Text

Initial Exam 2013-302 Draft SRO Written Exam
	ML14084A036
Person / Time
Site:	Oconee
Issue date:	03/21/2014
From:	; NRC/RGN-II
To:	; Duke Energy Corp
References
	50-269/13-302, 50-270/13-302, 50-287/13-302 50-269/13-302, 50-270/13-302, 50-287/13-302
	Download: ML14084A036 (89)
	v • d • e

Oconee Nuclear Station Question: 1 ILT44 ONS SRO NRC Examination (1 point)

Given the following plant conditions:

Unit 1 has just reached 100% power following a refueling outage Unit 2 is at 100% power with 93 EFPD Which ONE of the following will result in the highest amount of Emergency Feedwater flow required to stabilize RCS temperature 5 minutes following the trip?

A. Both Main Feedwater Pumps ONLY trip on Unit 1 B. Both Main Feedwater Pumps ONLY trip on Unit 2 C. Loss of Offsite Power on Unit 1 D. Loss of Offsite Power on Unit 2 Page 1 of 100

Oconee Nuclear Station Question: 2 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Reactor power = 100%

1RC-66 (PORV) is leaking past its seat Pressurizer temperature = 648 0F Quench tank pressure = 5 psig Reactor Building pressure = 0 psig Which ONE of the following describes the expected tailpipe temperature (°F) downstream of 1RC-66?

A. 212 B. 228 C. 272 D. 648 Page 2 of 100

Oconee Nuclear Station Question: 3 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Reactor tripped from 100% power due to SBLOCA 1A HPI Pump failed Subcooling Margin = 0°F stable Which ONE of the following is the reason the EOP directs increasing SG levels to the Loss of Subcooling Margin Setpoint level?

A. Establish a large secondary side inventory in support of a rapid RCS cooldown.

B. Establish a large secondary side inventory to ensure that a loss of coupling will NOT occur if a momentary loss of EFDW occurs.

C. Ensure a secondary water level higher than the primary water level inside the SG tubes to establish boiler condenser mode heat transfer D. Ensure a secondary side level sufficient to minimize the consequences of a total loss of feedwater during boiler condenser mode heat transfer Page 3 of 100

Oconee Nuclear Station Question: 4 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Reactor power = 50% stable 1B2 RCP is OFF Which ONE of the following would require immediate entry into AP/1/A/1700/016 (Abnormal Reactor Coolant Pump Operation)?

A. OAC point O1A0061 (RCP 1A1 MTR INPUT POWER) in HI alarm B. OAC point O1A1579 (RCP 1A2 MTR LOWER AIR TEMP) in HI alarm C. 1SA-15/A5 (RC PUMP MOTOR 1B1 OIL POT LOW LEVEL) in alarm D. 1SA-6/D5 (PUMP 1B2 CAVITY PRESS HI/LOW) in alarm Page 4 of 100

Oconee Nuclear Station Question: 5 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial conditions:

Reactor power = 100%

Current conditions:

LPI Flow Train A = 1800 gpm stable LPI Flow Train B = 1780 gpm stable Rule 2 (Loss of SCM) in progress.

IMAs complete

1) The SRO will direct actions from the __ (1) __ tab of the EOP.

2) In accordance with Rule 2, performance of Rule 3 (Loss of Main or Emergency FDW) __ (2) __ required.

Which ONE of the following completes the statements above?

A. 1. LOSCM

2. is B. 1. LOSCM

2. is NOT C. 1. ICC

2. is D. 1. ICC

2. is NOT Page 5 of 100

Oconee Nuclear Station Question: 6 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial conditions:

Normal LPI decay heat removal in service Current conditions:

Loss of offsite power occurs Power restored via CT-4 1A and 1B LPI Pumps NOT available Which ONE of the following describes the requirements to start the 1C LPI Pump?

Manual reset of Load Shed is __(1)___ and starting of 1C LPI Pump is allowed after a MINIMUM of ___(2)__ seconds.

A. 1. NOT required

2. 5 B. 1. required

2. 5 C. 1. NOT required

2. 30 D. 1. required

2. 30 Page 6 of 100

Oconee Nuclear Station Question: 7 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Reactor trip has just occurred Total RCP seal injection flow = 0 gpm Running Component Cooling pump tripped Standby CC pump did not start Which ONE of the following describes the procedure whose performance is directed by the EOP and why?

Initiate A. AP/20 (Loss of CC) to restore Component Cooling B. AP/20 (Loss of CC) to ensure letdown is isolated C. AP/25 (SSF EOP) to align an alternate letdown flowpath D. AP/25 (SSF EOP) to align an alternate source of seal injection Page 7 of 100

Oconee Nuclear Station Question: 8 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial conditions:

Reactor power = 90%

1B Main Feedwater pump trips Current conditions:

Reactor power = 70% decreasing RCS pressure = 2165 psig slowly decreasing Pressurizer level = 228 inches slowly decreasing Pressurizer temperature = 640°F slowly decreasing Pressurizer heater bank 1 (Group A and K) is ON Pressurizer heater banks 2, 3, and 4 are in AUTO and are OFF The pressurizer is ___(1)____ AND the pressurizer heater bank 2 ___(2)____.

Which ONE of the following completes the statement above?

A. 1. subcooled

2. will energize at 2145 psig B. 1. subcooled

2. should be energized C. 1. saturated

2. will energize at 2145 psig D. 1. saturated

2. should be energized Page 8 of 100

Oconee Nuclear Station Question: 9 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 2 conditions:

Loss of all sources of Feedwater has occurred RCS Pressure = 2250 psig increasing Pressurizer level = 294 inches increasing ALL SCMs = 24°F slowly decreasing What is the:

1) lowest RCS pressure (psig) that will require Rule 4 (Initiation of HPI Forced Cooling) to be performed?

2) PRIMARY reason for reducing the number of operating RCPs in accordance with Rule 4?

A. 1. 2300

2. Reduce the heat input to the RCS B. 1. 2300

2. Provide the ability to recover from HPI forced cooling and re-establish a Pressurizer bubble.

C. 1. 2255

2. Reduce the heat input to the RCS D. 1. 2255

2. Provide the ability to recover from HPI forced cooling and re-establish a Pressurizer bubble.

Page 9 of 100

Oconee Nuclear Station Question: 10 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 3 conditions:

Reactor power = 100%

Which ONE of the following will result in a Tech Spec LCO being NOT met?

A. 3A SGTL rate = 160 gpd B. 3B Core Flood Tank level = 12.69 feet C. 3B Core Flood Tank pressure = 622 psig D. 4 gpm RCS leak identified as being through valve stem packing of 3HP-1 Page 10 of 100

Oconee Nuclear Station Question: 11 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

ALL sources of feedwater have been lost Rule 4 (Initiation of HPI Forced Cooling) is complete with outstanding IAATs 1A HPI pump has failed HPI flow parameters are as indicated below In accordance with Rule 4, __(1)__ RCP(s) is/are operating and HPI flow __(2)__

required to be throttled.

Which ONE of the following completes the statement above?

A. 1. 1

2. is B. 1. 1

2. is NOT C. 1. 2

2. is D. 1. 2

2. is NOT Page 11 of 100

Oconee Nuclear Station Question: 12 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 Conditions:

Initial conditions:

Reactor Power = 100%

ACB-4 closed Current conditions:

Reactor trip CT-1 Locks out KHU-2 Emergency Lockout occurs Assuming no additional failures, which ONE of the following describes the first method used to restore power to Unit 1 MFBs?

A. Automatically through ACB-3 B. Automatically through SL1 and SL2 C. Manually through ACB-3 in accordance with Enclosure 5.38 (Restoration of Power) of the EOP D. Manually through SL1 and SL2 in accordance with Enclosure 5.38 (Restoration of Power) of the EOP Page 12 of 100

Oconee Nuclear Station Question: 13 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Unit shutdown in progress Reactor power = 38% slowly decreasing LOOP (Switchyard Isolation) occurs

1) Based on the conditions above, the status of the Main Turbine will be __(1)__ 5 minutes following the LOOP?

2) ANYTIME the Main Turbine is tripped, ICS uses __(2)__ to control the Turbine Bypass Valves?

Which ONE of the following completes the statements above?

A. 1. tripped

2. Turbine Header Pressure B. 1. tripped

2. Steam Generator Outlet Pressure C. 1. NOT tripped

2. Turbine Header Pressure D. 1. NOT tripped

2. Steam Generator Outlet Pressure Page 13 of 100

Oconee Nuclear Station Question: 14 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial conditions Reactor Power = 100%

SASS in Manual while SPOC repairs Pressurizer Level 3 level transmitter 1HP-120 in AUTO selected to Pressurizer Level 1 Current conditions:

Vital Power to ICCM Train A fails Which ONE of the following describes Pressurizer level control with 1HP-120?

A. Selecting Pressurizer Level 2 and depressing the AUTO pushbutton on 1HP-120 are required to restore automatic control at setpoint B. Selecting Pressurizer Level 2 ONLY will restore automatic control at setpoint C. Manual control using 1HP-120 Bailey controller is all that is available D. Additional actions are NOT required since Automatic control at setoint is retained Page 14 of 100

Oconee Nuclear Station Question: 15 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial conditions:

Reactor power = 100%

Instrument Air pressure = 85 psig decreasing AP/22 (Loss of Instrument Air) has been initiated Which ONE of the following

1) is the higher Instrument Air pressure (psig) that would require an immediate manual Reactor trip in accordance with AP/22?

2) states the reason AP/22 directs tripping the Main FDW pumps immediately after tripping the Reactor as described above?

A. 1. 70

2. Controlling FDW valves fail as is B. 1. 65

2. Controlling FDW valves fail as is C. 1. 70

2. Controlling FDW valves fail closed D. 1. 65

2. Controlling FDW valves fail closed Page 15 of 100

Oconee Nuclear Station Question: 16 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial conditions:

AP/34 (Degraded Grid) in progress Generator output = 850 MWe and 450 MVARs Generator Hydrogen Pressure = 60 psig Generator Output Voltage = 18.2 KV

1) The Generator output __ (1) __ within the limits of the Generator Capability Curve.

2) If the generator exceeds the Underfrequency Maximum Allowable Time given in AP/34 (Degraded Grid) the Main Turbine __ (2) __ automatically trip.

Which ONE of the following completes the statements above?

REFERENCE PROVIDED A. 1. is NOT

2. will B. 1. is NOT

2. will NOT C. 1. is

2. will D. 1. is

2. will NOT Page 16 of 100

Oconee Nuclear Station Question: 17 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 3 conditions:

A brief loss of power has occurred Unit auxiliaries are being supplied from the switchyard via CT-3 Subsequent Actions tab in progress

1) Subsequent Actions directs restarting __(1)__.

2) The __(2)__ RCP will provide the best Pressurizer Spray.

Which ONE of the following completes the statements above?

A. 1. one RCP ONLY

2. 3A1 B. 1. one RCP ONLY

2. 3B1 C. 1. one RCP per loop

2. 3A1 D. 1. one RCP per loop

2. 3B1 Page 17 of 100

Oconee Nuclear Station Question: 18 ILT44 ONS SRO NRC Examination (1 point)

In accordance with the EOP, which ONE of the following describes the instruments that are used when initially stabilizing RCS temperature following a Main Steam Line Break and states one of the reasons why they are used?

A. Tcolds are used since Tech Specs specifies that Tcold is RCS temperature B. Tcolds are used since they are the coldest temperature and therefore most indicative of PTS issues C. CETCs are used since the resultant RCS cooldown may result in Tcold being off scale low D. CETCs are used since they are qualified instruments and are therefore more reliable in the hostile containment environment Page 18 of 100

Oconee Nuclear Station Question: 19 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial conditions:

Time = 1200 Reactor power = 100%

1A steam generator tube leak = 2.1 gpd stable RCS activity = 0.25 Ci/ml DEI increasing Current conditions:

Time = 1400 NO change in 1A SG tube leak rate RCS activity = 0.65 Ci/ml DEI increasing Which ONE of the following describes the response of the radiation monitors between 1200 and 1400?

