ML13295A416
| ML13295A416 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 10/17/2013 |
| From: | Division of Nuclear Materials Safety III |
| To: | Carolina Power & Light Co |
| References | |
| 50-400/13-301, ES-401, ES-401-8 | |
| Download: ML13295A416 (214) | |
Text
ES-401 Site-Specific SRO Written Examination Form ES-401-8 Cover Sheet US. Nuclear Regulatory Commission 2013 HNP NRC Site-Specific SRO Written Examination Applicant Information Name:
Date: Facility/Unit: Harris Nuclear Plant Region: I LI lllll
[] IV [] ReactorType:W CEE]BWE]GEEI Start Time: Finish Time:
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 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.
Alicants Sicinature Results RO/SRO-OnlylTotal Examination Values 75 I 25 / 100 Points Applicants Scores / / Points Applicants Grade I I Percent V
44 1 P Answers
- ID Points Type 0 1 2013NRCRO1 1.00 MCS C 2 2013 NRC R02 1.00 MCS A 3 2O13NRCRO3 1.00 MCS A 4 2013 NRC R04 1.00 MCS A 5 2013 NRC R05 1.00 MCS D 6 2013 NRC R06 1.00 MCS A 7 2013 NRCRO7 1.00 MCS B 8 2013 NRCRO8 1.00 MCS C 9 2013 NRCRO9 1.00 MCS B 10 2O13NRCRO1O 1.00 MCS A 11 2O13NRCROI1 1.00 MCS A 12 2O13NRCRO12 1.00 MCS D 13 2O13NRCRO13 1.00 MCS C 14 2013 NRCRO 14 1.00 MCS C 15 2OI3NRCRO15 1.00 MCS C 16 2O13NRCRO16 1.00 MCS A 17 2O13NRCRO17 1.00 MCS C 18 2O13NRCROI8 1.00 MCS B 19 2O13NRCRO19 1.00 MCS D 20 2013 NRC RO 20 1.00 MCS A 21 2013 NRC RO 21 1.00 MCS C 22 2013 NRC RO 22 1.00 MCS C 23 2013 NRC RO 23 1.00 MCS A 24 2013 NRC RO 24 1.00 MCS B 25 2013 NRC RO 25 1.00 MCS D 26 2013NRCR026 1.00 MCS 27 2013 NRC RO 27 1.00 MCS D 28 2013 NRC RO 28 1.00 MCS C 29 2013 NRC RO 29 1.00 MCS A 30 2O13NRCRO3O 1.00 MCS B 31 2013 NRC RO 31 1.00 MCS C 32 2013 NRC RO 32 1.00 MCS B 33 2013 NRC RO 33 1.00 MCS C 34 2013 NRC RO 34 1.00 MCS B 35 2013NRCR035 1.00 MCS A 36 2013 NRC RO 36 1.00 MCS D 37 2013NRCR037 1.00 MCS A 38 2013NRCR038 1.00 MCS C 39 2013 NRC RO 39 1.00 MCS B 40 2013 NRC RO 40 1.00 MCS C 41 2013NRCR041 1.00 MCS C 42 2013 NRC RO 42 1.00 MCS C 43 2013 NRC RO 43 1.00 MCS A 44 2013 NRC RO 44 1.00 MCS B 45 2013 NRC RO 45 1.00 MCS B 46 2013 NRC RO 46 1.00 MCS C 47 2013 NRC RO 47 1.00 MCS D 48 2013 NRC RO 48 1.00 MCS C
e/K P2oP3 Answers
- ID Points Type 0 49 2013 NRC RU 49 1.00 MCS B 50 2O13NRCRU5O 1.00 MCS B 51 2013 NRC RU 51 1.00 MCS C 52 2013 NRC RU 52 1.00 MCS A 53 2013 NRC RU 53 1.00 MCS A 54 2013 NRC RU 54 1.00 MCS B 55 2OI3NRCRU55 1.00 MCS A 56 2OI3NRCRU56 1.00 MCS A 57 2O13NRCRU57 1.00 MCS B 58 2O13NRCRU58 1.00 MCS B 59 2013 NRC RU 59 1.00 MCS C 60 2013 NRC RU 60 1.00 MCS A 61 2013 NRCRO 61 1.00 MCS B 62 2013 NRCRU 62 1.00 MCS A 63 2013 NRCRU 63 1.00 MCS A 64 2013 NRCRU 64 1.00 MCS D 65 2013 NRC RU 65 1.00 MCS B 66 2013 NRC RU 66 1.00 MCS D 67 2013 NRC RU 67 1.00 MCS C 68 2013 NRC RU 68 1.00 MCS B 69 2O13NRCRU69 1.00 MCS B/f1 70 2013 NRCRU 70 1.00 MCS C 71 2013 NRC RU 71 1.00 MCS B 72 2013 NRC RU 72 1.00 MCS C 73 2013 NRC RU 73 1.00 MCS D 74 2013 NRC RU 74 1.00 MCS B 75 2013 NRC RU 75 1.00 MCS D 76 2013 NRC SRU 1 1.00 MCS C 77 2013 NRC SRU2 1.00 MCS A 78 2013 NRC SRU3 1.00 MCS A 79 2013 NRC SRU4 1.00 MCS C 80 2013 NRC SRU5 1.00 MCS C 81 2013 NRC SRU6 1.00 MCS A 82 2013 NRC SRU7 1.00 MCS B 83 2013 NRC SRU 8 1.00 MCS C 84 2013 NRC SRU9 1.00 MCS A 85 2O13NRCSRU1O 1.00 MCS 86 2O13NRCSRUI1 1.00 MCS D 87 2O13NRCSRU12 1.00 MCS B 88 2O13NRCSRU13 1.00 MCS A 89 2013 NRC SRU 14 1.00 MCS A 90 2013 NRC SRU 15 1.00 MCS B 91 2O13NRCSRU16 1.00 MCS D 92 2O13NRCSRU17 1.00 MCS D 93 2013 NRC SRU 18 1.00 MCS A 94 2013 NRC SRU 19 1.00 MCS C 95 2013 NRC SRU 20 1.00 MCS A 96 2013 NRC SRU 21 1.00 MCS B
O Answers
- ID Points Type 0 97 2O13NRCSRO22 1.00 MCS A 98 2013 NRC SRO 23 1.00 MCS D 99 2013 NRC SRO 24 1.00 MCS A 100 2O13NRCSRO25 1.00 MCS C SECTION 1 ( 100 items) 100.00
- 1. Given the following plant conditions:
- A Reactor Trip occurs due to lowering RCS Pressure
- A Reactor Trip breaker is OPEN
- B Reactor Trip breaker is CLOSED
- The crew is implementing E-0, Reactor Trip or Safety Injection to stabilize the plant when RCS pressure reaches the low RCS pressure safety injection setpoint Which ONE of the following completes the statements below?
When directed to reset safety injection in E-0, the operator must wait a MINIMUM of (1) seconds after the SI signal actuation.
Based on the current conditions, safety injection reset AND automatic block can be performed on (2)
A. (1)150 (2) A Train ONLY B. (1)150 (2) A AND B Train C. (1)60 (2) A Train ONLY D. (1)60 (2) A AND B Train Thursday, September 05, 2013 7:36:57 PM 1 Rev. FINAL
- 2. Given the following plant conditions:
- A LOCA occurs through a stuck open PZR Safety Valve
- The crew transitions to ES-i .2, Post LOCA Cooldown and Depressurization WHICH ONE of the following completes BOTH of the statements below?
In accordance with ES-I .2, Pressurizer heaters (1)
The basis for this restriction on heater operation is that (2)
A. (I) are NOT allowed to be energized until a TSC evaluation is provided (2) PZR level instruments may have measurement errors B. (I) are NOT allowed to be energized until a TSC evaluation is provided (2) heater elements may have been previously damaged C. (I) CAN be energized without a TSC evaluation if PZR level is at least 25%
(2) PZR level instruments may have measurement errors D. (I) CAN be energized without a TSC evaluation if PZR level is at least 25%
(2) heater elements may have been previously damaged Thursday, September 05, 2013 7:36:57 PM 2 Rev. FINAL
- 3. Given the following plant conditions:
- All RCPs are running
- RCS pressure is 920 psig and slowly LOWERING
- Slflowis 100 GPM
- Containment pressure is 3.2 psig and slowly RISING
- SG pressures are 1120 psig Which ONE of the following completes the statements below, in accordance with E-1, Loss of Reactor Or Secondary Coolant?
RCPs (1) be tripped.
(2) is the MINIMUM pressure above which the crew will transition to ES-1.2, Post LOCA Cooldown and Depressurization, where the SGs will be required for RCS cooldown.
A. (1) must NOT (2) 230 psig B. (1) must NOT (2) 360 psig C. (1) must (2) 230 psig D. (1) must (2) 360 psig Thursday, September 05, 2013 7:36:57 PM 3 Rev. FINAL
- 4. Given the following plant conditions:
- The crew is implementing E-1, Loss Of Reactor Or Secondary Coolant
- Intermediate range flux is 3x10 -11 amps and lowering
- Containment pressure is 26.5 psig and lowering
- RCS pressure is 675 psig and lowering
- SG pressure is 950 psig and lowering
- SI flow is 630 gpm Which ONE of the following predicts the status of the Source Range Detectors and identifies the required RHR pump alignment in accordance with E-1?
A. are energized; Leave RHR Pumps running B. are energized; Stop RHR Pumps C. are de-energized; Stop RHR Pumps D. are de-energized; Leave RHR pumps running Thursday, September 05, 2013 7:36:57 PM 4 Rev. FINAL
- 5. Given the following conditions:
- The Reactor is at 45% power
- RCP B trips
- ALB-010, 6-3A, RCS Loop A Tavg Hi/Lo Dev, is in alarm Given the above conditions, which of the following completes the statements below?
SG B Level will initially (1)
In accordance with APP-ALB-01 0 the crew will (2)
A. (1)rise (2) trip the Reactor and Go to E-0, Reactor Trip or Safety Injection B. (1) rise (2) commence a Reactor shutdown using GP-006, Normal Plant Shutdown from Power Operation to Hot Standby C. (1)lower (2) trip the Reactor and Go to E-0, Reactor Trip or Safety Injection D. (1)lower (2) Commence a Reactor shutdown using GP-006, Normal Plant Shutdown from Power Operation to Hot Standby Thursday, September 05, 2013 7:36:57 PM 5 Rev. FINAL
- 6. Given the following plant conditions:
- The unit is in Mode 6
- Auto makeup to the VCT is unavailable
- VCT level is currently 19% and slowly lowering Which ONE of the following is required in accordance with AOP-003, Malfunction of Reactor Makeup Control, Attachment 5, Manual Makeup in Modes 5 & 6?
A. From the MCB: Open ICS-291 & 292, CSIP Suctions From RWST AND close 1CS-165 & 166 VCT Outlet valves B. Locally: Open I CS-278, Emergency Boric Acid Addition AND I CS-274, Manual Blend From RMWST lsol valve C. From the MCB: Start one Boric Acid pump, open I CS-283 (FK-1 13 Borc Acid Flow),
ICS-156 (FCV-1 13B, Makeup to CSIP Suction) and ICS-151 (FCV-1 14, RWMU To Boric Acid Blender)
D. Locally: Open 1CS-287, Alt Emergency Boration Manual Isol AND ICS-274, Manual Blend From RMWST Isol valve Thursday, September 05, 2013 7:36:57 PM 6 Rev. FINAL
- 7. Given the following plant conditions:
- The unit is operating at 100% power
- CCW Surge Tank level is 50% and lowering Which ONE of the following completes both statements below?
The FIRST level at which annunciator ALB-005, 6-1, CCW Surge Tank High-Low Level, will alarm while level lowers is (1)
In accordance with AOP-014, Loss Of Component Cooling Water, an action required for this condition is (2)
A. (1) 38%
(2) SHUT 1CC-299, RCP Bearing Oil Coolers Return.
B. (1) 40%
(2) SHUT ICC-299, RCP Bearing Oil Coolers Return.
C. (1) 38%
(2) SHUT ICC-252, RCP Thermal Barriers Flow Control.
D. (1) 40%
(2) SHUT ICC-252, RCP Thermal Barriers Flow Control.
Thursday, September 05, 2013 7:36:57 PM 7 Rev. FINAL
- 8. Given the following plant conditions:
- The unit is operating at 100% power BOL conditions
- Steam Dumps are in the Tavg mode
- A Turbine trip occurs
- The Reactor does NOT trip Which ONE of the following completes both statements?
(Assuming NO operator actions)
Reactor Delta T indications Tl-412A, 422A, and 432A, RCS Loop Prot Delta Ts will (1)
SG Safety valves will (2)
A. (1) rise (2) lift B. (1) rise (2) not lift C. (1)lower (2) lift D. (1) lower (2) not lift Thursday, September 05, 2013 7:36:57 PM 8 Rev. FINAL
- 9. Given the following plant conditions:
- The unit is in Mode 3
- GP-007, Normal Plant Cooldown Mode 3 to Mode 5, is in progress
- PRZ LO PRESS TRAIN A and B SI BLOCKED status lights are illuminated
- STM LINE ISOL TRAIN A and B SI BLOCKED status lights are illuminated
- RCSTavgi5485°F
- RCS pressure is 1875 psig
- All SG pressures are 625 psig A fault on the A SG occurs inside Containment and the following conditions exist:
- Containment is 2.6 psig and rising
- A SG pressure has lowered to 450 psig in the last 30 seconds Which ONE of the following identifies (1) the ESFAS signal(s) that has (have) automatically initiated AND (2) the reason for the initiation?
A. (1) MSL Isolation ONLY (2) A SG pressure has lowered below the low pressure actuation setpoint.
B. (1) MSL Isolation ONLY (2) A SG pressure has exceeded the rate actuation setpoint.
C. (1) MSL Isolation AND MFW Isolation (2) A SG pressure has lowered below the low pressure actuation setpoint.
D. (1) MSL Isolation AND MFW Isolation (2) A SG pressure has exceeded the rate actuation setpoint.
Thursday, September 05, 2013 7:36:57 PM 9 Rev. FINAL
- 10. Given the following plant conditions:
- The unit is operating at 100% power Which ONE of the following predicts the Main EW Pump response, if any, to an inadvertant actuation of Train B Safety Injection?
A. Both Main FW pumps immediately trip B. No Main FW pump trip is initially generated; Both MEW pumps will trip when Tavg lowers to < 564°E C. ONLY B Main EW pump will trip D. B Main EW pump will trip, A Main EW pump continues to run until Tavg lowers to <564°E Thursday, September 05, 2013 7:36:57 PM 10 Rev. FINAL
- 11. Given the following plant conditions:
- The unit is operating at 100% power when the following annunciators are reported to the CRS:
- ALB-022-1-2, Start Up XFMR-A Both 230KV Bkrs Open ALB-022-9-2, Start Up XFMR-B Both 230KV Bkrs Open
- ALB-018-1-3, Turbine Trip Reactor Trip P4
- ALB-025-3-3, Diesel Generator B Start Failure
- ALB-002-2-4A, Condsr Pre Trip Low Vacuum
- The crew is implementing ES-0.1, Reactor Trip Response Based on the above conditions, (1) which AOP is required to mitigate the current conditions AND (2) what is the status of FW isolation valves?
1 FW-1 59, Main FW A Isolation 1FW-277, Main FW B Isolation 1FW-217, Main FW C Isolation A. (1) AOP-025, Loss of One Emergency AC Bus (6.9KV) or One Emergency DC Bus (125V)
(2) OPEN B. (1) AOP-025, Loss of One Emergency AC Bus (6.9KV) or One Emergency DC Bus (125V)
(2) CLOSED C. (1) AOP-039, Startup And Auxiliary Transformer Trouble (2) OPEN D. (1) AOP-039, Startup And Auxiliary Transformer Trouble (2) CLOSED Thursday, September 05, 2013 7:36:57 PM 11 Rev. FINAL
- 12. Given the following plant conditions:
- The unit is operating at 100% power
- ALB-015, 4-5, Channel Ill UPS Trouble has just alarmed
- Feed flows to all SGs have not changed
- The S-Ill inverter static switch has shifted to the bypass alignment Which ONE of the following completes both statements below in accordance with ALB-015, 4-5?
The 7.5 KVA Instrument Bus Ill INVERTER (1)
Instrument Bus Ill is currently powered from (2)
A. (1) has lost DC power ONLY (2) the 7.5KVA Instrument Bus III Inverter B. (1) has lost DC power ONLY (2) 1A21 C. (1) has lost AC and DC power (2) 1D21 D. (1) has lost AC and DC power (2) 1A21 Thursday, September 05, 2013 7:36:57 PM 12 Rev. FINAL
- 13. Given the following plant conditions:
- The plant is operating at 100% power
- B Train Safety Equipment is in service
- Both ESW Pumps are running to support surveillance testing The following indications and annunciators are observed:
- ALB-02-4-5, SERV WTR LEAKAGE
- ALB-02-5-5, SERV WTR HEADER A HIGH/LOW FLOW
- ALB-02-6-1, SERV WTR SUPPLY HEADER A LOW PRESS
- CNMT Sump level is increasing on ERFIS The crew enters AOP-022, Loss of Service Water and secures the A ESW Pump.
Which ONE of the following actions in accordance with AOP-022, identifies (1) the possible location of the rupture AND (2) the action required by the procedure?
A. (1) CNMT Fan Coil Units (2) Shut 1SW-231, NNS CNMT Fan CLRS Inlet Isol, AND ISW-242, NNS CNMT Fan CLRS Outlet lsol B. (1) CNMT Fan Coil Units (2) Shut 1SW-231, NNS CNMT Fan CLRS Inlet lsol, AND 1SW-276, Headers A&B Return to Normal Service Water C. (1) CNMT Fan Coolers (2) Shut ONLY AH-2/3 ESW Supply and Return Valves D. (1) CNMT Fan Coolers (2) Shut AH-1/2/3/4 ESW Supply and Return Valves Thursday, September 05, 2013 7:36:57 PM 13 Rev. FINAL
- 14. Given the following plant conditions:
- The crew is currently implementing E-3, Steam Generator Tube Rupture
- The OAC reports Train A Phase A valves will not open after resetting Phase A Based on the above conditions, which ONE of the following completes the statements below?
The required RCS depressurization will be accomplished using (1)
The E-3 RCS depressurization termination criteria, when using the PZR Spray Valves, is (2) the termination criteria when the PZR PORVs are used to depressurize the RCS.
A. (1) PZR Spray Valves (2) different than B. (1) PZR Spray Valves (2) exactly the same as C. (1)PZRPORVs (2) different than D. (1) PZR PORVs (2) exactly the same as Thursday, September 05, 2013 7:36:57 PM 14 Rev. FINAL
- 15. Given the following plant conditions:
- The unit is operating at 100% power
- Grid frequency is begining to lower Which ONE of the following completes the following statements in accordance with AOP-028, Grid Instability?
The highest frequency, below which entry into AOP-028 will be required is (1)
The highest frequency at which an automatic Reactor trip, as well as a trip of all RCPs, will occur is (2)
A. (1) 59.0 Hz (2) 57.5 Hz B. (1) 59.0 Hz (2) 58.4 Hz C. (1) 59.5 Hz (2) 57.5 Hz D. (1) 59.5 Hz (2) 58.4 Hz Thursday, September 05, 2013 7:36:57 PM 15 Rev. FINAL
- 16. Given the following plant conditions:
- The unit was operating at 100% power
- A LOCA has occurred in the RAB and the crew is implementing ECA-1 .2, LOCA Outside Containment, step 6 check break isolated Which ONE of the following identifies (1) a parameter trend, which is used to confirm that the break is isolated, AND (2) the reason for the trend?
A. (1) RCS pressure rising (2) SI flow is filling up the RCS B. (1) RCS pressure rising (2) Main Steam Lines are isolated C. (1) PZR level rising (2) SI flow is filling up the RCS D. (1) PZR level rising (2) Main Steam Lines are isolated Thursday, September 05, 2013 7:36:57 PM 16 Rev. FINAL
- 17. Given the following plant conditions:
- Bleed & Feed is in progress in accordance with FR-H.1, Response to Loss of Secondary Heat Sink
- Main Feedwater is now available
- No AFW Pumps are available
- Core Exit Thermocouple temperatures are stable
- All SG wide range levels are 10%
Which ONE of the following completes the statements below in accordance with FR-H.1, Attachment 1, Guidance on Restoration of Feed Flow?
Feed one intact SG at no more than (1)
Feed flow may be raised to maximum rate as soon as SG Wide Range level rises to greater than (2)
A. (1)5OKPPH (2) 15%
B. (1) 50 KPPH (2) 25%
C. (1) the lowest controllable rate (2) 15%
D. (1) the lowest controllable rate (2) 25%
Thursday, September 05, 2013 7:36:57 PM 17 Rev. FINAL
- 18. Given the following plant conditions:
- A LOCA has occurred
- Containment pressure is 15 psig and LOWERING
- Due to a failure of A and B train, CNMT Sump to RHR Pump suction valves, the crew has transitioned from E-1, Loss of Reactor Or Secondary Coolant to ECA-1 .1, Loss Of Emergency Coolant Recirculation
- Two CSIPs, two Containment Fan Coolers, both CT pumps and both RHR pumps are running
- RWST level is approximately 30% and lowering
- Wide Range Containment Sump level is 140 inches Which ONE of the following identifies (1) the reason why A CT pump is required to be secured AND (2) another required action if RWST level lowers to 3% while the crew continues with ECA-1.1?
(Reference provided)
A. (1) Preserve RWST inventory (2) secure the other Containment Spray pump when Containment pressure is less than 10 psig B. (1) Preserve RWST inventory (2) establish makeup to the RCS from an alternate source C. (1) Preclude unnecessary entry into FR-Z.2, Reponse To Containment Flooding (2) secure the other Containment Spray pump when Containment pressure is less than 10 psig D. (1) Preclude unnecessary entry into FR-Z.2, Reponse To Containment Flooding (2) establish makeup to the RCS from an alternate source Thursday, September 05, 2013 7:36:57 PM 18 Rev. FINAL
- 19. Given the following plant conditions:
- A Reactor startup is in progress
- The OAC withdraws CBD from 20 steps to the next doubling in accordance with GP-004, Reactor Startup (Mode 3 To Mode 2)
- The OAC releases the Rod Motion switch, but CBD rods continue to withdraw
- The MCB Rx Trip Switch #1 is taken to Trip
- The Reactor Trip Breaker indications change as indicated in the pictures below (NOTE: the light bulbs are not blown)
Before Rx Trip Switch # I taken to Trip After Rx Trip Switch # I taken to Trip z
0 0 j;j t,u c._
t Which ONE of the following completes the statement below?
-T( u/I The current status of the Reactor is (1) AND the indicationpfthe Reactor Trip Breakeçs on the MCB indicates a failure of the (2) Trip coil. A A. (1)tripped (2) UV B. (1)tripped (2) Shunt C. (1) NOT tripped (2) UV D. (1) NOT tripped (2) Shunt Thursday, September 05, 2013 7:36:57 PM 19 Rev. FINAL
- 20. A traverse drive system (roller chain) failure has occurred on the fuel transfer system conveyor while the cart was in the horizontal position and loaded with a fuel bundle inside Containment.
Which ONE of the following identifies (1) the back-up method of returning the fuel transfer cart to the Fuel Handling Building (FHB) in accordance with FHP-020, Refueling Operations AND (2) where the equipment is operated?
A. (1) Emergency pull-out cable (2) Inside the Fuel Handling Building B. (1) Emergency pull-out cable (2) Inside the Containment Building C. (1) Auxiliary Crane Traverse (2) Inside the Fuel Handling Building D. (1) Auxiliary Crane Traverse (2) Inside the Containment Building Thursday, September 05, 2013 7:36:57 PM 20 Rev. FINAL
- 21. Given the following plant conditions:
- The plant is operating at 100%
- One SG has developed a tube leak and the crew is implementing AOP-016, Excessive Primary Plant Leakage
- Chemistry has been directed to perform CRC-804, Primary to Secondary Leak Rate Monitoring, to quantify the leak rate Which ONE of the following instrument(s) is/are used to determine the primary to secondary leak rate in accordance with AOP-016?
A. SG Blowdown Radiation Monitor, REM-Ol BD-3527 B. Turbine Building Vent Stack Effluent Monitor, RM-1TV-3536-1 C. Condenser Vacuum Pump Effluent Monitor, REM-O1TV-3534 D. Main Steam Line Radiation Monitors RM-01 MS-3591 SB, 3592 SB, or 3593 SB Thursday, September 05, 2013 7:36:57 PM 21 Rev. FINAL
- 22. Given the following plant conditions:
- The unit was operating at 60% power when air leakage into the Condenser resulted in entry in AOP-01 2, Partial Loss of Condenser Vacuum
-A load reduction was initiated in accordance with AOP-038, Rapid Downpower Time Power Control Bank C Control Bank D 0800 60% 225 steps 130 steps 0830 50% 223 steps 95 steps 0900 45% 213 steps 85 steps 0930 40% 198 steps 70 steps 1000 35% 178 steps 50 steps Which ONE of the following identifies the EARLIEST time that the LCO for Technical Specification 3.1.3.6, Control Rod Insertion Limits was not met?
(Reference provided)
A. 0830 B. 0900 C. 0930 D. 1000 Thursday, September 05, 2013 7:36:57 PM 22 Rev. FINAL
- 23. Given the following plant conditions:
- An RWST leak has occurred
- REM-Ol MD-3530, Tank Area Drain Transfer Pumps Monitor, is in HIGH alarm
- Contaminated water is filling the retention dike area Which ONE of the following completes BOTH statements below?
As a result of this radiation alarm, (1) automatically.
In accordance with AOP-008, Accidental Release of Liquid Waste, a leak from the Refueling Water Storage Tank requires manual operation to (2)
A. (1) the Tank Area Drain Transfer Pump stops (2) shut 1 FD-1 09, FD Tank Area Drain Pump 1X Discharge to Storm Drain Valve B. (1) the Tank Area Floor Drain Sump Pump stops (2) shut I FD-1 09, FD Tank Area Drain Pump IX Discharge to Storm Drain Valve C. (1) 1 FD-1 09, FD Tank Area Drain Pump 1X Discharge to Storm Drain Valve shuts (2) secure the Tank Area drain pump D. (1) IFD-109, FD Tank Area Drain Pump IX Discharge to Storm Drain Valve shuts (2) secure the Tank Area Floor Drain Sump pump Thursday, September 05, 2013 7:36:57 PM 23 Rev. FINAL
- 24. Given the following plant conditions:
- The unit was operating at 100% power
- An 86 Lockout occurs on the A and B SUTs
- Sixty minutes later, the following plant conditions exist:
- RVLIS Full Range 63% and lowering
- Core Exit Thermocouples 745°F and rising
- Containment Pressure 3.5 psig and rising
- Pressurizer Level 0%
- SG NR level A 38%
- SG NR level B 44%
- SG NR level C 23%
Based on these conditions, which ONE of the following completes the statement below?
The Core Cooling Critical Safety Function Status Tree requires entry into (1)
AND the crew will depressurize the SGs to 130 psig using (2)
A. (1) FR-C.2, Response To Degraded Core Cooling (2) steam dumps B. (1) FR-C.2, Response To Degraded Core Cooling (2) SG PORVs C. (1) FR-C.1, Response To Inadequate Core Cooling (2) steam dumps D. (1) FR-C.1, Response To Inadequate Core Cooling (2) SG PORVs Thursday, September 05, 2013 7:36:57 PM 24 Rev. FINAL
- 25. Given the following plant conditions:
- A LOCA has occurred
- The crew is implementing ES-i .2, Post LOCA Cooldown and Depressurization
- Safety Injection has NOT been terminated Which ONE of the following identifies (1) the parameter used by the operator to determine whether the CLAs are required to be isolated AND (2) the reason the accumulators are isolated under these conditions?
A. (1) RCS Cold Leg Temperature (2) To allow minimum subcooling to be established B. (1) RCS Cold Leg Temperature (2) To prevent gas binding of the S/G U-tubes C. (1) RCS Hot Leg Temperature (2) To allow minimum subcooling to be established D. (1) RCS Hot Leg Temperature (2) To prevent gas binding of the S/G U-tubes Thursday, September 05, 2013 7:36:57 PM 25 Rev. FINAL
- 26. The crew has transitioned to E-1, Loss of Reactor or Secondary Coolant and is presently evaluating the RHR System capable of Cold Leg Recirculation.
The following conditions exist:
- Offsite Power has been lost
- EDG B has tripped
- CNMT Pressure is 17 psig and rising
- CNMT High Range Rad Monitors are in alarm
- CNMT Wide Range Sump Level is reading 211 inches
- RVLIS Full Range Level is reading 60%
- RCS Cold Leg Temperature is reading 265°F
- RCS Wide Range Pressure is reading 225 psig
- Core Exit Thermocouples are reading 740°F
- Containment Spray pump A has tripped Which ONE of the following is the procedure that the crew is required to implement at this time?
A. FR-Z.1, Response to High Containment Pressure B. FR-Z.2, Response to Containment Flooding C. FR-C.2, Response to Degraded Core Cooling D. FR-P.1, Response to Imminent Pressurized Thermal Shock Thursday, September 05, 2013 7:36:57 PM 26 Rev. FINAL
- 27. Which ONE of the following identifies the sources of water, in accordance with the WOG Background Document for FR-Z.2, Response To Containment Flooding, that are the basis for the maximum anticipated containment water level?
A. Condensate Storage Tank, Emergency Service Water, Reactor Coolant System B. Refueling Water Storage Tank, Emergency Service Water, Reactor Coolant System C. Condensate Storage Tank, Emergency Service Water, Refueling Water Storage Tank D. Refueling Water Storage Tank, Reactor Coolant System, Condensate Storage Tank Thursday, September 05, 2013 7:36:57 PM 27 Rev. FINAL
- 28. Given the following plant conditions:
- A plant cooldown is in progress following a planned shutdown in accordance with GP-007, Normal Plant Cooldown Mode 3 To Mode 5 to repair the Reactor Vessel Head
- The following conditions exist for RCP B Time Upper Thrust Bearing Temperature # I Seal Differential Pressure 0800 154°F 253 psig 0805 159°F 237 psig 0810 165°F 223 psig 0815 174°F 209 psig 0820 183°F 198 psig Which ONE of the following completes the statements below?
The (1) is the first RCP parameter outside the normal limit.
In accordance with GP-007 the action required under these conditions is (2)
A. (1) Upper Thrust Bearing Temperature (2) stop RCP B B. (1) Upper Thrust Bearing Temperature (2) open the RCP # 1 Seal Bypass C. (1) # 1 Seal Differential Pressure (2) stop RCP B D. (1) # I Seal Differential Pressure (2) isolate the RCP B Seal Water Return Thursday, September 05, 2013 7:36:57 PM 28 Rev. FINAL
- 29. Given the following plant conditions:
- A Large Break LOCA has occurred
- RWST level indicates 22% and continues to lower
- IRH-1, RCS LoopAto RHR Pump A-SA is CLOSED In accordance with ES-I .3, Transfer to Cold Leg Recirculation, which ONE of the following actions completes the statement below to establish the A CSIP alignment for long term operation?
The operator must FIRST (1) ICS-746 AND then (2) must be OPENED.
1CS-746, CSIP A Alternate Miniflow 1 RH-25 SA, Suction From RHR Heat Exchanger A-SA lSl-340, Safety Injection A train to Cold Leg A. (1) CLOSE (2) 1RH-25 B. (1) CLOSE (2) lSl-340 C. (1) OPEN (2) 1RH-25 D. (1) OPEN (2) ISI-340 Thursday, September 05, 2013 7:36:57 PM 29 Rev. FINAL
- 30. Given the following plant conditions:
- The unit was operating at 100% power
- ALB-007-4-3, VCT High-Low Level is in Alarm
- VCT level transmitter LI-I 15 has failed high
- VCT level transmitter Ll-112 reads 14%
Which ONE of the following completes the statements below?
In accordance with AOP-003, Malfunction Of Reactor Makeup Control, the HIGHEST VCT level below which gas binding of the running CSIP is a concern is (1)
Given these conditions, RWST suction valves AND VCT Outlet valves will (2)
ICS-291, Suction from RWST LCV-li5B 1CS-292, Suction from RWST LCV-115D I CS-i 65, VCT Outlet LCV-l 150 1CS-I66, VCT Outlet LCV-I 1 SE A. (1) 5%
(2) automatically realign B. (1) 5%
(2) require manual realignment C. (1) 10%
(2) automatically realign D. (1) 10%
(2) require manual realignment Thursday, September 05, 2013 7:36:57 PM 30 Rev. FINAL
- 31. Which ONE of the following identifies (1)the MINIMUM Containment wide range sump level required to place the RHR system in Cold Leg Recirculation in accordance with ES-I .3, Transfer To Cold Leg Recirculation AND (2) the basis for this level?
