 - C 71 WE04EK3.4 1 r C 72 WE05EK1.32 D i WE06EK3.4 3 A
\ WE08EK2.11
- B 75 WE09EK1.21 4.
A 76 022AA2.022
" B 77 027A02.1.321 027AG2.l.321 C 78 029E02.2.22 1 029EG2.2.22 A 79 058AA2.031 A-80 062AA2.021 A 81 WE11E02.4.2 WE11EG2.4.2 1 C 82 003A02.4.35 1 003AG2.4.35 C 83 028AA2.081 D 84 069A02.4.822 069AG2.4.8 <-
A 85 WE03EA2.222 WE03EA2.2 C 86 008A2.071 C 87 01202.4.2011 012G2.4.20 A 88 013A2.041 A 89 05902.4.32 059G2.4.32 D 90 061A2.031 C 91 05502.4.11 055G2.4.11 2 A 92 068A2.021 A 93 071A2.092 A 94 02.1.15 1 G2.1.151 B 02.2.221 G2.2.221 C
" J 02.2.381 G2.2.381 C 2of3 Page: 20f3 6/19/2009
Answers 97 G2.3.4 2 B i('~~
"I( G2.3.5 1 G2.3.51 B G2.4.20 3 C 100 G2.4.472 C l
Page: 30f3 3 of3 6119/2009 6/19/2009
Date: 06-26-2009 Date: 06-26-2009 VEGP NRC VEGP NRC SRO SRO Examination Examination Answer Answer Sheet Sheet Name:
Student Name:
Student
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I.I. 26.
- 26. 5I.
5I. 76.
76.
2.
- 2. 27.
- 27. 52.
- 52. 77.
77.
3.
- 3. 28.
- 28. 53.
- 53. 78.
78.
4.
- 4. 29.
- 29. 54.
- 54. 79.
79.
5.
- 5. 30.
- 30. 55.
- 55. 80.
80.
6.
- 6. 3I.
3I. 56.
- 56. 8I.
8I.
- 7. 32. 57. 82.
- 8. 33. 58. 83.
- 9. 34. 59. 84.
- 10. 35. 60. 85.
II. 36. 6I. 86.
- 12. 37. 62. 87.
- 13. 38. 63. 88.
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- 14. 39. 64. 89.
- 15. 40. 65. 90.
- 16. 4I. 66. 9I.
- 17. 42. 67. 92.
- 18. 43. 68. 93.
93.
19.
- 19. 44.
- 44. 69.
- 69. 94.
94.
20.
- 20. 45.
- 45. 70.
- 70. 95.
95.
2I.
2I. 46.
- 46. 7I.
7I. 96.
96.
22.
- 22. 47.
- 47. 72.
- 72. 97.
97.
23.
- 23. 48.
- 48. 73.
- 73. 98.
98.
24.
- 24. 49.
- 49. 74.
- 74. 99.
99.
25.
- 25. 50.
- 50. 75.
- 75. 100.
100.