A. 1RIA-59 (N-16 monitor) and 1RIA-40 (CSAE Off-gas) increased.

B. 1RIA-16 (Main Steam Line Monitor) and 1RIA-40 increased.

C. 1RIA-59 increased while1RIA-40 remained constant.

D. 1RIA-16 increased while 1RIA-40 remained constant.

Page 19 of 100

Oconee Nuclear Station Question: 20 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 2 condition:

Initial conditions:

Time = 0900 Reactor Startup in progress NI 1 & 2 = 370 cps NI 3 & 4 = 0 cps (out of service)

ALL WR NIs = ~ 2.7 E-4%

Current conditions:

Time = 0901 NI 1 & 2 are inoperable Which ONE of the following describes:

1) immediate actions required by Tech Spec 3.3.9 (Source Range Neutron Flux)?

2) the reason for the actions described above?

A. 1. Insert Control Rods to Group 1 at 50% withdrawn

2. Prevents power increases when the primary power indication for the operator is not available.

B. 1. Insert Control Rods to Group 1 at 50% withdrawn

2. 2 dpm Startup Rate Control Rod Out Inhibit is no longer available C. 1. Fully insert all Control Rods

2. Prevents power increases when the primary power indication for the operator is not available.

D. 1. Fully insert all Control Rods

2. 2 dpm Startup Rate Control Rod Out Inhibit is no longer available Page 20 of 100

Oconee Nuclear Station Question: 21 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Reactor power = 92% decreasing Unit shutdown in progress per the SGTR tab

1) In accordance with the SGTR tab and Enclosure 5.5 (Pzr and LDST Level Control),

RCS makeup and letdown will be adjusted to maintain Pressurizer level between

__ (1) __ inches.

2) The reason for this Pzr level band is to provide adequate inventory to __ (2) __.

Which ONE of the following completes the statements above?

A. 1. 140 - 180

2. ensure Pzr heaters will remain covered if a subsequent reactor trip occurs B. 1. 140 - 180

2. accommodate system shrinkage during shutdown/cooldown from 18%

power C. 1. 220 - 260

2. ensure Pzr heaters will remain covered if a subsequent reactor trip occurs D. 1. 220 - 260

2. accommodate system shrinkage during shutdown/cooldown from 18%

power Page 21 of 100

Oconee Nuclear Station Question: 22 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Reactor power = 100%

Which ONE of the following will result in an AUTOMATIC trip of the Main Turbine?

A. Bearing Oil Pressure = 5.5 psig B. Main Turbine speed = 1955 RPM C. EHC Hydraulic Oil pressure = 1210 psig D. EITHER Steam Generator Level = 93% OR Page 22 of 100

Oconee Nuclear Station Question: 23 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial conditions:

Reactor power = 100%

1A GWD tank release in progress 1RIA-38 OOS Current conditions:

Maintenance activities in the area result in an inadvertent loss of power to RM-80 skid of 1RIA-37 1SA8/B9 RM PROCESS MONITOR RADIATION HIGH in alarm 1SA8/B10 RM PROCESS MONITOR FAULT in alarm

1) 1GWD-4 (A GWD TANK DISCHARGE) should __(1)__.

2) The required Completion Time in SLC 16.11.3 (Radioactive Effluent Monitoring Instrumentation) for suspending the release by this pathway if not already isolated and both 1RIA-37 and 1RIA-38 become inoperable is __(2)__.

Which ONE of the following completes the statements above?

A. 1. remain open

2. immediately B. 1. automatically close

2. immediately C. 1. remain open

2. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months D. 1. automatically close

2. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Page 23 of 100

Oconee Nuclear Station Question: 24 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Reactor power = 2%

1SA2/B11 (ICS AUTO POWER FAILURE) actuated 1SA2/B13 (ICS HAND POWER FAILURE) actuated Which ONE of the following describes:

1) the level at which SGs will be maintained?

2) how decay heat removal from the core is controlled?

A. 1. 25 inches SUR

2. ADVs B. 1. 30 inches XSUR

2. ADVs C. 1. 25 inches SUR

2. TBVs D. 1. 30 inches XSUR

2. TBVs Page 24 of 100

Oconee Nuclear Station Question: 25 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial conditions:

Switchyard isolation occurs Current conditions:

Shutdown of KHUs is desired Which ONE of the following states:

1) if a Load Shed has occurred?

2) the procedure that will be used to perform a remote shutdown of the KHUs?

A. 1. Yes

2. OP/0/A/2000/041 (Keowee Modes of Operations)

B. 1. No

2. OP/0/A/2000/041 (Keowee Modes of Operations)

C. 1. Yes

2. OP/0/A/1106/019 (Keowee Hydro At Oconee)

D. 1. No

2. OP/0/A/1106/019 (Keowee Hydro At Oconee)

Page 25 of 100

Oconee Nuclear Station Question: 26 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

ES 1-8 have actuated LOCA CD tab in progress RCS pressure = 423 psig slowly decreasing 1A LPI Pump operating in the Piggyback alignment Which ONE of the following describes the:

1) operational limitations on the operating LPI pump?

2) pump(s) being protected by the above limitation?

A. 1. Maximized to < 3100 gpm

2. LPI B. 1. Maximized to < 3100 gpm

2. HPI C. 1. Maximized to < 2900 gpm

2. LPI D. 1. Maximized to < 2900 gpm

2. HPI Page 26 of 100

Oconee Nuclear Station Question: 27 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Reactor trip from 100% due to a SBLOCA Reactor building pressure has peaked at 1.7 psig Subcooled margins are stable as indicated below Which ONE of the following describes how Feedwater will be used to mitigate this event?

Steam Generator levels will be controlled at ________?

A. 240 inches using Emergency Feedwater B. 240 inches using Main Feedwater C. Loss of Subcooling Margin setpoint using Emergency Feedwater D. Loss of Subcooling Margin setpoint using Main Feedwater Page 27 of 100

Oconee Nuclear Station Question: 28 ILT44 ONS SRO NRC Examination (1 point)

Which ONE of the following describes:

1) the design purpose of extending RCP coast down time with the flywheel?

2) an expected core delta T (°F) 30 minutes following a lockout of 1TA and 1TB?

A. 1. Helps prevent the core from reaching DNBR limits

2. 35 B. 1. Helps prevent the core from reaching DNBR limits

2. 47 C. 1. Reduces the likelihood of a Reactor trip following a RCP trip at power

2. 35 D. 1. Reduces the likelihood of a Reactor trip following a RCP trip at power

2. 47 Page 28 of 100

Oconee Nuclear Station Question: 29 ILT44 ONS SRO NRC Examination (1 point)

1) The Letdown Storage Tank contains approximately __(1)__ gallons of water per inch of level.

2) The HIGHER Letdown Storage Tank level that will automatically open 1HP-24 and 1HP-25 is __(2)__ inches.

Which ONE of the following completes the statements above?

A. 1. 24

2. 38 B. 1. 24

2. 54 C. 1. 31

2. 38 D. 1. 31

2. 54 Page 29 of 100

Oconee Nuclear Station Question: 30 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 2 conditions:

RCS Cooldown in progress 2B LPI cooler isolated due to cooler leak Which ONE of the following states the:

1) LPI Decay Heat Removal mode that will be used for the INITIAL transition to LPI cooling?

2) HIGHER RCS pressure (psig) that will allow aligning LPI in the Normal Decay Heat Removal alignment?

A. 1. Switchover

2. 220 B. 1. High Pressure

2. 220 C. 1. Switchover

2. 115 D. 1. High Pressure

2. 115 Page 30 of 100

Oconee Nuclear Station Question: 31 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Rule 3 initiated Loss of Heat Transfer tab in progress Efforts underway to re-establish Steam Generator cooling 1SA-18/D1 (RC SYSTEM APPROACHING SATURATED CONDTIONS) in alarm 1SA-2/D3 (RC PRESS HIGH/LOW) in alarm Pressurizer level = 380 slowly increasing RCS pressure = 2240 psig slowly increasing SCM = 0°F Which ONE of the following states which additional EOP Rules (if any) should be initiated?

A. NO additional rules required B. Rule 2 (Loss of SCM) ONLY C. Rule 4 (Initiation of HPI Forced Cooling) ONLY D. Rule 2 AND Rule 4 Page 31 of 100

Oconee Nuclear Station Question: 32 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 2 condition:

Initial conditions:

Unit startup in progress RCS temperature = 310°F slowly increasing Maintenance in progress in the area of 2DIB panelboard Current conditions:

2DIB breaker #24 (2RC-66 Pilot Valve DC solenoid power supply) is inadvertently opened Which ONE of the following describes:

1) a Tech Spec Limiting Condition of Operation that is NOT met?

2) the position of 2RC-66?

A. 1. 3.4.9 (Pressurizer)

2. Open B. 1. 3.4.9 (Pressurizer)

2. Closed C. 1. 3.4.12 (LTOP)

2. Open D. 1. 3.4.12 (LTOP)

2. Closed Page 32 of 100

Oconee Nuclear Station Question: 33 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial conditions Loss of all Feedwater HPI forced cooling initiated Quench Tank pressure = 40 psig increasing RCS activity indicates no fuel failures present Current conditions Quench Tank pressure = 3 psig stable Which ONE of the following describes the:

1) reactor building RIAs response to the above conditions?

2) valve(s) that will automatically close anytime 1RIA-49 reaches its HIGH alarm setpoint?

A. 1. increases

2. 1LWD-1 AND 1LWD-2 B. 1. remains constant

2. 1LWD-1 AND 1LWD-2 C. 1. increases

2. 1LWD-2 ONLY D. 1. remains constant

2. 1LWD-2 ONLY Page 33 of 100

Oconee Nuclear Station Question: 34 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

1SA-08/B-9 (PROCESS MONITOR RADIATION HIGH) 1RIA-50 in HIGH alarm CC Surge Tank level increasing

1) The CC Surge tank __(1)__.

2) If the RCS leakage threatens to overflow the associated waste tank, AP/1/A/1700/002 (Excessive RCS Leakage) will direct __(2)__.

Which ONE of the following completes the statements above?

A. 1. will overflow to the LAWT

2. tripping the Reactor B. 1. will overflow to the LAWT

2. initiating a shutdown using AP/1/A/1700/029 (Rapid Unit Shutdown)

C. 1. will overflow to a floor drain which drains to the MWHUT

2. tripping the Reactor D. 1. will overflow to a floor drain which drains to the MWHUT

2. initiating a shutdown using AP/1/A/1700/029 (Rapid Unit Shutdown)

Page 34 of 100

Oconee Nuclear Station Question: 35 ILT44 ONS SRO NRC Examination (1 point)

Which ONE of the following states the automatic OPEN setpoints (psig) for 1RC-1 (Pzr Spray) and 1RC-66 (PORV) in Mode 1?

1RC-1 1RC-66 A. 2205 2450 B. 2205 2500 C. 2255 2450 D. 2255 2500 Page 35 of 100

Oconee Nuclear Station Question: 36 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Reactor power = 100%

1D RPS channel in Manual Bypass 1A RPS Thot RTD fails low Which ONE of the following:

1) describes ALL 1A RPS functions affected by the failure?

2) states if OP/1/A/1105/014 (Control Room Instrumentation Operation And Information) requires tripping the 1A RPS channel?

A. 1. RCS High Outlet Temperature ONLY

2. yes B. 1. RCS High Outlet Temperature ONLY

2. no C. 1. RCS High Outlet Temperature and RCS Variable Low Pressure

2. yes D. 1. RCS High Outlet Temperature and RCS Variable Low Pressure

2. no Page 36 of 100

Oconee Nuclear Station Question: 37 ILT44 ONS SRO NRC Examination (1 point)

Which ONE of the following describes how RCS Pressure signals are used to provide control signals to the Integrated Control System?

A. Median Selected from one wide range pressure and two channels of RPS narrow range pressure (A and B)

B. Median Selected from three channels of RPS narrow range pressure (A, B, and E)

C. 2nd Max Selected from RPS narrow range pressures (A, B, C, & D)

D. 2nd Min Selected from RPS narrow range pressures (A, B, C, & D)

Page 37 of 100

Oconee Nuclear Station Question: 38 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 2 conditions:

Initial conditions:

Reactor power = 100%

Current conditions:

MSLB occurs RCS pressure = 1580 psig slowly increasing RB peak pressure = 2.8 psig Which ONE of the following describes a valve that has received a signal to CLOSE?

A. 2CC-7 B. 2HP-24 C. 2LWD-2 D. 2LPSW-1062 Page 38 of 100

Oconee Nuclear Station Question: 39 ILT44 ONS SRO NRC Examination (1 point)

Given the following plant conditions:

Time = 1200 Unit 1 Reactor power = 100%

Unit 2 Reactor power = 100%

ACB-4 closed Time = 1201 LOCA occurs on Unit 1 Switchyard Isolation occurs Which ONE of the following states the source of power being used to energize 1DIA at Time = 1202?