A. (1) 137.5 inches (2) ensures the recirculation sump strainers are completely submerged B. (1) 137.5 inches (2) ensures the recirculation sump pH level is acceptable C. (1) 142 inches (2) ensures the recirculation sump strainers are completely submerged D. (1) 142 inches (2) ensures the recirculation sump pH level is acceptable Thursday, September 05, 2013 7:36:57 PM 31 Rev. FINAL
- 32. Which ONE of the following completes both statements in accordance with OP-i 07, CVCS, Attachment 5, Replacing B CSIP with C CSIP?
To align the C CSIP to i B-SB, a transfer switch located in the RAB, on elevation (1) , must be operated.
First, the B Train Kirk Key Lock Switch must be rotated, then (2) must be closed.
A. (1) 236 just south of the A CSIP room (2) the transfer switch, which is a knife switch, B. (1) 236 just south of the A CSIP room (2) a handle must be placed into the handle casting and the transfer switch C. (1) 286 Switchgear room (2) the transfer switch, which is a knife switch, D. (1) 286 Switchgear room (2) a handle must be placed into the handle casting and the transfer switch Thursday, September 05, 2013 7:36:57 PM 32 Rev. FINAL
- 33. Given the following plant conditions:
- The unit is operating at 100% power
- ALB-009-8-1, Pressurizer Relief Tank High-Low Level Press Or Temp, Alarms
- PRT temperature indicates 105°F
- PRT pressure indicates 8 psig
- PRT level indicates 73%
Which ONE of the following (1) identifies the cause of the alarm AND (2) describes the operator response for this alarm in accordance with the Annunciator Panel Procedure and OP-i 00, Reactor Coolant System?
A. (1) PRT level is high (2) Drain the PRT to the Reactor Coolant Drain Tank B. (1) PRT level is high (2) Drain the PRT to the Waste Hold Tank C. (1) PRT pressure is high (2) Vent the PRT to the Waste Gas Vent Header D. (1) PRT pressure is high (2) Drain the PRT to the Waste Hold Tank Thursday, September 05, 2013 7:36:57 PM 33 Rev. FINAL
- 34. Which ONE of the following completes both statements in accordance with OP-i 00, Reactor Coolant System?
Per the OP-I 00, Precautions and Limitation, the MAXIMUM temperature below which the Pressurizer Relief Tank (PRT) should be maintained is (I)
A rapid cool down of the PRT can be performed by draining the PRT and providing makeup water to the spray header from the (2)
A. (1) 120°F (2) RCDT B. (1) 120°F (2) RMWST C. (1) 150°F (2) RCDT D. (1) 150°F (2) RMWST Thursday, September 05, 2013 7:36:57 PM 34 Rev. FINAL
- 35. Given the following plant conditions:
- The unit is at 100% Reactor power
- A Reactor trip and Safety Injection has occurred
- Phase A fails to actuate Which ONE of the following CCW System loads are isolated from the CCW System?
(Assume NO Operator actions)
A. Primary Sample Panel AND Gross Failed Fuel Detector B. RCDT heat exchanger AND Excess Letdown heat exchanger C. RCDT heat exchanger AND Gross Failed Fuel Detector D. Primary Sample Panel AND Excess Letdown heat exchanger Thursday, September 05, 2013 7:36:57 PM 35 Rev. FINAL
- 36. Given the following plant conditions:
- The unit is operating at 100% power
- PZR Pressure Channel (PT-445) fails high Which ONE of the following completes the statement below describing the response of the PZR Pressure Control System to this failure?
(1) PZR PORV(s) will OPEN AND remain OPEN until the (2) setpoint is reached.
A. (1) ONE (2) Safety Injection B. (1) ONE (2) P-i 1, PZR High Pressure C. (1) TWO (2) Safety Injection D. (1) TWO (2) P-li, PZR High Pressure Thursday, September 05, 2013 7:36:57 PM 36 Rev. FINAL
- 37. Given the following plant conditions:
- The unit is at 100% power
- The PZR pressure master controller, PK-444A, is in AUTOMATIC
- A PZR pressure master controller malfunction causes the setpoint to slowly drift to 61% over 10 minutes Which ONE of the following is the expected plant response to the drifting of the setpoint?
(Assume NO Operator Actions)
A. Both spray valves will open B. The control heaters will be at maximum output C. Pressure will stabilize at 2280 psig D. One PZR PORV will cycle Thursday, September 05, 2013 7:36:57 PM 37 Rev. FINAL
- 38. Given the following plant conditions:
- The unit was operating at 8% power when the following parameters are indicated prior to the Reactor automatically tripping:
- P1-455, RCS Pressure is 2380 psig
- P1-456, RCS Pressure is 2390 psig
- P1-457, RCS Pressure is 2400 psig
- Ll-459, PRZ Level is 92%
- Ll-460, PRZ Level is 90%
- Ll-461, PRZ Level is 93%
Which ONE of the following (1) identifies the condition that caused the automatic Reactor trip AND (2) the associated basis for the automatic trip?
A. (1) PZR High Level (2) provides protection against over pressurizing the RCS.
B. (1) PZR High Level (2) precludes water relief through the Pressurizer safety valves.
C. (1) PZR High Press (2) provides protection against over pressurizing the RCS.
D. (1) PZR High Press (2) precludes water relief through the Pressurizer safety valves.
Thursday, September 05, 2013 7:36:57 PM 38 Rev. FINAL
- 39. Given the following plant conditions:
- The crew is responding to a Large Break LOCA in E-1, Loss Of Reactor Or Secondary Coolant
- Both RHR pumps are running
-The following actions have been taken:
- SI and Phase A have both been reset
- Instrument Air and Nitrogen have been restored to Containment Subsequently, a Loss of Off-site power occurs.
Which ONE of the following completes the statement below?
The sequencers will run in (1) after the Loss of Off-site power AND the RHR pumps (2)
A. (1)ProgramA (2) will automatically start in load block 2 B. (1)ProgramA (2) must be manually started after load block 9 C. (1)ProgramB (2) will automatically start in load block 2 D. (1) Program B (2) must be manually started after load block 9 Thursday, September 05, 2013 7:36:57 PM 39 Rev. FINAL
- 40. Which ONE of the following completes the statement below?
Instrument Buses (1) AND (2) provide power to the ESFAS Slave Relays.
A. (1)Sl (2) SlI B. (1)SII (2) SIll C. (1)S1 (2) Sly D. (1) SIll (2) SIV Thursday, September 05, 2013 7:36:57 PM 40 Rev. FINAL
- 41. Given the following plant conditions:
- The unit was operating at 100% power
- Containment Fan Coolers are in the Normal Cooling mode
- A steam leak into Containment occurs
- Containment pressure is 2.6 psig and rising
- Containment temperature is 135°F and rising Which ONE of the following completes the statement below?
Containment Fan Coolers are running in (1) speed with the post-accident dampers (2)
(Assume NO Operator actions)
A. (1)SLOW (2) SHUT B. (1)SLOW (2) OPEN C. (1)HIGH (2) SHUT D. (1)HIGH (2) OPEN Thursday, September 05, 2013 7:36:57 PM 41 Rev. FINAL
- 42. Which ONE of the following completes the statement below?
Following a Containment spray actuation signal, the HIGHEST Containment spray additive tank level at which Containment spray chemical addition valves 1 CT-I 1 and ICT-12 will auto-close is A. 23.4%
B. 10%
C.2%
D. 0%
Thursday, September 05, 2013 7:36:57 PM 42 Rev. FINAL
- 43. Given the following plant conditions
- The unit was operating at 100% power
- A LOCA has occurred and the crew is implementing E-1, Loss Of Reactor Or Secondary Coolant
- The CT Pump A tripped while aligned to the RWST When RWST level reaches the Low-Low level setpoint, which ONE of the following identifies (1)the recirc sump suction valve(s) will automatically open AND (2) after the recirc suction valve(s) reach(es) the full-open position, RWST suction valve(s) which will automatically close?
ICT-105, Containment Sump To CNMT Spray Pump A-SA ICT-102, Containment Sump To CNMT Spray Pump B-SB 1CT-26, RWST To CNMT Spray Pump A-SA ICT-71, RWST To CNMT Spray Pump B-SB A. (1)ICT-IO2ONLY (2) 1CT-71 ONLY B. (1) ICT-102 AND 1CT-105 (2) ICT-71 ONLY C. (1)ICT-IO2ONLY (2) 1CT-26 AND ICT-71 D. (1) 1CT-102 AND 1CT-105 (2) ICT-26 AND 1CT-71 Thursday, September 05, 2013 7:36:57 PM 43 Rev. FINAL
- 44. Given the following plant conditions:
- The unit is in MODE 2 at 1% power
- Tavg is at the NO load reference value
- A failure of an SG PORV results in the following:
- Steam Generator pressures at 1028 psig Which ONE of the following completes BOTH statements below?
Operation of the Condenser Steam Dumps is (1) at this time.
In accordance with GP-004, Reactor Startup (Mode 3 to Mode 2) the operator has 15 minutes t restore temperature to above a MINIMUM of (2)
(Assume NO operator action)
A. (1)blocked (2) 553°F B. (1) blocked (2) 551°F C. (1)NOTblocked (2) 553°F D. (1) NOT blocked (2) 551°F Thursday, September 05, 2013 7:36:57 PM 44 Rev. FINAL
- 45. Given the following plant conditions:
- The unit is operating at 91% power
- A Loss of Main Feedwater Pump B occurs
- The crew enters AOP-010, Feedwater Malfunctions Which ONE of the following describes (1)the plant response AND (2)the action required in accordance with AOP-010?
A. (1) Automatic turbine runback is initiated (2) Isolate Steam Generator Blowdown B. (1) Automatic turbine runback is initiated (2) Trip the Reactor and go to E-0 C. (1) Automatic turbine runback is NOT initiated (2) Isolate Steam Generator Blowdown D. (1) Automatic turbine runback is NOT initiated (2) Trip the Reactor and go to E-0 Thursday, September 05, 2013 7:36:58 PM 45 Rev. FINAL
- 46. Given the following plant conditions:
- The unit was operating at 100% Reactor power when a station black out occurs
- The crew is implementing ECA-0.0, Loss of All AC Power
- The TDAFW has been running in automatic with the controller setpoint at 31% for several minutes
- NO operator actions have been taken on the AFW system
- All SG NR levels are approximately 9% and lowering
- AFW flow is currently 160 kpph Which ONE of the following identifies the action(s) required to be taken for these conditions?
A. Transition to FR-H.1, Response to Loss of Secondary Heat Sink.
B. Open 1 SG PORV (on the SG with the highest level) to lower SG pressure.
C. Place Aux FW Turbine SPD PDK-2180.1 in MAN and depress the output RAISE pushbutton.
D. Depress the RAISE pushbutton(s) on the TDAFW FCV(s).
Thursday, September 05, 2013 7:36:58 PM 46 Rev. FINAL
- 47. Given the following plant conditions:
- The unit is operating at 100% power
- Annunciator ALB-014, 7-4, SG A, B, C Backleakage High Temp, has alarmed
- An NLO has been dispatched to verify local temperatures Which ONE of the following completes BOTH of the statement below?
The reason this condition occurred is because a (1) is leaking.
In accordance with the AOP-010, under these conditions with the TDAFW piping local temperature>212°F, the FIRST action required is (2)
A. (1) TDAFW pump steam supply piping check valve (2) start the TDAFW pump to flush the line through the exhaust B. (1) TDAFW pump steam supply piping check valve (2) isolate the TDAFW pump discharge header C. (1) AFW feed water piping check valve (2) start the TDAFW pump to flush the line to the SGs D. (1) AFW feed water piping check valve (2) isolate the TDAFW pump discharge header Thursday, September 05, 2013 7:36:58 PM 47 Rev. FINAL
- 48. Given the following plant conditions:
- The unit is operating at 100% power
- Aux Bus I E deenergizes and is locked out Which ONE of the following describes an effect on the unit?
A. RCP C is deenergized B. CSIP A is momentarily deenergized C. CSIP B is momentarily deenergized D. CTMU Pump lx is deenergized Thursday, September 05, 2013 7:36:58 PM 48 Rev. FINAL
- 49. Which ONE of the following completes the statements below in accordance with OP-156.01, DC Electrical Distribution, Section 8.2. Rotation of 125 VDC NNS Battery Chargers?
When placing a 125VDC battery charger in service, its (1) breaker is closed first.
A Low DC Volt alarm (2) expected after this first breaker is closed.
A. (1) DC output (2) is NOT B. (1) DC output (2) is C. (1) AC input (2) is NOT D. (1) AC input (2) is Thursday, September 05, 2013 7:36:58 PM 49 Rev. FINAL
- 50. Given the following plant conditions:
- The unit is currently in MODE 3
- DP-1A-SA has lost power Which ONE of the following completes the statement below for the A MDAFW Pump?
Breaker control from the MOB (1) AND the control switch indication on the MOB will (2)
A. (1) remains available (2) extinguish B. (1) is not available (2) extinguish
- 0. (1) remains available (2) remain illuminated D. (1) is not available (2) remain illuminated Thursday, September 05, 2013 7:36:58 PM 50 Rev. FINAL
- 51. Given the following EDG Fuel Oil Data:
- Both Fuel Oil Day Tanks Specific gravity: 0.835
- Fuel Oil Day Tank A: 47%
- Fuel Oil Storage Tank A: 90,000 gallons
- Fuel Oil Day Tank B: 42%
- Fuel Oil Storage Tank B: 110,000 gallons Which ONE of the following identifies the status of the EDGs in accordance with Technical Specification 3.8.1 .1, Electrical Power Systems AC Sources?
(Reference provided)
EDGA EDGB A. OPERABLE OPERABLE B. OPERABLE INOPERABLE C. INOPERABLE OPERABLE D. INOPERABLE INOPERABLE Thursday, September 05, 2013 7:36:58 PM 51 Rev. FINAL
- 52. Which ONE of the following identifies an RAB radiation monitor that requires entry into AOP-032, High RCS Activity, when a valid HIGH alarm condition exists?
A. RM-1 RR-3600, Recycle Evaporator Valve Gallery B. RM-2ICR-3578A, Recycle Monitor Tank 1A & 2A C. RM-IRR-3605A, Sample Room 1A Elev. 236 D. RM-1 RR-361 1, Recycle Holdup Tank Area Thursday, September 05, 2013 7:36:58 PM 52 Rev. FINAL
- 53. Given the following plant conditions:
- The unit is in Mode 4
- A Train safety equipment is in service
- The B NSW pump is tagged out for maintenance
- A Loss of Off-site power occurs
- The A EDG failed to start Which ONE of the following completes the statement below?
ESW is providing flow to (1) CCW Heat Exchanger(s) with ESW return header flow aligned to the (2)
A. (1) ONLYB (2) Auxiliary Reservoir B. (1) ONLYB (2) Cooling Tower Basin C. (1) AANDB (2) Auxiliary Reservoir D. (1) AANDB (2) Cooling Tower Basin Thursday, September 05, 2013 7:36:58 PM 53 Rev. FINAL
- 54. Given the following plant conditions:
- The unit is operating at 100% power
- An Instrument Air leak is occurring
- Instrument Air pressure is currently 85 psig and stable Which ONE of the following predicts the plant response for the current condition?
A. All FW flow control valves will CLOSE.
B. RCS letdown flowpath valves drift to mid-position.
C. PZR Spray valves drift to mid-position.
D. Gland Steam Seal Spillover Regulator Valve will OPEN.
Thursday, September 05, 2013 7:36:58 PM 54 Rev. FINAL
- 55. Which ONE of tbe following completes the statements below?
There are (1) Primary Shield Cooling Fans.
The Primary Shield Cooling Fans are located in the Containment Building at elevation (2)
A. (1)Two (2) 221T B. (1)Two (2) 236 C. (1) Four (2) 221 D. (1) Four (2) 236 Thursday, September 05, 2013 7:36:58 PM 55 Rev. FINAL
- 56. Given the following plant conditions:
- A Loss of Off-site Power occurs while the unit was operating at 100% power
- EDG A-SA failed to start
- Load Block 9 has been verified complete on EDG B-SB
- RCS pressure is 2180 psig Assuming NO operator action has been taken, which ONE of the following identifies the PZR Heaters group(s) that are currently energized, if any?
A. None B. B only C. C only D. Band C only Thursday, September 05, 2013 7:36:58 PM 56 Rev. FINAL
2013 HNPNRCSRO
- 57. Given the following plant conditions:
- The unit is operating at 8% power
- Intermediate Range (lR) N35 is inoperable
- N35 Level Trip Switch is in BYPASS in accordance with OWP-RP-21, Reactor Protection The following occur:
- At 12:00 N35 Instrument Power fuses blow
- At 12:15 N35 Control Power fuses blow Which ONE of the following identifies (1) the status of the Reactor Trip Breakers AND (2)the reason for the status of the Reactor Trip Breaker?
A. (1)OPENatI2:00 (2) N35 Instrument Power fuses blew B. (1)OPENatI2:15 (2) N35 Control Power fuses blew C. (1) CLOSED at 12:15 (2) N35 is BLOCKED in accordance with GP-005, Power Operations D. (1)CLOSEDat 12:15 (2) N35 is in BYPASS in accordance with OWP-RP-21 Thursday, September 05, 2013 7:36:58 PM 57 Rev. FINAL
- 58. Given the following plant conditions:
- The unit is operating at 100% power
- ALB-015-1-5, 7.5 KVA UPS Trouble, alarms Which ONE of the following identifies (1) the uninterruptible power supply that is potentially affected AND (2) the action taken, if this power supply is lost, per AOP-024, Loss Of Uninterruptible Power Supply?
A. (1) UPP-IB (2) Locally control Steam Dumps B. (1)UPP-IB (2) Locally control Condensate Booster pumps C. (1) UPP-1 (2) Locally control Steam Dumps D. (1)UPP-1 (2) Locally control Condensate Booster pumps Thursday, September 05, 2013 7:36:58 PM 58 Rev. FINAL
- 59. Given the following plant conditions:
- At 0200, the unit was operating at 100% power
- The crew is implementing E-1, Loss of Reactor or Secondary Coolant
- The hydrogen monitoring system .has been aligned
- At 0400 Containment hydrogen concentration was 0.35% and slowly rising
- At 0500 Containment hydrogen concentration was 0.52% and the Hydrogen Recombiner IA was placed in operation in accordance with OP-I 25, Post Accident Hydrogen Systems
- At 1800 Containment hydrogen concentration has increased to 3.14%
Based on these conditions, which ONE of the following actions is required in accordance with OP-I 25?
A. Start the 1 B recombiner ONLY when Containment hydrogen concentration exceeds 3.5%, then operate both recombiners.
B. Start the I B recombiner ONLY when Containment hydrogen concentration exceeds 4%, then operate both recombiners.
C. Start the I B recombiner NOW and operate both recombiners.
D. Do NOT start the 1 B recombiner. Increase the IA recombiner power by 4 KW.
Thursday, September 05, 2013 7:36:58 PM 59 Rev. FINAL
- 60. Given the following plant conditions:
- Refueling is in progress.
- A spent fuel assembly is being moved in the Fuel Handling Building (FHB) when it is damaged.
- Spent Fuel Pool area radiation monitor RM-1 FR-3566A-SA is in HIGH alarm.
- Spent Fuel Pool area radiation monitor RM-1 FR-3567B-SB is in ALERT.
Which ONE of the following completes BOTH of the statements below?
(1) train(s) of Fuel Handling Building Ventilation Emergency Exhaust has(have) received an automatic start signal.
RM-1FR-3566A-SA radiation monitor (2) sound an alarm locally.
A. (1)ONLYA (2) will B. (I) BOTH A and B (2) will C. (1)ONLYA (2) will NOT D. (1) BOTH A and B (2) will NOT Thursday, September 05, 2013 7:36:58 PM 60 Rev. FINAL
- 61. Which ONE of the following completes the statement below concerning the Waste Gas System in accordance with Technical Specification 3.11.2.5, Radioactive Effluents -
Explosive Gas Mixture?
The concentration of oxygen in the GASEOUS RADWASTE TREATMENT SYSTEM downstream of the hydrogen recombiners shall be limited to less than or equal to (1) by volume whenever the hydrogen concentration exceeds (2) by volume.
A. (1)2%
(2) 2%
B. (1)2%
(2)4%
C. (1)4%
(2) 2%
D. (1)4%
(2)4%
Thursday, September 05, 2013 7:36:58 PM 61 Rev. FINAL
- 62. Given the following plant condition:
- The unit is operating at 100% Reactor power
- S-IA, Airborne Radioactivity Removal fan is in AUTO
- Subsequently, CVI rad monitors indicate as follows:
RC-1CR351A-SA RC-1CR.3561C4A Rc1c-356I&-S8 RC-ICR-3561D-SB tMIT ISOLATION SYS CNMT ISOLATIO# SYS CNMT ISOLATION SYS CNMT [5QL4TON SYS
- E-5, Containment Pre-entry Purge Fan failed to trip Which ONE of the following identifies the system response during these conditions?
A. Containment Vacuum Relief dampers (CB-D51 SA and CB-D52 SB) receive a CLOSE signal.
B. Airborne Radioactivity Removal fan S-lA will Auto START.
C. Containment Isolation Phase A isolation valves receive a CLOSE signal.
T D. Containment Pre-entry Purge Makeup fans AH-8 lA/B receive a TRIP signal.
Thursday, September 05, 2013 7:36:58 PM 62 Rev. FINAL
- 63. Given the following plant conditions:
- The unit is operating in Mode 4 preparing to start up
- CWP A is running
- NSW Pump A is running Which ONE of the following completes the statements below?
Opening 1CW-77, Cooling Tower Bypass Valve #1, will (1) the back pressure on the NSW system.
To prevent CTMU Pump run out, the MAXIMUM total flow allowed is (2) gpm.
A. (1) reduce (2) 30,000 B. (1) reduce (2) 22,000 C. (1) NOT affect (2) 30,000 D. (1) NOT affect (2) 22,000 Thursday, September 05, 2013 7:36:58 PM 63 Rev. FINAL
- 64. Given the following plant conditions:
- The plant is operating at 100% power
- Air pressure on P1-9751 .1, Instrument Air Header Pressure is 80 psig Which ONE of the following completes the statement below?
I SA-506, Service Air Header Isol. Valve, is (1) AND ALB-002, 8-1, Instrument Air Low Pressure Annunciator, is (2)
A. (1) OPEN (2) in Alarm B. (1) OPEN (2) NOT in Alarm C. (I) CLOSED (2) in Alarm D. (1) CLOSED (2) NOT in Alarm Thursday, September 05, 2013 7:36:58 PM 64 Rev. FINAL
- 65. Given the following plant conditions:
- The Motor Driven Fire pump has just been stopped per FPT-3001, Motor Driven Main Fire Pump Operability Test Monthly Interval Modes: All
- A fire occurs
- Fire header pressure lowers to 90 psig
- Fire header pressure is now 125 psig and stable Which ONE of the following completes the statements below?
The Motor-Driven Fire Pump is (1)
The Diesel Driven Fire Pump is (2)
A. (1)OFF (2) OFF B. (1) RUNNING (2) OFF C. (1)OFF (2) RUNNING D. (1) RUNNING (2)RUNNING Thursday, September 05, 2013 7:36:58 PM 65 Rev. FINAL
- 66. Which ONE of the following completes the statement below describing the location and control of the Security Master Key in the control room that provides access to plant vital areas?
The Security Master Key is located in a (1)
The keys to this Box/Cabinet are controlled by (2)
A. (1) locked box in the SM desk (2) the SM B. (1) break-glass cabinet (2) the SM C. (1) locked box in the CRS desk (2) theCRS D. (1) break-glass cabinet (2) Security Thursday, September 05, 2013 7:36:58 PM 66 Rev. FINAL
- 67. Which ONE of the following completes BOTH of the statements below in accordance with OPS-NGGC-1 314, Communications?
Standing Instructions (1) contain items of long term significance.
During shift turnover, in accordance OPS-NGGC-1 314, it is REQUIRED that the crew review (2)
A. (1) normally (2) ONLY the NEW standing instructions since the last watch B. (1) normally (2) ALL current standing instructions C. (1) should NOT (2) ONLY the NEW standing instructions since the last watch D. (1) should NOT (2) ALL current standing instructions Thursday, September 05, 2013 7:36:58 PM 67 Rev. FINAL
- 68. Which ERFIS quality code (AND Color) indicates that an in-core thermocouple has failed due to an open circuit?
A. REDU (Red)
B. OPEN (White)
C. LWRN (Yellow)
D. DALM (Green)
Thursday, September 05, 2013 7:36:58 PM 68 Rev. FINAL
- 69. Which ONE of the following identifies an ACCEPTABLE example of a troubleshooting activity in accordance with AP-929, Troubleshooting Guide?
A. Installing gags on valves B. Pulling an annunciator card C. Replacing failed components on circuit boards D. Temporary M&TE Test point /jack connections Thursday, September 05, 2013 7:36:58 PM 69 Rev. FINAL
- 70. Given the following plant conditions:
- The unit is operating at 100% power
- Makeup to the C SI Accumulator has just been completed
- C SI Accumulator parameters are as follows:
Boron Concentration 2419 ppm Pressure 670 psig Level 68%
1SI-248, Accum C Disch Iso Valve OPEN Breaker 1A21-SA-3D, 1SI-248 Accum C Dish OFF Based on the current conditions of the C SI Accumulator, which ONE of the following describes the action required in accordance with Technical Specifications 3.5.1, Emergency Core Cooling System Accumulators?
A. Restore Level to within limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
B. Restore Boron concentration to within limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
C. Restore Pressure to within limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
D. Restore Disch Iso Valve Breaker to ON within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Thursday, September 05, 2013 7:36:58 PM 70 Rev. FINAL
- 71. Which ONE of the following completes the statement below in accordance with OP-I 20.07, Waste Gas Processing?
The MAXIMUM allowed total curie content for Two Gas Decay Tanks cross-tied together is less than curies.
A. 10,000 B. 20,000 C. 86,825 D. 105,000 Thursday, September 05, 2013 7:36:58 PM 71 Rev. FINAL
- 72. Given the following plant conditions:
- A Refueling Outage is in progress
- You have been assigned to hang a clearance in the RCA, have been briefed, and are preparing to sign on to the RWP
- The survey map records the radiation levels as 1750 mRem/hour in the general area Which ONE of the following completes the statements below?
The classification for this area in accordance with HPS-NGGC-0003, Radiological Posting, Labeling and Surveys, would be a (1) High Radiation Area.
In accordance with OPS-NGGC-1 301, Equipment Clearance, independent verification requirements may be waived by the (2) if excessive radiation exposure would result.
A. (1) Very (2) Control Room Supervisor B. (1) Very (2) Radiation Control Supervisor C. (1) Locked (2) Control Room Supervisor D. (1) Locked (2) Radiation Control Supervisor Thursday, September 05, 2013 7:36:58 PM 72 Rev. FINAL
- 73. Given the following plant conditions:
- The Reactor has tripped and Safety Injection has actuated due to a Large Break Loss of Coolant Accident (LOCA).
- The crew is implementing E-1, Loss of Reactor Or Secondary Coolant
- The OAC reports the following for Critical Safety Function Status Trees:
- Containment Orange
- Subcriticality Orange
- Heat Sink Red
- Integrity Red
- All others are Green Which ONE of the following identifies the required procedure transition AND what it is based on?
A. FR-P.1, Response to Imminent Pressurized Thermal Shock, based on a Severe Challenge to the RPV Intergity B. FR-Z.1, Response to High Containment Pressure, based on an Severe Challenge to the Containment C. FR-S.2, Response to Loss of Core Shutdown, based on an Severe Challenge to the Subcriticality D. FR-H.1, Response to Loss of Secondary Heat Sink, based on an Severe Challenge to the Secondary Heat Sink Thursday, September 05, 2013 7:36:58 PM 73 Rev. FINAL
- 74. Given the following plant conditions:
- AOP-036, Safe Shutdown Following a Fire, is being implemented
- MCB level indicators Ll-9010A1 SA & Ll-9010B1 SB, CST Level, are not available Which ONE of the following completes the statement below?
In accordance with AOP-036.02, Fire Area 1-A-BAL-A, 1-A-BAL-G, 1-A-BAL-H, the alternate method of checking CST level greater than 10% is to use A. the local CST level indicator Ll-901 I B. a graph of AFW Pump suction pressure vs CST level C. a graph of Condensate Transfer Pump suction pressure vs CST level D. the annunciator ALB-01 7, 5-5, Condensate Storage Tank Low Minimum Level Thursday, September 05, 2013 7:36:58 PM 74 Rev. FINAL
- 75. Which ONE of the following completes the statements below in accordance with PEP-230, Control Room Operations?
During an event including an Alert or higher all NLO watch stations should report to the (1) promptly after putting work in a safe conditions.
The (2) must be informed when assigning additional duties to people who were already dispatched to perform another duty and have not yet returned from the first duty assignment.
A. (1) Operations Support Center (2) Site Emergency Coordinator B. (1) Operations Support Center (2) Plant Operations Director C. (1) Main Control Room (2) Site Emergency Coordinator D. (1) Main Control Room (2) Plant Operations Director Thursday, September 05, 2013 7:36:58 PM 75 Rev. FINAL
- 76. Given the following plant conditions:
- The unit is in Mode 6
- Refueling Cavity Level is at 23 6
- Both A and B RHR pumps were in operation in Shutdown Cooling mode when B RHR pump trips on overcurrent Which ONE of the following completes the statement below in accordance with Technical Specification 3.9.8, Residual Heat Removal and Coolant Circulation?
The MINIMUM RHR flowrate for the above conditions is (1) gpm AND the basis forthisflowrequirementisto (2)
A. (1) 900 (2) minimize the effect of a boron dilution incident and prevent boron stratification.
B. (1) 900 (2) preclude cavitation during RHR pump operation.
C. (1) 2500 (2) minimize the effect of a boron dilution incident and prevent boron stratification.
D. (1) 2500 (2) preclude cavitation during RHR pump operation.
Thursday, September 05, 2013 7:36:58 PM 76 Rev. FINAL
- 77. Which ONE of the following completes BOTH of the statements below?
The loss of feedwater ATWS is the limiting ATWS event for the (1) fission product barrier.
In accordance with the FR-S.1 Background Document, for the loss of Feedwater ATWS event, the analysis assumes that theturbine is tripped within a MAXIMUM of (2) seconds.
A. (1) Reactor Coolant System (2) 30 B. (1) Reactor Coolant System (2) 60 C. (1) Containment (2) 30 D. (1) Containment (2) 60 Thursday, September 05, 2013 7:36:58 PM 77 Rev. FINAL
- 78. Given the following plant conditions:
- The plant was operating at 100%
- 0600, C SG develops a 15 gpm tube leak and CRS directs a plant shutdown in accordance with AOP-016, Excessive Primary Plant Leakage
- 0610, IMS-45, MS Line C Safety relief valve, opens and cannot be shut
- 0630, C SG tube leakage degrades and a Reactor Trip and Safety Injection are initiated
- 0645, Chemistry confirms an offsite release is in progress Which ONE of the following identifies (1) the FIRST required classification for the conditions above AND (2) the EARLIEST required time the State and Counties must be hotified?
(Reference provided)
A. (1)FUI.1 (2) 0625 B. (1)FUI.1 (2) 0710 C. (1) SU8.1 (2) 0615 D. (1) SU8.1 (2) 0700 Thursday, September 05, 2013 7:36:58 PM 78 Rev. FINAL
- 79. Given the following plant conditions:
- The crew transitioned from E-1, Loss of Reactor or Secondary Coolant to FR-Hi, Loss of Secondary Heat Sink
- RCS Bleed and Feed was NOT initiated
- Core exit TCs are stable
- Containment pressure is 4.5 psig
- A CT Pump running, B CT Pump is under clearance
- Aux Feedwater flow has just been established at 250 KPPH
- SG levels are as follows:
- A 39% Narrow range
- B 24% Narrow range
- C 29% Narrow range Which ONE of the following is (1) required in accordance with FR-H.1 AND (2) the reason?
A. (1) Remain in FR-H.1 (2) Because NONE of the SG Narrow range levels are greater than the minimum required for heat sink.
B. (1) Remain in FR-H.1 (2) Because NOT ALL SG Narrow range levels are greater than the minimum required for heat sink.
C. (1) Transition to E-1, and step in effect (2) Adequate heat sink has been restored.
D. (1) Transition to FR-Z.1, Response to High Containment Pressure (2) Only one CT Pump is running with adverse Containment conditions.
Thursday, September 05, 2013 7:36:58 PM 79 Rev. FINAL
- 80. Given the following plant conditions:
- The unit was operating at 100% power when a loss of Offsite power occurred
- 6.9 KV Emergency Bus 1 B-SB 86 lockout actuates
- EDGA fails to start
- The ASI system is supplying RCP seal injection
- The crew is implementing ECA-0.0, Loss of All AC Power
- ECA-0.0, step 29 to initiate a cooldown to control PZR level using the SG PORVs is in progress Which ONE of the following completes the statements below in accordance with ECA-0.0?
(1) SG PORV(s) can be operated from the MCB.
The RCS cooldown is required to be stopped when (2)
A. (1)Allthree (2) all cold leg temperature are <400°F B. (1)Allthree (2) the RCS pressure is < 700 psig C. (1)ONLYtheC (2) all cold leg temperature are <400°F D. (1) ONLY the C (2) the RCS pressure is < 700 psig Thursday, September 05, 2013 7:36:58 PM 80 Rev. FINAL
- 81. Given the following plant conditions:
- The Unit is operating at 100% power.
- The following PlC-i loads have lost power
- TE-41 3 RCS Hot Leg Temp Loop A
- TE-423 RCS Hot Leg Temp Loop B
- TE-433 RCS Hot Leg Temp Loop C Which ONE of the following completes the statements below?
(consider each statement separately)
Based on the event above, in accordance with OST-1 020, Remote Shutdown Monitoring And Accident Monitoring Instrumentation Channel Check Monthly Interval Modes 1-2-3, the RCS Subcooling Margin Monitor (1)
If at any time the subcooling monitor becomes inoperable, in accordance with Technical Specification, 3.3.3.6, Accident Monitoring Instrumentation, an acceptable backup method of calculating subcooling margin is to calculate it using (2) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
(Reference provided)
A. (1) remains operable (2) the CSFST graph B. (1) remains operable (2) the OSI/PI computer C. (1) is inoperable (2) the CSFST graph D. (1) is inoperable (2) the OSI/PI computer Thursday, September 05, 2013 7:36:58 PM 81 Rev. FINAL
- 82. Given the following plant conditions:
- The unit is operating at 100% power
- At 1200, Sept 13, 2013 a load rejection occurs
- TheQ C ett&4ka4one group of control rods in Bank D failed to move and aemisaligned by a.ppxizu.ate1y 20 steps
- ALB-013-7-1, Rod Control Urgent Alarm, is in alarm
- At 1215, Sept 13, 2013 all rods have been verified above the Rod Insertion limits Which ONE of the following completes the statements below?
In accordance with Technical Specification 3.1.3.1, Movable Control Assemblies -
Group Height, the MOST limiting action required is to place the unit in Hot Standby prior to (Reference provided)
A. 1900, Sept 13, 2013 B. 1800, Sept 13, 2013 C. 0000, Sept 15, 2013 D. 0600, Sept 15, 2013 Thursday, September 05, 2013 7:36:58 PM 82 Rev. FINAL
- 83. Given the following plant conditions:
- The unit is operating at 100% power
- A release of WGDT E is in progress
- ALB-01 0-4-5, Rad Monitor System Trouble, alarms
- The RM-1 1 status display screens are provided as a reference
- The actions for AOP-005, Radiation Monitoring System, have been completed
- The CRS has entered AOP-009, Accidental Release of Waste Gas
- It is desired to continue with the Waste Gas Decay Tank release Which ONE of the following (1) describes the status of REM-3546 PIG (4GG793), AND (2) in accordance with ODCM 3.3.3.11, Radioactive Gaseous Effluent Monitoring Instrumentation, what are the MINIMUM actions required?
(Reference provided)
A. (1) Inoperable equipment failure monitor loss of isokinetic flow is present.
(2) samples, release rate calcs, and the valve line-up are Independently Verified B. (1) Inoperable equipment failure monitor loss of isokinetic flow is present.
(2) samples, release rate calcs, and the valve line-up are Independently Verified AND once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> grab samples are analyzed for radioactivity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> C. (1) Inoperable - operate failure monitor loss of sample flow is present.
(2) samples, release rate calcs, and the valve line-up are Independently Verified D. (1) Inoperable operate failure monitor loss of sample flow is present.
(2) samples, release rate calcs, and the valve line-up are Independently Verified AND once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> grab samples are analyzed for radioactivity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Thursday, September 05, 2013 7:36:58 PM 83 Rev. FINAL
- 84. Given the following plant conditions:
- The unit is operating at 100% power
- One of the MSL becomes inoperable r1nr4.
Which ONE of the following identifies (1) the normal indication for MSL Radiation Monitors AND (2) the pre-planned alternate method of montioring the Main Stean lines in accordance with Technical Specification, 3.3.3.6, Accident Monitoring Instrumentation, and OWP-RM-09, Radiation, Effluent, And Explosive Gas Monitoring?
(Reference provided)
A. (1) 0.35 mRem/hour (2) TB Vent Stack and CVPETS Rad monitor B. (1) 0.35 mRem/hour (2) SGBD rad monitor C. (1) 1.05 mRem/hour (2) TB Vent Stack and CVPETS Rad monitor D. (1) 1.05 mRem/hour (2) SGBD rad monitor Thursday, September 05, 2013 7:36:58 PM 84 Rev. FINAL
- 85. Given the following plant conditions:
- The plant is operating in Mode 3
- At 0900 on Sept 1 st, the Personnel Air Lock (PAL), Inner door seal fails
- At 0800 on Sept 3 rd, the Emergency Air Lock (EAL), Outer door seal fails Which ONE of the following completes the statements below in accordance with Technical Specification 3.6.1.3, Containment Air Locks, and its Bases?
The latest day/time that either of the air locks can be use for entry/exit under administrative controls is (1) .
In accordance with the Technical Speci1iation 3.6.1 .3, Bases, during this period of time, the use of the air lock to performnon-Technical Specification required activites or repairs onq-vital plant equipment is (2) in Containment.
(Reference provided)
A. (1) 0900 on September 8 th (2) allowed B. (1) 0900 on September 8 th (2) not allowed C. (1) 0800 on September 10 th (2) allowed D. (1) 0800 on September 10 th (2) not allowed Thursday, September 05, 2013 7:36:58 PM 85 Rev. FINAL
- 86. Given the following plant conditions:
- A MDAFW pump is unavailable due to a motor problem
- B Main Steam Line radiation monitor is in HIGH alarm
- The crew trips the Reactor and actuates Safety Injection due to lowering PZR level
- After the Reactor Trip, one B SG safety valve stuck open
- An 86 lockout occurs on the I B-SB 6.9KV Emergency Bus
- MSIVs will not close
- B SG narrow range level is 30% and rising following isolation of feed Which ONE of the following completes the statement below?
I MS-70, Main Steam B To Aux Fw Turbine, (1) , AND (2) will direct this for the given conditions.
A. (1) must remain open (2) E-2, Faulted Steam Generator Isolation B. (1) is required to be closed (2) E-2, Faulted Steam Generator Isolation C. (1) must remain open (2) E-3, Steam Generator Tube Rupture D. (I) is required to be closed (2) E-3, Steam Generator Tube Rupture Thursday, September 05, 2013 7:36:58 PM 86 Rev. FINAL
- 87. Given the following plant conditions:
- The unit is operating at 100% power
- A LOCA occurs
- The crew is implementing E-1, Loss of Reactor or Secondary Coolant
- RCS pressure is 675 psig and slowly lowering
- ALB-004, 2-2, Refueling Water Storage Tank Low Level, is in alarm
- ALB-004, 2-3, Refueling Water Storage Tank Low Low Level Alert, is NOT in alarm
- Safety Injection has been reset
- Subsequently, a loss of offsite power occurs Which ONE of the following (1) identifies the required transition AND (2) the attachment used to verify proper configuration of safeguards equipment following the loss of offsite power?
A. (1) ES-1.3, Transfer To Cold Leg Recirculation (2) E-0, Attachment 6, Safeguards Equipment Realignment Following A Loss Of Offsite Power B. (1) ES-1.2, Post LOCA Cooldown And Depressurization (2) E-0, Attachment 6, Safeguards Equipment Realignment Following A Loss Of Offsite Power C. (1) ES-I .3, Transfer To Cold Leg Recirculation (2) E-0, Attachment 8, Response To Loss of Offsite Power to AC Emergency After SI Actuation D. (1) ES-1.2, Post LOCA Cooldown And Depressurization (2) E-0, Attachment 8, Response To Loss of Offsite Power to AC Emergency After SI Actuation Thursday, September 05, 2013 7:36:58 PM 87 Rev. FINAL
- 88. Given the following plant conditions:
- The unit is operating at 100% power
- At 0800 the I B-SB Emergency Battery has been declared inoperable due to a failure of the 1 B-SB battery charger
- At 0830 an electrician performing the weekly maintenance surveillance test for the 1A-SA Emergency Battery reports following pilot cell indications:
- electrolyte level is midway between the minimum and maximum marks
- float voltage is 2.10 volts
- specific gravity is 1.198 Which ONE of the following completes BOTH of the statements below?
In accordance with Technical Specification 3.8.2.1, D. C. Sources Operating, the IA-SA battery is (1)
Based on the conditions provided above, the MINIMUM required action is to (2)
(Reference provided)
A. (1) operable (2) place the 1 A-SB battery charger in service prior to 1000 B. (1)operable (2) place the lA-SB battery charger in service prior to 1030 C. (1) inoperable (2) enter Technical Specification 3.0.3 at 0930 D. (1) inoperable (2) enter Technical Specification 3.0.3 at 1000 Thursday, September 05, 2013 7:36:58 PM 88 Rev. FINAL
- 89. Given the following plant conditions:
- The unit is operating at 65% power
- The following annunciator is received in the Control Room:
- ALB-002-7-2, Serv Wtr Pumps Discharge Low Press
- The BOP notes that Cooling Tower Basin Level is lowering rapidly
- Service Water header pressure is 50 psig and lowering One minute later
- Service Water header pressure is 35 psig and continues to lower
- CTMU cannot maintain Cooling Tower Basin level
- The Cooling Tower Basin Level continues to lower
- The RAB AO reports that a large volume of water is gushing from the downstream flange of 1 SW-276, Headers A & B Return To Normal SW Header valve Which ONE of the following completes the statements below?
The leak is located in the (1) system.
In accordance with Technical Specification 3.7.4, Emergency Service Water, the bases for the Limiting Condition of Operation is to ensure that sufficient cooling capacity is available for continued operation of safety related equipment during (2) conditions.
A. (1) Normal Service Water (2) normal AND accident B. (1) Normal Service Water (2) ONLY accident C. (1) Emergency Service Water (2) normal AND accident D. (1) Emergency Service Water (2) ONLY accident Thursday, September 05, 2013 7:36:58 PM 89 Rev. FINAL
- 90. Given the following plant conditions:
- The unit is operating at 100% power
- A loss of power to Safety Bus 1 B-SB occurs
- The B EDG fails to start Which ONE of the following describes (1) the effect on the plant AND (2) the Technical Specification requirements that currently apply?
(Reference provided)
A. (1) A Containment Ventilation Isolation Signal will be generated (2) Restore the B Train Containment vacuum breaker in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
B. (1) A Containment Ventilation Isolation Signal will be generated (2) Be in at least HOT STANDBY within the next 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.
C. (1) A Containment Ventilation Isolation Signal will NOT be generated (2) Restore the B Train Containment vacuum breaker in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
D. (1) A Containment Ventilation Isolation Signal will NOT be generated (2) Be in at least HOT STANDBY within the next 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.
Thursday, September 05, 2013 7:36:58 PM 90 Rev. FINAL
- 91. Given the following plant conditions:
- The crew is implementing E-1, Loss Of Reactor Or Secondary Coolant
- Plant conditions are as follows:
- CNMT pressure 12.6 psig
- RCS Hot leg temperature 650°F-
- The five hottest core exit thermocouples are:
A08 1201°F B05 1208°F G02 857°F H15 753°F L14 734°F
- RCS pressure 200 psig
- RVLIS Full Range level 40%
- The SPTOP and CSFST displays are NOT available on ERFIS Which ONE of the following identifies (1)the requirement for FR-C.1, Response to Inadequate Core Cooling AND (2)the status of the Fuel Clad Barrier in accordance with EP-EAL?
(Reference provided)
A. (1) required to be implemented (1) Loss of Fuel Clad Barrier B. (1) required to be implemented (2) Potenitial Loss of Fuel Clad Barrier C. (1) NOT required to be implemented (2) Loss of Fuel Clad Barrier D. (1) NOT required to be implemented (2) Potential Loss of Fuel Clad Barrier Thursday, September 05, 2013 7:36:58 PM 91 Rev. FINAL
- 92. Given the following plant conditions:
- 1000 A General Emergency has been declared due to a LOCA
- 1015 RVLIS Full Range is 35% and lowering
- RCS Pressure is 100 psig
- Core Exit Theromcouple temperature is 694°F
- 1115 The hydrogen monitoring system and recombiners were placed in service in accordance with E-1 and OP-I 25, Post Accident Hydrogen System
- 1200 Due to a malfunction of the recombiners, the containment hydrogen concentration is now 6%
- RVLIS Full Range is 34%
- RCS Pressure is 80 psig
- Core Exit Theromcouple temperature is 712°F Which ONE of the following completes the statements below regarding the hydrogen in containment?
The containment hydrogen monitoring system is designed with an intermittent cycle of hydrogen indication for (1) different sample points in containment.
The required Protective Action Recommendation is to evacuate a (2) mile radius.
(Reference Provided)
A. (1)Three (2) 2 B. (1)Three (2) 5 C. (1)Six (2) 2 D. (1)Six (2) 5 Thursday, September 05, 2013 7:36:58 PM 92 Rev. FINAL
- 93. Given the following plant conditions:
- A batch release of the Secondary Waste Sample Tank is in progress
- A HIGH ALARM is received on REM-21WS-3542, Secondary Waste Sample Tank Pump discharge radiation monitor; however the release failed to AUTO terminate Which ONE of the following completes the statements below?
In accordance with ODCM 3.11 .1.1, Liquid Effluents Concentration, the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS shall be limited to (1) times the concentrations specified in 10 CFR Part 20.
In accordance with the ODCM 3.3.3.10, Monitoring Instrumentation Radioactive Liquid Effluent Monitoring Instrumentation, with REM-21WS-3542 inoperable, this release may continue from this pathway provided that (2)
(Reference provided)
A. (1)10 (2) samples, release rate calcs, and the valve line-up are Independently Verified B. (1)10 (2) once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> grab samples are analyzed for radioactivity at a LLD C. (1)20 (2) samples, release rate calcs, and the valve line-up are Independently Verified D. (1)20 (2) once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> grab samples are analyzed for radioactivity at a LLD Thursday, September 05, 2013 7:36:58 PM 93 Rev. FINAL
- 94. Given the following plant conditions:
- A core off load is in progress to support a refueling outage in accordance with FHP-014, Fuel and Insert Shuffle Sequence
- A fuel assembly has just been latched and raised for serial number verification
- The serial number on Attachment 2, Core Offload/Reload Fuel Transfer Data Sheet, does NOT match the serial number on the fuel assembly Which ONE of the following identifies the action(s) required by FHP-014, with regard to the latched fuel assembly?
A. Lower the fuel assembly in the location it was removed from AND unlatch B. Move the fuel assembly to the temporary storage location AND unlatch C. Lower the fuel assembly in the location it was removed from but do NOT unlatch D. Move the fuel assembly to the temporary storage location but do NOT unlatch Thursday, September 05, 2013 7:36:58 PM 94 Rev. FINAL
- 95. Which ONE of the following choices completes the statements below?
OPS NGGC-1 301, Equipment Clearance, requires that the ground checklist be authorized by a(the) (1)
(2) verification is required for ground installation.
A. (1) Senior Reactor Operator (2) Concurrent B. (1) Senior Reactor Operator (2) Independent C. (1) Electrical Maintenance Supervisor (2) Concurrent D. (1) Electrical Maintenance Supervisor (2) Independent Thursday, September 05, 2013 7:36:58 PM 95 Rev. FINAL
- 96. Given the following plant conditions:
- The unit at 100% power
- At 09:00 on Sept 8, 2013, the the A-SA EDG Fuel Oil Transfer pump was placed under clearance to repair a fuel oil leak
- At 1 1:00 on the same day, a fault in the control power circuit for the B-SB Containment Spray pump causes the control power fuses to blow Assuming no additional changes to equipment operability which ONE of the following identifies, when the unit must enter Mode 3 in accordance with Technical Specifications?
(Reference provided)
A. l800onSept8,2013 B. 2200 on Sept 8,2013 C. l500onSeptll,2013 D. l700onSeptll,2013 Thursday, September 05, 2013 7:36:58 PM 96 Rev. FINAL
- 97. Given the following plant conditions:
- An employee was injured and contaminated
- The employee was transported to Western WakeMed for treatment before he was de-contaminated
- Duke Energy Progress is planning a news release for this event Which ONE of the following completes the statements below?
In accordance with AP-617, Reportability Determination And Notification, the EARLIEST required NRC notification of this event is within (1) hours.
In accordance with AOP-01 3 (2) is the primary radiological concern for fuel off-loaded more than 6 months ago because it will NOT be detected by personal dosimetry or area radiation monitors.
(Reference provided)
A. (1)4 (2) Krypton-85 B. (1)4 (2) lodine-131 C. (1)8 (2) Krypton-85 D. (1)8 (2) lodine-131 Thursday, September 05, 2013 7:36:58 PM 97 Rev. FINAL
- 98. Which ONE of the following completes the statements below in accordance with PEP-330, Radiological Consequences, Attachment 1, Limitations for Lifesaving and Emergency Reentry/Repair Actions?
Emergency worker exposures during life saving missions should be limited to (1)
Exposures in excess of 5 REM TEDE shall not be permitted unless specifically authorized by the (2)
A. (1) 15 (2) Emergency Response Manager B. (1) 15 (2) Site Emergency Coordinator C. (1) 25 (2) Emergency Response Manager D. (1) 25 (2) Site Emergency Coordinator Thursday, September 05, 2013 7:36:58 PM 98 Rev. FINAL
- 99. Which ONE of the completes the statements below in accordance with PEP-230, Control Room Operations?
The Emergency Response Organization (ERO) accountability process must be completed within a MAXIMUM of (1) from the time the Site Area Emergency was declared.
The SEC-MCRs task of making Offsite Protective Action Recommendations (PARs)
(2) be delegated to the TSC.
A. (1)30 minutes (2) can NOT B. (1)30 minutes (2) can C. (1)60 minutes (2) can NOT D. (1)60 minutes (2) can Thursday, September 05, 2013 7:36:58 PM 99 Rev. FINAL
2013 HNP NRC SRO 100. Given the following plant conditions:
- The plant is operating at 100% power
- The OSl/Pl and ERFIS server are Out of Service for a software update
- At 0800 the following occurs:
- ALB-026 /1-4, Annun Sys I Power Supply Failure
- ALB-003 / 4-5, Annunciator System 2 Power Supply Failures
- The QAC reports 23 of the 30 Main Control Board ALBs have lost annunciators
- The AO reports the both Annuciator System I and Annunciator System 2 have lost multiple 125 VDC power supplies
- At 0900 the following occurs:
- ALB-019, 3-2A, HTR DRN Pump B 0/C TRIP-GND
- The HTR DRN Pump B trips Which ONE of the following completes both statements?
At 0800 the HIGHEST required classification is (1)
At 0900 the HIGHEST required classification is (2)
(Reference provided)
A. (1) Unusual Event (2) Unusual Event B. (1) Unusual Event (2) Alert C. (1)Alert (2) Alert D. (1)Alert (2) Site Area Emergency You have completed the test!
Thursday, September 05, 2013 7:36:58 PM 100 Rev. FINAL
2013 ILC NRC Exam DUKE ENERGY PROGRESS HARRIS TRAINING SECTION EXAM NUMBER: 2013 NRC LESSON/COURSE CODE: SO6CO3H SUBJECT/CATEGORY: SRO Written EXAM POINT VALUE: 100 STUDENT NAME (PLEASE PRINT):
DATE: SSN:
Prepared by: Archie Lucky! JR. Horton DATE: 9/25/2013 Exam Validation by: Mike Matheny and Kyle Kelly DATE: 9/04/2013 APPROVED BY: Simon Schwindt DATE: 9/06/2013 SUPERVISOR OR DESIGNEE ALL WORK DONE ON THIS EXAM (INCLUDING CORRECTIONS) IS MY OWN. I HAVE NEITHER GIVEN NOR RECEIVED AID.
I AGREE THAT I WILL NOT DIVULGE ANY INFORMATION WITH REGARDS TO THE CONTENT OF THIS EXAMINATION TO ANY UNAUTHORIZED PERSONNEL.
SIGNATURE: DATE GRADE: GRADED BY: DATE GRADE VERIFICATION: DATE References and/or tools provided for use with the examination include: (list below)
- Calculator
- PEP-hO, Rev. 22
- Steam Tables / Mollier Diagram
- RM-1 1 Status Display Screen print (2)
- Curve No. F-18-1, Rev. 0
- T.S. 3.1.3.1 pg 3/4 1-l4thru 1-16
- Curve No. F-X-20, Rev. 1
- T.S. 3.3.3.6 pg 3/4 3-66 thru 3-69
- AP-617, Rev. 33
- T.S. 3.6.1.3 pg 3/4 6-4 thru 6-5
- EAL Matrix, Rev. 10
- T.S. 3.6.2.1 pg 3/4 6-11
- ECA-1.1 step 7.b, pg 6, Rev. 0
- T.S. 3.6.5 pg 3/4 6-32
- ODCM 3.3.3.10 pg D-2 thru D-4
- T.S. 3.8.1.1 pg 3/4 8-1 thru 8-4
- ODCM 3.3.3.11 pg D-7 thru D-9
- T.S. 3.8.2.1 pg 3/4 8-12 thru 8-14 QA/VITAL RECORD
2013 NRC SRO Question 81(6) Reference TRUN TAT TON ACCI DENT )NJ TORTN(i INSTRUNENTAT IUN LTtilTI CCJNO[flONOR OPERATION 3 3t 7 wnt:crIro I r trwnnt.dtorI dn1 ywn in TL: 3iQ shH be GPERABLE.
jjJT r400ES 1. 2. arid 3.
TON
- a. Witi the number of OPERABLE accident monitoring instrumentation charmels except Tn Core Tiermocouplec and Reactr Vessel tevel less than the Total frequired Number of Chainels reuireinents cIrvn in Teble 3.340 restore the iOpeTab1e channel(s) to OPEPJBLE status witniri 7 days or be in at 1ast HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> nd in t least HOT S%1UTOOt within the foilcing 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- b. it.h the number of OPERABLE accident monitoring instrwiientation chaiiels e<cept tie rait ion rronitos tne Pressurler Safety tjlvci Pi1Ion Jnditor the RaC1ur Cooant System jthcoiinq Margin Monitor. In Core [hernocoupies or Reactor Vessel Level, iess than the Miniiiir Chne1s flPEPBLE i Tabe 3 0 re I oi th irhl chnrFls to 0ELL latis n1n 4 hnur OL e ir a: les HO SUNOBr i:1ur the rxt t rours ard in t least Jr SHUTW iiiiin the falling 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
. With the numbor of 0PERBLF aoident monltorinq itrjmnt.ation channels fr the rdiition rncnitar(s). the Pressurizer Safety Valve Ps tiOri 1rditor c-r th Recto Coclnt Ssten Subcoolirg Magi Men tor# less than the 1ininxirn Chnnes GLKA&E reotmements of Table 3.340. initiate the prepianned alternate method of monitoring the pproprite paramneteri ithln 7 Iicurs and eir[er restore th noperble c[orinel Is) tn uPERAI3L[ status within day3 or prepare and submit a Special Rporz to the Conmnsson pursuant to Specification 6 9 2 ithin thc next 14 days tht prondes actions taken cause o the inoperab.illty md the plris anJ schedule for restoring the channeis) to OPERABLE status,
- d. With the numebor of OPERABLE accident monitoring instrumentation channels fa- In Core Thniiocouples or Reactor Vessel Level leos than tie toLl required nuniher OL chnelr shown in 3 10 restore the inoprcible chaqnel(;) to LABLE status within 30 days or subu t a SoEcial Report pursuant to oecifca ton 9 2 withui the 1ulloirtg 14 ds romi the iie The dcton s rcuired f reper s[ al outl ire :t e prcplan1e.J al ter9ate ri hod of cnon torq e cause of te inoperbilit id the plans and scneule ton estorrq the instrurrentation ciannels to operable status.
- e. With the iiti-crber of OPEPARLE accident rronitoriiig instrument ehannels o Ti Ccre Therrnocoupl.s r Remr. Vecel Le& les thri nuninurn iarrels OPEPA F rr1jeInert of Lible 1J e t9r restore one- channel to OPtiABLE status within 7 days or be in at SHEARON HARRTS UNIT L 3/4 3-66 Amendment No. 110
2013 NRC SRO Question 81(6) Reference
[41f1 CONDITION OR OP[PJJION least. HOT StANO3r in tne next 6 hcus in at least HOT SHUTO2 withm the oliainçj 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
f Th proisins f Spcifiction 3O.4 ot appiicabi.
- Thi alternate ethad shall be a ctieck of safety valve piping tereratures arid ealuation to detrrnine position.
Thi alternate method shall be Le iflit1ati]m .hC backup wthod os reqi1 red by Specification SURVEIaAhOE RRUIREME NTS
& 3,3 6 Eacfi accident morn toring instruinentaticri channel shall be dencwistrated OPERABLE by perfor1narcP of tne CbA1EL U1EC and CHANrWL CAU BRA ION at the frequencies shown in Table 4.37.
SHERO HPRR1S - UNIT 3/$ 0-67 Amendrr,ent No, 110
2013 NRC SRO Question 81(6) Reference T&LE 3J1O TOTAL RFIRED OF L Continn eu
- a. )arrow Range 2 1 b., Jide Rang 2 1 2, karn Cociant HtLe TempraturWide Range 2 1
- 3. R Caia+/-t C1d-L Tenprtire- -Wide Ran,e 2 1 4 Reacttt Cuo1mt rui 44id mg 2 1 f Preurizer Wtt Lt 2 1
- 6. Steait Line Pr* 2ta generator 1fst.e &enerator Steaa Ceat iater Le1arr Rnge 4,A, 1/steaiu gera.
Steam Cemeratr Water LeviWide Raigr Nd. 1/ci generatnr 9, Refuaiing iatr Stcrag Tat* 1It Level I lO Aux:iliary FeIwaer flc Rate I/sten gereratcr i1 Raaetr Ciant Syte Subcoo1ie Xarin &1tr
- 12. ?GRV Psjtc Indiatr* 1/valve.
13, PORV Block Valve $kivtt Indicator** LA. 1/valve I4 kfry V1 Tttr
- 16. Ctalent W3te Leve -Wide Range 1
2013 NRC SRO Question 81(6) Reference TL[ Contiued)
ACCi)ENTjTOR11$TiUMENTATTh)
TOIAL
- EQUIR[D IiilMU1 NO (4 MtELS iI1WUtT Afj OPEP4L
- 17. In Core hennocciupes fcore idrnt qLr nt Pint Vent Stack--Hh Range No1e Gas Raitftn onitcr I I. Miin Line dtcw onitcrs NA. 1/sLpni linE 20 Cntaintnnt Wgh 1argo Radi ati cn Honitor NA
- 21. Reuictr ese1 1eel 2 1
- 22. Co1Ld [ML Spiiy t1i)H tank Lev1 1 Turb1e iJi1cIHlg efl Stack MgI- Inge Nob[e Uds Kddlatlon rA 1 Mon tr 24, tte [rceiflg Bul 1dirg Vert stack IIi9h Range Noble G-s Rd1tion rnitors nt st N b Vefit Sad NA. 1 25 CoIeiste Storg Tank Level 2 1
- N plicb1e ir the asoc1ated blcck valve Is tn the cThed position.
Not piKable ii t[a blk vlv i vri (id tJi lL pi t un nd 1JMW iON 1RIS UNIT 1 3/4 369 AHerJwIL Nu. 35
2013 NRC SRO Question 82 (7) Reference REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING GONDITION FOR OPERATION 3.1.3.1 All shutdown and control rods shall be OPERABLE and positioned within
+/- 12 steps (indicated position) of their group step counter demand position.
APPLICABILITY: NODES f and 2.
ACT I ON:
- a. With one or more rods inoperable due to being immovable as a result of excessive friction or inecharncal interference or knon to be untrippable determine that the SHUTDOWN MARGIN requirerneni of Specification 3.Ll.i is satisfied within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b With more than one rod misaligned from the group step counter demdnd position by more than +/- 12 steps (indicated position) be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,
- c. With more than one rod inoperable, due to a rod control urgent failure alarm or obvious electrical problem in the rod control system existing for greater than 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> be in HOT STANDBY within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- d. With one rod trippable but inoperable due to causes other than addressed by ACTION a above or misaligned from its group step counter demand height by more than +/- 12 steps (rndicated position). POWER OPEftATION may continue provided that within 1 hour:
- 1. The rod is restored to OPERABLE status within the above alinmeot requiernents. or
- 2. The rod is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within
+/- 1? steps of the inoperable rod while maintaining the rod sequence and insertion limits of Specification 3.1.36. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation. or
- 3. The rod is declared inoperable and the SHUTDOI4N MARGIN requirement of Specification 3 1. 1 1 is satisfied POWER OPERATION may then continue provided that:
a) A reevaluation of each accident analysis of Table 3.14 is performed within 5 days: this reevaluation shall confirm that the previously analyzed results of these accidents See Special Test Exceptions Specifications 3.10.2 and 3.10.3..
SHEARON HARRIS UNIT 1 3/4 144 Amendment No. 25
2013 NRC SRO Question 82 (7) Reference REACT I V t Y CCM ROt S STFMS LIMITING COiDION FOR OPFiTfJN ACTiON (Cant irued) remain va 1 d for the dura Li on of cuerat i on under these conditions:
b) The SHUTDOWN MARGIN requirement of Specification 3Li.i is determined at least once per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />s:
c) A power distribution map is obtained from the movable incore detectors and FZ) and are veniHed to be within their limits within 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />s: and d) The ThERMAL POWER level is reduced to less than or equal to 7 t, of RAT LO II IER POvF R wit H n [w next hou a rid w i tn i n the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85 of RATED HER1AL DOWER SLRVE ILLANCE REOU I RENEW IS 4i.31.1 The position of each rod shall be determined to be within toe grouu demand limit by verifying te indviduai rod positions at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during tine intervals when the rod position deviation monitor is inoperable, then verify the group positons at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
41312 Each rod rot fully inserted in the core shall be determined to be OPERABI E by movement of at least 10 steps in any one direction at least once per 92 days.
SHEARON HARRIS Aiendnent No
2013 NRC SRO Question 82 (7) Reference ASL! 3. 11 ACCIOENT ANAL ESRE[LR1NG REEVALIJTION IN Th EVENT OF N INQPRA8E R Rd Cluster Control A5seinbly tn5ertjfl Charatertstls Rod Clstar Contro1,Asseby MisaHgrent Loss f Reactor Cool ant from Small Ruptured Pipes a from Crac)s in Large Ptpes Which Actuates the E erercy Core Cooling System single Rod Cluster Control Assely Wittidrawal at Fl Power Mi1o ctoi Coolant Syst Pipe P.uptures (Loss fCooant Acel dent) 4&Jor Secondary Cool ant Systam P1e Ruptire Rupture of a Controt Rod Drive !echanfs Hoslng (od Cluster Conti Asm1y Ejecticn) 5HEARQN HARRIS UNIT 1
2013 NRC SRO Question 83 (8) Reference 2013 NRC SRO Question 83 (8) Reference A
1 1US DtSPL S4i WP HMOM LOW */2j3 OIffiNNL ID ø# 2*
DtSCR!PTXON HAS RHWWt po,. STATUS MONflOR OF7LIN IIU COHt4UNIOATIONS MONItOR OOtIMUNIOA?EONS r*
I a:::
O ANN L UT W SERVtCaasaaa,,
CH*NNtL PZL?E NOT NOVZNGa a a a. a a CHANNEL ZLflR OLOGEDa a a a a a aaa a CHANNEL NO PULSES RECEIUEDa a a a a a a a a CHANNEL ONEOR SOUROE TEST r*xuo. a a a CHANNEL LOSS or SANPLE LOWa. a a a a a a
- CHANNEL NIGH TPR*TURE ODNOITION a a CHANNEL OPERATE FALURE. a a a a a a a a a a a CHANNEL HIGH ALARM CHANNEL ZN NIGH ALARM. a a a a a a a a
- CHANNEL. ALERT ALARM: CHANNEL ZN ALERT ALARM. a a a a a a a a a a a a a a EUZPHENT FAILURE MONITOR LOSS OF PROCESS FLOW. a a a. a a MONITOR IN SCAN OVERLOAD. a a MONITOR LOSS OF FLOW CONTROL MONITOR LOSS OF ISOKENETIC CONTROL....