A. Control Batteries B. KHU-1 C. KHU-2 D. CT-5 Page 39 of 100

Oconee Nuclear Station Question: 40 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 2 conditions:

Time = 1200 Reactor in MODE 5 LPI aligned to normal Decay Heat Removal mode 2RIA-35 (Combined LPSW effluent leaving the Reactor Building and the Auxiliary Building) is isolated for repair Time = 1300 Large LPI Cooler leak in 2A LPI Cooler occurs Time = 1500 Actions to isolate the 2A LPI Cooler are initiated RCS temperature slowly increasing

1) At Time = 1330 there __(1)__ an RIA alarm indicating the LPI Cooler leak?

2) Entry into MODE 4 will occur when RCS temperature exceeds __(2)__°F?

Which ONE of the following completes the statements above?

A. 1. is

2. 200 B. 1. is

2. 250 C. 1. is NOT

2. 200 D. 1. is NOT

2. 250 Page 40 of 100

Oconee Nuclear Station Question: 41 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Reactor is in MODE 5 RB Purge in progress Unit 1 vent activity increasing 1RIA-45 HIGH alarm fails to actuate at setpoint

1) Automatic termination of RB Purge operation due to increasing activity __(2)__

available?

2) Purge operation __(1)__ be allowed if the unit were in MODE 4.

Which ONE of the following completes the statements above?

A. 1. is

2. would B. 1. is

2. would NOT C. 1. is NOT

2. would D. 1. is NOT

2. would NOT Page 41 of 100

Oconee Nuclear Station Question: 42 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

1A LPSW Pump trips Standby LPSW pump fails to start Which ONE of the following will begin to increase in temperature?

ASSUME NO MANUAL ACTIONS ARE TAKEN A. Letdown B. Spent Fuel Pool C. Main Feedwater Pump oil temperature D. Primary Instrument Air Compressor discharge air temperature Page 42 of 100

Oconee Nuclear Station Question: 43 ILT44 ONS SRO NRC Examination (1 point)

Which ONE of the following is the power supply for the Unit 2 Auxiliary Instrument Air System compressor?

A. 2XD B. 2XF C. 2XP D. 2XS1 Page 43 of 100

Oconee Nuclear Station Question: 44 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial conditions:

Reactor Power = 100%

1A MSLB inside containment Current conditions:

Core SCM = 18°F stable RB Pressure = 17 psig slowly decreasing Which ONE of the following sets of actions is required by Enclosure 5.1 (ES Actuation)

A. Take ES Channel 1 to manual AND open 1HP-20 B. Take ES Channel 1 to manual AND open 1HP-3 C. Override Odd Voters AND open 1HP-20 D. Override Odd Voters AND open 1HP-3 Page 44 of 100

Oconee Nuclear Station Question: 45 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial conditions:

Reactor Power = 100%

Which ONE of the following:

1) describes the LOWER SG Operating Range level (%) that will result in an automatic trip of both Main Feedwater pumps?

2) states if SG Operating Range level indications are temperature compensated?

A. 1. 87

2. No B. 1. 87

2. Yes C. 1. 97

2. No D. 1. 97

2. Yes Page 45 of 100

Oconee Nuclear Station Question: 46 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Reactor power = 50% slowly decreasing OAC Unavailable Computer Reactor Calculation Package NOT running Which ONE of the following:

1) is the HIGHER power level (% Power) at which Tech Spec limits on Reactor Power Imbalance do NOT apply?

2) describes how OP/1/A/1105/014 (Control Room Instrumentation Operation And Information) directs the operator to determine if Imbalance limits specified in the COLR have been exceeded?

A. 1. 35

2. by use of CR gages for Power Range NIs B. 1. 35

2. by performing PT/1/A/1103/019 (Backup Incore Detector System)

C. 1. 15

2. by use of CR gages for Power Range NIs D. 1. 15

2. by performing PT/1/A/1103/019 (Backup Incore Detector System)

Page 46 of 100

Oconee Nuclear Station Question: 47 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial conditions:

Mode 6 REFUELING is in progress All four SR NIs in service SR 1NI-1 and SR 1NI-3 are the designated NIs for Fuel Handling Current conditions:

Power supply to SR 1NI-1 fails (0 vdc)

Which ONE of the following describes the impact on refueling activities in accordance with OP/1/A/1502/007 (Operations Defueling/Refueling Responsibilities)?

A. Allowed to continue because two other SR NIs remain in service B. Allowed to continue because SR NI-3 is still in service C. Required to be stopped until another SR NI is designated because other NIs are procedurally allowed to be designated D. Required to be stopped and cannot be resumed until SR 1NI-1 is returned to service because other NIs are NOT procedurally allowed to be designated Page 47 of 100

Oconee Nuclear Station Question: 48 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Time = 1200 Reactor Power = 40% stable following an instrument failure Turbine Header Pressure = 860 psig stable Feedwater, Reactor, and Main Turbine in Manual Time = 1300 ICS in Automatic Turbine Header Pressure = 860 psig stable Time = 1301 Reactor Trips prior to any additional Turbine Header Pressure setpoint adjustments Which ONE of the following is the pressure (psig) where the Turbine Bypass Valves will automatically control Steam Generator pressure?

A. 885 B. 910 C. 985 D. 1010 Page 48 of 100

Oconee Nuclear Station Question: 49 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Time = 1200:00 Reactor power = 80% stable 1A and 1B CBP operating Time = 1201:00 1A CBP trips Feedwater Pump suction pressure = 225 psig slowly decreasing Time = 1203:00 Feedwater Pump suction pressure = 220 slowly increasing Which ONE of the following describes the:

1) runback rate (%/min) inserted at Time = 1201:00 to ICS?

2) procedure that will be directed by the CRS at Time = 1203:00?

A. 1. 15

2. AP/1/A/1700/001 (Unit Runback)

B. 1. 15

2. EOP C. 1. 20

2. AP/1/A/1700/001 (Unit Runback)

D. 1. 20

2. EOP Page 49 of 100

Oconee Nuclear Station Question: 50 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Reactor power = 100%

Primary to Secondary leakage of 10 gpd has just been detected AP/1/A/1700/031 (Primary to Secondary Leakage) has been initiated

1) In accordance with AP/31, opening the Turbine Building Sump (TSP) pump breakers prior to being ready to hang White Tags on the TBS pump breakers

__(1)__ allowed.

2) A sustained loss of power to 1RIA-54 will trip BOTH Turbine Building Sump Pumps

__(2)__.

Which ONE of the following completes the statements above?

A. 1. is NOT

2. after a 2 minute timer B. 1. is NOT

2. immediately C. 1. is

2. after a 2 minute timer D. 1. is

2. immediately Page 50 of 100

Oconee Nuclear Station Question: 51 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial conditions:

Time = 1200 1A GWD tank pressure = 68 psig stable Current conditions:

Time = 1205 1A GWD tank pressure = 18 psig rapidly decreasing Various Aux Building RIAs in alarm 1RIA-1 (Control Room Monitor) in HIGH alarm 1RIA-39 (CNTL RM Gas) in HIGH alarm AP/1/A/1700/018 (Abnormal Release of Radioactivity) in progress A and B Outside Air Booster Fans have been started Which ONE of the following:

1) states if 1RIA-1 has a local alarm (do not count associated statalarm(s))?

2) describes the areas being provided outside air via the Outside Air Booster Fans?

A. 1. Yes

2. Control Room ONLY B. 1. No

2. Control Room ONLY C. 1. Yes

2. Control Room, Cable Rooms, and the Equipment Rooms D. 1. No

2. Control Room, Cable Rooms, and the Equipment Rooms Page 51 of 100

Oconee Nuclear Station Question: 52 ILT44 ONS SRO NRC Examination (1 point) 1RIA-59 setpoints are set by __(1)__ and the MINIMUM power level at which 1RIA-59 is used to determine SGTL rate is __(2)__ (% power) in accordance with the EOP.

Which ONE of the following completes the statement above?

A. 1. I&E

2. 20 B. 1. I&E

2. 40 C. 1. ROs

2. 20 D. 1. ROs

2. 40 Page 52 of 100

Oconee Nuclear Station Question: 53 ILT44 ONS SRO NRC Examination (1 point)

Which ONE of the following states all of the switchgear that can supply power to the B LPSW pump?

A. 1TD ONLY B. 2TC ONLY C. 1TC or 2TD D. 1TD or 2TD Page 53 of 100

Oconee Nuclear Station Question: 54 ILT44 ONS SRO NRC Examination (1 point)

Based on the graph above, which ONE of the following describes the EARLIEST time at which SA-141 (SA to IA Controller) will automatically open?

A. 1207 B. 1210 C. 1212 D. 1215 Page 54 of 100

Oconee Nuclear Station Question: 55 ILT44 ONS SRO NRC Examination (1 point)

Which ONE of the following describes what should be used in the case of a large Hydrogen leak to maintain Hydrogen concentration below the lower flammability limit in accordance with OP/1/A/1106/017 (Hydrogen System)?

A. CO2 B. Water C. Halon D. Foam fire retardant Page 55 of 100

Oconee Nuclear Station Question: 56 ILT44 ONS SRO NRC Examination (1 point)

Which ONE of the following is the LOWER limit on RCS activity that would require entry into AP/21 (RCS Activity)?

A. Xe-133 = 0.25 µCi/gm B. Xe-133 = 1.0 µCi/gm C. DEI = 0.25 µCi/gm D. DEI = 1.0 µCi/gm Page 56 of 100

Oconee Nuclear Station Question: 57 ILT44 ONS SRO NRC Examination (1 point)

Which ONE of the following activities complies with guidance contained in SOMP 1-2 (Reactivity Management)?

A. Manual rod withdrawal during a Feedwater transient to stop a temperature decrease caused by an instrument failure B. Manually increasing Feedwater flow to stop an RCS pressure increase caused by an RCS temperature increase C. Manually raising one Loop FDW demand while lowering the other Loop FDW demand to control Tcold following an RCP trip D. Manually increasing turbine demand to adjust RCS temperature Page 57 of 100

Oconee Nuclear Station Question: 58 ILT44 ONS SRO NRC Examination (1 point)

Which ONE of the following tags would be used ONLY for configuration control of 1HP-409 in accordance with NSD-500 (Red Tags/Configuration Control Tags)?

A. White Tag B. MORT Tag C. OORT Tag D. CORT Tag Page 58 of 100

Oconee Nuclear Station Question: 59 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 condition:

Reactor in MODE 1 Which ONE of the following is the MINIMUM Pressurizer level (inches) that would require declaring Tech Spec 3.4.9 (Pressurizer) LCO NOT met in accordance with PT/1/A/0600/001 (Periodic Instrument Surveillance)?

A. 240 B. 260 C. 285 D. 340 Page 59 of 100

Oconee Nuclear Station Question: 60 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Reactor trip due to loss of both Main FDW pumps Instrument Air pressure = 0 psig Auxiliary Instrument Air pressure= 0 psig Which ONE of the following describes the status of 1FDW-315 and 1FDW-316?

A. Available for Manual operation ONLY once the air supply was lost B. Will be available for Automatic operation for a MINIMUM of 30 minutes from the loss of air supply C. Will be available for Automatic operation for a MINIMUM of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from the loss of air supply D. Will be available for Automatic operation for a MINIMUM of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months from the loss of air supply Page 60 of 100

Oconee Nuclear Station Question: 61 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 2 conditions:

Initial conditions:

Time = 1200 RCS temperature = 92°F stable RB Purge in progress 2RIA-45 HIGH alarm setpoint = 1520 cpm 2RIA-45 = 1342 cpm stable Current conditions:

Time = 1205 2RIA-45 = 1520 cpm increasing Which ONE of the following describes:

1) ALL valves that will CLOSE?

2) 2RIA-46 reading (cpm) at time = 1200?

A. 1. 2PR-1 through 2PR-6

2. Zero B. 1. 2PR-1 through 2PR-6

2. 1342 C. 1. 2PR-2 through 2PR-5 ONLY

2. Zero D. 1. 2PR-2 through 2PR-5 ONLY

2. 1342 Page 61 of 100

Oconee Nuclear Station Question: 62 ILT44 ONS SRO NRC Examination (1 point)

Given the following plant conditions:

Venting the 1C LPI Pump in progress using the following RWP information:

o Dose Alarm : 25 mrem o Dose Rate Alarm: 200 mrem/hr o Dose Alarm: Stop work - Exit Area - Notify RP o Unanticipated Dose Rate Alarm: Stop Work - Exit Area - Notify RP Which ONE of the following states the MAXIMUM time work can continue before complying with the RWP will require exiting the area?

SEE PLAN VIEW PROVIDED Do NOT consider dose received while traveling to or from the job.

A. 15 minutes B. 30 minutes C. 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months D. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Page 62 of 100

Oconee Nuclear Station Question: 63 ILT44 ONS SRO NRC Examination (1 point)

Of the two tabs below, the __(1)__ tab of the EOP has a higher priority because

__(2)__.

Which ONE of the following completes the statement above?

A. 1. LOSCM

2. prompt actions are required to ensure core cooling is maintained B. 1. LOSCM

2. actions to initiate HPI injection are required prior to the RCS void fraction reaching 70%.

C. 1. SGTR

2. a Reactor trip with a SGTR results in a direct release path for radionuclides to the environment D. 1. SGTR

2. actions to depressurize RCS to minimize SCM during a SGTR is a Time Critical Action that may not otherwise be met Page 63 of 100

Oconee Nuclear Station Question: 64 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Reactor Power = 70%

Which ONE of the following would require entry into the EOP?

A. Condenser vacuum = 22.3 hg B. 1RIA-59 = 31.4 gpm C. 1B Main FDW pump trips D. 1A1 RCP trips Page 64 of 100

Oconee Nuclear Station Question: 65 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial conditions:

Reactor power = 100%

1KVIA Panelboard de-energized Current conditions:

MSLB inside the Reactor Building occurs Lowest RCS pressure = 1137 psig Reactor Building Pressure peaked at 32 psig Which ONE of the following describes ALL ES Actuation Logic Channels that have actuated?

A. 1, 3, 5, 7 B. 2, 4, 6, 8 C. 1, 5, 7 ONLY D. 2, 6, 8 ONLY Page 65 of 100

Oconee Nuclear Station Question: 66 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial conditions:

Time = 1200 Reactor Power = 100%

1A MSLB inside the Reactor Building Current conditions:

Time = 1201 Reactor Building Pressure = 3 psig increasing Which ONE of the following describes the operation of 1LPSW-18?

A. It is NORMALLY fully open however it will receive a signal to open from ES-5 at 1201 B. It is NORMALLY throttled and will go fully open when it receives a signal to open from ES-5 at 1201 C. It is NORMALLY fully open however it will receive a signal to open from ES-5 at 1204 D. It is NORMALLY throttled and will go fully open when it receives a signal to open from ES-5 at 1204 Page 66 of 100

Oconee Nuclear Station Question: 67 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial conditions:

Reactor power = 50%

Current conditions:

LBLOCA occurs 1TD de-energized 1B RBCU switch in OFF Which ONE of the following describes the status of the below listed Reactor Building Cooling Units five (5) minutes after ES actuates?

ASSUME NO OPERATOR ACTIONS 1B RBCU 1C RBCU A. LOW LOW B. LOW OFF C. OFF LOW D. OFF OFF Page 67 of 100

Oconee Nuclear Station Question: 68 ILT44 ONS SRO NRC Examination (1 point)

Given the following plant conditions:

SBLOCA has occurred on Unit 1 Reactor Building Pressure = 11.2 psig slowly decreasing Unit 2 Reactor Power = 100%

Which ONE of the following describes the actions directed (if any) by Enclosure 5.1 (ES Actuation) to ensure the required LPSW flow exists in the 1A LPI cooler?

A. Place 1LPSW-251 in Failed Open ONLY B. Place 1LPSW-251 in Failed Open AND fully open 1LPSW-4 C. Place 1LPSW-251 in Failed Open AND Throttle LPSW flow to approximately 3000 gpm using 1LPSW-4 D. Place 1LPSW-251 in Failed Open AND Throttle LPSW flow to approximately 5200 gpm using 1LPSW-4 Page 68 of 100

Oconee Nuclear Station Question: 69 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial conditions:

Reactor power = 100%

Loss of offsite power occurs Current conditions:

Main Feeder Buses remain de-energized

1) The position of 1MS-112 (SSRH Control) is __(1)__.

2) 1MS-77 (MS to MSRH) __(2)__ be operated from the control room switch.

Which ONE of the following completes the statements above?

A. 1. open

2. can B. 1. closed

2. can C. 1. open

2. can NOT D. 1. closed

2. can NOT Page 69 of 100

Oconee Nuclear Station Question: 70 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Reactor Power = 50%

1A Turbine Bypass Valve fails OPEN Which ONE of the following describes the plant response?

ASSUME NO OPERATOR ACTIONS Reactor power will...

A. Increase then return to pre-transient level.

B. Increase and stabilize at a higher power level.

C. Decrease then return to pre-transient level.

D. Decrease and stabilize at a lower power level.

Page 70 of 100

Oconee Nuclear Station Question: 71 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Reactor Power = 80% stable ICS in Manual 1B Main Feedwater Pump trips Which ONE of the following is the MAXIMUM power level allowed in accordance with AP/1 (Plant Runback).

A. 74%

B. 65%

C. 60%

D. 55%

Page 71 of 100

Oconee Nuclear Station Question: 72 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 3 conditions:

Initial conditions:

Reactor tripped from 35% power due to 3TA lockout 3A Main FDW pump operating 3FDW-35 & 3FDW-44 (3A and 3B Startup FDW Control) in MANUAL 3A and 3B SG levels = 38" SU and stable Current conditions:

3FDW-35 & 44 are placed in Automatic Which ONE of the following describes the response of 3FDW-35 & 44?

A. Travel open to increase SG levels to 240" XSUR.

B. Travel open to increase SG levels to 50% on Operating level.

C. Travel closed to decrease SG level to 30" on XSUR.

D. Travel closed to decrease SG level to 25" on SU level.

Page 72 of 100

Oconee Nuclear Station Question: 73 ILT44 ONS SRO NRC Examination (1 point)

Which ONE of the following describes the:

1) primary concern at ONS regarding Main Feedwater backleakage into the EFDW discharge piping?

2) method used to determine if Main Feedwater backleakage into the EFDW discharge piping is occurring?

A. 1. Vapor binding of the EFDW pumps

2. locally monitoring EFDW pump discharge piping for increasing temperature B. 1. Vapor binding of the EFDW pumps

2. Monitoring EFDW temperature OAC points for increasing temperature C. 1. Overpressurizing the EFDW system piping

2. locally monitoring EFDW pump discharge piping for increasing temperature D. 1. Overpressurizing the EFDW system piping

2. Monitoring EFDW temperature OAC points for increasing temperature Page 73 of 100

Oconee Nuclear Station Question: 74 ILT44 ONS SRO NRC Examination (1 point)

Given the following plant conditions:

No Keowee Units are operating ACB-3 closed

1) KHU 1X switchgear is being powered from __ (1) __.

2) Keowee control power will be available for a MINIMUM of approximately __ (2) __

hour(s) following a loss of ALL AC power.

Which ONE of the following completes the statements above?

A. 1. 1TC

2. one B. 1. 1TC

2. four C. 1. the 230 KV switchyard

2. one D. 1. the 230 KV switchyard

2. four Page 74 of 100

Oconee Nuclear Station Question: 75 ILT44 ONS SRO NRC Examination (1 point)

Given the following plant conditions:

3CA Battery Charger fails - output voltage = 0 VDC 3CA Battery voltage = 120 VDC 3DCB Bus voltage = 123 VDC Unit 1 DCA/DCB Bus voltage = 125 VDC Unit 2 DCA/DCB Bus voltage = 127 VDC Which ONE of the following will automatically supply power to 3DIA panelboard?

A. 3CA Battery B. Unit 1 DC Bus C. 3DCB Bus D. Unit 2 DC Bus Page 75 of 100

Oconee Nuclear Station Question: 76 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial conditions:

Reactor power = 100%

Quench Tank temperature, level, and pressure are slowly increasing 1RC-66 tailpipe temperature slowly increasing Current conditions:

1RC-4 has been closed to verify the leak source Which ONE of the following states:

1) if ALL equipment credited in the UFSAR to prevent the RCS from exceeding the Tech Spec Safety Limit on RCS pressure is still available?

2) the LOWER Quench Tank pressure that will result in blowing the Quench Tank rupture disc?

A. 1. No

2. 45 psig B. 1. No

2. 55 psig C. 1. Yes

2. 45 psig D. 1. Yes

2. 55 psig Page 76 of 100

Oconee Nuclear Station Question: 77 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Time = 1200 RCS temperature = 274°F stable 1/0 Pump Ops in progress 1A1 RCP operating 1A2 RCP NOT available 1B1 RCP available 1B2 RCP NOT available LPSW to the 1A1 RCP ONLY is lost Time = 1202 1A1 RCP secured

1) In accordance with AP/16, the 1A1 RCP must be tripped as soon as its stator temperature reaches __(1)__ degrees F.

2) At Time = 1205, Tech Spec 3.4.5 (RCS Loops - MODE 3) Condition(s) __(2)__

apply.

Which ONE of the following completes the statements above?

REFERENCE PROVIDED A. 1. 260

2. A B. 1. 260

2. A and C C. 1. 295

2. A D. 1. 295

2. A and C Page 77 of 100

Oconee Nuclear Station Question: 78 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Reactor in MODE 5 LPI aligned to normal DHR mode Low Range Cooldown Pressure = 5 psig Which ONE of the following:

1) are the MINIMUM conditions that would indicate a TOTAL loss of LPI flow?

2) actions are taken to control RCS pressure in accordance with AP/26 (Loss of Decay Heat Removal) once a total loss of LPI flow has occurred?

NOTE: 1SA-3/C8 = LP Injection Loop A Flow High/Low 1SA-3/C9 = LP Injection Loop B Flow High/Low A. 1. 1SA-3/C8 ONLY with Train A LPI flow = 0 gpm

2. Initiate Enclosure 5.18 (SSF Operation For Loss of DHR Events)

B. 1. 1SA-3/C8 ONLY with Train A LPI flow = 0 gpm

2. Cycle 1RC-66 C. 1. 1SA-3/C8 AND 1SA-3/C9 actuated with Train A AND Train B LPI flow = 0 gpm

2. Initiate Enclosure 5.18 (SSF Operation For Loss of DHR Events)

D. 1. 1SA-3/C8 AND 1SA-3/C9 actuated with Train A AND Train B LPI flow = 0 gpm

2. Cycle 1RC-66 Page 78 of 100

Oconee Nuclear Station Question: 79 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial conditions:

Reactor power = 100%

BOTH channels of AMSAC disabled Current conditions:

Both Main Feedwater Pumps Tripped Reactor power = 60% and slowly decreasing

1) 1A MDEFWP, 1B MDEFWP AND the TDEFWP __(1)__ automatically start.

2) In accordance with B&W analysis, a MINIMUM of __ (2)__ gallons per minute of Emergency Feedwater flow is required to limit the RCS pressure increase to below the design standard for this event.

Which ONE of the following completes the statements above?

A. 1. will

2. 750 B. 1. will NOT

2. 750 C. 1. will

2. 375 D. 1. will NOT

2. 375 Page 79 of 100

Oconee Nuclear Station Question: 80 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Time = 1200:

Reactor power = 100%

1SA6/B2 INVERTER 1DID SYSTEM TROUBLE actuated Time = 1205 NEO reports:

o 1SA13/A8 INVERTER 1DID INPUT VOLTAGE LOW actuated o Inverter 1DID output voltage low

1) The status of 1KVID at Time = 1205 is __(1)__.

2) The MINIMUM action(s) required to restore the 1DID inverter to OPERABLE in accordance with Tech Spec 3.8.6 (Vital Inverters-Operating) is/are to restore DC input voltage __(2)__.

Which ONE of the following completes the statements above?

A. 1. NOT energized

2. ONLY B. 1. NOT energized

2. AND re-connect to 1KVID C. 1. Energized

2. ONLY D. 1. Energized

2. AND re-connect to 1KVID Page 80 of 100

Oconee Nuclear Station Question: 81 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Time = 1200 Reactor power = 100%

1RIA-60 = 35 gpm stable Time = 1215 Main Turbine manually tripped due to high bearing vibration Time = 1245 Steam Line Break in 1A Steam Generator

1) The __(1)__ tab will be used to stabilize the plant following the reactor trip.