MONITOR LOSS OF RM23 COMMUNICATIONS...
MONITOR INSTRUMENT FAILURE. . .. .
LOSS OF ASSOCIATED FLOW.
MONITOR HIGH PRESSURE ALARM. . .
CHANNEL EXCESSIVE NEGATIVE DELTAS.
ISOKINETIC VALVE POSITION FAILURE CONTROL FUNCTIONS MONITOR PURGING. a a a a a a CHANNEL PURGING a a a a a a CHANNEL CHECK SOURCE ENERGIZEC CHANNEL FILTER ADVANOEt4Ga a a I NORN OPERATIONS NORMAL OPERATING CONDITION a a a a a PILTEfi vczw CHECKSOURCE GRIDS
2013 NRC SRO Question 83 (8) Reference hearon Harris Nuclear Power Plant .TPP: December l8 Offeite Dse Calculation Nanual (QtN Rev. 11 3/4 .3 3 MONITORING IN.cTRtJMTATION 3/4.3.3.11 Radioactive Gaseous Effluent Monitorino Instrumentation OPERATIONAL REQUIREMEBT 3.3.3.11 The radioactive gaseous effluent monitorinq instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their Alarm/Trip getpoiata set to ensure that the limits of Operational P.equirements 3,11.2.1 are not exceeded The Alarm/Trip &etpoints of these channels meeting Operational Requirement 3 .11.2.1 shall be determined and adjusted in accordance with the etbodolc-v and parameters in the ODCN.
APPLIcABILITY As shown in Table 3.3-13 ACTION
- a. With a radioactive gaseous effluent monitoring instrumentation channel Alarm/Trip £etpoint less conservative than required by the above Operational Requirement, immediately 1) suspend the release of radioactive gaseous effluents monitored by the affected channel or declare the channel inoperable and take ACTION as directed by h. below.
- b. With th number of OPERABLE radioactive gaseous effluent monitoring instrumentation channels less than the t4iniTnum Channels OPERABLE, take the ACTION shown in Table 3.3-13. Exert best efforts to return the instrument to OPERABLE status within 30 days. If unsuccessful, explain in the next Annua.1 Radioactive Effluent Release Report pursuant to ODCT*1, Appendix F, section F.2 why this inoperability was not corrected in a timely manner.
!JRVEILLANCE REQrJIRE4EtrT 4.3.3.11 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECR, gOtThj2E CHECK, CHANNEL CALIBRATION and a DIGITAL CHANNEL OPERATIONAL TE&T or an A1JjD CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-9.
Each surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25% of the specified surveillance interval.
2013 NRC SRO Question 83 (8) Reference harot 1tariis Uu:1ar Fei Pan: (rPP August 195 1t LOSE a1ilat1n iar.ua: t)LJ1:
TL .-12 i:1:Tri GACirS PLURtT rcNrx?.rN3 IT1LEUTATIt NtLl.
INSTRYI4WT CJ{1NNL AP ?..!ILIT1 ACrIQT CPILE
- 1. cz-crjD wzcr c-Ycrzi iRccr :.NI oicT zp.c pE:iEicaIon ia nt 1ise in 2 TTRTNV. P1TTr1TTh VPT C9r}
- a. Noble Gas ActLvitv tonitr 1
- b. Icdin sampler 1
- c. Pariuiat rn1er 1 *
- d. Plow .ate r4onLtor j *
- e. ampJ.r L]xw aze xonLtor i -
- 3. p.rr VENV srAc}:
- a. Uu1i1 LiviL cij.LL_i 1
- b. cdin anipler 1.
- r..rtiu1t rp1cr 1. *
- d. Plow at 1onLor 1 JLplr ]:w Ncnitor 1
- WASTE PROCESIIIG MflLDING VEI1T 3TICK 5 1 Nnh1 rtj,itv ?r,i1-,,r (pm) 1 * .i a..2 Uobi Gas lctLvitv t4onitcr 1 NC3E 1, 2, 52
- h. Idn 3an1pler 1 C. Par1u1at arnp1er 1
- d. Plow sate ronLtor 1 Sd1.j.1L Flw M.iLui 1 5 ATE FRCCESIflG !UILO!NG ThCK 51
- a. Uub1 ILLviLv JLLLL 1
- b. Iodine riip1er 1 C. Crr1Cr 1
- d. Plow ate FIonLt:r 1 2.rtp1r ]r NcnLtor 1 TALE ICTTICN FL all. J,iii.
D-B
2013 NRC SRO Question 83 (8) Reference hLon Ki,:..r ;lar CT; L2S ff site rose CalDuatin nia]. ccrii: Rev. 6 T3LE l-l1 Ccntinued)
ACTIOT TATEI1EllT AcrIt 45 - With :he number channels Wft-3LE less than required by the Ninirnus Ohaxrn1 OPRL r 1imIt, thG otnt of te ga day tank s aay be released to the environment provided tha: prior to irititirg the rele At 1et :wo indepnieiit eaiip1e oE the tank
- s cnteat re analysed, and k t least :wo technically qualified rnenters of the fa:ility staff 1niepenc1et.Ly verity tie release rate caLculations aa scbarge valve lineup.
therise, suspend release of cadjoactive efflueits via this pathway.
46 - With :he number of 2harjels OP2RLE less than required by the ttinimun Channels OPER?BLE requilernent, effluent releases via this pathway say contiaue provided the flow rate is estimated at least once per 4 hjurs.
(9TN a - with -h vnirAhi- f hqrnR1 PZPPTV 1 th;i, rrjnirr by frh reinimur.
Channels OPEAPLE requli-ennent, effluent releases via this pathway say cortiii p.idd grab atrip1s ax taken at ].as: ozo per 12 kour ad these samples are analysed for radioactivity witnin 2i hours ACTION 48 - Not Used in :he oocii.
ICTION 49 - With :he number of channels OPZJt.ELE less than required by the r-cininnin CILdLjL1 OEEBLE Le4 LIL,jLL. LfluiiL yi Uj LL)l &Lwdy nay continue prcvided samples are continuus]v ccllected with auDcilialy aamp.L1n equipment as required in Table 411-2 ACTIiN ES Not used in :he ODfl.
ATIU El - With :he number of chanflels OPZLE less than required by the r-inimiin Channels C LE requiLemnent Eor both the P]G aid RGN. effluent releases via this pathway nay continue provided grat samples a:e taker. at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samrles are analysed or rad.oacivitj withi:i 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
kTION E2 - With the number of OPEEBLE accident sonitoring instrumentation dhsnnels for tte radiation monitDr) lse thai tha Minieiam Channe1 ORL requirements of Technical pecLfication Table 1.-lC, initiate the preplanned Ltrnte methc of ritrin.g the prorie rneCeL- s idthia 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />, and either restore the inoperable channel (s to CPERLE etatue within 13 d,n:a 0L prepcre end eubmit Speia1 nepit to the Commissicn, pursuant to Technical pecification 5 2, *itbin the iext 14 1yb _ljL juv L3 dL jL1 tk-n, eua uL L1L i.LUpL1JL1iLy, ni Lh pans and schedule for Lestori1g the channel 5:1 cPERLE status
2013 NRC SRO Question 84 (9) Reference llTRUMENTA1JO ACC I DENT ON I TORI NG I NSFRUNENTAT I ON LiMITINf CONDITION OR OPERAtION Th cijn irn tm1r I r trwin.aflci chinneis iCwri in Tae 3 -1Q shall be OPERAE3LE ijJjJ: MODEES L 2, and 3.
AOTION:
- i. With the nuruber of O?EPJBLE accident monitoring instrunentation cbaiv-1s except In Ccii Thericouple ard reac:r Vsl frll 1os than tn To frqurerJ tftnbr ot hnrc1s rruruients r[rn in Table .3 3 O estor he ircperbe cnannls to OPER?LE siaL itnir dais r b in at 1 east HOT S1AND ithin the next 6 hur and ir t least: HT S1UTIXWN within the follng 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,
- b. With the nuniber of OrERABtE ccident nioriltoring instrwiientatiori channels exret the radiation (ranitors the Presurier Sfet Jd1vc Ps1tin Indcator th Reattor CrotarU System Subcoiinq Marg n Morntcr n Lcr thernocouptes or RectDr Ves3el eel ass than the Mirnmjir Chrne1s OPFPARIE quwiients o Tab e 10 recire me inp to OPE°.LE stat thln-18 hnir o ze in at 1est HO ST flBi ithir tie r xt 6 hojrs and in at least WIF SHUTW wtthin th ffih1ownçj 6 [fours.
- c. With the number of OPERABLE ceident monitorinq thstruent.ation channels for the r.adiitioi nionitcr.s) bie pressurizer Safet 1 Vahe Pos ticr lid c1or or th Reator Cod ant S eiu ubcool iry MdrQln Mon tor# less t9an thF Mirnuumi Cnne s C[KA&E rernrrernts of Table 3.3I0, initiate the preplanned alternate method of nionitoning the proor-Iat paranter si ifl thin 72 r and ii rsror th noperle c iil1 tn PERA13L[ statLs tJ1R UdV or prepare ano submit a Special Report to the Coninssioq pursuarr tu Specification 6 9 2 ithin the next 14 days that proyioes act inns taken cddse o the inoperabiHL, md the glanS ann chcdul for restoring the channei(s to OPERAbLE stdtu5,
- 1. With the number of QPEPBLE accident monitoring instrurrntjtion channels for Iii Core Therniocmpies or Reactor Vessel Level less than the total reopi red number of channels shown in Table 33-iO. restore Ui inoperb1e chnrel(s to cPEP?.6LE tus ithir 30 days or subi t a SpEcial Report oursuant to speincarion 6 9 2 within the rolloing 14 dais roii the tue the ictmon is reeulred roe report halI oLJtHn tt preplarineci alter1dte rethod of rnonitorng the cause of the in operalii 11 ty d1d the p1 arms and scnedul e frn restoring the irtstrumrntation channels to operable status,
- a. With the nunber of OPERfBLE accident monitoring instrument chanftels lot Ti Cure TlLrmocouls r Recor Vesel LeeI t tan n raininuru Ln1nrelm OPE°P1 F *jt Table .3 !Jl einwr restore one c.honnei to OPtERAi3LE status within 7 days or be in at SHEARON HPRRIS UNIT L 3/4 363 Aiirdment No 110
2013 NRC SRO Question 84 (9) Reference L14TJ CONDITiON OR OP[PJJION 1est HOT STArOWf in tie rxt 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> aria in at least HUT SHU[COJN within the tol1ainçj 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- f. Th pvisions f Spcifictiori 3O.4 ar r. ppiicabl.
- The al nate iiethxi shi1 he i check of safety va! ye pipi ng tewieratures arid eaJuation to dtrmin positior.
The alternate method shall be he niitation of th bmckup wthcd as rquir by Specificticn 64 SURV[ILIAMDE RF(jiREME kTS
&. 3.3 6 Eac!i acifenh oiol itoring inshruin tat&m charmel hai 1 he ckncrnstrt.ed OPERAI3LE by perforwarice of the CAmEL CiI[CK ici cirri,. cuRAiIarJ at the freencies shoi in Table 4.7.
SHEARO4 HARRiS - UNIT 1 3/4 Z-i7 Aiendrnent No. liii
2013 NRC SRO Question 84 (9) Reference ThBLE 3,:-1O AUWT,. ?WN TOThL XiflRED t4c OF flELS Q1FRAL
- 1. Containment eiiura
- a. Narrow Rrg 2
- b. Wide Range 2
- 2. Rt Co1artt i{otLeg Temperatur ice Range 2 1 3, Racr Cpit.t Co1d-L Twprtire--Wide Range 2 1
- 4. Raact Coo1t ,r4ur 14dc 2 1
- i. Presurizer Uter Leet 2 1.
- 6. Steam Line Pr$ure 2ftta gertator l/ste gererator
- 7. Steaa Gnrair W:at vlro Range L/steant genetato Steam Generatr Wtcr Levi Wide Raflgn 1/se geieratr
- 9. Refiing atr Stragc tnk atr Level 2 1
- 10. Auxiliary F ewter Flaw R.at 1/sten nerator
- 11. Rctr Coolant Syste Subeoo1ig Xargin&itor A.
12 ORV Piti Inioatr*
l3 P0KV Block Valve ?otan Ind,lcator** 1/va1e 14 P 1 kfRty V1 P* Tr
- 15. Cntairient W;tr Lvt (EGGS Sump)Narrov Range 2 1
- 16. ctaitient Wte L e.W1de Range 2 1
2013 NRC SRO Question 84 (9) Reference 1LLLNO CCnt9ui?d)
ACC. D[N1 VC ITO I kG HSTRU$ENTATrIN TO1AL EQUIRfD NO cir :HAEL s JTRU1ET PE4Ei. E
- 17. In Core Thermoccup.5 Licore qudrnt ?/cre idrnt:
. Pnt Vent, Stc--Hq Range Notle G Radiation rit NA. I iq. Main in NA 1?srPnfl llnF 20.. Ciitainimnt Wh aic Rditicn Honlto I
- 21. flector ft5e1 ieei 2 1 22 *CoLd1IIIiL S)rdy t4dN dnk LevEl 1 TuIbne uIicTh1q tent tlgr inge Noble s Kaiation 1
- 24. PrGC3jm Bui1dir Vert tack IIfth R4r Noble Gs Rd1tioii r4nitor.
fl4 S*;( ç N 1 gent Sad 5 A. 1
- 25. Con1sat;e Storage Tank Lei 2 1 Not ppThcaNe if rhe soc1ated blcck vavc is 1 the c1oed o1tlon tot apfrl icb1e i f Uw bl,;k i1 vr 1i,d in Lli rl wU pr hUn and nwr ivei.
1Ri)N 1RRS - UNr: 1 AIirJaTInL 3/4 ,39 Nu.
2013 NRC SRO Question 85 (10) Reference CONTAINMENT sYSTEM CONTAIN4ENT AH COCKS LINFrING COtDITI0N FOR OPERJIJIOI4 3.6,1.3 Two containment air iock shaH be OPERABLE:
APPLIILIT t400ES 1. 2, 3 nd 4.
ACT 1 ON:
Entry and exit is pormissble to porform rpairs oi the affected air lock coponnts.
A separate ACTION is allowed fm- each air lock.
ter 3611 LCQ for Coritinnent Tntegrity tien he air lQck lease results in exceeding the overall containment leakage rate, Specification 3.6.1 .2.a, 4 Laek,n a Peonnel Ar Lock door thut Conlt of l&kln the associated manual pumping stations and deactivating the electronic necimarnsms used to open a Personnel Ar Lock door once the associated y lock cjçr t LQckng an Emer.gency Air Lcck door ijt cQ1at of lacking the mechanical operator.
- a. One or more containment air locks with one containment air lock door inoperab1e:
Within one hour. verify the OPERABLE door is closed in the affctd air hock, and
- 2. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, leek th WEPMtE door c.Thsed in the affected air lack, and
- 3. Once per 31 days, verify the OPERABLE dcr is locked closed in the affected air lock*, r 4, 0thrwis, be in at least 1-eCT STAI4D6Y within the next hus and in COLD SHUTDOWN ithin th following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
1 ACTiONS 3 6 1 3 a 1 3 6 1 3 a 2 3 6 1 3 a 3, and 3 6 1 3 a 4 are not applcable if both doors in the same ir lock are noperab]e nd ACTION 3 6 1 3 c is entered 2 Entry and exit is permissible for 7 days under adniinistrativ controls if both air locks are tnoperabie.
Air lock doors in high radiation areas nay be verified closed by administrative means.
SHEARON HARRiS UNIT 1 3/4 64
2013 NRC SRO Question 85 (10) Reference COTMNMENT SThTEMS COtFfi MNT iU I.
LIhITING CODITION*0R OPERTDN One or nore containment ir 1ock with containment air lock tiok chnii iprbl#
- 1. Within onE hour, ify n OPERL[ dG:r iS 1oed in the affected air lock, and
- 2. Within 24 hnois, Tck an OPERM3LE dGor closed in the loek, and por 31 days., verify the 9PERL door is locked cloed i the affeted air lock, or
- 4. Othrise, be in ilOT STNOBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN witirn the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
One or more cnntrnent air lotks inopera&le for reasons othor than 3.L3..a or 3.L3b.
- 2. Ioniedateiy Initiate action to evaluate twerall containment lakaçe rate er Lftt and
- 2. WIthin one hour, verify a dor is cosed in the affected air lock, ard 3 Within 24 iours restore air lock tc OPERkB.LE status or 4 Otherwise e in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> md in COIl) SHUTbO14N within the foflowirig O hours.
1 ACTIONS 3.6L3b.1. 3.,6i3.b.2, 36,L3ii3 1 and 3.613.b4 are not dpplicabTh if both doors n the same ir lock are inoperbl and ACTTCN 3,&L3.c is ritered.
- 2. Entry and exit of containment is permissible under the control of a dedicted individual ir lock doors ii high radiation are rma be erf,d closed by administrative moans.
SHEARON HARt5 - ur4:T I 34 6$a mendrnent t4OO
2013 NRC SRO Question 85 (10) Reference cOul MNT S S tE.
- 0NTPIEN I AIR CC.KS SIJRVEILLAbCF RFQURErIENTS 4,6, 1 .3 Each cürit.a nirrt a 1 r 1 ock shall b demonstrated OPERABLE by:
Performirici repuirec air lock leakage raze tes:inci in accordance with 1 (1 c rid i x 1 d inc. 1 f od t i roved < p ion The acceptance criteria fdr air lock testing are:
- 1. lvcrall ir lock icakgc ratc is O L hon tested it P.
- 2. Fo each door. leakaqe ra:e s .01 L, when zesten at.
5 I çJflt 1 n TO it b, Inn i t un L one l ) r th diC lOck Cd be opened a: t1rTlc?*.
- 1.AT1 I noperabi di r lock door does not I nvai i date the prevIous jcesful nrcjrinr t the eal1 dr1tc eakcie tet RcsuIt3 st hi: ci1uvo qint rtr fi 1 2 accordarce with 10 CFR 5. Appendix J. os modified by approvcd exn1pL1on..
0 ii ii r d he hr r . med upc ri e t r j u e 1 t 9 r ocqb the on: i iirer L o r 1yk . II S ir n 1 1 anci. Requ i r eun 4 6 1 . Li ias lot bee .er formed in the last e months thui perform Idni C I1qu1r11Crit 4 6 b dur rg the next contanment eritry through the osocia ted air lock, SHERO FIARRI S UNIT I Amend:nrit No
2013 NRC SRO Question 88 (13) Reference ELECTRICAL POWER SYSTEMS 3/4.8.2 D.C SOURCES OPERATING LIMITING CONDITION FO OPERATION 3,8.2.1 As a minimun, the foflowing D.C. electrical sources shall be OPERABLE:
- b. 125-volt Emergency Battery Bank 18-SB and either full capacity charger, 1ASB or lB-SB.
LtAB,jIJy MODES 1, 2, 3. and 4.
ACTION:
With one of the required D.C. electrical sources inoperable, restore the nDpeable U C electrical source to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in dt least HOT STANDBY within the next 6 nour and in COLD SllJTDc1N witrnn the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REOU1REMENTS 4.8.2.1 Each i25volt Emergency Battery and charger shall be demonstrated OPERABLE:
- a. At least once per 7 days by verifying tiat:
- 1. The parameters in Table &82 meet the Category A lnits.
a rid
- 2. The total battery terminal voltage is greater than or equal to 129 volts on float charge.
- b. At least once per 92 days and within 7 days after a battery discharge with battery terminal voltage below 110 volts, or battery overcharge with battery terminal voltage above 150 volts, by verifying that:
- 1. The parameters in Table 4.8-2 meet the Category B limits,
- 2. There is no visible corrosion at either terminals or connectors, or the connection resistance of these items is less than 150 x IO ohm, and
- 3. The average electrolyte temperature of 10 connected cells is above lOb F.
SHEARON HARRIS - UNIT 1 3/4 8-12
2013 NRC SRO Question 88 (13) Reference ELECTRICAL POWER SYSTEHS SURVEILLANCE REQUIREMEITS (continued)
- c. At least once per 18 months by verifying that:
- 1. The cells, cell plates and battery racks show no visual indication of physical damage or abnorrnal deterioration,
- 2. The cell-to-cell and terminal connections are clean, tiqht, and coated with ariticorro:sion material,
- 3. The resistance of eacfr cellto-cell and terminal corinectiori is less than or equal to 150 iO ohm, and
- 4. The battery charger will supply at least 150 amperes at greater than or equal to 125 volts for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
d, At least once per 18 months, during shutdown, by verifying that the battery capacity is adequate to supply and antain ir OPERABLE status all of the actual or simulated emergency loads for the design duty cycle when the battery is subjected to a battery serVice test:
At least once per 60 months, during shutd;own. by verifyin that the battery capacity is at least B0 of the nianuldcture- s rating when subjected to a perfornarice discharge test Once per 60 *nonth interval ttiis performance discharge test may be performed in lieu of the battery service test required by Specification 4.8:2jd.:
and
- f. At least once per 18 months, during shutdown, by giving performarce discharge tests of battery capacity to any battery that shows signs of degradation or has reac1ed 85 ot the service life expected for the application Degradation is indicated when the battery capacity arops more than 10 of rated capacity from its average on previous performance tests or is below 90 of the manufacturer s rating.
SHEARON HARRIS UNiT 1 314 8-13
2013 NRC SRO Question 88 (13) Reference TABLE 4.32 BATTERY SURVEILLANCE C
CATEGORY CATEGORY (2)
LINUS FOR EACH ALLOWA&E t3 DESIGNATED PILOT UNITS FOR EACH VALUE FOR EACH PARAMETER CELL CONNECTED CELL CONNECTED CELL Electrolyte )t4inimum level 44inin level Above top of Level indIcation mark, indication marks plates, S and < ? above and < 1/4 above arid not maximum level maximum level overflowing indication mark on mark 4
lndicat Float volta9eI> 2.13 volts > 2.13 volts > 2.07 volts Not re than 0.020 below the average of all Specific > 1.193 connected cells 4
aravity > I 20O
- Average of all Average of all connected cells connected cells
> 1.205
> L1.9S 5
TABLE NOTATIONS (1) For any Category A parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> all the Cate gory B measuresents are taken and found to be within their allowable values, aM provi dad all Category A and 3 p41aeter(s) are restored to within limits within the next 6 days.
(2) For any Category 3 parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that the Category B paraseten are within their allowable values arid provided the Category B parameter(s) are restored to within limits wIthin 7 days.
(3) Any Category B parameter not within its allowable value indicates an in operable battery.
(4) Corrected for electrolyte temceraturt and level.
(5) Or battery charging current is less than 2 amps when on charqe (5) Corrected for average electrolyte temperature.
SHEARON HARRIS - UNIT I 3/4 814
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2013 NRC SRO Question 92 (17) Reference Progress Energy REFEENCE HARRIS NUCLEAR PLANT PLANT OPERATING MANUAL VOLUME 2 PART 5 PROCEDURE TYPE: PLANT EMERGENCY PROCEDURE NUMBER:
PEP-hO TITLE:
Emergency Classification and Protective Action Recommendations PEP-lb Rev. 22 Page 1 of 31J
2013 NRC SRO Question 92 (17) Reference Table of Contents Section Page 1.0 PURPOSE 3
- 2. 0 INITIATING CONDITIONS 3
- 3. 0 GENERAL 4 3.1. General Guidelines for Use of the EAL Matrix 4 3.2. Specific Rules for Use of the EAL Matrix 6 3.3. Protective Action Recommendations (PARs) General Guidance 7
- 4. 0 PROCEDURE STEPS 10 4.1. Emergency Classification 10 4.2. Plant-based Protective Action Recommendations (PARs) 11 4.3. Dose Assessment Based Protective Action Recommendations (PARs) 13 4.4. Downgrading the Emergency Classification Level 15 4.5. Emergency Termination and Transition to Recovery 16
- 5. 0 REFERENCES 17 5.1. PLP-201, Emergency Plan 17 5.2. Referenced Plant Emergency Procedures 17 5.3. Other References 17
- 6. 0 SPECIAL TOOLS AND EQUIPMENT 18
- 7. 0 DIAGRAMS AND ATTACHMENTS 18 Attachment 1 Intentionally blank 19 Attachment 2 Intentionally blank 20 Attachment 3 Protective Action Recommendation Process 21 Attachment 4 Event Information Worksheet 24 Attachment 5 Termination Checklist 26 Attachment 6 Dose-Assessment-Based Protective Action Recommendations Background 28 PEP-lb Rev. 22 Page2of3l
2013 NRC SRO Question 92 (17) Reference 10 PURPOSE
- 1. The purpose of this procedure is to provide guidance on the use of Emergency Action Levels (EAL5) for classifying an emergency. This implements Section 4.1 of PLP-201.
- 2. This procedure provides guidelines for determining Protective Action Recommendations (PARs) to be made to offsite authorities during a General Emergency. This implements Section 4.5 of PLP-201.
- 3. This procedure provides guidance for summarizing events and actions taken during an event for use during facility turnover and facility briefings. This implements Section 2.3 of PLP2O1.
- 4. This procedure provides guidance for event termination and entry into Recovery.
This implements Section 6.7 of PLP-201.
2.0 INITIATING CONDITIONS
- 1. Entry into the Emergency Action Level (EAL) Matrix has been directed by any of the Emergency Operating Procedures, Fire Protection Procedures, Abnormal Operating Procedures, or any other procedure.
- 2. A Critical Safety Function Status Tree (CSFST) on the Safety Parameter Display System has produced a valid red or orange output and monitoring of the CSFSTs has been authorized in accordance with an approved procedure.
- 3. Notification has been received from a member of the Security Organization that a Security Condition, Threat, or Hostile Action has occurred.
- 4. Conditions exist which, in the judgment of the Shift Manager (SM), could be classified as an emergency.
PEP-hO Rev.22 Page3of3l
2013 NRC SRO Question 92 (17) Reference 3.0 GENERAL NOTE: The Current revision of the EAL Matrix is located in EP-EAL. Large print versions of the EAL Matrix are located in the Main Control Room, Technical Support Center and Emergency Operations Facility.
3.1. General Guidelines for Use of the EAL Matrix All emergency classifications shall be made within 15 minutes following indications that conditions have reached an EAL threshold, based upon valid indications, reports or conditions. An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.
- 2. Dose projections are used during the evaluation of EALs. When the dose assessment is complete the clock starts. This is the Run date/time on the Dose Assessment Summary Report. It is up to the Dose Assessment Team Leader, Dose Assessment Team, and the RCM to validate assumptions and results, and report those results to the ERM within this 15-minute period, so they may communicate to the SEC for emergency event classification.
- 3. The primary tool for determining the emergency classification level is the EAL Matrix.
EP-EAL, Emergency Action Levels is used in conjunction with this procedure and the EAL Matrix when classifying events. EP-EAL provides the technical basis and additional explanatory material to correctly classify events.
- 4. Many of the EALs derived from the NEI methodology are fission product barrier based. That is, the conditions that define the EALs are based upon loss or potential loss of one or more of the three fission product barriers. Loss and Potential Loss signify the relative damage and threat of damage to the barrier.
- a. Loss means the barrier no longer assures containment of radioactive materials.
- b. Potential loss infers an increased probability of barrier loss and decreased certainty of maintaining the barrier.
- 5. To the extent possible, the EALs are symptom-based. That is, the action level threshold is defined by values of key plant operating parameters that identify emergency or potential emergency conditions. This approach is appropriate because it allows the full scope of variations in the types of events to be classified as emergencies. However, a purely symptom-based approach is not sufficient to address all events for which emergency classification is appropriate. Particular events to which no predetermined symptoms can be ascribed have also been utilized as EALs since they may be indicative of potentially more serious conditions not yet fully realized.
PEP-i 10 Rev. 22 Page 4 of 31
2013 NRC SRO Question 92 (17) Reference 3.1 General Guidelines for Use of the EAL Matrix (continued)
- 6. The requirement is that emergency classifications are to be made as soon as conditions are present for the classification, but within 15 minutes in all cases of conditions present.
- 7. Where possible, the EALs have been made consistent with and utilize the conditions defined in the HNP Emergency Operating Procedure (EOP) network. While the symptoms that drive operator actions specified in the EOP5 are not indicative of all possible conditions which warrant emergency classification, they define the symptoms, independent of initiating events, for which reactor plant safety and/or fission product barrier integrity are threatened.
- 8. Planned evolutions to test, manipulate, repair, perform maintenance or modifications to systems and equipment that result in an EAL value being met or exceeded are not subject to classification and activation requirements as long as the evolution proceeds as planned and is within the operational limitations imposed by the specific operating license. However, these conditions may be subject to the reporting requirements of 10 CFR 50.72.
- 9. Existing guidance for classifying transient events addresses the period of time of event recognition and classification (15 minutes). However, in cases when EAL declaration criteria may be met momentarily during the normal expected response of the plant, declaration requirements should not be considered to be met when the conditions are a part of the designed plant response, or result from appropriate Operator actions.
- 10. There may be cases in which a plant condition that exceeded an EAL was not recognized at the time of occurrence but is identified well after the condition has occurred (e.g., as a result of routine log or record review), and the condition no longer exists. In these cases, an emergency should not be declared. Reporting requirements of 10 CFR 50.72 are applicable and the guidance of NUREG-1 022, Event Reporting Guidelines 10 CFR 50.72 and 50.73, should be applied.
ii. The highest emergency class for which an Emergency Action Level was exceeded shall be declared.
- a. Only one Emergency Action Level (EAL) classification shall be made at a time.
- b. If two EAL5 are clearly met, then choose the EAL of highest classification level as determined by the EAL matrix.
PEP-i 10 Rev. 22 Page 5 of 31
2013 NRC SRO Question 92 (17) Reference 3.1 General Guidelines for Use of the EAL Matrix (continued)
- 12. If the plant condition degrades and a higher classification emergency is declared before the notifications are made for the lesser emergency declaration, update the notification to reflect the higher emergency classification and complete the updated notifications within the 15 minutes of the lesser emergency declaration.
[RIS 2007-02]
- 13. If the notification cannot be updated and completed within 15 minutes of the lesser emergency declaration, make the notification for the lesser emergency declaration within 15 minutes of its declaration with a caveat that explains a change in classification is forthcoming. [RIS 2007-02]
- 14. In parallel, prepare the notification for the higher emergency classification and make the notification for the higher emergency classification within 15 minutes of the classification time of the higher emergency declaration. [RIS 2007-02]
3.2. Specific Rules for Use of the EAL Matrix
- 1. The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action is initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred.
- 2. For some events, the condition may be corrected before a declaration has been made. The key consideration in this situation is to determine whether or not further plant damage occurred while the corrective actions were being taken. In some situations, this can be readily determined, in other situations, further analyses may be necessary (e.g., coolant radiochemistry following an ATWS event, plant structural examination following an earthquake, etc.). Classify the event as indicated and terminate the emergency once assessment shows that there were no consequences from the event and other termination criteria are met.
- 3. Although the majority of the EALs provide very specific thresholds, the Site Emergency Coordinator (SEC) must remain alert to events or conditions that lead to the conclusion that exceeding the EAL threshold is imminent, If, in the judgment of the SEC, an imminent situation is at hand, the classification should be made as if the threshold has been exceeded.
- 4. The EAL Matrix should be read from left to right and top to bottom.
PEP-lb Rev.22 Page6of3l
2013 NRC SRO Question 92 (17) Reference 3.3. Protective Action Recommendations (PARs) General Guidance
Additionally, if in the opinion of the Emergency Response Manager, or the SEC-CR if the EOF is not yet activated, conditions warrant the issuance of PARs, a General Emergency will be declared (HNP will not issue PARs for any accident classified below a General Emergency).
- 2. PARs provided in response to a radioactive release include evacuation, taking shelter and consideration of the use of KI.
- a. Evacuation is the preferred action unless external conditions impose a greater risk from the evacuation than from the dose received.