2) The __(2)__ tab will direct the unit cooldown to LPI.

Which ONE of the following completes the statements above?

A. 1. Subsequent Actions

2. SGTR B. 1. SGTR

2. SGTR C. 1. Subsequent Actions

2. Forced Cooldown D. 1. SGTR

2. Forced Cooldown Page 81 of 100

Oconee Nuclear Station Question: 82 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Time = 1200 Reactor Power = 100%

Control Rod Group 5 Rod 4 ONLY has partially inserted and indicates 90%

withdrawn Time = 1330 Reactor Power = 55% stable Time = 1400 Control Rod Group 3 Rod 1 has dropped and indicates 0% withdrawn At Time = 1330 Control Rod Group 5 Rod 4 is considered __(1)__ in accordance with Tech Spec 3.1.4 (Control rod Group Alignment Limits),.

At Time = 1400 the CRS will direct the ROs to __(2)__.

Which ONE of the following completes the statements above?

A. 1. misaligned ONLY

2. notify SPOC to reduce RPS trip setpoints B. 1. misaligned ONLY

2. trip the Reactor C. 1. Inoperable

2. notify SPOC to reduce RPS trip setpoints D. 1. Inoperable

2. trip the Reactor Page 82 of 100

Oconee Nuclear Station Question: 83 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial conditions Reactor Power = 100%

A Fire has been identified in the Reactor Building AP/43 (Fire Brigade Response Procedure) is in progress AP/50 (Challenging Plant Fire) has been initiated Current conditions:

Reactor in MODE 3 Steam Generator level indications in the control room are erratic and suspected to be inaccurate due to the fire Which ONE of the following:

1) states the definition of a Challenging Fire in accordance with AP/43?

2) should be used to validate the Steam Generator level indication in accordance with AP/50?

A. 1. A fire that is burning cables (bundles/ trays which have the potential to affect additional equipment)

2. SSF Steam Generator level B. 1. A fire that is burning cables (bundles/ trays which have the potential to affect additional equipment)

2. Aux Shutdown Panel Steam Generator level C. 1. A fire in the plant that is NOT extinguished within 15 minutes of Control Room notification

2. SSF Steam Generator level D. 1. A fire in the plant that is NOT extinguished within 15 minutes of Control Room notification

2. Aux Shutdown Panel Steam Generator level Page 83 of 100

Oconee Nuclear Station Question: 84 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial Conditions:

Reactor Power = 50% stable Toxic gas is entering the Control room Current Conditions:

The Reactor has been manually tripped KHUs have been Emergency Started 1HP-24 has been opened Which ONE of the following describes the:

1) initial actions directed to control Pressurizer level in accordance with AP/8?

2) Pressurizer level instrumentation available at the alternate location INITIALLY used in accordance with AP/8 (Loss of Control Room)?

A. 1. At ASDP, take control of the 1B HPI pump and 1HP-120 to control level

2. Temperature compensated Pzr level B. 1. At SSF, start the SSF-RCMUP and adjust RC makeup and letdown to control level

2. Temperature compensated Pzr level C. 1. At ASDP, take control of the 1B HPI pump and 1HP-120 to control level

2. Uncompensated Pzr level D. 1. At SSF, start the SSF-RCMUP and adjust RC makeup and letdown to control level

2. Uncompensated Pzr level Page 84 of 100

Oconee Nuclear Station Question: 85 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial conditions:

Time = 1200 Reactor power = 80% stable 1B2 RCP breaker fails causing the 1B2 RCP to trip Current conditions:

Time = 1300 Control Rod #2 in Group 4 drops into core Which ONE of the following states the:

1) abnormal procedure entered at Time = 1200?

2) procedure that will be used to reach the final power level required by the dropped rod?

A. 1. AP/1 (Unit Runback)

2. AP/1 (Unit Runback)

B. 1. AP/1 (Unit Runback)

2. OP/1/A/1102/004 (Operation at Power)

C. 1. AP/16 (Abnormal RCP Operations)

2. AP/1 (Unit Runback)

D. 1. AP/16 (Abnormal RCP Operations)

2. OP/1/A/1102/004 (Operation at Power)

Page 85 of 100

Oconee Nuclear Station Question: 86 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Time = 1200:

Reactor Power = 100%

1A HPI Pump operating 1B HPI Pump in AUTO Time = 1202:

A seal supply filter problem results in total Reactor Coolant Pump seal injection flow decreasing to 25 gpm

1) At time 1202 the 1B HPI pump is __(1)__.

2) The low seal injection flow Auto Start of the 1B HPI pump __(2)__ required to be operable for the 1B HPI pump to be Operable in accordance with Tech Spec 3.5.2 (HPI).

Which ONE of the following completes the statements above?

Assume NO operator actions A. 1. running

2. is B. 1. running

2. is NOT C. 1. NOT running

2. is D. 1. NOT running

2. is NOT Page 86 of 100

Oconee Nuclear Station Question: 87 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Time = 1215 Unit shutdown in progress 1A1 RCP in operation Tcold = 220°F slowly decreasing RCS Pressure 325 psig slowly decreasing LPI Cooler outlet temperature = 110°F stable Time = 1230 Tcold = 200°F slowly decreasing RCS Pressure 275 psig slowly stable LPI Cooler outlet temperature = 110°F stable 1A LPI Pump is started Time = 1240 Tcold = 197°F stable 1A1 RCP secured Time = 1245 LPI Cooler outlet temperature = 180°F stable

1) The RCS cooldown rate __(1)__ violate the maximum cooldown rate allowed per Tech Specs?

2) When the 1A LPI pump is started at Time = 1230, the temperature transient that results from the difference in LPI Cooler Outlet temperature and Tcold __(2)__

outside the bounds of the Reactor Vessel stress analysis?

Which ONE of the following completes the statements above?

A. 1. does

2. is B. 1. does

2. is NOT C. 1. does NOT

2. is D. 1. does NOT

2. is NOT Page 87 of 100

Oconee Nuclear Station Question: 88 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial Conditions Reactor power = 100%

BOTH Main FDW pumps trip Current Conditions Reactor power = 28% decreasing RULE 1 (ATWS/UNPP) is complete 1RC-66 is failed open Pressurizer level = 400 stable

1) The UNPP tab __(1)__ direct closing 1RC-4.

2) The HIGHEST RCS pressure that will allow a retransfer to Subsequent Actions is less than __(2)__ psig.

Which ONE of the following completes the statements above?

A. 1. will

2. 2450 B. 1. will NOT

2. 2450 C. 1. will

2. 2300 D. 1. will NOT

2. 2300 Page 88 of 100

Oconee Nuclear Station Question: 89 ILT44 ONS SRO NRC Examination (1 point)

Unit 1 initial conditions:

RCS temperature = 310°F stable 1A2 and 1B2 RCPs operating Unit Auxiliaries powered from CT-1 Current conditions:

RC Makeup Flow = 150 gpm PCB-17 AND PCB-18 trip OPEN Which ONE of the following describes the:

1) status of Reactor Coolant Pumps

2) MAXIMUM cooldown rate allowed in accordance with the EOP once cooldown to LPI has commenced?

A. 1. Off

2. 50°F per hour B. 1. Off

2. 50°F per 1/2 hour C. 1. On

2. 50°F per hour D. 1. On

2. 50°F per 1/2 hour Page 89 of 100

Oconee Nuclear Station Question: 90 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Time = 1200 Reactor power = 100%

BOTH AFIS channels in OFF Time = 1205 Large MSLB inside the RB occurs ES Channels 1-8 have actuated Time = 1207 Component status as pictured below

1) Reactor Building Cooling Units __(1)__ operating as designed.

2) Systems and/or equipment required to ensure Reactor Building Pressure is maintained below the design pressure of the Reactor Building __(2)__ available.

Which ONE of the following completes the statements above?

ASSUME NO OPERATOR ACTIONS A. 1. are

2. is B. 1. are NOT

2. is C. 1. are

2. is NOT D. 1. are NOT

2. is NOT Page 90 of 100

Oconee Nuclear Station Question: 91 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 3 plant conditions:

Time = 1200 Reactor in MODE 1 RBNS level 20 inches increasing LPSW leak into RB Begin pumping RBNS using 1A and 1B RBNS Pumps Time = 1205 RBNS level as indicated below Time = 1230 RBNS level indication unchanged from Time = 1205 Which ONE of the following states:

1) if Condition A of Tech Spec 3.4.15 must be entered?

2) which RIA can be used to satisfy the requirements of TS 3.4.15 LCO?

REFERENCE PROVIDED A. 1. Yes

2. 1RIA-47 B. 1. Yes

2. 1RIA-49 C. 1. No

2. 1RIA-47 D. 1. No

2. 1RIA-49 Page 91 of 100

Oconee Nuclear Station Question: 92 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Time = 1200 Reactor power = 100%

Relative Position Indication (RPI) inoperable for ALL Control Rods Time = 1230 Absolute Position Indication (API) inoperable for Group 1 Rod 7 Control Rod Current Conditions:

Tech Spec 3.1.4 Required Action A.2.1.1 (SDM Verification) has just been completed and shutdown margin requirements of the COLR have been determined to be NOT met Which ONE of the following:

1) is the LATEST time that Group 1 Rod 7 Control Rod must be declared inoperable in accordance with Tech Specs?

2) should be used to restore shutdown margin requirements in accordance with Tech Spec bases?

REFERENCE ATTACHED A. 1. 1230

2. OP/1/A/1103/004A Enclosure 4.1 (RCS Boration from CBAST With CBAST Pump) ONLY B. 1. 1230

2. OP/1/A/1103/004A Enclosure 4.1 (RCS Boration from CBAST With CBAST Pump) OR OP/1/A/1103/004 Enclosure 4.4 (RCS Makeup From 1A BHUT C. 1. 1330

2. OP/1/A/1103/004A Enclosure 4.1 (RCS Boration from CBAST With CBAST Pump) ONLY D. 1. 1330

2. OP/1/A/1103/004A Enclosure 4.1 (RCS Boration from CBAST With CBAST Pump) OR OP/1/A/1103/004 Enclosure 4.4 (RCS Makeup From 1A BHUT Page 92 of 100

Oconee Nuclear Station Question: 93 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Time = 0900 Reactor trip due a sheared RCP shaft on 1A1 RCP Control Room has indications that the 1A1 RCP seals have failed.

Time = 0904 ES 1 & 2 actuate on low RCS Pressure.

The 1A HPI pump fails to Auto start and cannot be started manually.

Time = 0910 RCS saturated and stable at 1000 psig. Rule 2 is complete Time = 0919 1RIA-57 reads 350 R/HR and stable.

RB Pressure = 2.6 psig slowly increasing Time = 0920 RB Pressure = 0.2 psig decreasing Which ONE of the following states the:

1) position of 1HP-410?

2) Emergency Classification in accordance with RP/0/A/1000/001 (Emergency Classification)?

REFERENCE PROVIDED DO NOT USE EMERGENCY COORDINATOR JUDGEMENT WHEN DETERMINING CLASSIFICATION A. 1. Open

2. General Emergency B. 1. Open

2. Site Area Emergency C. 1. Closed

2. General Emergency D. 1. Closed

2. Site Area Emergency Page 93 of 100

Oconee Nuclear Station Question: 94 ILT44 ONS SRO NRC Examination (1 point)

In accordance with AD-OP-ALL-1000 (Conduct of Operations),

1) The CRS __(1)__ allowed to correct archived log entries when the original watchstander will not be available in the near future.

2) the MINIMUM level of management with overall responsibility for the approval of the quality of Unit Log entries is the __(2)__.

Which ONE of the following completes the statements above A. 1. is

2. SM B. 1. is

2. CRS C. 1. is NOT

2. SM D. 1. is NOT

2. CRS Page 94 of 100

Oconee Nuclear Station Question: 95 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Reactor power = 100%

Main T/G Journal Bearing #7 vibration as described below:

Which ONE of the following describes:

1) The EARLIEST time that the Main Turbine must be manually tripped in accordance with Limits and Precautions of OP/1/A/1106/001 (Turbine Generator)?

2) The procedure that would be used to take the Main Turbine off line without tripping the Reactor?