- b. HNP personnel do not have the necessary information on external factors to determine whether offsite conditions would require sheltering instead of an evacuation. Therefore, an effort to base PARs on external factors (such as road conditions, traffic/traffic control, weather or offsite emergency worker response) should not be attempted.
- c. Sheltering may be an appropriate action for controlled releases of radioactive material from the containment, if there is assurance that the release is short term (puff release) and the area near the plant cannot be evacuated before the plume arrives.
- d. KI should be a recommendation if dose assessment or projection results indicate offsite radioactive iodine dose 5 Rem CDE to the adult thyroid.
- 3. At a minimum, a plant condition driven PAR to evacuate a 2 mile radius and 5 miles downwind, and shelter all other Subzones, is issued at the declaration of a General Emergency. Depending on plant conditions, evacuation of a 5 mile radius and 10 miles downwind, and shelter all other Subzones, may be issued instead of the minimum PAR.
- a. PARs are included with the initial and follow-up notifications issued at a General Emergency.
- b. The PAR must be provided to the State within 15 minutes of (1) the classification of the General Emergency or (2) any change in recommended actions.
- c. The PAR must be provided to the NRC as soon as possible and within 60 minutes of (1) the classification of the General Emergency or (2) any change in recommended actions.
PEP-lb Rev. 22 Page 7 of 31
2013 NRC SRO Question 92 (17) Reference 3.3 Protective Action Recommendations (PARs) General Guidance (continued)
- 4. The Emergency Response Manager, or the SEC-MCR if the EOF is not yet activated, may elect to specify PARs for any combinations of Subzones or the entire EPZ (or beyond) regardless of plant and dose based guidance.
- 5. Evacuation and shelter PARs should not be extended based on the results of dose projections unless the postulated release is likely to occur within a short period of time. Plant-based PARs are inherently conservative such that expanding the evacuation zone as an added precaution may result in a greater risk from the evacuation than from the radiological consequences of a release. It also would dilute the effectiveness of the offsite resources used to accommodate the evacuation.
- 6. Protective actions taken in areas affected by plume deposition following the release are determined and controlled by offsite governmental agencies.
- a. HNP is not expected to develop offsite recommendations involving ingestion or relocation issues following plume passage.
- b. HNP may be requested to provide resources to support the determination of post plume protective actions.
- 7. Throughout the duration of a General Emergency, assess plant conditions and effluent release status to ensure the established PARs are adequate.
- 8. The Site Emergency Coordinator (SEC) is the decision maker on determining if a radiological emergency release is in progress. An emergency release is defined as any unplanned quantifiable discharge of radioactive material to the environment that causes, or is due to, a declared emergency event. A radiological emergency release is in progress if:
- a. Any radiation monitor listed in Table R-1 of the EAL Matrix shows an increase in activity.
- b. Primary-to-secondary leakage causes an emergency declaration.
- c. A known unmonitored release path exists from an area that contains radioactive material.
- d. Environmental Monitoring Team surveys detect an increase in background radiation levels outside the site boundary
- e. Any alternate methods is used to determine a release is in progress.
EXAMPLE The Plant Vent Stack radiation monitor (RM-21AV-3509-1SA) is out of service and compensatory Lmeasures indicate a release is in progress.
PEP-lb Rev.22 Page8of3l
2013 NRC SRO Question 92 (17) Reference 3.3 Protective Action Recommendations (PARs) General Guidance (continued)
- 9. Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment Performance Indicator Guideline states: (1) Protective Action Recommendations (PARs) are made consistent with the goal of 15 minutes once data is available, and (2) Dose assessment and PAR development are expected to be made promptly following indications that the conditions have reached a threshold in accordance with the licensees PAR scheme. The 15 minute goal from data availability is a reasonable period of time to develop or expand a PAR. Plant conditions, meteorological data, field monitoring data, and/or radiation monitor data should provide sufficient information to determine the need to change PARs. If radiation monitor readings provide sufficient data for assessments, it is not appropriate to wait for field monitoring to become available to confirm the need to expand the PAR. The 15 minute goal should not be interpreted as providing a grace period in which the licensee may attempt to restore conditions and avoid making the PAR recommendation.
- a. Time is of the essence when conducting and approving dose projections.
Dose projection results may escalate or preclude emergency declarations.
- b. The clock starts when you have indications that a PAR threshold is exceeded. This could be radiation level readings via installed instrumentation (e.g., ERFIS, OSI/PI, local monitors, etc.), radiation level readings from field teams, or when you complete a dose assessment.
- c. When the dose assessment is complete the clock starts. This is the Run dateltime on the Dose Assessment Summary Report. It is up to the Dose Assessment Team Leader, Dose Assessment Team, and the RCM to validate assumptions, results, and recommend PARs for approval by the ERM within this 15-minute period.
PEP-lb Rev. 22 Page9of3l
2013 NRC SRO Question 92 (17) Reference 4.0 PROCEDURE STEPS 4.1. Emergency Classification NOTE: The expectation is that emergency classifications are to be made as soon as conditions are present for the classification, but within 15 minutes in all cases of conditions being present.
NOTE: The All Conditions EAL matrix must be evaluated for all plant conditions (hot or cold).
NOTE: Use a marker on the EAL matrix to aid in place-keeping and EAL applicability.
CAUTION The highest emergency classification for which an Emergency Action Level (EAL) was exceeded shall be declared.
- 1. EVALUATE the All Conditions EAL Matrix.
- a. READ the EAL Matrix from left to right and top to bottom
- b. READ the EAL Category
- c. READ the EAL subcategory
- d. READ the Initiating Condition
- e. READ the Mode Applicability bar
- f. READ the category number criterion
- g. READ any applicable notes or tables
- h. DETERMINE EAL classification threshold applicability
- 2. IF the Reactor Coolant System temperature is greater than 200°F, THEN EVALUATE the Hot Conditions EAL Matrix.
- a. READ the EAL Matrix from left to right and top to bottom
- b. READ the EAL Category
- c. READ the EAL subcategory
- d. READ the Initiating Condition
- e. READ the Mode Applicability bar
- f. READ the category number criterion
- g. READ any applicable notes or tables
- h. DETERMINE EAL classification threshold applicability PEP-lb Rev. 22 Page lOof 31
2013 NRC SRO Question 92 (17) Reference 4.1 Emergency Classification (continued)
- 3. IF the Reactor Coolant System temperature is less than or equal to 200°F, THEN EVALUATE the Cold Conditions EAL Matrix.
- a. READ the EAL Matrix from left to right and top to bottom
- b. READ the EAL Category
- c. READ the EAL subcategory
- d. READ the Initiating Condition
- e. READ the Mode Applicability bar
- f. READ the category number criterion
- g. READ any applicable notes or tables
- h. DETERMINE EAL classification threshold applicability
- 4. IDENTIFY the highest applicable emergency classification level.
- 6. IMPLEMENT requirements in PEP-230 and/or PEP-240, as appropriate.
4.2. Plant-based Protective Action Recommendations (PARs)
- 1. Use Attachment 3, Protective Action Recommendation Process as an aid in determining the proper PAR.
- 2. At a minimum, evacuation of a 2 mile radius and 5 miles downwind (with sheltering of all other Subzones) will be recommended for a General Emergency declaration.
- 3. Evacuation of a 5 mile radius and 10 miles downwind (with sheltering of all other Subzones) will be recommended for plant conditions in which damage is imminent or has occurred for all three fission product barriers as indicated by all three conditions below (a., b. and c.):
- a. Substantial core damage is imminent or has occurred as indicated by any of the following conditions:
(1) Core damage estimations >1% melt.
(2) Core Exit Thermocouple readings 2300° F.
(3) Core uncovered > 30 minutes.
PEP-lb Rev. 22 Page 11 of 31
2013 NRC SRO Question 92 (17) Reference 4.2 Plant-based Protective Action Recommendations (PARs) (continued)
- b. A significant loss of reactor coolant is imminent or has occurred are indicated by any of the following conditions:
(1) Containment Radiation Monitors reading:
- >10,000 R/Hr with no containment spray.
- >4,000 R/Hr with containment spray on.
(2) Containment hydrogen gas concentration >1%.
(3) Rapid vessel depressurization.
(4) A large break loss of coolant accident.
- c. Containment Barrier failure (primary or SIC) is imminent or has occurred as indicated by:
(1) A release of radioactivity greater than the projected dose of either:
- 1000 mRem TEDE at or beyond the site boundary.
- 5000 mRem Thyroid CDE at or beyond the site boundary.
OR a measured dose rate of either:
- >1000 mRemlhr at or beyond the site boundary.
- 1-131 equivalent concentration> 3.9 E-6 pCi/cc at or beyond the site boundary.
(2) Primary containment pressure can not be maintained below design basis pressure of 45 psig.
(3) Primary containment H 2 gas concentration can not be maintained below combustible limit of 4% by volume.
(4) Faulted/Ruptured S/C with a relief valve open.
- 4. Containment monitors may provide indication of both core damage and loss of RCS.
Monitor values used to determine a specific amount of core damage are dependent on plant conditions, power history, and time after shutdown. Monitor readings used to quantify an amount of damage or coolant leakage should be complimented by other indications and engineering judgment.
PEP-lb Rev. 22 Page 12 of 31
2013 NRC SRO Question 92 (17) Reference 4.2 Plant-based Protective Action Recommendations (PARs) (continued)
- 5. Acceptable changes in initial PARs includes expanding evacuation but does not allow a change from evacuation of zones to sheltering of those zones.
NOTE: A direct release is defined as a pathway from the containment to any environment outside the containment when containment or system isolation is required due to a safety injection signal, containment pressure greater than 3 psig, or a valid containment ventilation isolation signal and the pathway cannot be isolated from the Main Control Room.
- 6. If a release is in progress:
- a. Perform dose assessment as soon as possible to determine if PAGs are exceeded and if additional Subzones require evacuation. Add any Subzones requiring evacuation as determined by dose assessment to the plant-based PARs.
- 7. If no release is in progress:
- a. Perform dose projections on possible conditions as time permits to determine if PAGs could be exceeded. Consider adding any Subzones requiring evacuation as determined by dose projection to the plant-based PARs.
- 8. If either the dose assessment or dose projection indicate that the KI PAG (5 REM CDE to the adult thyroid) is or could be exceeded, then the KI consideration PAR should be added (line 5D on ENF).
4.3. Dose Assessment Based Protective Action Recommendations (PARs)
NOTE: Dose projections are not required to support the decision process in Attachment 3, Protective Action Recommendation Process.
NOTE: Many assumptions exist in dose assessment calculations, involving both source term and meteorological factors, which make computer predictions over long distances highly questionable.
- a. Issue an initial ENF to state and Counties that include a statement similar to the following:
Dose assessment results indicate PAGs are exceeded X miles from the Harris Nuclear Plant. Environmental Monitoring Teams have been dispatched to verify dose assessment results.
- b. Dispatch Environmental Teams to downwind areas to verify the calculated exposure rates.
PEP-lb Rev. 22 Page l3of 31
2013 NRC SRO Question 92 (17) Reference 4.3 Dose Assessment Based Protective Action Recommendations (PARs) (continued)
- c. IF the dose assessment data is verified, THEN issue an initial ENF to State and Counties that includes a statement similar to the following:
Environmental Monitoring Teams have verified PAGs are exceeded X miles from the Harris Nuclear Plant. Recommend expanding evacuation zones X miles downwind from the plant.
- d. IF dose assessment data is NOT verified, THEN issue a follow up ENF to State and Counties that includes a statement similar to the following:
Environmental Monitoring Teams were unable to verify PAG5 are exceeded beyond the 10 mile Emergency Planning Zone. No additional protective actions are recommended at this time.
NOTE: Refer to Attachment 6, DOSE-ASSESSMENT-BASED PROTECTIVE ACTION RECOMMENDATIONS BACKGROUND for background information on Dose Assessment Based PARs.
- 2. From the Main Control Room: If a release is in progress and time permits, perform offsite dose assessment in accordance with PEP-340 to determine whether the plant-based protective actions of Attachment 3 are adequate.
- 3. From the Emergency Operations Facility: Conduct offsite dose assessment in accordance with EMG-NGGC-0002 to determine whether the plant-based protective actions of Attachment 3 are adequate using the following methods as applicable:
- a. Monitored Release:
(1) If a release is in progress, assess the calculated impact to determine whether the plant-based PARs of Attachment 3 are adequate.
(2) If a release is not in progress, use current meteorological and core damage data to project effluent monitor threshold values which would require 2, 5, and 10 mile evacuations (Attachment 3). Reestablish threshold values whenever meteorological conditions or core damage assessment values change.
PEP-hO Rev. 22 Page l4of 31
2013 NRC SRO Question 92 (17) Reference 4.3 Dose Assessment Based Protective Action Recommendations (PARs) (continued)
- b. Conta,nment Leakage/Failure:
(1) If a release is in progress, assess the calculated impact to determine whether the plant-based PARs of Attachment 3 are adequate.
(2) If a release is not in progress, use current meteorological and core damage data on various scenarios (design leakage, failure to isolate, catastrophic failure) to project the dose consequences.
(3) Determine whether the plant-based PARs of Attachment 3 are adequate.
(4) Reestablish scenario values whenever meteorological conditions or core damage assessment values change.
- c. Field Survey Analysis: Actual field readings from Environmental Teams should be compared to dose assessment results and used as a dose projection method to validate calculated PARs and to determine whether the plant or release based protective actions of Attachment 3 are adequate.
- d. Release Point Analysis: Actual sample data from monitored or unmonitored release points should be utilized in conjunction with other dose assessment and projection methods to validate calculated PARs and to determine whether the plant-based protective actions of Attachment 3 are adequate.
- 4. The Emergency Response Manager and the Radiological Control Manager shall discuss dose assessment and projection analysis results and evaluate their applicability prior to issuing PARs to the State if possible.
4.4. Downgrading the Emergency Classification Level NOTE: The preferred method during plant recovery concerning EALs is to terminate the declared event when the plant has recovered from the effects of the initiating events rather than reducing the EAL level as recovery is completed. It is not required that emergency declarations be reduced and lower EALs declared as plant conditions improve.
If the action level currently has abated to a lower declaration or the situation has been resolved prior to completion of off-site reporting:
- a. Declare the highest classification for which an Emergency Action Level was exceeded, if not already done, and
- b. Evaluate downgrading to the emergency classification appropriate for the present conditions.
PEP-lb Rev.22 Pagel5of3l
2013 NRC SRO Question 92 (17) Reference 4.4 Downgrading the Emergency Classification Level (continued)
- 2. Downgrading of an emergency is performed by issuing a notification to a lower emergency classification level whenever plant conditions improve to satisfy the affected Emergency Action Levels. However, the following guidelines apply:
- a. If the Emergency Response Manager (ERM) position is activated, they shall be consulted before downgrading occurs.
- b. If the NRC Director of Site Operations position is activated, they should be consulted before downgrading occurs.
- c. If offsite Protective Action Recommendations have been made, the SEC-TSC shall consult with the ERM and with State and County authorities, prior to downgrading. It is recommended that any off-site Protective Action Recommendations be completed prior to downgrading of a General Emergency.
- d. Where lasting damage has occurred to the fission product barriers or to safety systems, the ERM should transition to PEP-500 rather than a simple downgrade of the emergency.
- e. For Alert or higher classifications, unless the conditions causing emergency action levels are very quickly resolved (less than approximately 30 minutes),
downgrading should not occur until after the TSC and EOF are activated.
4.5. Emergency Termination and Transition to Recovery
- 1. If entering Recovery from an Unusual Event, determine the need for a Recovery Plan and support organization.
- a. Generally, the activities following an Unusual Event will not require the formation of a Recovery Organization or a transition period prior to event termination and entry into Recovery.
- b. Refer to PEP-500 for further guidance if recovery efforts following an Unusual Event extend beyond offsite notification and the generation of required reports.
- 2. Complete the Termination Checklist (Attachment 5).
- a. If conditions will allow for the termination of the emergency and entry into Recovery, exit this procedure and enter PEP-500, Recovery.
- b. If conditions do no support termination of the emergency and entry into Recovery, continue following the guidance provided in Section 3.1.
PEP-lb Rev.22 Pagel6of3l
2013 NRC SRO Question 92 (17) Reference
5.0 REFERENCES
5.1. PLP-201, Emergency Plan
- 1. Section 4.1, Emergency Classification
- 2. Section 4.5.1, Protective Action Guides
- 52. Referenced Plant Emergency Procedures
- 1. PEP-230, Control Room Operations
- 2. PEP-240, Activation and Operation of the Technical Support Center
- 3. PEP-270, Activation and Operation of the emergency Operations Facility
- 4. PEP-31 0, Notifications and Communications
- 5. PEP-344, HNP Offsite Dose Assessment Based on Monitored Releases
- 6. PEP-500, Recovery 5.3. Other References
- 1. EMG-NGGC-0002, Off-site Dose Assessment
- 2. State of North Carolina Radiological Emergency Response Plan for Nuclear Power Facilities
- 3. EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents
- 4. NUREG-0654 Supplement 3, Criteria for Protective Action Recommendations for Severe Accidents
- 5. NUREG-1 022, Event Reporting Guidelines: 10 CFR 50.72 and 50.73
- 6. NUREG/BR-0150, Vol. 4, Rev.4, US NRC, RTM-96 Response Technical Manual
- 7. Regulatory Guide 1.101 Emergency Planning and Preparedness for Nuclear Power Plants
- 8. EPPOS No.1 Emergency Preparedness Position (EPPOS) on Acceptable Deviations to Appendix ito NUREG-0654/FEMA-REP-i
- 9. Harris Nuclear Plant Development of Evacuation Time Estimates, KLD Associates Final Report August 23, 2007 PEP-lb Rev.22 Pagel7of3l
2013 NRC SRO Question 92 (17) Reference 5.3 Other References (continued)
- 10. NRC Bulletin 2005-02, Emergency Preparedness and Response Actions for Security-Based Events
- 11. EP-EAL, Emergency Action Levels
- 12. NEI 1999-02, Regulatory Assessment Performance Indicator Guideline
- 13. NRC Regulatory Issue Summary 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events.
6.0 SPECIAL TOOLS AND EQUIPMENT
- 2. PAR Boards: PAR boards, based on Attachment 3, are maintained in the Main Control Room, TSC and EOF
- 3. EP-EAL: Copies of the Emergency Action Levels are maintained in the Main Control Room, TSC and EOF 7.0 DIAGRAMS AND ATTACHMENTS See Table of Contents.
PEP-lb Rev. 22 Page l8of 31
2013 NRC SRO Question 92 (17) Reference Attachment I Intentionally blank Sheet I of 1 PEP-hO Rev.22 Pagel9of3l
2013 NRC SRO Question 92 (17) Reference Attachment 2 Intentionally blank Sheet 1 of 1 PEP-lb Rev. 22 Page 20 of 31
2013 NRC SRO Question 92 (17) Reference Attachment 3 Protective Action Recommendation Process Sheet 1 of 3 General Emergency No No PARs Required Declared?
Yes Does an approved dose projection or assessment Yes Recommend consideration indicate 5 REM Adult Thyroid ODE? of the use of KI 1,4a No 14 1
Substantial core damage is imminent or has occurred? No Yes 2,4b
[
A significant loss of reactor coolant is imminent or has No occurred?
Yes 3
Containment barrier failure (Primary or SIG) is No imminent or has occurred?
4, Yes Evacuate 5 Mile Radius and 10 Miles Evacuate 2 Mile Radius and 5 Miles Downwind. Downwind.
Shelter all other Subzones. Shelter all other Subzones.
Refer to PEP-ho, Section 3.3 if a short Refer to PEP-hO, Section 3.3 if a short puff release is anticipated or external puff release is anticipated or external conditions impose a greater risk from the conditions impose a greater risk from the evacuation than from the dose received. evacuation than from the dose received.
4, 4, 5 Mile Radius and 10 Miles Downwind 2 Mile Radius and 5 Miles Downwind.
Wind Direction Evacuate Shelter Wind Direction Evacuate Shelter (From 0) Subzones Subzones (From 0) Subzones Subzones 348° 0100 A,B,C,D,H,I,K,L E,F,G,J,M,N 327° 010° A,D,K
- B,C,E,F,G,H,I,J,L,M,N 0110 0340 A,B,C,D,H,l,K,L E,F,G,J,M,N 011° - 056° A,K B,C,D,E,F,G,H,I,J,L,M,N 035° - 079° A,B,C,D,I,J,K,L,M E,F,G,H,N 057° - 124° A,K,L B,C,D,E,F,G,H,I,J,M,N 080° - 101° A,B,C,D,J,K,L,M E,F,G,H,I,N 125° - 191° A,B,L C,D,E,F,G,H,I,J,K,M,N 102° - 124° A,B,C,D,K,L,M E,F,G,H,I,J,N 192° - 214° A,B C,D,E,F,G,H,I,J,K,L,M,N 125° - 146° A,B,C,D,K,L,M,N E,F,G,H,I,J 215° - 259° A,B,C D,E,F,G,H,I,J,K,L,M,N 147° - 191° A,B,C,D,E,K,L,M,N F,G,H,I,J 260° - 281° A,C B,D,E,F,G,H,I,J,K,L,M,N 192° - 214° A,B,C,D,E,K,L F,G,H,I,J,M,N 282° - 304° A,C,D B,E,F,G,H,I,J,K,L,M,N 215° - 236° A,B,C,D,E,K,L F,G,H,I,J,M,N 305° - 326° A,D B,C,E,F,G,H,I,J,K,L,M,N 237° - 259° A,B,C,D,E,F,K,L G,H,I,J,M,N 260° - 326° A,B,C,D,F,G,H,K,L E,I,J,M,N 327° - 347° A,B,C,D,H,K,L E,F,G,I,J,M,N PEP-lb Rev. 22 Page 21 of 31
2013 NRC SRO Question 92 (17) Reference Attachment 3 Protective Action Recommendation Process Sheet 2 of 3 Acceptable changes in initial PARS would include expanding evacuation but would not allow a change from evacuation of zones to sheltering of those zones.
Indications that substantial core damage is imminent or has occurred include:
a) Core damage> 1% melt.
b) Core Exit Thermocouple readings 2300° F.
c) Core uncovered > 30 minutes.
- 2. Indications that a significant loss of reactor coolant is imminent or has occurred include:
a) Containment radiation reading> 10,000 R/Hr without spray or >4,000 R/Hr with spray.
b) Containment hydrogen gas concentration> 1%.
c) Rapid vessel depressurization.
d) A large break loss of coolant accident.
- 3. Indications that containment barrier failure (primary or SIG) is imminent or has occUrred are indicated by:
a) A release of radioactivity greater than the projected dose of either:
- 1000 mRem TEDE at or beyond the site boundary.
- 5000 mRem Thyroid CDE at or beyond the site boundary.
Or a measured dose rate of either:
- >1000 mRem/hr at or beyond the site boundary.
- 1-131 equivalent concentration > 3.9 E-6 pCi/cc at or beyond the site boundary.
b) Primary containment pressure can not be maintained below design basis pressure of 45 psig.
c) Primary containment H 2 gas concentration can not be maintained below combustible limit of 4% by volume.
d) Faulted/Ruptured S/G with a relief valve open.
NOTE: A direct release is defined as a pathway from the containment to any environment outside the containment when containment or system isolation is required due to a safety injection signal, containment pressure greater than 3 psig, or a valid containment ventilation isolation signal and the pathway cannot be isolated from the Main Control Room.
- 4. Accidents which result in a direct release pathway to the environment will most likely be thyroid dose limiting. For a faulted and ruptured SIG, water level must be below the tube bundles (S/G Narrow Range <25% normal containment conditions or < 40% adverse containment conditions) with a relief valve open before it is considered a direct release pathway to the environment. For circumstances involving a direct release pathway to the environment:
a) Consider any loss of Fuel sufficient to warrant the determination that substantial core damage has occurred.
b) Consider any loss of RCS sufficient to warrant the determination that a significant loss of reactor coolant has occurred.
- 5. PARs due to Spent Fuel Pool releases are determined using Attachment 6, Dose Assessment Based Protective Action Recommendations.
- 6. Containment monitors can provide indication of a loss or potential loss of both core damage and loss of RCS.
Monitor readings used to quantify an amount of damage or coolant leakage should be complimented by other indications and engineering judgment.
PEP-i 10 Rev. 22 Page 22 of 31
2013 NRC SRO Question 92 (17) Reference Attachment 3 Protective Action Recommendation Process Sheet 3 of 3 If a release is in progress:
- Perform dose assessment as soon as possible to determine if PAGs are exceeded and if additional Subzones require evacuation.
- Add any Subzones requiring evacuation as determined by dose assessment to the plant-based PARs.
If no release is in progress:
- Perform dose projection on possible conditions as time permits to determine if PAGs could be exceeded.
Consider adding any Subzones requiring evacuation as determined by dose projection to the plant-based PARs.
PEP-i 10 Rev. 22 Page 23 of 31
2013 NRC SRO Question 92 (17) Reference Attachment 4 Event Information Worksheet Sheet 1 of 2 Date/Time: (Use ERFIS time)
EVENT INFORMATION WORKSHEET A) Emergency Classification D) Radiological Release Time Declared: (24 hr) Li None Li Controlled Li Unusual Event Li Alert Li Is Occurring Li Uncontrolled LI Site Area LI General Li Has Occurred Li Below PAGs Provide a brief summary of the event and LI Above PAGs mitigating actions in progress:
Time Started:
(24 hr)
EAL:
Noble Gas: Ci/sec lodines: Ci/sec Projected Duration: hours Environmental Monitoring Team activities:
B) Fission Product Barrier Status E) Personnel Status Fu& RCS Cnmt Missions in plant: Li No Li Yes Intact: Li Li Li Location of in-plant teams/personnel:
Potential Loss: Li Li Li Loss: Li Li Li Injuries (No. ): Li No Li Yes C) Plant Conditions Contamination(s): Li No Li Yes Li On-Line Over Exposure(s): Li No Li Yes Li At Power: %
LI Minor Li Major Li Off-Line Li Cooling Down Details (names of injured, status of family notification):
LI Cold Shutdown Time of Rx Shutdown: (24 hr)
Li Stable Li Improving F) Facility Activation Status Li Degrading Li TSC: (24 hr)
Describe plant and recent activities LI OSC: (24 hr)
Li EOF: (24 hr)
Li JIC: (24 hr)
If TSC is not ready for activation can the TSC Describe equipment, instrument, or other accept responsibility for:
problems: Notification to NRC: Li N/A Li No Li Yes If EOF is not yet ready for activation can the EOF accept responsibility for:
Emergency Communicator Communications to ERDS Status: On-Line Li Off-Line State and Counties (ENF must still be approved ERFIS Status: J On-Line Li Off-Line by SEC) Li No Li Yes OSI/PI Status: J On-Line Li Off-Line Dose Assessment Li No Li Yes PEP-hO Rev. 22 Page24 of 31
2013 NRC SRO Question 92 (17) Reference Attachment 4 - Event Information Worksheet Sheet 2 of 2 EVENT INFORMATION WORKSHEET G) Offsite Assistance Requested J) PARs Li None Li None Issued, or Li Medical (24 hr) OEvac: ABCDEFGHIJKLMN O Ambulance O Helicopter OShelter: ABCDEFGHIJKLMN Li Fire Department (24 hr) (Circle the affected subzones)
O Holly Springs O Apex o Consideration of the use of KI Li Law Enforcement (24 hr) K) Offsite Facility Activation Status O Local O State Li Chatham County EOC: (24 hr)
H) Onsite Protective Actions Li Harnett County EOC: (24 hr)
Li None Li Lee County EOC: (24 hr)
Li Assembly/Accountability Li Wake County EOC: (24 hr)
Li Local Area(s) Evacuated Li State EOC: (24 hr)
Li Protected Area Evacuated Li NRC Incident Response Center (24 hr)
Li Exclusion Area Evacuated L) Offsite ActionslResponse Li Potassium Iodide Issued Li None Issued, or Li Employee Info Phone #: O Schools O Daycare I) Offsite Notifications (last issued) O Hospitals O Rest Homes State/County Time: (24 hr) O Lake Evacuations NRC Time: (24hr) o Other:
News Release Time: (24 hr)
Hospital Time: (24 hr) OEvac: ABCDEFGHIJKLMN INPO Time: (24hr) OShelter: ABCDEFGHIJKLMN ANI Time: (24 hr) (Circle the affected subzones) 0 KI administered to the General Public Li Sirens Activated: (24 hr)
Li Tone Alerts Activated: (24 hr)
Li EAS Activated: (24 hr)
Any applicable incomplete items from previous pages of PEP-230, Attachment 1 - SITE EMERGENCY COORDINATOR - CR checklist?
Any assistance needed?
Comments PEP-lb Rev.22 Page25of3l
2013 NRC SRO Question 92 (17) Reference Attachment 5 Termination Checklist Sheet 1 of 2 TERMINATION CHECKLIST True False
- 1. Conditions no longer meet an Emergency Action Level and it appears Li unlikely that conditions will deteriorate.
List any EAL(s) which is/are still exceeded and a justification as to why a state of emergency is no longer applicable:
- 2. Plant releases of radioactive materials to the environment are under control (within Tech Specs) or have ceased and the potential for an uncontrolled radioactive release is acceptably low.
- 3. The radioactive plume has dissipated and plume tracking is no longer J LI required. The only environmental assessment activities in progress are those necessary to determine the extent of deposition resulting from passage of the plume.
- 4. In-plant radiation levels are stable or decreasing, and acceptable given the plant conditions.
- 5. The reactor is in a stable shutdown condition and long-term core cooling is available.
- 6. The integrity of the Reactor Containment Building is within Technical Specifióation limits.
- 7. The operability and integrity of radioactive waste systems, decontamination facilities, power supplies, electrical equipment and plant instrumentation including radiation monitoring equipment is acceptable.
- 8. Any fire, flood, earthquake or similar emergency condition or threat to security no longer exists.
PEP-hO Rev.22 Page26of3l
2013 NRC SRO Question 92 (17) Reference Attachment 5 Termination Checklist Sheet 2 of 2 TERMINATION CHECKLIST True False 9 Any contaminated injured person has been treated and/or transported to a medical care facility.
- 10. All required notifications have been made. 1J U
- 11. The NRC Senior Resident Inspector has been notified that the event will be U U terminated.
- 12. Offsite conditions do not unreasonably limit access of outside support to the 1J UI station and qualified personnel and support services are available.
- 13. Discussions have been held with Federal, State and County agencies and agreement has been reached and coordination established to terminate the emergency.
It is not necessary that all responses listed above be TRUE; however, all items must be considered prior to event termination and entry into Recovery.. For example, it is possible that some conditions remain which exceed an Emergency Action Level following a severe accident but entry into Recovery is appropriate. Additionally, other significant items not included on this list may warrant consideration such as severe weather.
Comments:
Approved:
Site Emergency Coordinator Date Time PEP-hO Rev. 22 Page 27 of 31
2013 NRC SRO Question 92 (17) Reference Dose-Assessment-Based Protective Action Recommendations Background Sheet 1 of 2 Protective Action Guides
- The evacuation of the general public will usually be justified when the projected TEDE dose to an individual is one Rem or greater or the projected CDE thyroid dose is five Rem or greater.
- 2. EPZSubzones
- The objective of the dose assessment calculations is to allow for the determination of protective actions. Protective actions may affect any portion of or the entire Emergency Planning Zone (EPZ).
- a. The EPZ extends out to ten miles from the plant. The EPZ is then divided radially into three rings (0-2, 2-5, and 5-10 miles) and axially into sixteen 22.5° sectors. This allows for the implementation of protective actions within specifically affected areas, or Key Holes rather than across the entire EPZ.
- b. Subzones are used to define the areas within the Harris EPZ. Radial distances are maintained approximately equal to the standard EPZ, but axial sector based areas have been abandoned. The fourteen Subzones which make up the HNP EPZ are divided using geopolitical, natural, and man-made boundaries.
- c. The Subzones are divided into three groups.
(1) Subzone A encompasses the inner ring which extends out to approximately two miles from the plant.
(2) Subzones B, C, 0, K, and L compose the middle ring, at about two to five miles from the plant.
(3) Subzones E, F, G, H, I, J, M, and N make up the outer ring, five to ten miles from the plant.
- 3. Subzone Evacuation Groups
- a. The combination of Subzones which compose a group was determined as follows:
PEP-lb Rev. 22 Page28 of 31
2013 NRC SRO Question 92 (17) Reference Dose-Assessment-Based Protective Action Recommendations Background Sheet 2 of 2
- b. For any given wind direction a combination of Subzones will be affected by the plume. By defining the maximum horizontal dispersion at ten miles in the crosswind axis for a plume of stability class A (most unstable) and solving for the angle a plume with a 320 footprint is created.
= (-0.0234
- ln(x) + 0.350)x = 2.96 miles tan(x) = ,a 5
Ix = 16°
- c. The wind direction band that will affect each Subzone is ascertained by transcription of worse case plume onto a United States Geological Survey map. By accounting for the overlapping of wind directions, fifteen distinct Subzone combinations (groups) are established.
- d. Taking into account that not all Subzones out to the EPZ boundary need be evacuated in all cases, additional subarea groups are generated. From this, twenty five possible evacuation combinations exist for all possible wind directions. A further adjustment was included to align the sub-area groups to agree with the ETE data. The combinations of subarea groups are as follows:
Evacuation SubzonelGroups W.D. 0-2 Miles 0-5 Miles 0-10 Miles (0
From) Sub-zone Group Sub-zones Group # Sub-zones Group 01i°-034° A 1 A, K 2 A, K, H, I, J 11 035°-056° A 1 A, K 2 A, K, I, J, M 12 057°-079° A 1 A, K, L 3 A, K, I, J, L, M 13 080°-i01° A 1 A, K, L 3 A, K, J, L, M 14 102°-i 24° A 1 A, K, L 3 A, K, J, L, M, N 15 125°-146° A 1 A, B, L 4 A, B, L, M, N 16 147°-191° A 1 A, B, L 4 A, B, E, L, M, N 17 192°-214° A 1 A, B 5 A, 8, E, N 18 215°-236° A 1 A, B, C 6 A, B, C, E, F 19 237°-259° A 1 A, B, C 6 A, B, C, E, F, G 20 26002810 A 1 A, B, C, D 7 A, B, C, D, F, G, H 21 282°-304° A 1 A, C, D 8 A, C, D, F, G, H 22 305°-326° A 1 A, C, D, K 9 A, C, D, F, G, H, K 23 327°-347° A 1 A, D, K 10 A, D, G, H, I, K 24 348°-010° A 1 A, D, K 10 A, D, H, I, K 25 PEP-i 10 Rev. 22 Page 29 of 31
2013 NRC SRO Question 92 (17) Reference Revision Summary Revision 22 Summary Rev. 22 processed with PRR: 409023 PRRs Incorporated: 409023 CRs Incorporated: 501711-05 (CORR), 569001-18 (ENHN), CR 589380-05 (CORR)
ECs Incorporated: NA 3.3 Step 8 [CR 501711-05, CR 569001-18]
From: The Site Emergency Coordinator (SEC) is the decision maker on determining if an emergency release (radioactive) is in progress. An emergency release is defined as any unplanned quantifiable discharge to the environment of radioactive effluent attributable to a declared emergency event. To assist in this determination, the following are gaseous and liquid release in-progress definitions:
- a. A gaseous (airborne) emergency release (radioactive) is in progress if any of the following conditions exist (1) An approved monitored release was occurring AND the reading on the radiation monitor designated to monitor this release increases due to the emergency event.
(2) Any release due to the emergency event that was not previously approved.
(3) Any primary-to-secondary leak which causes an emergency declaration.
- b. A liquid emergency release (radioactive) is in progress if any of the following conditions exist:
(1) An approved monitored release was occurring AND the reading on the radiation monitor designated to monitor this release increases but does not isolate on an alarm signal (2) The rupture of a system, that releases radioactive liquids into an area which affects or has the potential to affect an offsite environment.
- c. A direct release is defined as a pathway from the containment to any environment outside the containment when containment or system isolation is required due to a safety injection signal, containment pressure greater than 3 psig, or a valid containment ventilation isolation signal and the pathway cannot be isolated from the Main Control Room.
To: The Site Emergency Coordinator (SEC) is the decision maker on determining if a radiological emergency release is in progress. An emergency release is defined as any unplanned quantifiable discharge of radioactive material to the environment that causes, or is due to, a declared emergency event. A radiological emergency release is in progress if:
- a. Any radiation monitor listed in Table R-1 of the EAL Matrix shows an increase in activity.
- b. Primary-to-secondary leakage causes an emergency declaration.
- c. A known unmonitored release path exists from an area that contains radioactive material from commercial nuclear power plant operations.
- d. Environmental Monitoring Team surveys detect an increase in background radiation levels outside the site boundary
- e. Any alternate methods is used to determine a release is in progress.
EXAMPLE The Plant Vent Stack radiation monitor (RM-21AV-3509-1SA) is out of service and compensatory measures indicate a release is in progress.
4.2 step 3.c, From: Containment failure (primary or SIG) . . . sh 1 To: Containment barrier failure (primary or S/G) .
flowchart, and sh 2 step 3 PEP-lb Rev. 22 Page 30 of 31
2013 NRC SRO Question 92 (17) Reference Revision 22 Summary 4.2 step 4, and [PRR 409023, CR 501711-05] sh 2 Changed first sentence step 6 From: Containment monitors may provide indication of both core damage and RCS breach To: Containment monitors may provide indication of both core damage and loss of RCS.
4.3 step 1 [CR 589380-05]
From: In the event dose assessment results indicate the need to recommend actions beyond the outer EPZ boundaries that is past 10 miles:
Dispatch Environmental Teams to downwind areas to verify the calculated exposure rates prior to issuing PARs outside the EPZ.
Many assumptions exist in dose assessment calculations, involving both source term and meteorological factors, which make computer predictions over long distances highly questionable.
To:
NOTE: Many assumptions exist in dose assessment calculations, involving both source term and meteorological factors, which make computer predictions over long distances highly questionable.
IF dose assessment results exceed PAGs at the outer boundary of the 10 mile EPZ, THEN:
- a. Issue an initial ENF to state and Counties that include a statement similar to the following:
Dose assessment results indicate PAGs are exceeded X miles from the Harris Nuclear Plant. Environmental Monitoring Teams have been dispatched to verify dose assessment results.
- b. Dispatch Environmental Teams to downwind areas to verify the calculated exposure rates.
- c. IF the dose assessment data is verified, THEN issue an initial ENF to State and Counties that includes a statement similar to the following:
Environmental Monitoring Teams have verified PAGs are exceeded X miles from the Harris Nuclear Plant. Recommend expanding evacuation zones X miles downwind from the plant.
- d. IF dose assessment data is NOT verified, THEN issue a follow up ENF to State and Counties that includes a statement similar to the following:
Environmental Monitoring Teams were unable to verify PAGs are exceeded beyond the 10 mile Emergency Planning Zone. No additional protective actions are recommended at this time. sh 2 [PRR 409023, CR 501711-05]
step 4.a From: Consider any Fuel Breach sufficient...
To: Consider any loss of Fuel sufficient... sh 2 [PRR 409023, CR 501711-05]
step 4.b From: Consider any RCS Breach sufficient...
To: Consider any loss of RCS sufficient sh 2 [PRR CR 501711-05] new step step 5 PARs due to Spent Fuel Pool releases are determined using Attachment 6, Dose Assessment Based Protective Action Recommendations.
Throughout Incorporated formatting and word processing features, such as consistent use of auto step numbering, indentations, note boxes and cross referencing. These are editorial corrections per PRO-NGGC-0204 and do not need to be addressed further.
PEP-lb Rev. 22 Page 31 of 31
2013 NRC SRO Question 93 (18) Reference 2haron IaEri riuo1ar war 1art (L]fo )4y OCI.
Off si:e Dose Cal2ulaticn Nanual DDCM ]ev 1 D .. IPTftUNTAT:ON 2/ .3.3 INLTCRIMG IN .t11EITAI O 3/4.3.3.10 adioactiv Liquid Effueat 1onioring Inrition OEE1tI0{TL ItEU :p.irr
- 2. .310 The radio ive liqdd efE].uent nonitoring ir.etnimentation channels shown in rafle 3.-12 ehail be OPELE with their Alarm/Trip etpo.nt set to t1t tht liTait of cprtion. tq1iirrnxLt 3.11. a.i re nt exceeded. The Alarit/Trip $etpcints of these channels shall be detrnned and aduated in aeccrdance with the uetoth].cgy and erametere i the OFP3tI5 DOSE C C3L?.TI0t IL.WLL oDct)
APFLICAPILITI At all. times.
- a. cith a radioactive liquid effluent monitorina instinentation chanr1 alarm/Trip etpint lees Conservative than required hq the above Jpeacionai teu1reTtent, LLTIned.latelv suaefl4 te release o radioactive liuid effueats monitored by the affete cliazne. oi declare the channel inoperable and take ACTICN as directed by b. belo,
- b. with less than the itin:.mui n.imher of raicactive lLquid effluent monitoring instrumer.tatioi cbaiuieis OPBtE. take the AT]O shown in TabLe 3 . -12 Ezert test ettort to :estore to tIie nanitrLlrn nunter c iaditiv liujd ffaiat withi. 20 day2 nd if uoful.
explain in the next ?nuuaI. Rsdicactive f fluent ReLease Reprt plramnt
- o )D31 Appendix P. gection P.2 hv this incperability as not :otscted ui tim1y mnnr.
CURVE LLVCE 5IRHT
- 4. .3.10 ac1 xadioactie liuid ef].ient non.torinj instruElentatior. clan2el shall e Ierronscrate y erformance of Cte iieii LNEVL Ltc.
54VK, r3N1rPT TTTflv T1TT, rPZ1T3VT, (P ZTTiIT rv.r M- ih Erejuencles shown in Table 4.3-8, aoh urvi1Thn R rnent ha21 t rnd Lthin th 2poifid sureiliance interval vita a mmcirium allowable extension not to exceed 2E of tri.e speDttaec surve11Laae interval.
2013 NRC SRO Question 93 (18) Reference h.qrnn W-1ia flk1.r P4,Ir P1.rt- c4N i-nt- iq Cffsite Lo5e CalDuIat:c.n Nu11 ot.1) RY TL L-12 RD1D.CIIV LtQTJtD PLUENT 4ONI1t)NG INETTUt4NTATION MflhI4UM x2rT,tJt4EtTr DEEI.BLE IC1f
- 1. aadiatirit: tbnit.:r Pridin. Aaru and i,1ucmat1 Te:rnlnatic.n of tiea
- a. Lquid R.waste EffluEnt Lire a: Trtod Ltunry ozd Uot bDwr 1 tan1 LiachargE Mnitor 2 Waste {nitr lanka nd 1 Evaporator Cnderiate Tanka Di scharg Mrtitor
- ornndvy Wc rp Tnk 1 Di Echarga M:nitor
- b. Turhint ti1d.in PLi: DraLrz ff1unt 1 L ,ne 2, riati-it Mnitar Prviaing ],arn an Aitcratic stop gicna. to Divhaie Fillip
- a. O.itdoox Tan.k ta D rin Tzan. Ear Pnp 1 Noni:or
- tcactivit. r*rnlcors iro.r11nc tarn mit Pr..iriimg Ziicriir mi ti.ri nf
- a. Norinl eivic Water yat ]t.?turn From 1 Wata aain.J riding to ha C,rcilatir,g at.r yteTI
- b. Norrn1 aia Water 3y.tam reti.arn From tb ,eartor Au,dlir Euildinr to tba crcalatiLg water yten
- 4. Flcw Nea rErnEt DicS
- a. Liquid Radwaate Ett1unt Lines Traated Laundrj aiid ot Ixwar 1 Tani. DiachargE 2: Waste 4cnitor lanks and Waste 1 Evaporator C.nensatt Tanks Di aharga ecandaiy WastE arnpe Taok 1
- b. coolir.i TOCr 1ow4r
- Whn th woto cyct 10 ,n th: ocaatiruu rclzaoo ndo md r:1a.zcz ars cccti-ring, Action 2 shall bE taken wieri tke nrnitor L inoperable. In th9 batch reLeas a-ode. Action E La applicable.
2013 NRC SRO Question 93 (18) Reference LdLUL H.LLL 1i.L.1dL EvL E1uL 1FE c,fE:t r:. ca1u1atici 1u1 :cr:
TLL omir.ued:
ACI0N iirr; With the ither f h.nr.e1 oPansLs ]e thLn rquid by the Tiuirnu Channels 0PEA!LE rcruizernen:. effluer.t releases iia this pathway may cor.tin.ue prcvide+/- that rir to ini:iating a release:
- a. at least cwo itdeendent samples ars anal.y:ed in accordance wLct CperaticnaJ. uiierent 4.Ii1.1.L, rd
- 1. .t least two teclviical1 qualified mel+/-CLa of the facility staff independenti erifv the release rate caLculations and dis&iarge line alviri 0tbrci$e, .ipr.d 1e iotive ff1ien .yia ACrION 16 - With the Lurn.e: of :h3nr.e1s 0PL ]ess than required by the PEiaiwum Channels 0PEME requirerren:. effluent releases iia this pathway may cor.ttnue provld.e qral. sarrples are analyzed or ra1cactLvIty a: a lowsr
- iit of t:tic f ic rrrc thr LE 07 aci,ma1:
- a. t least ccce per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when te scLfic activity of the secondary coolant is greater than 0.01 1jCicmram POSE JIVALETT I -
J..-J. or, h at rnr. rr a hramr w3ii t pri fir )r rf I-hz secondary coolant is )ess than r ecual :o 0.01 ac/gram DGE cUIV1LNT 1-111.
r0N 37 - With the Luef hrne1 OPEEI! 2e th reguied by the Iinimurn chonsi9 rqIixrn. ffluer.t ;ia thi pathy my contiwe providel that, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, qrab samples are collected. an analyzed for radioactLvity at a lower limi: of de:ection of no more than lL-C7 iCjiTLl.
Ort0U 28 With thc urror cf hanro1o ODLPJLI. ic thcai by tb tirimun Channels CPE-LE r iirernen:, efflier.t releases via this pathway may cor.tiriue providel the flow rats is estimated at least once per hours durinj actual re)eases. Pump pafformnance curves zienerated in pLace may he used to estjnate flow.
(9TN .Q .. With r1lrr ef 1P3PZRT 1 th;r i iii i1 by th inirmmn Channels 0PE.?LE requiremnen:, effluent releases via this pathway may cor.tiriue provided the weekly Cooling 7ower loedown weir surveillance is pertormued as required by Operational. .egtrenent . U 1. .L. .. O:flerwias, io1lc*c th s:r:or .,eciflei i ?Crt01T 37 al;e, D- 4
2013 NRC SRO Question 96 (21) Reference
(% [AIrMEk I cH; (11)! ItJri:YFr1S rijj;M1I tkA pj i I\ r:so] row FQP O2FOPION 3.6.2 To inropriJn CJrtd ITonL Spi i ii L PERABLE r Sprsy Syro oapab.e of tkino suction rcrr tnc RiSI no torrorrinq suction Li 1 hr c irtci i nmeiL Jij) 1 I
5 L 1CAU!L1ILj HULLS P a nJ I fl lJiJ4 Mlii Vntii rnI Sorov ytr nioMLe irsi v fle incoorible Sjwc, S,sVr u OPE4Ui .3 sL:L us v Inn U hours M & r n I lea U H U I7NUE3I ui the no>: t hoj. restore tIe iropeacic bor* Systen to uFdlASLs status i V in V nest ai he irs or b in SIlL 3 SIB Ill IWL vU I h n 3e In by rq Jfl hours.
Rn rilsu Lu 3purfuu.u 3 0 203 AU UI HnOJ vrPu J 3 :i 1: 5 I LU 1
2013 NRC SRO Question 96 (21) Reference 3)4 8 [ FCTRICAL PCER i[M 3/4 S AiL. SDUES rPFpKrri LIMI 1IU CLNUIUUN Ur UFLRJ I{,N
- a. Iwo physically ndepenthnt circuits between the oils te trarisrnissicn net.wok and the onsite Cass 1E distribution ysen,
- b. Two separate and lidopendeit diesel çererators. each wth:
- 1. oaroe thy tank oontainiiy 0 ThlrflrPLm of 147 oi1or oF fuel,
- 2. A sear;e am n fuel oil stoae tan conta hir a Iflifl IrniJill o 1DO0O3 gllon of fue, nc
- 3. A senarate iie1 ol transfer pLrfl.
- c. Autoiratc :oad Segeners tor rato 1 vci Trn RLICbILI1: FDOES 1, 2. 3 arid $
a, 4iLh one o1fste circuit of 38,1,a inperab1e; 1 Peftrm Surveilanze Requirerent $8,1,,la within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> tiorearter: and
- 2. Retcre the otfcite circuit to OPRARI F .çtti c itoir 7? or be in at least HOT SIANBY witlin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> ard in OD 5HUULMU within th fol owirlq 30 flCurs; ani
- 3. Veify reqircd features rcered fron th OPEFL&$L[ oflste A,C.
source are OPERABLE.. I required feature(s) perecI frijin the OPERBLE offslt.e circuit are dicovereU Lu be iriuprcbI dL dfly tinie whle in this ccndtton, restrO ho required aturer,s to OPRBLE status within 24 iours from discovery of inopeab1e required features) or declare the redundant required teature(S) powered frrn th toperhl P.C. source as inoperablo, 3HLRC hAr,R:s - UNIT )9errn. 111
2013 NRC SRO Question 96 (21) Reference ELECTRICAL POEl SYSIEIIS A.C. SOIRC{S OPE RAT Lfl4iTs CONDITION FDR OPLATION INQn.inue:
- b. WLh uric diee1 ienaLor of 3.i.1.b inoperab1o
- 1. Perftrrn urvai11sn:e Reurenient 4 11i. within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> nd Dri:e per hous tnereat:er: and Within 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br />. determine the QPEALE diesei 9eflertr is not noprabie due to a con cause faiure or perorri Surveillance equirement 4.I.?.a4#: and
, Reatare :he diesel generator to OPtRABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Or be in Lear kOT STANDB th n the next 6 ocrs nd in COLO SHUTDOWN within he ollcAing 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />s: and
- 4. IeriIy required feature(s) powered rocu the OPERLE diesel ero-atQr re OPERABLE I reluired fature() 1 owrec frcm te OPERABLE: diesel oenerator are discovered to be iror,erable at ary
- ime hile in this coithlion retor tne requirtd fcature(s) to OPERABLE status wiiin 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from discovery cf iroperabl required features or decaro the rethindant rquirec featur(s) powered roi the inoperable LC source as roerable.
- c. th one oifsite circuit and one desel oriror cf 3.84.1 ircperale:
t1OE: Enter apIicabe Candltionis and F:equired Act.ior(s) of icc J ONS TU POWEP O1S11iOr4 OFEATI iT conthtlon is entered wt)i no AC. power to ore train, Restore one of the inoperable AC. sources to CPEiVBi1 status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or be n at edst HOT ST?NDB ithin tie next hours and n COLD ShUTDOWN wthin the foflowin O hours,
- 2. Follownq restoration of one A.C ource (offsite circuit cr aiesl qeneratar restore the nncln1ng 1roerdtlc ! 1 scurcc to OPR1E status pursuant to requirnents of either ACTIOIi a ci b based on the t.me of initial loss of the reoi2lrirg AC. source..
Th1s CION i req rd i be cvmpletrd rjarcle.,s ci hen tnc ineperable ELO is restored to 0PERAB:L :i.
- Actvt&.s that nornaly supoort 1.estnq pursLart. to which der :r dise 1nupbi ( q . r roll sh& I ro: h porormed for oesina requi red by tIns hCT1CN statencTt SHAROIJ HARRIS - UNI 31d 8-2 Nrendwent No.?u
2013 NRC SRO Question 96 (21) Reference CLtCRi1M POWF_SYSTH AC. SOJRCES OFERATI IG LtMETING COIDITI0N FOR OPEPATIO IIQN Continuedj:
- d. With t of the required offsitc t,C 5ource inopcrob1e
- 1. Restore one offsite circuit to OPER8LE s:atus wi:hin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or te in t least HOT TAD2Y ithir the 6 hourc ard in CUD SHUTDON wlthii the fo11oirg 30 nours: ind 2 rify rrqniri ftur(c) rp OPFRtR[ F If rquird fea:ure()
re discovered to be nooerb1e aL any tine hiie in this cridit or resLore he rquiro1 fatue(s to OPLRDL[ status 3 rro discvtr at iriorj1e required feamre( )
?ithln 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> cr decLare the redundant recuirecJ teature) inoperabe.
- 3. Folowinj restoration of on offit A C sorc. rctrr th remaining otfsite A C sirce n Icor dance ith the provisions o CT1O a with the time requirenent of that A3TION besee on the tinie a imtia) loss o the renaiiin mo rb1e AC source.
- e. 4-th to of the required :iesel generators inoperale:
L Iarlorai Surveillance Requirement 4&IJ. a within 1 rour and once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereatter arid
- , Restore one of tti ciesel aenordLor Lu OPERABLE LdLu wiLhin 2 hriir nr Iii in 1 lat Ru STMDRY ithn t[ nxL hours and in COLI SHUTDOWN iithin the foilowir 30 3, Followinq restoraticn of one diesel qererator. restore the renidining d esel gererato in .tcordance the orvmsions or MUOF4 b with -na timO reuiiornont cf that .CTIOI baec on te time of inittel loss of the reniainir iriopcrbie diesel generator
- f. With threG o riior of the requied AC. scorces inoperable:
- 1. Inintiitely en:er Technical Specification 3. .3.
- 2. Followng restoraticn of one or more A.C. sources, restore te rcmning inpvrablc A C oucc in accord ice with te provisions ot ACTJOt. a b d and/or e cS oolcab e itn the tine reurrnenr of that ION based on trc tlrr OT nitlal loss (if the renarnnq inoporabe. A.C. ouces.
J With LtJII1jLiiiUS even:s if either n nffsite cr vn5ite A C. curce bcorrir inoperable and resuting in failure to meet the [DO:
- 1. Within 6 days. rcstcrc 1i A.C. sourc requircd by 3.0.1 1 r be in ct least HOI SIADBY w thin tile next o hours and in COW SHUTDOWN within tne fo1lowirj 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
AL lvii s Ut nuril ly upurL. Ls pursuanL lo 4.. 1. i.2.i .4, whirh n&c rnder diel inapuruble g air rofl shall not be perforried for t.esinçj required by this ACTIOJ statement Si+[A HARRIS U(dT 1 3/4 8-3 Aeriment No. 75 1
2013 NRC SRO Question 96 (21) Reference LIBJCAL jQLYSiMS A!C. SQVRCES oERAT1NG LmnTINGcoNnmoFoRoPERAnoN F AGTION (Contfu):
- h. With one aiitonatIc oa:d sequencer inoperable; 1.. Restore the autorntft load seauencer to OPERABlE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least NOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUThOW within the following 30 hors.
SLEARG RARRIS INIT 3/4 84 Aenwet No. Si
2013 NRC SRO Question 97 (22) Reference
, Progress Energy INFORMATION USE HARRIS NUCLEAR PLANT PLANT OPERATING MANUAL VOLUME 1 PART 1 PROCEDURE TYPE: ADMINISTRATIVE PROCEDURE (AP)
NUMBER:
AP-617 TITLE:
REPORTABILITY DETERMINATION AND NOTIFICATION AP-617 Rev. 33 Page 1 of 47
2013 NRC SRO Question 97 (22) Reference Table of Contents Section Page 1.0 PURPOSE 3
2.0 REFERENCES
4 3.0 DEFINITIONS/ABBREVIATIONS 5 4.0 RESPONSIBILITIES 6 5.0 PROCEDURE 5.1 Immediate Reportability 7 5.2 Other Reports 11 6.0 DIAGRAMS/ATTACHMENTS Attachment 1 - Immediate Notification Requirements 12 Attachment 2 - Technical Specification and ODCM Special Reports 20 Attachment 3 - Routine Reports 24 Attachment 4 - Event Reports (Other than LERs) 27 Attachment 5 - One Hour Notifications - Sample Wording 36 Attachment 6 SAMPLE Reactor Plant Event Notification Worksheet 38 Attachment 7 - Reactor Plant Event Notification Worksheet 39 Attachment 8 Reportability Evaluation (REW) Worksheet 41 Revision Summary 47 AP-617 Rev. 33 Page 2 of 47
2013 NRC SRO Question 97 (22) Reference 1.0 PURPOSE R 1. This procedure provides guidance in determining NRC Reportability in the following areas:
- a. Events requiring verbal notification to the NRC via Emergency Telecommunication System (ETS) within one, four, eight, or twenty-four hours.
- b. Events requiring a written follow-up report to the NRC as a Licensee Event Report (LER) or as a Special Report.
- c. Scheduled Routine Reports required by Title 10 of the Code of Federal Regulations (CFR) or by the Operating License Technical Specifications; PLP-114, Relocated Technical Specifications and Design Basis Requirements; and the Offsite Dose Calculation Manual (ODCM).
- d. Event Reports (other than LER5) that are prepared on an as needed basis.
- e. Reportability evaluation for non-routine reports will be based on Condition Reports per Reference 2.3.
- 2. The reports listed in this procedure are regulatory requirements. The specific reference for each report is identified with each report on the applicable attachment.
- 3. Several specific reporting requirements are also addressed by other procedures:
- a. Immediate notifications of safeguards (security) related events related to the Physical Security of Special Nuclear Material as required by §73.71 (Reporting of Safeguards Events) will be classified per Reference 2.7.
- b. Reporting of events which result in the declaration of an emergency classification shall be in accordance with Emergency Plan and implementing procedures.
- c. Reporting of events regarding fish kills, hazardous substance releases, and oil spills shall be in accordance with Reference 2.17 for notification to appropriate Corporate, Local, State and Federal (non-NRC) agencies.
- d. Reporting of events regarding environmental violations shall be in accordance with this procedure and References 2.30 and 2.31 for notification to appropriate Corporate, Local, State and Federal (non-NRC) agencies.
- e. Reporting of events to insurers regarding certain fires or other losses shall be in accordance with Reference 2.32.
- f. Reporting of events regarding non-routine radioactive releases shall be in accordance with Reference 2.37 for notification to appropriate Corporate, Local, State and Federal (non-NRC) agencies.
AP-617 Rev. 33 Page 3 of 47
2013 NRC SRO Question 97 (22) Reference I0 PURPOSE (continued)
- 4. This procedure provides instructions for the immediate notification to the NRC using the Emergency Telecommunication System (ETS) phone for non-emergency events that require such reporting according to §50.72, Technical Specifications and other § requirements.
2.0 REFERENCES
- 1. SEC-NGGC-2120, Use Storage and Protection of Safeguards and Other Limited Access Information
- 2. AP-611, Regulatory Correspondence (superseded by REG-NGGC-0016)
- 3. CAP-NGGC-0200, Condition Identification and Screening Process
- 4. REG-NGGC-0013, Evaluating and Reporting of Defects and Noncompliance in Accordance With 10 CFR 21
- 5. PEP-310, Notifications and Communications
- 6. AP-620, Licensee Event Report Development and Approval (superseded by REG-NGGC-0016)
- 7. SEC-NGGC-2 147, Reporting of Safeguards and Fitness for Duty Events
- 8. SHNPP Operating License and Technical Specifications
- 9. NUREG 1022, Licensee Event Report System
- 10. NRC Inspection Procedure 61706, Core Thermal Power Evaluation
- 11. SP-014, Additional Surveillance/Compensatory Security Measures
- 12. ADM-NGGC-0201, Nuclear Task Management
- 13. RDC-NGGC-0001, NGG Standard Records Management Program
- 15. PLP-1 14, Relocated Technical Specifications and Design Basis Requirements
- 17. PLP-500, Fish Kill Reporting, Hazardous Substances Release Notification, and Oil Spill Notification
- 18. NUREG 1460, Guide to NRC Reporting and Recordkeeping Requirements
- 19. Regulatory Guide 1.16, Reporting of Operating Information -Appendix A Technical Specifications AP-617 Rev. 33 Page 4 of 47
2013 NRC SRO Question 97 (22) Reference
2.0 REFERENCES
(continued)
- 20. Regulatory Guide 10.1, Compilation of Reporting Requirements for Persons Subject to NRC Regulations
- 21. IE Information Notice No. 83-34, Event Notification Information Worksheet
- 22. IE Information Notice No. 85-62, Backup Telephone Numbers to the NRC Operations Center
- 23. lE Information Notice No. 85-78, Event Notification
- 24. Letter HELD-H-278, Zimmerman to Beatty et al., March 31, 1987
- 25. SAF-S UBS-00033, Employee Incident I nvestigatiohs
- 26. PLP-201, Emergency Plan
- 27. EPM-400, Public Notification and Alerting System
- 29. U.S. Department of Transportation, Federal Aviation Administration Advisory Circular AC 70/7460-1 K
- 30. EMP-001, NPDES Permit Monitoring
- 31. National Pollutant Discharge Elimination System (NPDES) Permit Number NC0039586, North Carolina Department of Environment and Natural Resources, Division of Water Quality
- 32. PLP-105, Insurance Programs at Harris Nuclear Plant
- 33. NEI Position Statement: Guidance to Licensees on Complying with the Licensed Power Limit (NRC ADAMS Accession No. ML081750537)
- 34. NRC Memorandum titled Discussion of Licensed Power Level, Jordan, E.L.,
Division of Reactor Operations Inspection, Aug. 22, 1980
- 35. PLP-300, Process Control Program
- 36. SEC-NGGC-2140, Fitness For Duty Program
- 37. CHE-NGGC-0057, Groundwater Protection Program
- 38. PLP-717, Equipment Important To Emergency Preparedness and ERO
Response
- 39. REG-NGGC-0016, Regulatory Correspondence & LER Development
- 40. CR 580945, Oct. 3, 2012 Notice of Violation for EOF
- 41. FPP-013, Fire Protection-Minimum Requirements, Mitigating Actions and Surveillance Requirements 3.0 DEFINITIONSIABBREVIATIONS
- 1. Code of Federal Regulations CFR
- 2. Equipment Inoperable Record EIR-
- 3. Emergency Response Facility Information System ERFIS AP-617 Rev. 33 Page 5 of 47
2013 NRC SRO Question 97 (22) Reference 3.0 DEFINITIONSIABBREVIATIONS (continued)
- 4. Emergency Telecommunication System - ETS
- 5. Engineered Safety Feature - ESF
- 6. Licensee Event Report (LER) A written report conforming to the format and content requirements of §50.73 and NUREG 1022.
- 7. National Oceanic and Aeronautic Administration - NOAA
- 8. Nuclear Regulatory Commission - NRC
- 10. Operating License - CL
- 11. Reactor Protection System - RPS
- 13. Solid State Protection System SSPS
- 14. Emergency Notification System ENS
- 15. Health Physics Network - HPN 4.0 RESPONSIBILITIES
- 1. The Shift Manager (SM):
- a. Determining immediate NRC reportability, and
- b. Making appropriate notifications.
- 2. The Supervisor Licensing/Regulatory Programs:
- a. Confirming the correctness of immediate reportability determinations.
- b. Determining need for reports to other outside agencies.
- c. Generating related reports as required.
- 3. The Superintendent Security (or On-duty Security Supervisor):
- a. Evaluating security related events in accordance with Reference 2.7
- b. Informing the SM when a security related event must be reported to the NRC.
AP-617 Rev. 33 Page 6 of 47
2013 NRC SRO Question 97 (22) Reference 5.0 PROCEDURE 5.1 Immediate Reportability The Shift Manager (SM) determines that an event requires immediate notification (per Attachment 1), or the Superintendent Security (or On-duty Security Supervisor) informs the SM that an event requires notification to the NRC.
- 2. The SM prepares or assigns an individual to prepare the Reactor Plant Event Notification Worksheet (Attachment 7). Event Notification Worksheets for Safeguard/Security events are normally prepared by the Security Organization.
- 3. The Event Description narrative should be as short and concise as possible while conveying a clear description of the event. Attachment 5 may be used for developing one-hour notifications. Attachment 6 is a completed sample Worksheet.
- 4. The initial part of the Event Description should state:
- a. The initial conditions of the plant or affected systems prior to event occurrence.
- b. The actual event and direct cause, if known.
- c. The current conditions of the plant or affected systems Example Plant was in Mode 1 at 50% reactor power and increasing load. At 1000, the reactor was manually tripped following a loss of both main feedwater pumps caused by feedwater regulating valve oscillations. The plant is stable in Mode 3 at normal temperature and pressure.
- 5. The balance of the Event Description should contain known specific details of precursor events which led to the reportable event, including the time of each event. Report only known facts; do not speculate.
Example Feedwater regulating valve oscillations occurred when placing valves into automatic control. The A and B feedwater regulating valves had been successfully placed in automatic at 0950 and were controlling normally. When the C valve was placed in automatic at 0958, large oscillations were noted in the C valve followed by oscillations in the A and B valves. During the oscillation, the condensate booster pump tripped on low flow resulting in tripping of the feed pump. The reactor was manually tripped prior to receipt of a steam generator low water level signal.