A. 1. 1216

2. AP/29 (Rapid Unit Shutdown)

B. 1. 1216

2. OP/1/A/1106/001 (Turbine Generator)

C. 1. 1224

2. AP/29 (Rapid Unit Shutdown)

D. 1. 1224

2. OP/1/A/1106/001 (Turbine Generator)

Page 95 of 100

Oconee Nuclear Station Question: 96 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 2 conditions:

Reactor in MODE 5 Maintenance is required on 2FDW-105 (2A SG SAMPLE)

Multiple electrical valve manipulations of 2FDW-105 are required as part of the repairs A check valve needs to be used as one of the isolation boundary valves Which ONE of the following describes the requirements in accordance with NSD 500 (Red Tags/Configuration Control Tags) for:

1) the lowest level of Operations Management approval required for the tagout?

2) who must approve the use of wax string to attach a Red Tag to its associated component?

A. 1. ANY Licensed SRO

2. Shift Manager B. 1. Any Operations Manager

2. Shift Manager C. 1. ANY Licensed SRO

2. Operational Control Group supervisor D. 1. Any Operations Manager

2. Operational Control Group supervisor Page 96 of 100

Oconee Nuclear Station Question: 97 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Reactor = Mode 3 B shift is performing a plant startup following a refueling outage in progress PT/0/A/0711/001 (Zero Power Physics Testing) is to be performed Which ONE of the following is the LOWEST level of management that meets the expectations provided in NSD 213 (Risk Management Process) regarding who can provide the 91-01 IPTE briefing associated with ZPPT?

A. B shift Shift Technical Advisor B. Reactor Engineering Supervisor C. B shift Shift Manager (Operations Shift Manager)

D. Operations Manager (Superintendant of Operations)

Page 97 of 100

Oconee Nuclear Station Question: 98 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Initial conditions:

Mode 6 Defueling in progress 1RIA-3 (Fuel Transfer Canal Monitor) = 4 mr/hr stable Main Fuel Bridge Area Monitor = 6 mr/hr stable Current conditions:

1RIA-3 local reading = 0 mr/hr 1RIA-3 View Node indication is magenta The Refueling SRO will determine that Fuel Handling activities in the Reactor Building may ______ in accordance with OP/1/A/1502/007 (Operations Defueling/Refueling Responsibilities)?

Which ONE of the following completes the statement above?

A. NOT continue until a portable area monitor with local alarm capability is in place.

B. continue because only the Main Fuel Bridge Area Monitor is required.

C. NOT continue until continuous RP coverage is present on the Main Fuel Bridge.

D. continue provided the audible alarm associated with 1RIA-49 (RB Normal Gas) is operable.

Page 98 of 100

Oconee Nuclear Station Question: 99 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Reactor trip has occurred Main Turbine status as indicated below SRO directs performance of Immediate Manual Actions

1) The Reactor Operator performing IMAs __(1)__ expected to depress the Turbine Trip pushbutton.

2) IF the Main Turbine is not tripped during performance of IMAs, the basis behind requiring the EHC pumps being secured is to prevent __(2)__.

Which ONE of the following completes the statements above?

A. 1. is

2. water from reaching the Main Turbine during a SG overfeed B. 1. is

2. excessive RCS cooldown C. 1. is NOT

2. water from reaching the Main Turbine during a SG overfeed D. 1. is NOT

2. excessive RCS cooldown Page 99 of 100

Oconee Nuclear Station Question: 100 ILT44 ONS SRO NRC Examination (1 point)

Given the following Unit 1 conditions:

Reactor Power = 100%

Operating CC pump trips and standby fails to Auto start CC Surge tank level = 28 stable Letdown Temperature = 138°F increasing The following alarms have actuated and associated actions are being performed:

o 1SA-9/B-1 (CC CRD Return Flow Low).

o 1SA-2/C-1 (HP Letdown Temperature High) o 1SA-9/C-1 (CC Component Cooling Return Flow Low)

Which ONE of the following states:

1) The SETPOINT (degrees F) at which 1SA-2/C1 actuated?

2) The Procedure that will be entered FIRST if actions directed in the associated Alarm Response guides fail to restore normal system parameters?

A. 1. 130

2. AP/20 (Loss of Component Cooling)

B. 1. 130

2. AP/32 (Loss of Letdown)

C. 1. 135

2. AP/20 (Loss of Component Cooling)

D. 1. 135

2. AP/32 (Loss of Letdown)

Page 100 of 100

RCS Loops - MODE 3 3.4.5 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.5 RCS Loops - MODE 3 LCO 3.4.5 Two RCS loops shall be OPERABLE and at least one RCS loop shall be in operation.

NOTE------------------------------------------------

All reactor coolant pumps (RCPs) may not be in operation for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period for the transition to or from the Decay Heat Removal System, and all RCPs may not be in operation for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period for any other reason, provided:

a. No operations are permitted that would cause reduction of the RCS boron concentration; and

b. Core outlet temperature is maintained at least 10°F below saturation temperature.

APPLICABILITY: MODE 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RCS loop A.1 Restore RCS loop to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable. OPERABLE status.

B. Required Action and B.1 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A not met.

(continued)

OCONEE UNITS 1, 2, & 3 3.4.5-1 Amendment Nos. 300, 300, & 300

RCS Loops - MODE 3 3.4.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Two RCS loops C.1 Suspend all operations Immediately inoperable. involving a reduction of RCS boron OR concentration.

Required RCS loop not AND in operation.

C.2 Initiate action to restore Immediately one RCS loop to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify required RCS loop is in operation. In accordance with the Surveillance Frequency Control Program SR 3.4.5.2 Verify correct breaker alignment and indicated In accordance with the power available to the required pump that is Surveillance Frequency not in operation. Control Program OCONEE UNITS 1, 2, & 3 3.4.5-2 Amendment Nos. 372, 374, 373

RCS Leakage Detection Instrumentation 3.4.15 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.15 RCS Leakage Detection Instrumentation LCO 3.4.15 The following RCS leakage detection instrumentation shall be OPERABLE:

a. One containment normal sump level indication; and

b. One containment atmosphere particulate radioactivity monitor.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS

NOTE--------------------------------------------------------

LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. Required containment A.1 -----------NOTE-----------

sump level indication Not required until 12 inoperable. hours after establishment of steady state operation.

Perform SR 3.4.13.1. Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months AND A.2 Restore required 30 days containment sump level indication to OPERABLE status.

(continued)

OCONEE UNITS 1, 2, & 3 3.4.15-1 Amendment Nos. 359, 361, & 360

RCS Leakage Detection Instrumentation 3.4.15 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required containment B.1.1 Analyze grab samples Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months atmosphere of the containment radioactivity monitor atmosphere.

inoperable.

OR B.1.2 -----------NOTE-----------

Not required until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Perform SR 3.4.13.1. Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months AND B.2 Restore required 30 days containment atmosphere radioactivity monitor to OPERABLE status.

C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met. AND C.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months D. Both required D.1 Enter LCO 3.0.3. Immediately instrument functions inoperable.

OCONEE UNITS 1, 2, & 3 3.4.15-2 Amendment Nos. 300, 300, & 300

RCS Leakage Detection Instrumentation 3.4.15 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.15.1 Perform CHANNEL CHECK of required In accordance with the containment atmosphere radioactivity monitor. Surveillance Frequency Control Program SR 3.4.15.2 Perform CHANNEL FUNCTIONAL TEST of In accordance with the required containment atmosphere radioactivity Surveillance Frequency monitor. Control Program SR 3.4.15.3 Perform CHANNEL CALIBRATION of In accordance with the required containment sump level indication. Surveillance Frequency Control Program SR 3.4.15.4 Perform CHANNEL CALIBRATION of In accordance with the required containment atmosphere radioactivity Surveillance Frequency monitor. Control Program OCONEE UNITS 1, 2, & 3 3.4.15-3 Amendment Nos. 372, 374, 373

CONTROL ROD Group alignment Limits 3.1.4 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 CONTROL ROD Group Alignment Limits LCO 3.1.4 Each CONTROL ROD shall be OPERABLE and aligned to within 6.5% of its group average height.

APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One trippable A.1 Restore CONTROL 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months CONTROL ROD ROD alignment.

inoperable, or not aligned to within 6.5% OR of its group average height, or both. A.2.1.1 Verify SDM is within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the limit specified in the COLR. AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter OR A.2.1.2 Initiate boration to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore SDM to within limit.

AND (continued)

OCONEE UNITS 1, 2, & 3 3.1.4-1 Amendment Nos. 300, 300, & 300

CONTROL ROD Group alignment Limits 3.1.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.2 Reduce THERMAL 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months POWER to 60% of the ALLOWABLE THERMAL POWER.

AND A.2.3 Reduce the nuclear 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months overpower trip setpoints, based on flux and flux/flow imbalance, to 65.5%

of the ALLOWABLE THERMAL POWER.

AND A.2.4 Verify the potential 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ejected rod worth is within the assumptions of the rod ejection analysis.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time for Condition A not met.

C. More than one trippable C.1.1 Verify SDM is 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months CONTROL ROD within the limit inoperable, or not specified in the COLR.

aligned within 6.5% of its group average OR height, or both.

(continued)

OCONEE UNITS 1, 2, & 3 3.1.4-2 Amendment Nos. 300, 300, & 300

CONTROL ROD Group alignment Limits 3.1.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.1.2 Initiate boration to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore SDM to within limit.

AND C.2 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months D. One or more rods D.1.1 Verify SDM is 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months untrippable. within the limit specified in the COLR.

OR D.1.2 Initiate boration to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore SDM to within limit.

AND D.2 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months OCONEE UNITS 1, 2, & 3 3.1.4-3 Amendment Nos. 300, 300, & 300

CONTROL ROD Group alignment Limits 3.1.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify individual CONTROL ROD positions In accordance with the are within 6.5% of their group average height. Surveillance Frequency Control Program SR 3.1.4.2 Verify CONTROL ROD freedom of movement In accordance with the (trippability) by moving each individual Surveillance Frequency CONTROL ROD that is not fully inserted by Control Program an amount in any direction sufficient to demonstrate the absence of thermal binding.

SR 3.1.4.3 Verify the rod drop time for each CONTROL Prior to reactor criticality ROD, from the fully withdrawn position, is after each removal of the 1.66 seconds at reactor coolant full flow reactor vessel head conditions or 1.40 seconds at no flow conditions from power interruption at the CONTROL ROD drive breakers to 3/4 insertion (25% withdrawn position).

OCONEE UNITS 1, 2, & 3 3.1.4-4 Amendment Nos. 372, 374, 373

Duke Energy Procedure No.

Oconee Nuclear Station 0 RP/ /A/1000/001 Emergency Classification Revision No.

000 Electronic Reference No.

OP009A63 Reference Use PERFORMANCE

* * * * * * * *

UNCONTROLLED FOR PRINT * * * * * * * * * *

(ISSUED) - PDF Format

RP/0/A/1000/001 Page 2 of 6 Emergency Classification NOTE: This procedure is an implementing procedure to the Oconee Nuclear Site Emergency plan and must be forwarded to Emergency Planning within seven (7) working days of approval.

1. Symptoms 1.1 This procedure describes the immediate actions to be taken to recognize and classify an emergency condition.

1.2 This procedure identifies the four emergency classifications and their corresponding Emergency Action Levels (EALs).

1.3 This procedure provides reporting requirements for non-emergency abnormal events.

1.4 The following guidance is to be used by the Emergency Coordinator/EOF Director in assessing emergency conditions:

1.4.1 Definitions and Acronyms are italicized throughout procedure for easy recognition. The definitions are in Enclosure 4.10 (Definitions/Acronyms).

1.4.2 The Emergency Coordinator/EOF Director shall review all applicable initiating events to ensure proper classification.

1.4.3 The BASIS Document (Volume A, Section D of the Emergency Plan) is available for review if any questions arise over proper classification.

1.4.4 IF An event occurs on more than one unit concurrently, THEN The event with the higher classification will be classified on the Emergency Notification Form.

A. Information relating to the problem(s) on the other unit(s) will be captured on the Emergency Notification Form as shown in RP/0/A/1000/015A, (Offsite Communications From The Control Room),

RP/0/A/1000/015B, (Offsite Communications From The Technical Support Center) or SR/0/B/2000/004, (Notification to States and Counties from the Emergency Operations Facility).

1.4.5 IF An event occurs, AND A lower or higher plant operating mode is reached before the classification can be made, THEN The classification shall be based on the mode that existed at the time the event occurred.

2

RP/0/A/1000/001 Page 3 of 6 1.4.6 The Fission Product Barrier Matrix is applicable only to those events that occur at Mode 4 (Hot Shutdown) or higher.