AP-617 Rev. 33 Page 7 of 47
2013 NRC SRO Question 97 (22) Reference 5.1 Immediate Reportability (continued)
- 6. The Event Description should include a statement of the proper functioning or failure to function of safety systems and the safety significance of an event, if such a determination is possible. If possible, also include two or three compensatory actions taken to assure safety.
The Shift Technical Advisor may assist the SM in making such a determination.
Example All safety systems functioned as expected (or list equipment which failed to function as expected). AFW automatically actuated to provide continued decay heat removal. Compensatory actions to assure safety include...
- 7. The SM:
- a. Reviews the Event Notification Worksheet.
- b. Makes changes if necessary.
- c. Approves it for release.
- 8. If time permits, the SM shall contact the General Manager Harris Plant and the NRC Resident Inspector, and Licensing/Regulatory Programs and provide them the information contained in the notification.
- 9. The SM notifies the NRC by giving the approved Event Notification Worksheet to an available individual to telecopy the Worksheet to the NRC via the fax number (301-816-5151, may be confirmed via EPL-001).
NOTE: The NRC electronically records notifications.
- 10. When the approved Event Notification Worksheet has been sent, contact the NRC Operations Center Duty Officer by performing either of the following:
- a. Pick up the receiver on the Emergency Telecommunication System Telephone and dial the NRC Operations Center Duty Officer via one of the numbers located on the phone label, in EPL-001, or on the Event Notification Worksheet.
- b. If desired, use a normal telephone line to call the NRC Operations Center Duty Officer via one of the numbers located on the phone label, in EPL-001, or on the Event Notification Worksheet.
AP-617 Rev. 33 Page 8 of 47
2013 NRC SRO Question 97 (22) Reference 5.1 Immediate Reportability (continued)
- 11. When the Duty Officer responds:
- a. Caller says, THIS IS THE HARRIS NUCLEAR PLANT. THIS IS A NOTIFICATION OF (appropriate event classification from worksheet). HAVE YOU RECEIVED A TELECOPY OF THIS NOTIFICATION?
- b. If response is No, have an available individual perform Step 5.1.9 again while continuing with this step.
- c. The caller gives the information on the Event Notification Worksheet and repeats information when requested.
- d. The notification should be read in its entirety.
- e. The caller should respond to any requests for additional information that can be answered accurately, or if the caller is not able to accurately respond to the Duty Officers requests, the caller shall write down the request and inform the Duty Officer that this information will be delivered in a follow up notification.
- f. The caller should record questions asked, responses provided and if follow up is necessary on a separate sheet of paper and attach it to the Event Notification Worksheet.
- 12. If the Duty Officer has not received a telecopy after the notification has been completed, the caller shall request the Duty Officer to read back the notification and, if necessary, correct any errors.
- 13. The caller records the Event Notification Number, name of the individual contacted and time of contact on the Event Notification Worksheet.
- 14. The caller informs the Duty Officer that the caller is signing off. The Duty Officer may request to stay on and leave the line open. If this occurs, the caller should comply. A replacement caller may be necessary to stay on the phone.
- 15. If additional information is provided to the Duty Officer beyond the initial Event Notification Worksheet, notify the General Manager Harris Plant, the NRC Resident Inspector, and Licensing/Regulatory Programs of the additional information provided.
L-617 Rev. 33 Page 9 of 47
2013 NRC SRO Question 97 (22) Reference 5.1 Immediate Reportability (continued)
- 16. Follow up Notifications.
In addition to making the required initial notifications, during the course of the event IMMEDIATELY report:
- a. Any further degradation in the level of safety of the plant or other worsening plant conditions, including those that require the declaration of any of the Emergency Classes, if such a declaration has not been previously made.
- b. The results of ensuing evaluations or assessments of plant conditions.
- c. The effectiveness of response or protective measures taken.
- d. Information related to plant behavior that is not understood.
- 17. Notify Site Communications, or if there is no response, Corporate Communications Media Line of this Reactor Plant Event Notification Worksheet. If after hours leave a message for the on call person. (See EPL-0O1 for contact numbers).
- 18. Event Notification Worksheets which have been designated Safeguards Information in accordance with the provision of References 2.1 and 2.7 shall be returned to Security after the notification has been made with no further dissemination.
- 19. The Event Notification Worksheet should be sent to Licensing/Regulatory Programs; this does not apply to Security Notifications.
- 20. Licensing/Regulatory Programs will also evaluate if follow up notification is required for clarification, retraction or other. The below criteria should be considered:
- a. Clarity for public docket (not extremely technical)
- b. Minimize inflammatory jargon be precise and factual
- c. Provides perspective and mitigating conditions
- 21. Licensing/Regulatory Programs will initiate ARs to track generation of follow up reports.
- 22. Licensing/Regulatory Programs will forward a copy of the Event Notification Worksheet to the ICES Coordinator within five working days for entry into the ICES database.
AP-617 Rev. 33 Page 10 of 47
2013 NRC SRO Question 97 (22) Reference 5.2 Other Reports Licensing/Regulatory Programs shall perform the following:
- 1. Reportability determinations for Steps 2 through 6 below shall be completed expeditiously. Reports should be confirmed as tracked by an AR.
- 2. Evaluate the condition for reportability as a Special Report under Technical Specification Section 6.9.2 per Attachment 2.
- 3. Evaluate the condition for reportability as an LER using Reference 2.9.
Development of the LER is per Reference 2.6. As indicated in
§50.73(a)(1), invalid actuations, other than Reactor Protection System actuations when the reactor is critical, may be reported by telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of a written LER.
CAUTION Evaluation for §21 Reportability may not be substituted for reporting pursuant to §50.73.
Actual reporting per §21 may be performed using an LER per §50.73 and Ref. 2.6.
- 4. Evaluate the condition for potential reportability under §21 per Reference 2.4.
- 5. Evaluate the condition for reportability via a Routine Report per Attachment 3.
- 6. Evaluate the condition for reportability via an Event Report (other than LER) per Attachment 4.
6.0 DIAGRAMSIATTACHMENTS Attachment 1 - Immediate Notification Requirements Attachment 2 - Technical Specification and ODCM Special Reports Attachment 3 - Routine Reports Attachment 4 - Event Reports (Other than LER5)
Attachment 5 - One Hour Notifications Sample Wording Attachment 6 SAMPLE Reactor Plant Event Notification Worksheet Attachment 7 - Reactor Plant Event Notification Worksheet Attachment 8 - Reportability Evaluation (REVV) Worksheet AP-61 7 Rev. 33 Page 1 1 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 1 Sheet 1 of 8 IMMEDIATE NOTIFICATION REQUIREMENTS The following tables are divided into sections based upon the time allowed for reporting the respective events as follows:
One Hour Notifications II Four Hour Notifications Ill Eight Hour Notifications IV Twenty-four Hour Notifications NOTE: The events listed in this attachment may be concurrent with conditions that result in a declared emergency. In the case of a declared emergency, the notification made under the Emergency Plan and implementing procedures satisfies the notifications required by this procedure (10 CFR 50.72(a)). Written reports will be based on §50.73 and Technical Specifications regardless of whether the initial notification is made under the Emergency Plan or this procedure.
I. ONE HOUR NOTIFICATIONS l.A. OPERATIONAL EVENTS -10 CFR 50.72 (b) (1)
- 1. Technical Specification Deviations (10 CFR 50.54x)
- 2. Safety Limit Violation (TS 6.7.1)
I.B. RADIOLOGICAL EVENTS
- 1. Radioactive Shipments (Note 1)
- 2. Loss or Theft of Licensed Material/Radiological Sabotage (Note 2)
- 3. Exposure to Individuals or Releases (Note 3)
- 4. Accidental Criticality (Note 4)
I.C. SECURITY EVENTS (Note 5)
- 1. Security Events per SEC-NGGC-2147.
- 2. International Atomic Energy Agency (IAEA) Representative I.D. FITNESS FOR DUTY (Note 6)
- 1. FFD - NRC Employee AP-617 Rev. 33 Page 12 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 1 Sheet 2 of 8 IMMEDIATE NOTIFICATION REQUIREMENTS II. FOUR HOUR NOTIFICATIONS OPERATIONAL EVENTS 10 CFR 50.72 (b) (2)
- 1. Initiation of any Nuclear Plant Shutdown required by Technical Specifications.
- 2. Unplanned Actuation of the reactor protection system (scram) when the reactor was critical and any event that results or should have resulted in ECCS discharge into the RCS.
- 3. Off-Site Notification Has Been or Will Be Made (Note 12)
Ill. EIGHT HOUR NOTIFICATIONS
- 1. Degraded or Unanalyzed Condition
- 2. Loss of Emergency Response Capability (Note 7)
- 3. Unplanned Actuation of selected ESF Systems Refer to NUREG 1022 System Actuation to identify applicable system actuations.
- 4. Loss of a Safety Function
- 5. Transport of a Potentially Contaminated Individual
- 6. Fatality or Hospitalization (Note 13)
IV. TWENTY-FOUR HOUR NOTIFICATIONS
- 1. EXPOSURE TO INDIVIDUALS OR RELEASES
- a. Radiological Exposure/Release (Note 8)
- b. Unusual or Important Environmental Events (Note 9)
- 2. VIOLATION OF OPERATING LICENSE CONDITIONS (Note 10)
- 3. FITNESS FOR DUTY PROGRAM EVENTS (Note 11)
AP-617 Rev. 33 Page 13 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 1 Sheet 3 of 8 NOTES IMMEDIATE NOTIFICATION REQUIREMENTS NOTIFICATION REFERENCE WRITTEN FOLLOW-UP RADIOACTIVE SHIPMENTS a) Removable contamination from a received §20.1906(d)(1) package containing radioactive material in excess §71.87(i) of the limits specified in §71.87(i). The involved RP Supervisor shall immediately notify the final delivery carrier.
b) Radiation levels from a received package of §20.1906(d)(2) radioactive material in excess of the limits §71.47 specified in §71.47. The involved RP Supervisor shall immediately notify the final delivery carrier.
c) Security related events with respect to the §73.71 (b)(2) transport of special nuclear material are handled §73 Appendix G via SEC-NGGC-2147. Security threats or theft of §73.71(a)(5) licensed material shall be reported to site Security personnel.
- 2. LOSS OR THEFT OF LICENSED MATERIAL! RADIOLOGICAL SABOTAGE Note: Theft and Sabotage are Security Events handled by SEC-NGGC-2147.
Any loss or theft or attempted theft of:
a) Licensed material in an aggregate quantity equal §20.2201 (a)(1 )(i) 30-Day Written Report to or greater than 1000 times the quantity §20.2201(d) required per §20.2201(b) specified in Appendix C to §20.1000-20.2401 under such circumstances that it appears that an exposure could result to persons in unrestricted areas, b) Any Special Nuclear Material or spent fuel, (theft See SEC-60-Day Written Report NGGC-2147) required per §73.71 (a) or
§74.11 (b) See SEC-NGGC
§150.16(b) 2147
§73.71 (a) 15-Day Written Report may be required per
§150 c) Recovery of or accounting for loss of any shipment §73.71(a) of Special Nuclear Material or spent fuel d) Greater than 10 curies of tritium at any one time or §30.55(c) 15-Day Written Report 100 curies in one calendar year, or required e) More than 15 pounds of uranium or thorium at any §40.64(c) 15-Day Written Report one time or more than 150 pounds in one calendar §150.17(c) required year.
AP-617 Rev. 33 Page 14 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 1 Sheet 4 of 8 NOTES IMMEDIATE NOTIFICATION REQUIREMENTS NOTIFICATION REFERENCE WRITTEN FOLLOW-UP
- 3. EXPOSURE TO INDIVIDUALS OR RELEASES Any event involving by-product, source or Special Nuclear Material that may have caused or threatens to cause:
a) An individual to receive: §20.2202(a)(1) LER required by
§50.73(a)(2)(viii)
- 1) A total effective dose equivalent of 25 Rem (a)(2)(ix) and §20.2203
- 2) An eye dose equivalent of 75 Rem
- 3) A shallow-dose equivalent to the skin or extremities of 250 Rad
- 4) An intake of 5 ALl in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> b) Release of radioactive material in excess of §20.2202(a)(2) LER required by Technical Specification Instantaneous Limits shall §50.72(b)(2)(iv) §50.73(a)(2)(viii),
be declared an emergency in accordance with (a)(2)(ix) and §20.2203 PEP-310. The reporting requirements of PEP-310 shall take precedence over the less restrictive times for reporting requirements of §20.2202 and
§50.72(b)(2) for releases.
- 4. ACCIDENTAL CRITICALITY Accidental criticality of special nuclear material. §70.52(a) None
- 5. SECURITY EVENTS Note: Reporting of Security Events (Including Safeguards and Fitness-For-Duty Events) is per SEC NGGC-2147. Notify site Security personnel.
INTERNATIONAL ATOMIC ENERGY AGENCY (IAEA) §75.7 None REPRESENTATIVE Individual claiming to be an IAEA representative who is not accompanied by an NRC employee and has no prior confirmation of credentials in writing.
Notification is by telephone to Director, Office of §75.6 and §75.7 Nuclear Reactor Regulation
- 6. FITNESS FOR DUTY NRC EMPLOYEE Notification of NRC employees unfitness for duty. §26.27(d) None Per §26.27(d), the appropriate Regional Administrator must be notified immediately by telephone. During other than normal working hours, the NRC Operations Center must be notified.
AP-61 7 Rev. 33 Page 15 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 1 Sheet 5 of 8 NOTES IMMEDIATE NOTIFICATION REQUIREMENTS NOTE: PLP-717, Equipment Important To Emergency Preparedness and ERO Response, Attachment 2, Essential ERO Equipment and Compensatory Measures, contains guidance on determining immediate reportability. Other Technical Specification and ODCM reporting requirements may apply and are located in Attachment 2 of this (AP-61 7) procedure. [CR 580945 CAPR]
NOTIFICATION REFERENCE WRITTEN FOLLOW-UP
- 7. LOSS OF EMERGENCY RESPONSE §50.72(b)(3)(xiii) None CAPABILITY Any event that results in a major loss of assessment capability, offsite response capability, or communications capability (e.g., significant portion of Control Room indication, Emergency Telecommunication System, or offsite notification system).
This may include loss of any of the following:
a) Emergency Response Facilities b) Radiation Monitors and Plant Equipment used in identification of Emergency Action Levels c) Computers and Telecommunications including:
- 1. Selective Signaling
- 2. NRC Emergency Telecommunication System
- 4. PABX telephone system
- 5. Plant PA System
- 6. Corporate Telephone Communication System (Voicenet) and the Commercial Telephone System
- 7. Satellite Phones
- 8. Sound Powered Phone System
- 9. HNP Emergency Notification (Everbridge) System
- 10. Emergency Response Facility Information System (ERFIS)
- 12. Dose Assessment Software (RASCAL) d) Sirens and Tone Alert Radios Use PLP-71 7, Attachment 2 in determination of immediate reportability.
AP-617 Rev. 33 Page 16 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 1 Sheet 6 of 8 NOTES IMMEDIATE NOTIFICATION REQUIREMENTS NOTIFICATION REFERENCE WRITTEN FOLLOW-UP
- 8. RADIOLOGICAL EXPOSU RE/RELEASE §20.2202(b) 30-Day Wriffen Report Required per §20.2203 Any event involving licensed material possessed by the licensee that may have caused or threatens to cause an individual to receive, in a period of 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s:
a) A total effective dose equivalent> 5 Rem; or b) An eye dose equivalent> 15 Rem; or c) A shallow-dose equivalent to the skin or extremities> 50 Rem; or d) Anintakeof>1 ALl.
- 9. UNUSUAL OR IMPORTANT ENVIRONMENTAL EVENTS Any event that indicates or could result in significant Env. Prot. Plan, Local, State & Federal environmental impact causally related to plant (Operating License Agency Notifications operation. Examples are: Appendix B) defined in PLP-500 Section 4.1, require NRC notification and PLP-500 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per 50.72 (See Note 12) a) Excessive bird impaction ESS determines If event is significant and threshold not reportable to a Local, State, or Federal b) Onsite plant or animal disease outbreak ESS determines agency, a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> NRC threshold notification may still be required per Env. Prot.
c) Mortality or unusual occurrence of Endangered ESS determines Plan.
Species threshold d) Fish Kills Local and State 30-Day Follow-up written Notifications defined report required per Env.
in PLP-500 Prot. Plan Subsection 5.4.2 e) Increase in nuisance organisms or conditions ESS determines threshold f) Unanticipated or emergency discharge of waste Local and State water or chemical substances Notifications defined in PLP-500 g) Damage to vegetation resulting from cooling tower ESS determines drift deposition threshold h) Station outage or failure of any cooling water ESS determines intake or service water system components due to threshold bio-fouling by Corbicula (Asiatic Clam)
AP-61 7 Rev. 33 Page 17 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 1 Sheet 7 of 8 NOTES IMMEDIATE NOTIFICATION REQUIREMENTS NOTIFICATION REFERENCE WRITTEN FOLLOW-UP
- 10. VIOLATION OF OPERATING LICENSE CONDITIONS a) Any event resulting in the plant operating in a OL Section 2.G 60-day LER required per manner which violates the SHNPP Facility 10 CFR 50.73 and 10 Operating License, Section 2.C: CFR 50.4(e)
(1) Intentionally raising power above 2948 MWt NEI Position (100%) for any period of time. Statement: Guidance to Licensees on (2) Failure to reduce thermal power to less than or Complying with the equal to 2948 MWt when the 2-hour average Licensed Power Limit exceeds 2948 MWt. (NRC ADAMS Accession No.
(3) Permitting the core thermal power 8-hour ML081750537) average to exceed 2948 MWt.
(4) Failure to take prudent action prior to a pre planned evolution that could cause a power increase to exceed 2948 MWt (example:
Scheduled securing of a Heater Drain Pump without first reducing power to accommodate the expected power increase. A short term increase in transient power above 2948 MWt following a boron dilution is not included if actions to reduce power are taken in a reasonable time following the dilution reactivity transient).
Note: No actions are allowed that would intentionally raise core thermal power above 2948 MWt for any period of time. Small, short-term fluctuations in power that are not under the direct control of a license reactor operator or result from actions taken for a different purpose (example: temperature control) are not considered intentional.
b) A failure to comply with the following OL Section 2.G 60-day LER required per administrative requirements (See Note 1): 10 CFR 50.73 and 10 CFR 50.4(e)
- 1) Deviation from the requirements of the OL Section 2.C.2 Environmental Protection Plan;
- 3) Failure to comply with new fuel storage OL Section 2.C.10 requirements.
AP-617 Rev. 33 Page 18 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 1 Sheet 8 of 8 NOTES IMMEDIATE NOTIFICATION REQUIREMENTS NOTIRCATION REFERENCE WRITTEN FOLLOW-UP
- 11. FITNESS FOR DUTY PROGRAM EVENTS Note: Reporting of Security Events (Including Safeguards and Fitness-For-Duty Events) is per SEC-NGGC 2147. Notify site Security personnel.
a) Sale, use, or possession of illegal drugs within the §26.73(a)(1) None protected area.
b) Any acts by any person licensed under §55, or by §26.73(a)(2) None any supervisory personnel assigned to perform duties within the scope of §26
- 1) Involving the sale, use, or possession of a controlled substance,
- 2) Resulting in a confirmed positive test on such persons,
- 3) Involving use of alcohol within the protected area, or
- 4) Resulting in a determination of unfitness for scheduled work due to the consumption of alcohol.
c) False positive error on a blind performance test App. A to Part 26 Non-docketed specimen when error is determined to be B.2.8(e)(5) correspondence to NRC administrative. Reference 2.35 FFD coordinator
- 12. OFF SITE NOTIFICATION HAS BEEN OR WILL BE MADE Any event or situation, related to the health and safety §50.72(b)(2)(xi) of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Such an event may include an onsite fatality or inadvertent release of radioactively contaminated materials.
- 13. FATALITY OR HOSPITALIZATION a) See SAF-SUBS-00033 and contact the Site Safety SAF-SUBS-00033 See Note 12 for NRC Representative required 4-hour notification b) OSHA must be notified within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of:
- 1) Workplace Fatality
- 2) Workplace incident with 3 or more personnel hospitalized c) North Carolina requires a call to the Dept. of Labor, Elevator Division, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of an injury or fatality related to elevators.
AP-617 Rev. 33 Page 19 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 2 Sheet 1 of 4 TECHNICAL SPECIFICATION AND ODCM SPECIAL REPORTS A Special Report may be identified from CRs, ElRs, declared emergencies or as the result of equipment inspections. The following steps shall be performed when it appears that a Special Report is required.
- 1. The Supervisor Licensing/Regulatory Programs shall be notified of the event if this has not already occurred via a CR or EIR.
- 2. The Supervisor Licensing/Regulatory Programs shall inform the General Manager Harris Plant and applicable unit manager(s) of the need for a special report.
- 3. The Supervisor Licensing/Regulatory Programs shall assign action items to the responsible units per Reference 2.12 to provide input for the required reports.
- 4. Completed reports shall be routed for approval per Reference 2.2.
- 5. A copy of the completed special report shall be provided to the Secretary PNSC for review at a subsequent PNSC meeting.
- 6. The special report shall be transmitted as a QA Record.
REPORTING REQUIREMENTS TS FODCM1 RESP TIMING OF SUBJECT REFERENCE UNIT RESPONSE Leak or boron deposit found during inspection O.L. NRC Order HESS 60 days after returning the dated 2/11/03 plant to operation Moderator Temperature Coefficient more 3.1.1.3 HESS 10 Days positive than specified limits Remote Shutdown Monitoring Instrument 3.3.3.5.a Maint 14 Days inoperable for greater than 60 days Radiation Monitors, Pressurizer Safety Valve 3.3.3.6 Maint or IT (and 14 Days Position Indicators or Subcooling Margin Engineering for Monitors inoperable for greater than 7 days. Rad. Monitors)
(Also see OWP-ERFIS)
PORV/RCS Vents used to mitigate RCS 3.4.9.4 Operations 30 Days pressure transient at low temperature An ECCS actuation and injection of water into 3.5.2, 3.5.3 Operations 90 Days (See Note 3) the Reactor Coolant System (See Note 2)
Change to Sample Plan Used for Snubber 3.7.8 HESS Before Implementation Functional Testing PLP-106 Att. 4 Attachment 2
[ AP-617 Rev. 33 Page 20 of 47
2013 NRC SRO Question 97 (22) Reference Sheet 2 of 4 TECHNICAL SPECIFICATION AND ODCM SPECIAL REPORTS TS [ODCM1 RESP TIMING OF SUBJECT REFERENCE UNIT RESPONSE Radioactive Material in Liquid Holdup Tanks 3.11.1.4 Environmental Include in Annual exceeding limits and Chemistry Radioactive Effluent Release Report Use of non-preferred Incore Detectors when 4.2.4.2 HESS 30 Days evaluating QPTR (Bases)
Abnormal degradation of Containment Vessel 4.6.1.6.2 HESS 15 Days structure detected during required inspections Sealed Source leakage test results 4.7.9.3 Radiation Annually if removable Protection contamination greater than 0.005 pCi detected Greater than 30 total rods or 10 rods per fuel 5.3.1 HESS 30 Days Following Start-up assembly replaced with filler rods or vacancies during any single refueling.
Safety Limit Violation. 6.7.1 Operations 14 Days Startup Report following: 6.9.1.1 HESS 90 Days Following Resumption of Commercial
- 1) License amendment to increase power level. Operations or Completion of Startup Test Program, or
- 2) Installation of fuel of different design or 9 months after Initial manufacturer. Criticality, whichever is earliest. Supplementary
- 3) Modifications that significantly alter the reports required each nuclear, thermal or hydraulic characteristics 3 months.
of the unit.
Change to Core Operating Limits Report 6.9.1.6.4 HESS Upon issuance; must be submitted no later than the date of implementation.
Steam Generator Tube Inspection Report 6.9.1.7 HESS Within 180 days after initial entry into Hot Shutdown following completion of an inspection performed in accordance with TS 6.8.4.1.
Special Reports 6.9.2 As Assigned As Requested Unreviewed Environmental Question T.S. Appendix B Environmental Before implementation of EPP Section 3.1 and Chemistry change Attachment 2 AP-617 Rev. 33 Page 21 of 47
2013 NRC SRO Question 97 (22) Reference Sheet 3 of 4 TECHNICAL SPECIFICATION AND ODCM SPECIAL REPORTS TS [ODCM} RESP TIMING OF SUBJECT REFERENCE UNIT RESPONSE Proposed changes/renewal application for T.S. Appendix B Environmental At time of submittal to the NPDES Permit EPP Section 3.2 and Chemistry permitting agency Changes to/renewal of NPDES Permit or State T.S. Appendix B Environmental 30 Days after Certification EPP Section 3.2 and Chemistry change/renewal Stay of NPDES Permit or State Certification T.S. Appendix B Environmental 30 Days following stay EPP Section 3.2 and Chemistry Unusual or Important Environmental Events T.S. Appendix B Environmental 30 Days after event. (Note EPP Sections 4.1 and Chemistry 4) See also 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 5.4.2 notifications Seismic Monitoring Instrument inoperable for PLP-1 14 Maint 10 Days greater than 30 days Actuation of Seismic Monitoring Instruments PLP-114 HESS 14 Days during seismic event greater than or equal to 0.Olg (See Note 1)
Meteorological Monitoring Instrument inoperable PLP-114 Maint 10 Days for greater than 7 days Metal Impact Monitoring System Channel(s) PLP-1 14 Maint 10 Days inoperable for greater than 30 days.
Explosive gas monitoring instrument inoperable PLP-114 Maint 30 Days for greater than 30 days. (Note: No time specified in PLP-1 14)
Area Temperatures exceeding PLP-114, PLP-114 HESS & 30 Days limits by more than 30°F, or for Operations greater than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> A calculated dose to a member of the public from [3.11.1.2] Environmental 30 Days the release of radioactive materials in liquid and Chemistry effluents to an unrestricted area exceeding limits Radioactive liquid waste being discharged [3.11.1.3] Environmental 30 Days without treatment and in excess of limits and any and Chemistry portion of the liquid radwaste treatment system & Operations not in operation Calculated air dose in gaseous effluent [3.11.2.2] Environmental 30 Days exceeding limits in areas at or beyond site and Chemistry boundary Attachment 2 Sheet 4 of 4 AP-617 Rev. 33 Page 22 of 47
2013 NRC SRO Question 97 (22) Reference TECHNICAL SPECIFICATION AND ODCM SPECIAL REPORTS TS FODCM1 RESP TIMING OF SUBJECT REFERENCE UNIT RESPONSE Calculated dose to a member of the public from [3.11.2.3] Environmental 30 Days a release of gaseous effluents containing and Chemistry Iodine 131, Iodine 133, tritium and radionuclides in particulate form with half-lifes greater than eight days exceeding the limits.
Untreated radioactive gaseous waste discharged [3.11.2.4] Environmental 30 Days in excess of limits and any portion of the and Chemistry gaseous radwaste treatment system not in & Operations operation.
Calculated dose from release of radioactive [3.11.4] Environmental 30 Days material in liquid or gaseous effluents, to a and Chemistry member of the public in excess of limits.
Level of radioactivity, as a result of plant effluent [3.12.1] Environmental 30 Days in a specified location, exceeding the reporting and Chemistry levels of ODCM O.R. Table 3.12-2 when averaged over any calendar quarter.
NOTES:
- 1. Refer to the following TS Surveillance Requirements following any seismic event:
4.3.3.3.2 and 4.4.5.3.c.2
- 2. Refer to the following TS Surveillance Requirements following any safety injection actuation: 4.4.5.3.c.1, 4.4.5.3.c.3, 4.4.5.3.c.4, 4.4.6.2.2.d, and 4.5.2.g.1
- 3. LER also required (see Reference 2.6). The information required by the Special Report exceeds the requirements of the LER.
- 4. Events requiring reports to other government agencies shall be reported per those requirements in lieu of the Environmental Protection Plan. A copy of the report shall be sent to the NRC.
AP-617 Rev. 33 Page 23 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 3 Sheet 1 of 3 ROUTINE REPORTS The SHNPP Technical Specifications, ODCM, §20 and §50 require that reports be provided to the NRC at routine intervals. The following steps apply to routine reports:
- 1. The Supervisor Licensing/Regulatory Programs shall establish action items per Reference 2.12 for these reports.
- 2. As applicable, the responsible unit and Licensing/Regulatory Programs shall establish standard formats for a routine report.
- 3. Routine reports shall be routed for approval per Reference 2.2.
- 4. Routine reports shall be transmitted as a QA record.
NOTE: TS = Technical Specification OR = ODCM Operational Requirement REGULATORY RESP TIMING OF SUBJECT REFERENCE UNIT RESPONSE Annual Operating Report T.S. 6.9.1.2 Lic/ Reg Prog Before 3/1 Annual Radiological T.S. 6.9.1.3 Environmental Before 5/1 Environmental Operating O.R. 3.12.1 and Chemistry Report O.R. Table 3.12-1 O.R. Table 4.12-1 O.R. 4.12.2 O.R. 3.12.3 O.R. 4.12.3 Annual Environmental Env. Prot. Plan 5.4.1 Environmental Before 5/1 Operating Report and Chemistry Annual Radioactive Effluent T.S. 3.11.1.4 Environmental Before 5/1 Release Report T.S. 6.9.1.4 and Chemistry T.S. 6.14.c ODCM App. F.3 O.R. Table 4.11-1 O.R. 3.12.1 O.R. Table 3.12-1 O.R. 3.12.2 O.R. 3.3.3.10 O.R. 3.3.3.11
§50.36a(a)(2)
AP-617 Rev. 33 Page 24 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 3 Sheet 2 of 3 ROUTINE REPORTS NOTE 1: This report does not need to be routed for approval per Reference 2.2 since it is not correspondence to a regulatory agency.
REGULATORY RESP TIMING OF SUBJECT REFERENCE UNIT RESPONSE Consolidated Data Entry (ODE) FSAR Section 1.8 Lic I Reg Prog By the end of the Report (Reg. Guide 1.16) month following each quarter Individual Worker Radiation §19.13(b) ESS Rad Annually Dose Report (To employee) Services (See Note 1 above) (Dosimetry)
Annual Exposure Report §20.2206(b) Lic I Reg Prog Annually, by for Individual Monitoring April 30th Semiannual Fitness for Duty §26.71(d) Nuclear Within 60 days of Program Performance Data Operations the end of each 6-Analysis Report month reporting period (Jan-June and July-Dec)
ECCS evaluation model §50.46(a)(3) NFM&SA Annually changes or errors where sum of absolute magnitudes of changes results in PCT change 50°F Changes to QA Program which §50.54(a)(3) NAS In accordance with do not reduce commitments §50.71(e)
Insurance and Financial §50.54(w)(3) Nuclear Annually, on April 1 Security Annual Report Operations In-service Inspection Summary §50.55a (ASME HESS 90 days after Section XI IWA- completion of 6230) inspections FSAR Update, facility changes, §50.59(b) Lic / Reg Prog Six months tests, and experiments §50.71(e) following each conducted without prior approval refueling outage.
Interval not to exceed 24 months between updates AP-617 Rev. 33 Page 25 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 3 Sheet 3 of 3 ROUTINE REPORTS REGULATORY RESP TIMING OF SUBJECT REFERENCE UNIT RESPONSE Annual Financial Report, §50.71(b) Nuclear Upon issuance of the including certified financial Operations report (Normally statements April 30)
Status of Decommissioning §50.75(f) Nuclear March 31, 1999 and Funding Operations at least once every two years thereafter (frequency becomes annual when the plant is within 5 years of projected end of operation or when the plant is involved in mergers or acquisitions).
Simulator Report of
- §55.45(b)(5)(ii) Training Every 4 years on uncorrected performance anniversary of test failures and schedule for certification correction Material Status Reports (old §74.13(a)(1) HESS 30 days after Forms DOE/NRC-742 and §70.53(a)(1) March 31 and 742(c)) §40.64(b) September 30
§150.17(b) NOTE: 40.64 and 150.17 require statement of foreign origin source material.
QAProgramfor §71.101 Nuclear Every5Years.
Transportation of Operations Docket 71-0345 Radioactive Material Packages Financial Protection - §140.21 Nuclear Annually, on Guarantee of payment of Operations anniversary date on deferred premiums which indemnity agreement is effective (Normally April 30)
AP-617 Rev. 33 Page 26 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 4 Sheet 1 of 9 EVENT REPORTS (Other than LERs)
Title 10 of the Code of Federal Regulations and other requirements require that reports be provided to the NRC and other regulatory agencies based on the occurrence of specific events. The following instructions apply to these events (unless otherwise noted):
- 1. The responsible unit shall determine if written procedures should be prepared to implement the reporting requirement.
- 2. When written reports are required to be submitted to the NRC, the General Manager Harris Plant and the Supervisor Licensing/Regulatory Programs shall be informed.
- 3. Completed reports shall be routed for approval per Reference 2.2.
- 4. The event report shall be transmitted as a QA Record.
10 CFR Reference RESP TIMING OF SUBJECT or (other) UNIT RESPONSE Report to former radiation worker of 19.13(c) Radiation Upon request; within 30 workers exposure to radiation Protection days from request or 30 (Note 7) days after exposure has been determined Radiation Exposure Data to 19.13(d) Radiation Upon overexposure Individual-Overexposure Protection (Note 5)
(Note 7)
Radiation Exposure Data to 19.13(e) ESS Rad Upon request at Terminating Employees Services termination (Note 7) (Dosimetry)
Bioassay Services to Determine 20.1204(c) Radiation As requested by NRC Exposure Protection Report of planned special exposure 20.1206(f) Radiation Within 30 days following 20.2204 Protection planned special exposure Report to individual of planned 20.1206(g) Radiation Within 30 days from special exposure (Note 7) Protection planned special exposure Respiratory Protection Program 20.1703(d) Radiation 30 Days prior to Equipment not certified by Protection equipment usage NIOSH/MSA Theft or loss of licensed material 20.2201(a)(1)(ii) Radiation 30 Days (by ETS) greater than 10 times Appendix C 20.2201(b) Protection quantities Radiation 30 Days (written Protection follow-up)
Additional information on theft or 20.2201(d) Radiation 30 Days loss Protection AP-617 Rev. 33 Page 27 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 4 Sheet 2 of 9 EVENT REPORTS (Other than LERs) 10 CFR Reference RESP TIMING OF SUBJECT or (other) UNIT RESPONSE Reports of Overexposures! 20.2203 Radiation 30 Days (Note 5)
Excessive Levels and 4OCFR1 90 Protection Concentrations ODCM OR.
3.11.4 Missing Waste Shipment Trace 20 App. G, Radiation 2 weeks after Investigation Sec. IlI.E Protection completion of Investigation Interim evaluation report of identified 21.21(a)(2) Lic/Reg Prog Within 60 days of deviation or failure to comply (when discovery evaluation cannot be completed within 60 days of discovery)
Failure to Comply or Existence of a 21.21(c)(1) & Lid Reg Prog 2 days, written follow-up Defect (Refer to AP-616) (d)(3) within 30 days FFD testing more conservative than 26 App. A, Human Res Within 60 days of
§26 requirements Sec. A.1.1(2) implementing change FFD Unsatisfactory performance
- 26 App. A, Human Res 30 days testing of a certified laboratory Sec. B.2.8 (e)(4)
Reports required as condition of 30.34(e)(4) Radiation As specified in license Parts 30-35 licenses (By-product Protection Material)
Renewal or non-renewal of Parts 30.36 Radiation 30 Days prior to 30-35 licenses Protection expiration Notification of byproduct (Part 30) or 30.36(d) Radiation Within 60 days SNM (Part 70) license expiration or 70.38(d) Protection cessation of principal activities Amendment to Part 30-35 License 30.38 Radiation As Required Protection Failure of or damage to shielding, 31 .5(c)(5) Radiation 5 Days (5 Day report on-off mechanism or indicator; Protection required per Byproduct detection of removable radioactive Materials License in lieu material of 30 day requirement)
Transfer of device to specific or 31 .5(c)(8) Radiation 30 Days general licensee 31 .5(c)(9)(i) Protection Leaking of sealed radiographic 34.27(d) Licensed 5 Days source Radiographer AP-617 Rev. 33 Page 28 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 4 Sheet 3 of 9 EVENT REPORTS (Other than LERs) 10 CFR Reference RESP TIMING OF SUBJECT or (other) UNIT RESPONSE Incidents involving radiographic 34.101 (a) Licensed Within 30 days of equipment: Radiographer occurrence
- unintentional disconnection of the source assembly from the control cable.
- inability to retract the source assembly to its fully shielded position and secure it in this position.
- failure of any component (critical to safe operation of the device) to properly perform its intended function.
Licensee identified information which 50.9(b) Lic / Reg Prog As necessary (Note 4);
has significant implications for public 30.9(b) §70.9(b) specifies within health and safety or the common 40.9(b) two working days of defense and security 70.9(b) identifying the 71.7(b) information Change to the LOCA analysis which 50.46(a)(3) NFM&SA If 50°F, then include in results in a change to the Peak Clad the annual report. If Temperature >50°F, then 30 days.
Changes in QA Program which 50.54(a) NAS Before implementation reduce commitments Request for Written Information 50.54(f) As Specified As Requested Changes to Operator Requalification 50.54(1-1) Training Before Implementation Program which decreases scope, time allotted or frequency of conducting portions of the program Changes in Security Plan, Guard 50.54(p) Security Before implementation if Training and Qualification Plan, or 70.32(g) changes reduce Safeguards Contingency Plan made effectiveness of plan; without prior approval otherwise, within two months after change.
NOTE: §70.32(g) specifies within 60 days for safeguards contingency plan.
AP-617 Rev. 33 Page 29 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 4 Sheet 4 of 9 EVENT REPORTS (Other than LERs) 10 CFR Reference RESP TIMING OF SUBJECT or (other) UNIT RESPONSE Changes in emergency plan or 50.54(q) Emerg Prep Before implementation if implementing procedures made 50 App E(V) changes reduce without prior approval effectiveness of plan, otherwise within 30 days after change Notification of safe and stable 50.54(w)(4)(ii) Emerg Prep After attaining safe and condition of reactor and no stable condition significant risk to public health and (following an accident safety with costs >$100 million)
Cleanup plan for decontaminating 50.54(w)(4)(ii) Emerg Prep Within 30 days of reactor to permit resumption of notification that reactor operation or commencement of is in safe and stable decommissioning condition (following an accident with costs
$100 million)
Plan for management of ?J2 50.54(bb) Lic I Reg Prog Within 2 years after notification of significant change in cessation of operation or the proposed waste management 5 years before program (of irradiated fuel at the expiration of OL and reactor, after expiration of license after making change in and until transferred to DOE) program Filing of petition for bankruptcy (Title 50.54(cc)(1) Lic I Reg Prog Immediately following 11 of US Code) 30.34(h) filing 40.41(f) 70.32(a)(9)(i)
Notification that conformance to a 50. 55a(f)(5) HESS After identifying problem certain Code required by Section Xl (iii) of the ASME B&PV Code and Addenda for inservice test is impractical Determination that a pump or valve 50. 55a(f)(5) H ESS No later than 12 months test required by Section XI of the (iv) after expiration of initial ASME B&PV Code and Addenda is 120-month period of impractical and not included in the operation, and each revised inservice test program subsequent 120-mo.
period of operation during which the test is determined to be impractical AP-61 7 Rev. 33 Page 30 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 4 Sheet 5 of 9 EVENT REPORTS (Other than LERs) 10 CFR Reference RESP TIMING OF SUBJECT or (other) UNIT RESPONSE Notification that conformance to a 50.55a(g)(5) H ESS After identifying problem certain Code required by Section Xl (iii) of the ASME B&PV Code and Addenda for inservice inspection is impractical Determination that a pump or valve 50.55a(g)(5) HESS No later than 12 months test required by Section XI of the (iv) after expiration of initial ASME B&PV Code and Addenda is 120-month period of impractical and not included in the operation, and each revised inservice inspection program subsequent 120-mo.
period of operation during which the test is determined to be impractical Updated assessment of projected 50.61(b)(1) HESS After change value for RTPTS for reactor vessel beltline materials after significant change in projected value of RTPTS or change in facilitys operating expiration date Plan for thermal annealing of reactor 50.66 H ESS 3 yrs prior to date when vessel fracture toughness criteria would be exceeded Reports required as a condition of 50.71(a) As Specified As Specified in License license Telephone report in lieu of LER for 50.73(a)(1) Operations 60 Days an invalid actuation of a system listed in 50.73(a)(2)(iv)(B), other than RPS actuation when critical Reassignment of Licensed Operator 50.74(a) Operations 30 Days to position not requiring license Termination of Licensed Operator 50.74(b) Operations 30 Days Hardware and software changes that 50 App. E, Sec. Nuclear Info Within 30 days after affect transmitted data points VI. 3. a Systems changes are completed identified in ERDS Data Point Library Hardware and software changes 50 App. E, Sec. Nuclear Info As soon as practicable (except data point modifications) that VI.3.b Systems and at least 30 days could affect transmission format and prior to making computer communication protocol to modification the ERDS AP-617 Rev. 33 Page 31 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 4 Sheet 6 of 9 EVENT REPORTS (Other than LERs) 10 CFR Reference RESP TIMING OF SUBJECT or (other) UNIT RESPONSE Test methods for supplemental 50 App. G Sec. IlI.B HESS Submitted and approved fracture toughness tests prior to testing Fracture Toughness 50 App G, Sec. HESS 3 years before date IV.A.1 when predicted fracture toughness will no longer satisfy App G Report of test results of specimens 50 App H, Sec. IV HESS Within One Year of withdrawn from capsules (fracture Withdrawal toughness tests)
Report of effluents released in 50 App I, Sec. IV.A Environmental Within 30 Days from end excess of one-half design objective and Chemistry of quarter (Special exposure Report required by T.S. 3.11.4 satisfies this requirement)
Reactor Building ILRT 50 App J, Option A, HESS 3 Months After Test Sec. V.B (No time specified in TS 4.6.1.2 regulation or TS)
Certification of Medical Fitness 55.23 Training Upon Application 55.31 Incapacitation Because of Disability 55.25 Operations 30 Days after learning of or Illness . 50.74(c) diagnosis Application for Operator License 55.31 Training As necessary Reapplication for Operator License 55.35 Training Two months after first denial, six months after second denial, two years after third and subsequent denials Conviction of a Felony for Licensed 55.53(g) Operations! 30 Days Operator 73.71(b) Security Application for Operator License 55.57 Training As necessary Renewal Reports of Conditions of Part 70 70.32(b)(5) Security As specified in license License 75.36 Changes in plan for Physical 70.32(d) Security 2 Months Protection of SNM in transit made without prior approval AP-617 Rev. 33 Page 32 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 4 Sheet 7 of 9 EVENT REPORTS (Other than LERs) 10 CFR Reference RESP TIMING OF SUBJECT or (other) UNIT RESPONSE Accident notification report required 71.5(a)(1)(iv) Radiation Carrier is to provide by DOT on transportation of licensed 49CFR171.15 & 16 Protection notice to DOT at the material earliest practicable moment and written follow-up within 30 days Transportation Package Information 71.12(c)(3) Radiation Before first use 71.101(f) Protection (Note 1)
Deviations related to Type B 71.95 Radiation Within 30 days (NOTE:
package for transport of radioactive Protection requirement is for material, specifically: licensee to report)
- significant reduction in effectiveness of Type B packaging during use
- defects with safety significance in Type B packaging after first use, or the means employed to repair the defects and prevent recurrence
- conditions of approval of certificate of compliance not observed in making a shipment Advance notification of shipment of 71.97(a) Radiation Before shipment Irradiated Fuel, Nuclear Waste, or 73.37(b)(1) Protection (Note 2) certain shipments of SNM 73.72(a) 73.73(a) 73.74(a)
Revision notice or cancellation 71.97(e) Radiation Upon change!
notice for shipment of irradiated fuel, 71.97(f) Protection cancellation nuclear waste, or certain shipments 73.37(f) (Note 2) of SNM 73.72(a)(5) 73.73(b) 73.74(b)
Advance Notice and Approval of 73.37(b)(7) Radiation Before shipment Routes for Shipment of Irradiated 73.37(f) Protection & (Note 2)
Fuel Nuclear Operations Theft or unlawful diversion or 73.71 Security As required (See also attempted theft or unlawful diversion 74.11 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> notifications and of SNM or spent fuel 150.16(b) SEC-NGGC-2 147)
AP-617 Rev. 33 Page 33 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 4 Sheet 8 of 9 EVENT REPORTS (Other than LERs) 10 CFR Reference RESP TIMING OF SUBJECT or (other) UNIT RESPONSE Results of Trace Investigation of 73.71 (a) Security 30 Days Lost or Unaccounted for Shipment of 74.11 SNM Threat to or reduced effectiveness of 73.71(d) Security 30 Days physical security (See also 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> notifications and SEC N GGC-2 147)
Nuclear Material Transfer 74.15(a) HESS Whenever transfer (old Form DOE/NRC-741) 40.64(a) occurs 70.54 (Note 6) 150.16(a) 150.17(a)
Earthquake exceeding Operating 100 App. A, Sec. HESS Prior to resuming Basis Earthquake values V(A)(2) operations Importorexportof nuclear 110 Radiation Varies; see Part 110 equipment or material Protection Bodily injury or property damage 140.6(a) Radiation As promptly as practical from possession or use of Protection radioactive material resulting in an indemnity claim Change in proof of financial 140.15(e) Treasury Promptly protection or other financial information filed with the NRC Termination of liability insurance 140.17(b) Treasury At least 30 days prior to policy used for financial protection termination (notification of renewal or other proof of financial protection)
Failure of High Integrity Container or PLP-300 Radiation Within 30 days of Notification of Misuse of a High Protection knowledge of the Integrity Container incident Cooling Tower Beacon Outage FAA Advisory Operations Upon Discovery greater than 30 minutes or Circular AC restoration from an outage greater 70/7460-1 K than 30 minutes. (Note 8)
Level of radioactivity in onsite CHE-NGGC-0057 Environmental Within 30 days of groundwater, exceeding the and Chemistry discovery reporting levels of ODCM OR. Table 3.12-2 for drinking water.
AP-617 Rev. 33 Page 34 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 4 Sheet 9 of 9 EVENT REPORTS (Other than LERs)
NOTES:
- 1. Development and use of a package will require special reporting requirements per §71.5, 71.95 and 71.101(f).
- 2. Notification to NRC received at least 10 days before transport of the shipment commences; see §73.72(a), 73.73(a), or 73.74(a) for additional details. State Governor(s) of states through which material is to be shipped shall also be notified by mail postmarked at least 7 days before shipment or by messenger 4 days prior to shipment. Notification of any subsequent schedule changes of greater than six (6) hours or cancellation of shipment shall also be made before the change (See §71.97(c) for additional notification to the Regional NRC Administrator).
- 3. Not used
- 4. Report not required if such reporting would duplicate information already submitted per other NRC reporting requirements.
- 5. When reporting exposure of an individual, the individual shall also be notified not later than the transmittal to the NRC.
- 6. §40.64(a) specifies next working day for transfers and 10 days for receipt of foreign origin source material. §150.16(a) and 150.17(a) specify within 10 days after material is received.
- 7. This report does not need to be routed for approval per Reference 2.2 since it is not correspondence to a regulatory agency.
- 8. Notify the DOT-FAA Flight Service Station at either 1-877-487-6867 or 1-800-992-7433.
The following information will be required:
- a. Harris Nuclear Plant, caller name and telephone number
- b. Which Hyperbolic Cooling Tower Beacon Warning Lights (by compass orientation) are inoperable
- c. Location Plant location is latitude 353801N, longitude 7W5723W and a distance of 16 miles Southwest of Raleigh
- d. Height The Cooling Tower is 523 feet above ground level. The height of the Cooling Tower above sea level is 784 feet.
- e. Estimated return to service date Determine from the Flight Service Station the name of the individual contacted and when a follow-up call should be made. Document the notification and any required follow-up in an NCR with Licensing as the responsible organization.
AP-617 Rev. 33 Page 35 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 5 Sheet 1 of 2 One Hour Notifications Sample Wording for the Description Field (Licensing will review notifications for follow-up clarification as needed.)
l.A. OPERATIONAL EVENTS -10 CFR 50.72 (b) (1)
- Technical Specification Deviations (10 CFR 50.54x)
DEVIATION FROM TECHNICAL SPECIFICATIONS PER 10 CFR 50.54(x)
At hrs license condition was deviated from per 10 CFR 50.54(x). This condition requires Discussion as to why the condition was not met, affect on the plant and when compliance was/will be restored.
The NRC Resident Inspector was notified.
- Safety Limit Violation (TS 6.7.1)
VIOLATION OF SAFETY LIMIT At Technical Specification Safety Limit was violated when Immediate corrective actions were Discussion as to the affect on the plant, additional planned actions, and any compensatory actions taken to assure safety.
The NRC Resident Inspector was notified.
I.B. RADIOLOGICAL EVENTS
- Radioactive Shipments Note: The time and date of the spent fuel shipment is safeguards information. Date and time of discovery is not safeguards but care should be taken not to link the event to arrival of the cask.
Example: SURFACE CONTAMINATION ABOVE LIMITS Smearable contamination on a radioactive materials package, a used fuel shipping cask transported by rail, exceeded the limits of 10 CFR 71.47. There is no evidence of personnel contamination or spread of contamination beyond the rail car. There is no indication of increased exposure to the public as a result of this event.
The NRC Resident Inspector was notified.
- Loss or Theft of Licensed Material/Radiological Sabotage (This example is for Loss only. Theft or Sabotage is reported using SEC-NGGC-2 147.)
LOCATION OF IS UNKNOWN How was loss of SNM discovered and what efforts to relocate are undeiway? What assurance is there that the lost SNM is under the control of a licensee (vs. the public) and that no personnel have been overexposed?
The NRC Resident Inspector was notified. Region II (name) and the State of North Carolina have also been notified.
AP-617 Rev. 33 Page 36 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 5 Sheet 2 of 2 Exposure to Individuals or Releases PERSONNEL OVEREXPOSURE A worker received (internal/external) contamination resulting in an estimated total effective dose equivalent (TEDE) of Rem ( mSv). The individual had been doing what, where. How detected? What location on the body. What decontamination was performed to what result?
What immediate corrective actions are being taken?
The NRC Resident Inspector was notified
- Accidental Criticality in the Fuel Handling Building No example. This event is extremely rare.
l.C. SECURITY EVENTS
- Security Events Reported per SEC-NGGC-2 147.
Security and Safeguards events are prepared by the Security Organization per SEC-NGGC 2147.
- International Atomic Energy Agency (IAEA) Representative No example. This event is extremely rare.
I.D. FITNESS FOR DUTY
- FFD - NRC Employee No example. This event is extremely rare.
AP-61 7 Rev. 33 Page 37 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 6 Sheet 1 of 1 SAMPLE WORKSHEET NRC FORM 361 U.S. NUCLEAR REGULATORY COMMISSION (12-2000) OPERATIONS CENTER REACTOR PLANT EVENT NOTIFICATION WORKSHEET EN #
NRC OPERATION TELEPHONE NUMBER; PRIMARY-- 301-816-5100 or 8005323469*, BACKUPS-- [1st] 301-951-0550 or 8004493694*,
[2nd] 301-415-0550 and [3rd] 301-415-0553 *Ucensees who maintain their own ETS are provided these telephone numbers.
NOTIFICATION TIME FACILITY OR ORGANIZATION UNIT NAME OF CALLER CALL BACK #
Harris Nuclear Plant I John Caves (919) 362-3636 15:43 EDT EVENT TIME & ZONE EVENT DATE POWER/MODE BEFORE POWER/MODE AFTER 14:33 EDT 0911812003 100% Power, Mode 1 100% Power, Mode I EVENT CLASSIFICATIONS 1-Hr. Non-Emergency 10 CFR 50.72(b)(1) (v)(A) SafeS/DCapability GENERAL EMERGENCY GEN/AAEC TS Deviation: :ADEV (v)(B) RHR Capability AINB SITE AREA EMERGENCY Hr Non-Emergency 10 CFR 50 72(b)(2) (v)(C) Control of Rad Release AINC SIT/AAEC -
ALERT (i) TS Required S/D ASHU (v)lD) Accident Mitigation AIND ALE/AAEC UNUSUAL EVENT UNU/AAEC (iv)(A) ECCS Discharge to RCS ACCS (xli) Offsite Medical AMED x 50.72 NON-EMERGENCY (see next columns) (iv)(B) RPS Actuation (scram)
ARPS x (xlN) Loss CommlAsmuResp ACOM PHYSICAL SECURITY (73.71) DDDD (xi) Offsite Notification APRE 60-Day Optional 10 CFR 50.73(a)(1)
MATERIAUEXPOSURE B??? 8-Hr. Non-Emergency I OCFR 50.72(b)(3) Invalid Specified System Actuation AINV FITNESS FOR DUTY HFIT (ii)(A) Degraded Condition ADEG Other Unspecified Requirement (identify)
OTHER UNSPECIFIED REQMT. (see last column) (ii)(B) Unanalyzed Condition AUNA NONR INFORMATION ONLY NNF (iv)(A) Specified System Actuation AESF NONR DESCRIPTION Include: Systems affected, actuations and their initiating signals, causes, effect of event on plant, actions taken or planned, etc. (Continue on back)
As of 2:33 PM, EDT, more than 20% of the offsite emergency sirens were inoperable for greater than one hour due to loss of power caused by Hurricane Isabel. Currently 27 of 81 sirens are out of service. The State of North Carolina and all four counties within the 10-mile emergency planning zone were notified and are in stand-by to implement mobile route alerting if needed. At this time, Harris cannot estimate the time of siren recovery. This requires an 8-hour non-emergency notification per 1 OCFR 50.72(b)(3)(xiii) due to the loss of a significant portion of the offsite notification system. The NRC Senior Resident Inspector was informed.
NOTIFICATIONS YES NO WILL BE ANYTHING UNUSUAL OR D YES (EXPLAIN ABOVE) X NO NRC RESIDENT X NOT UNDERSTOOD?
STATE(s) X DID ALL SYSTEMS X YES L1 NO LOCAL X FUNCTION AS REQUIRED?
OTHER GOV AGENCIES X ADDITIONAL INFO ON BACK MODE OF OPERATION I ESTIMATE FOR MEDIA/PRESS RELEASE I
X RESTART DATE:
UNTIL CORRECTED: 1 I 1YES X NO AP-617 Rev. 33 Page 38 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 7 Sheet 1 of 2 NRC FORM 361 US. NUCLEAR REGULATORY COMMISSION (12-2000) OPERATIONS CENTER REACTOR PLANT EVENT NOTIFICATION WORKSHEET EN #
NRC OPERATION TELEPHONE NUMBER: PRIMARY-- 301-816-5100 or 8005323469*, BACKUPS-- [1st] 301-951-0550 or 8004493694*,
r2nd] 301-415-0550 and [3rd] 301-415-0553 *Licensees who maintain their own ETS are provided these telephone numbers.
NOTIFICATION TIME FACILITY OR ORGANIZATION UNIT NAME OF CALLER CALL BACK#
Harris Nuclear Plant 1 919 -
EVENT TIME & ZONE EVENT DATE P0WERJMODE BEFORE POWER/MODE AFTER EVENT CLASSIFICATIONS 1-I-jr. Non-Emergency 10 CFR 5072(b)(1) (v)(A) SafeS/DCapability GENERAL EMERGENCY GEN/MEC TS Deviation ADEV (v)(5) RHR Capability AINB SITE AREA EMERGENCY 4-Hr. Non-Emergency 10 CFR 50.72(b)(2) (v)(C) ControlofRad Release AINC SIT/AAEC ALERT (I) TS Required S/D ASHU (v)(D) Accident Mitigation AIND ALE/AAEC UNUSUAL EVENT UNU/AAEC (iv)(A) ECCS Discharge to RCS ACCS (xii) Offsite Medical AMED 50.72 NON-EMERGENCY (see next columns) (iv)(B) RPS Actuation (scram) (xiii) Loss Comm/AsmtiResp ACOM ARPS PHYSICAL SECURITY (73.71) DDDD (xi) Offsite Notification APRE 60-Day Optional 10 CFR 50.73(a)(1)
MATERIAL/EXPOSURE B??? 8-Hr. Non-Emergency IOCFR 50.72(b)(3) Invalid Specitied System Actuation AINV FITNESS FOR DUTY HFIT (ii)(A) Degraded Condition ADEG Other Unspecified Requirement (identify)
OTHER UNSPECIFIED REQMT, (see last column) (ii)(B) Unanalyzed Condition AUNA NONR INFORMATION ONLY NNF (iv)(A) Specified System Actuation AESF NONR DESCRIPTION Include: Systems affected, actuations and their initiating signals, causes, effect of event on plant, actions taken or planned, etc. (Continue on back)
NOTIFICATIONS YES NO NRC RESIDENT WILL BE ANYTHING UNUSUAL OR LI YES (EXPLAIN ABOVE) LI NO NOT UNDERSTOOD?
STATE(s)
LOCAL DID ALL SYSTEMS LI YES LI NO FUNCTION AS REQUIRED?
OTHER GOV AGENCIES I I I LI ADDITIONAL INFO ON BACK MEDIA/PRESS RELEASE MODE OF OPERATION ESTIMATE FOR UNTIL CORRECTED: RESTART DATE: YES LI NO AP-617 Rev. 33 Page 39 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 7 Sheet 2 of 2 RADIOLOGICAL RELEASES: CHECK OR FILL IN APPLICABLE ITEMS (specific details/explanations should be covered in event description)
LIQUID RELEASE GASEOUS RELEASE UNPLANNED RELEASE PLANNED RELEASE ONGOING TERMINATED MONITORED UNMONITORED OFFSITE RELEASE T. S. EXCEEDED RM ALARMS AREAS EVACUATED PERSONNEL EXPOSED OR CONTAMINATED OFFSITE PROTECTIVE ACTIONS RECOMMENDED *State release path in description Release Rate (Ci/sec) % T. S. HOO GUIDE Total Activity (Ci) % T. S. LIMIT HOO GUIDE LIMIT Noble Gas 0.1 Ci/see 1000 Ci Iodine 10 uCi/sec 0.01 Ci Particulate 1 uCi/sec 1 mCi Liquid (excluding tritium and 10 uCi/min 0.1 Ci dissolved noble gases)
Liquid (tritium) 0.2 Ci/min 5 Ci Total Activity PLANT STACK CONDENSERIAIR MAIN STEAM SG OTHER EJECTOR LINE BLOWDOWN RAD MONITOR READINGS ALARM SETPOINTS
% T. S. LIMIT (if applicable)
RCS OR SG TUBE LEAKS: CHECK OR FILL IN APPLICABLE ITEMS: (specific details/explanations should be covered in event description)
LOCATION OF THE LEAK (e.g., SG #, valve, pipe, etc.)
LEAK RATE: UNITS: gpmlgpd IS. LIMITS: SUDDEN OR LONG-TERM DEVELOPMENT:
LEAK START DATE TIME COOLANT ACTIVITY PRIMARY SECONDARY AND UNITS:
LIST OF SAFETY RELATED EQUIPMENT NOT OPERATIONAL:
EVENT DESCRIPTION (continued from front)
NRC HEADQUARTERS DUTY OFFICER CONTACTED:____________________________ / I : AM/PM NAME DATE TIME LAD-617 Rev. 33 Page 40 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 8 Sheet lof6 Reportability Evaluation (REW) Worksheet Notification Requirements Per 10CFR5O.73 Reports NCR (Ref. NUREG-1022)
A Licensee Event Report (LER) is generally required for any event of the type described in this Attachment within 60 days after the discovery of the event. HNP shall report any applicable event if it occurred within three years of the date of discovery regardless of the plant mode or power level, and regardless of the significance of the structure, system, or component that initiated the event.
I. Description of the Event L1D617 Rev. 33 Page 41 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 8 Sheet 2 of 6 II. For each Reportability Evaluation, perform the following:
. Consult with or perform a pre-job briefing with Licensing.
- Using the details of the event, determine if the reporting category applies to this event.
- Mark the appropriate block Yes or No.
- If uncertain, gather more information to make the determination. Consult Licensing as needed.
- Reference NUREG-1 022 as needed since it contains many examples that may aid in the determination.
- Complete the Reportable Evaluation Section to justify the conclusion that was reached based on known facts.
Situation meets this Reportable Event condition?
Yes LI No LI Plant Shutdown Required by Technical Specifications?
The completion of any nuclear plant shutdown required by the HNP Technical Specifications.
Yes LI No LI Operation or Condition Prohibited by Technical Specifications?
Any operation or condition whicn was prohibited by the HNP Technical Specifications except when:
- a. The Technical Specification is administrative in nature;
- b. The event consisted solely of a case of a late surveillance test where the oversight was corrected, the test was performed, and the equipment was found to be capable of performing its specified safety functions; or
- c. The Technical Specification was revised prior to discovery of the event such that the operation or condition was no longer prohibited at the time of discovery of the event.
Yes LI No LI Deviation from Technical Specifications?
Any deviation from the HNP Technical Specifications authorized pursuant to section 50.54(x).
AP-61 7 Rev. 33 Page 42 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 8 Sheet 3 of 6 Situation meets this Reportable Event condition?
Yes LI No LI System Actuation?
Any event or condition that resulted in a manual or automatic actuation of any of the systems listed below, except when:
- a. The actuation resulted from and was part of a pre-planned sequence during testing or reactor operation; or
- b. The actuation was invalid and; (1) Occurred while the system was properly removed from service; or (2) Occurred after the safety function had been already completed.
The systems to which the requirements above apply are:
- a. Reactor protection system (RPS) including reactor trip.
- b. General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIV5).
- c. Emergency core cooling systems (ECCS) including: high-head and low-head injection systems and the low pressure injection function of residual (decay) heat removal systems.
- d. Auxiliary or Feedwater system.
- e. Containment heat removal and depressurization systems, including containment spray and fan cooler systems.
- f. Emergency ac electrical power systems, including emergency diesel generators (EDG5).
- g. Emergency service water systems.
Yes El No LI Common Cause Inoperability of Independent Trains or Channels?
Any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to:
- 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
- d. Mitigate the consequences of an accident.
AP-617 Rev. 33 Page 43 of
2013 NRC SRO Question 97 (22) Reference Attachment 8 Sheet 4 of 6 Situation meets this Reportable Event condtjon Yes El No El Event or Condition that Could Have Prevented Fulfillment of a Safety Function?
Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to:
- 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
- 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 available to perform the required safety function.
Yes El No El Single Cause that Could Have Prevented Fulfillment of the Safety Functions of Trains or Channels in Different Systems?
Any event or condition that as a result of a single cause could have prevented the fulfillment of a safety function for two or more trains or channels in different systems that are needed to:
- 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
- d. Mitigate the consequences of an accident.
Events covered above may include cases of procedural error, equipment failure, and/or discovery of design, analysis, fabrication, construction, and/or procedural inadequacy. However, HNP is not required to report an event above if the event results from:
- a. A shared dependency among trains or channels that is a natural or expected consequence of the approved plant design; or
- b. Normal and expected wear or degradation.
Yes El No El Degraded or Unanalyzed Condition?
Any event or condition that resulted in:
- a. The condition of the nuclear power plant including its principal safety barriers, being seriously degraded; or
- b. The nuclear power plant being in an unanalyzed condition that significantly degraded plant safety.
AP-61 7 Rev. 33 Page 44 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 8 Sheet 5 of 6 Situation meets this Reportable Event condition?
Yes El No El External Threat or Hampering?
Any natural phenomenon or other external condition that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear plant.
Yes LI No LI Internal Threat or Hampering?
Any event that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant including fires, toxic gas releases, or radioactive releases.
Yes LI No LI Radioactive Release?
Any airborne radioactive release that, when averaged over a time period of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, resulted in airborne radionuclide concentrations in an unrestricted area that exceeded 20 times the applicable concentration limits specified in appendix B to 10CFR2O, table 2, column 1.
Any liquid effluent release that, when averaged over a time period of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, exceeds 20 times the applicable concentrations specified in Appendix B to 1 OCFR2O table 2, column 2, at the point of entry into the receiving waters (i.e., unrestricted area) for all radionuclides except tritium and dissolved noble gases.
AP-61 7 Rev. 33 Page 45 of 47
2013 NRC SRO Question 97 (22) Reference Attachment 8 Sheet 6 of 6 III. Reportable Evaluation IV. Conclusion Based on the information above, this event:
(check one)
Does NOT meet the reportability requirements of 10 CFR 50.73.
IS reportable to the NRC per the requirements of 10.CFR 50.73.
AP-61 7 Rev. 33 Page 46 of 47
2013 NRC SRO Question 97 (22) Reference Revision Summary (PRR-609830)
General This revision adds a reference for CR 580945. This change is an editorial correction per PRO NGGC-0204.
Description of Changes Page Section Change Description 5 2.0 Added new reference #41 for CR 580945 AP-61 7 Rev. 33 Page 47 of 47