A. An event that is recognized at Mode 5 (Cold Shutdown) or lower shall not be classified using the Fission Product Barrier Matrix.

1. Reference should be made to the additional enclosures that provide Emergency Action Levels for specific events (e.g., Severe Weather, Fire, Security).

1.5 IF A transient event should occur, THEN Review the following guidance:

1.5.1 IF An Emergency Action Level (EAL) identifies a specific duration AND The Emergency Coordinator/EOF Director assessment concludes that the specified duration is exceeded or will be exceeded, (i.e.;

condition cannot be reasonably corrected before the duration elapses),

THEN Classify the event.

1.5.2 IF A plant condition exceeding EAL criteria is corrected before the specified duration time is exceeded, THEN The event is NOT classified by that EAL.

A. Review lower severity EALs for possible applicability in these cases.

NOTE: Reporting under 10CFR50.72 may be required for the following step. Such a condition could occur, for example, if a follow up evaluation of an abnormal condition uncovers evidence that the condition was more severe than earlier believed.

1.5.3 IF A plant condition exceeding EAL criteria is 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, THEN An emergency shall NOT be declared.

Refer to NSD 202 for reportability 3

RP/0/A/1000/001 Page 4 of 6 1.5.4 IF An emergency classification was warranted, but the plant condition has been corrected prior to declaration and notification THEN The Emergency Coordinator must consider the potential that the initiating condition (e.g.; Failure of Reactor Protection System) may have caused plant damage that warrants augmenting the on shift personnel through activation of the Emergency Response Organization.

A. IF An Unusual Event condition exists, THEN Make the classification as required.

1. The event may be terminated in the same notification or as a separate termination notification.

B. IF An Alert, Site Area Emergency, or General Emergency condition exists, THEN Make the classification as required, AND Activate the Emergency Response Organization.

1.6 Emergency conditions shall be classified as soon as the Emergency Coordinator/EOF Director assessment determines that the Emergency Action Levels for the Initiating Condition have been exceeded.

4

RP/0/A/1000/001 Page 5 of 6

2. Immediate Actions 2.1 Assessment, classification and declaration of any applicable emergency condition should be completed within 15 minutes after the availability of indications or information to cognizant facility staff that an EAL threshold has been exceeded.

2.2 Determine the operating mode that existed at the time the event occurred prior to any protection system or operator action initiated in response to the event.

2.3 IF The unit is at Mode 4 (Hot Shutdown) or higher AND The condition/event affects fission product barriers, THEN GO TO Enclosure 4.1, (Fission Product Barrier Matrix).

2.3.1 Review the criteria listed in Enclosure 4.1, (Fission Product Barrier Matrix) and make the determination if the event should be classified).

2.4 Review the listing of enclosures to determine if the event is applicable to one of the categories shown.

2.4.1 IF One or more categories are applicable to the event, THEN Refer to the associated enclosures.

2.4.2 Review the EALs and determine if the event should be classified.

A. IF An EAL is applicable to the event, THEN Classify the event as required.

2.5 IF The condition requires an emergency classification, THEN Initiate the following:

for Control Room - RP/0/B/1000/002, (Control Room Emergency Coordinator Procedure)

for TSC - RP/0/A/1000/019, (Technical Support Center Emergency Coordinator Procedure)

for EOF - SR/0/A/2000/003, (Activation of the Emergency Operations Facility) 2.6 Continue to review the emergency conditions to assure the current classification continues to be applicable.

5

RP/0/A/1000/001 Page 6 of 6

3. Subsequent Actions 3.1 Continue to review the emergency conditions to assure the current classification continues to be applicable.

4. Enclosures Enclosures Page Number 4.1 Fission Product Barrier Matrix 7 4.2 System Malfunctions 8 4.3 Abnormal Rad Levels/Radiological Effluents 10 4.4 Loss Of Shutdown Functions 12 4.5 Loss of Power 14 4.6 Fires/Explosions And Security Actions 15 4.7 Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety 17 4.8 Radiation Monitor Readings For Emergency Classification 20 4.9 Unexpected/Unplanned Increase In Area Monitor Readings 21 4.10 Definitions 22 4.11 Operating Modes Defined In Improved Technical Specifications 27 4.12 Instructions For Using Enclosure 4.1 28 4.13 References 30 6

Enclosure 4.1 RP/0/A/1000/001 Fission Product Barrier Matrix Page 1 of 1 DETERMINE THE APPROPRIATE CLASSIFICATION USING THE TABLE BELOW: ADD POINTS TO CLASSIFY. SEE NOTE BELOW RCS BARRIERS (BD 5-7) FUEL CLAD BARRIERS (BD 8-9) CONTAINMENT BARRIERS (BD 10-13)

Potential Loss (4 Points) Loss (5 Points) Potential Loss (4 Points) Loss (5 Points) Potential Loss (1 Point) Loss (3 Points)

RCS Leakrate 160 gpm RCS Leak rate that results in a loss Average of the 5 highest Average of the 5 highest CETC CETC 1200° F 15 minutes Rapid unexplained containment of subcooling. CETC 1200° F OR pressure decrease after increase 700° F CETC 700° F 15 minutes with a OR valid RVLS reading 0 containment pressure or sump level not consistent with LOCA SGTR 160 gpm Valid RVLS reading of 0 Coolant activity 300 µCi/ml DEI RB pressure 59 psig Failure of secondary side of SG OR results in a direct opening to the RB pressure 10 psig and no environment with SG Tube Leak RBCU or RBS 10 gpm in the SAME SG NOTE: RVLS is NOT valid if one or more RCPs are running OR if Entry into the PTS (Pressurized 1RIA 57 or 58 reading 1.0 R/hr LPI pump(s) are Hours RIA 57 OR RIA 58 Hours RIA 57 OR RIA 58 SG Tube Leak 10 gpm exists in Thermal Shock) Operation running AND taking Since SD R/hr R/hr Since SD R/hr R/hr one SG.

2 RIA 57 reading 1.6 R/hr suction from the LPI AND NOTE: PTS is entered under the other SG has secondary side either of the following: 2 RIA 58 reading 1.0 R/hr drop line. 0 - <0.5 300 150 0 - < 0.5 1800 860 failure that results in a direct

A cooldown below 400°F @ opening to the environment AND

> 100°F/hr. has occurred. 3RIA 57 or 58 reading 1.0 R/hr 0.5 - < 2.0 80 40 0.5 - < 2.0 400 195 is being fed from the affected unit.

HPI has operated in the 2.0 - 8.0 32 16 2.0 - 8.0 280 130 injection mode while NO RCPs were operating.

HPI Forced Cooling RCS pressure spike 2750 psig Hydrogen concentration 9% Containment isolation is incomplete and a release path to the environment exists Emergency Coordinator/EOF Emergency Coordinator/EOF Emergency Coordinator/EOF Emergency Coordinator/EOF Director Emergency Coordinator/EOF Emergency Coordinator/EOF Director judgment Director judgment Director judgment judgment Director judgment Director judgment UNUSUAL EVENT (1-3 Total Points) ALERT (4-6 Total Points) SITE AREA EMERGENCY (7-10 Total Points) GENERAL EMERGENCY (11-13 Total Points)

OPERATING MODE: 1, 2, 3, 4 OPERATING MODE: 1, 2, 3, 4 OPERATING MODE: 1, 2, 3, 4 OPERATING MODE: 1, 2, 3, 4 4.1.U.1 Any potential loss of Containment 4.1.A.1 Any potential loss or loss of the RCS 4.1.S.1 Loss of any two barriers 4.1.G.1 Loss of any two barriers and potential loss of the third barrier 4.1.U.2 Any loss of containment 4.1.A.2 Any potential loss or loss of the Fuel 4.1.S.2 Loss of one barrier and potential loss of either Clad RCS or Fuel Clad Barriers 4.1.G.2 Loss of all three barriers 4.1.S.3 Potential loss of both the RCS and Fuel Clad Barriers NOTE: An event with multiple events could occur which would result in the conclusion that exceeding the loss or potential loss threshold is IMMINENT (i.e., within 1-3 hours). In this IMMINENT LOSS situation, use judgment and classify as if the thresholds are exceeded.

7

Enclosure 4.2 RP/0/A/1000/001 System Malfunctions Page 1 of 2 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY

1. RCS LEAKAGE (BD 15)

===============================

OPERATING MODE: 1, 2, 3, 4 A. Unidentified leakage 10 gpm B. Pressure boundary leakage 10 gpm C. Identified leakage 25 gpm

Includes SG tube leakage

2. UNPLANNED LOSS OF MOST OR ALL 1. UNPLANNED LOSS OF MOST OR ALL SAFETY SYSTEM ANNUNCIATION/ SAFETY SYSTEM ANNUNCIATION/ 1. INABILITY TO MONITOR A INDICATION IN CONTROL ROOM INDICATION IN CONTROL ROOM SIGNIFICANT TRANSIENT IN FOR > 15 MINUTES (BD 16) (BD 20) PROGRESS (BD 22)

===================================== ====================================== ================================

OPERATING MODE: 1, 2, 3, 4 OPERATING MODE: 1, 2, 3, 4 OPERATING MODE: 1, 2, 3, 4 A. Unplanned loss of > 50% of the following A. Unplanned loss of > 50% of the following A. Unplanned loss of > 50% of the following annunciators on one unit for > 15 minutes: annunciators on one unit for > 15 minutes: annunciators on one unit for > 15 minutes:

Units 1 & 3 Units 1 & 3 Units 1 & 3 1 SA1, 2, 3, 4, 5, 6, 7, 8, 9, 14, 15, 16, & 18 1 SA1, 2, 3, 4, 5, 6, 7, 8, 9, 14, 15, 16, & 18 1 SA1, 2, 3, 4, 5, 6, 7, 8, 9, 14, 15, 16, & 18 3 SA1, 2, 3, 4, 5, 6, 7, 8, 9, 14, 15, 16, & 18 3 SA1, 2, 3, 4, 5, 6, 7, 8, 9, 14, 15, 16, & 18 3 SA1, 2, 3, 4, 5, 6, 7, 8, 9, 14, 15, 16, & 18 Unit 2 Unit 2 Unit 2 2 SA1, 2, 3, 4, 5, 6, 7, 8, 9, 14, 15, & 16 2 SA1, 2, 3, 4, 5, 6, 7, 8, 9, 14, 15, & 16 2 SA1, 2, 3, 4, 5, 6, 7, 8, 9, 14, 15, & 16 AND AND AND Loss of annunciators or indicators requires Loss of annunciators or indicators requires A significant transient is in progress additional personnel (beyond normal shift additional personnel (beyond normal shift complement) to safely operate the unit complement) to safely operate the unit AND AND Loss of the OAC and ALL PAM indications (CONTINUED)

Significant plant transient in progress AND OR Inability to directly monitor any one of the following functions:

Loss of the OAC and ALL PAM indications

1. Subcriticality (END) 2. Core Cooling

3. Heat Sink

4. RCS Integrity

5. Containment Integrity

6. RCS Inventory (END) 8

Enclosure 4.2 RP/0/A/1000/001 System Malfunctions Page 2 of 2 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY

3. INABILITY TO REACH REQUIRED SHUTDOWN WITHIN LIMITS (BD 17)

==========================

OPERATING MODE: 1, 2, 3, 4 A. Required operating mode not reached within TS LCO action statement time

4. UNPLANNED LOSS OF ALL ONSITE OR OFFSITE COMMUNICATIONS (BD 18)

===============================

OPERATING MODE: All

. Loss of all onsite communications capability (Plant phone system, PA system, Pager system, Onsite Radio system) affecting ability to perform Routine operations

. Loss of all onsite communications capability (Selective Signaling, NRC ETS lines, Offsite Radio System, AT&T line) affecting ability to communicate with offsite authorities.

5. FUEL CLAD DEGRADATION (BD 19)

===============================

OPERATING MODE: All:

. DEI - >5µCi/ml (END) 9

Enclosure 4.3 RP/0/A/1000/001 Abnormal Rad Levels/Radiological Effluent Page 1 of 2 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 1 ANY UNPLANNED RELEASE OF 1. ANY UNPLANNED RELEASE OF 1. BOUNDARY DOSE RESULTING FROM 1. BOUNDARY DOSE RESULTING FROM GASEOUS OR LIQUID RADIOACTIVITY GASEOUS OR LIQUID RADIOACTIVITY ACTUAL/IMMINENT RELEASE OF ACTUAL/ IMMINENT RELEASE OF TO THE ENVIRONMENT THAT TO THE ENVIRONMENT THAT GASEOUS ACTIVITY (BD 35) GASEOUS ACTIVITY (BD 39)

EXCEEDS TWO TIMES THE SLC EXCEEDS 200 TIMES RADIOLOGICAL ===========================================

LIMITS FOR 60 MINUTES OR LONGER TECHNICAL SPECIFICATIONS FOR 15 OPERATING MODE: All OPERATING MODE: All (BD 25) MINUTES OR LONGER (BD 30)

=========================================== =========================================== A. Valid reading on RIA 46 of > 2.09E+06 cpm or A. Valid reading on RIA 46 of > 2.09E+05 cpm or OPERATING MODE: All OPERATING MODE: All RIA 56 reading of > 175 R/hr or RP sample RIA 56 reading of > 17.5 R/hr or RP sample reading of 6.62E+01 uCi/ml Xe 133 eq for > reading of 6.62E+02uCi/ml Xe 133 eq for > 15 A. Valid indication on radiation monitor RIA 33 A. Valid indication of RIA-46 of > 2.09 E+ 04 15 minutes (See Note 2) minutes (See Note 3) of 4.06E+06 cpm for > 60 minutes cpm or RP sample reading of > 6.62 uCi/ml Xe (See Note 1) 133 eq for > 15 minutes. (See Note 1) B. Valid reading on RIA 57 or 58 as shown on B. Valid reading on RIA 57 or 58 as shown on Enclosure 4.8 (See Note 2) Enclosure 4.8 (See Note 3)

B. Valid indication on radiation monitor RIA-45 B RIA 33 HIGH Alarm of > 9.35E+05 cpm or RP sample reading of > C. Dose calculations result in a dose projection at C. Dose calculations result in a dose projection at 6.62E-2uCi/ml Xe 133 eq for > 60 minutes AND the site boundary of:

the site boundary of:

(See Note 1)

Liquid effluent being released exceeds 200 100 mRem TEDE or 500 mRem CDE adult 1000 mRem TEDE C. Liquid effluent being released exceeds two times the level of SLC 16.11.1 for > 15 minutes thyroid times SLC 16.11.1 for > 60 minutes as as determined by Chemistry Procedure OR determined by Chemistry Procedure C. Gaseous effluent being released exceeds 200 D. Field survey results indicate site boundary dose 5000 mRem CDE adult thyroid D. Gaseous effluent being released exceeds two times the level of SLC 16.11.2 for >15 minutes rates exceeding 100 mRad/hr expected to times SLC 16.11.2 for > 60 minutes as as determined by RP Procedure continue for more than one hour D. Field survey results indicate site boundary dose determined by RP Procedure rates exceeding 1000 mRad/hr expected to OR continue for more than one hour Analyses of field survey samples indicate adult OR (CONTINUED) thyroid dose commitment of 500 mRem CDE (3.84 E-7 µCi/ml) for one hour of Analyses of field survey samples indicate adult inhalation thyroid dose commitment of 5000 mRem CDE for one hour of inhalation NOTE 1: If monitor reading is sustained for the time period indicated in the EAL AND the required assessments (procedure NOTE 2: If actual Dose Assessment cannot calculations) cannot be completed within NOTE 3: If actual Dose Assessment cannot be completed within 15 minutes, then the this period, declaration must be made on the be completed within 15 minutes, then the valid radiation monitor reading should be valid Radiation Monitor reading. valid radiation monitor reading should be used for emergency classification.

used for emergency classification.

(CONTINUED)

(END) 10

Enclosure 4.3 RP/0/A/1000/001 Abnormal Rad Levels/Radiological Effluent Page 2 of 2 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 2 UNEXPECTED INCREASE IN PLANT 2. RELEASE OF RADIOACTIVE 2. LOSS OF WATER LEVEL IN THE `

RADIATION OR AIRBORNE MATERIAL OR INCREASES IN REACTOR VESSEL THAT HAS OR CONCENTRATION (BD 27) RADIATION LEVELS THAT IMPEDES WILL UNCOVER FUEL IN THE OPERATION OF SYSTEMS REQUIRED REACTOR VESSEL (BD 38)

OPERATING MODE: All TO MAINTAIN SAFE OPERATION OR TO ESTABLISH OR MAINTAIN COLD OPERATING MODE: 5, 6 SHUTDOWN (BD 32)

A. LT 5 reading 14 and decreasing with makeup ===========================================

not keeping up with leakage WITH fuel in the OPERATING MODE: All A. Loss of all decay heat removal as indicated by core the inability to maintain RCS temperature below 200° F A. Valid radiation reading 15 mRad/hr in CR, B. Valid indication of uncontrolled water decrease CAS, or Radwaste CR in the SFP or fuel transfer canal with all fuel AND assemblies remaining covered by water B. Unplanned/unexpected valid area monitor readings exceed limits stated in Enclosure 4.9 LT 5 indicates 0 inches after initiation of RCS AND makeup Unplanned Valid RIA 3, 6 or Portable Area 3. MAJOR DAMAGE TO IRRADIATED B. Loss of all decay heat removal as indicated by Monitor readings increase. FUEL OR LOSS OF WATER LEVEL the inability to maintain RCS temperature THAT HAS OR WILL RESULT IN THE below 200° F UNCOVERING OF IRRADIATED FUEL C. 1 R/hr radiation reading at one foot away from OUTSIDE THE REACTOR VESSEL AND a damaged storage cask located at the ISFSI (BD 33)

Either train ultrasonic level indication less than D. Valid area monitor readings exceeds limits 0 inches and decreasing after initiation of RCS stated in Enclosure 4.9. OPERATING MODE: All makeup A. Valid RIA 3*, 6, 41, OR 49* HIGH Alarm NOTE: This Initiating Condition is also located NOTE: This Initiating Condition is also in Enclosure 4.4., (Loss of Shutdown Functions). * - Applies to Mode 6 and No Mode Only located in Enclosure 4.4., (Loss of Shutdown High radiation levels will also be seen with this Functions). High radiation levels will also be condition. seen with this condition.

B. HIGH Alarm for portable area monitors on the main bridge or SFP bridge C Report of visual observation of irradiated fuel uncovered (END))

D. Operators determine water level drop in either (END) the SFP or fuel transfer canal will exceed makeup capacity such that irradiated fuel will be uncovered NOTE: This Initiating Condition is also located in Enclosure 4.4., (Loss of Shutdown Functions).

High radiation levels will also be seen with this condition.

(END) 11

Enclosure 4.4 RP/0/A/1000/001 Loss of Shutdown Functions Page 1 of 2 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY

1. FAILURE OF RPS TO COMPLETE OR 1. FAILURE OF RPS TO COMPLETE OR 1. FAILURE OF RPS TO COMPLETE INITIATE A Rx SCRAM (BD 44) INITIATE A Rx SCRAM (BD 50)

OPERATING MODE: 1, 2 OPERATING MODE 1, 2, 3 OPERATING MODE: 1, 2 (CONTINUE TO NEXT PAGE) A. Valid Rx trip signal received or required A. Valid reactor trip signal received or required A. Valid reactor trip signal received or required WITHOUT automatic scram WITHOUT automatic scram WITHOUT automatic scram AND AND AND DSS has inserted Control Rods Manual trip from the Control Room was OR DSS has NOT inserted Control Rods NOT successful in reducing reactor power to <

Manual trip from the Control Room is 5%

successful and reactor power is less AND and decreasing than 5% and decreasing Manual trip from the Control Room was NOT AND successful in reducing reactor power to less than 5% and decreasing Average of the 5 highest CETCs 1200° F on ICCM

2. INABILITY TO MAINTAIN PLANT IN 2. COMPLETE LOSS OF FUNCTION (END)

MODE 5 (COLD SHUTDOWN) (BD 46) NEEDED TO ACHIEVE OR MAINTAIN MODE 4 (HOT SHUTDOWN) (BD 51)

OPERATING MODE: 5, 6 OPERATING MODE: 1, 2, 3, 4 A. Loss of LPI and/or LPSW A. Average of the 5 highest CETCs 1200° F AND shown on ICCM Inability to maintain RCS temperature B. Unable to maintain reactor subcritical below 200° F as indicated by either of the following:

C. EOP directs feeding SG from SSF ASWP or station ASWP RCS temperature at the LPI Pump Suction OR (CONTINUED)

Average of the 5 highest CETCs as indicated by ICCM display OR Visual observation (CONTINUED) 12

Enclosure 4.4 RP/0/A/1000/001 Loss of Shutdown Functions Page 2 of 2 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY

1. UNEXPECTED INCREASE IN PLANT 3. MAJOR DAMAGE TO IRRADIATED 3. LOSS OF WATER LEVEL IN THE RADIATION OR AIRBORNE FUEL OR LOSS OF WATER LEVEL REACTOR VESSEL THAT HAS OR CONCENTRATION (BD 42) THAT HAS OR WILL RESULT IN THE WILL UNCOVER FUEL IN THE OPERATING MODE: All UNCOVERING OF IRRADIATED FUEL REACTOR VESSEL (BD 52)

OUTSIDE THE REACTOR VESSEL (BD 48) OPERATING MODE: 5, 6 A. LT 5 reading 14 and decreasing with makeup not keeping up with leakage WITH fuel in the core OPERATING MODE: All .1 A. Loss of all decay heat removal as indicated by the inability to maintain RCS temperature A. Valid RIA 3*, 6, 41, OR 49* HIGH Alarm below 200° F B. Valid indication of uncontrolled water decrease in the SFP or fuel transfer canal with all fuel AND assemblies remaining covered by water *Applies to Mode 6 and No Mode Only B. HIGH Alarm for portable area monitors on the LT-5 indicates 0 inches after initiation of RCS AND Makeup main bridge or SFP bridge Unplanned Valid RIA 3, 6 or Portable Area B. Loss of all decay heat removal as indicated by Monitor readings increase. C Report of visual observation of irradiated fuel the inability to maintain RCS temperature uncovered below 200° F C. 1 R/hr radiation reading at one foot away from AND a damaged storage cask located at the ISFSI D. Operators determine water level drop in either the SFP or fuel transfer canal will exceed makeup capacity such that irradiated fuel will Either train ultrasonic level indication less than D. Valid area monitor readings exceeds limits 0 inches and decreasing after initiation of RCS stated in Enclosure 4.9. be uncovered makeup NOTE: This Initiating Condition is also located NOTE: This Initiating Condition is also located in Enclosure 4.3, (Abnormal Rad in Enclosure 4.3., (Abnormal Rad Levels/Radiological Effluent). High radiation NOTE: This Initiating Condition is also located Levels/Radiological Effluent). High radiation levels will also be seen with this condition. in Enclosure 4.3, (Abnormal Rad levels will also be seen with this condition. Levels/Radiological Effluent). High radiation levels will also be seen with this condition.

(END)

(END) 13

Enclosure 4.5 RP/0/A/1000/001 Loss of Power {4} Page 1 of 1 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY

1. LOSS OF ALL OFFSITE POWER TO 1. LOSS OF ALL OFFSITE AC POWER AND 1. LOSS OF ALL OFFSITE AC POWER AND 1. PROLONGED LOSS OF ALL OFFSITE ESSENTIAL BUSSES FOR GREATER LOSS OF ALL ONSITE AC POWER TO LOSS OF ALL ONSITE AC POWER TO POWER AND ONSITE AC POWER THAN 15 MINUTES (BD 55) ESSENTIAL BUSSES (BD 57) ESSENTIAL BUSSES (BD 59) (BD 62)

OPERATING MODE: All OPERATING MODE: 5, 6 OPERATING MODE: 1, 2, 3, 4 OPERATING MODE: 1, 2, 3, 4 Defueled A. Unit auxiliaries are being supplied from A. MFB 1 and 2 de-energized A. MFB 1 and 2 de-energized Keowee or CT5 A. MFB 1 and 2 de-energized AND AND AND AND Failure to restore power to at least one MFB SSF fails to maintain Mode 3 Failure to restore power to at least one MFB within 15 minutes from the time of loss of (Hot Standby) {1}

Inability to energize either MFB from an offsite within 15 minutes from the time of loss of both both offsite and onsite AC power source (either switchyard) within 15 minutes. offsite and onsite AC power AND At least one of the following conditions exist:

2. AC POWER CAPABILITY TO 2. LOSS OF ALL VITAL DC POWER

2. UNPLANNED LOSS OF REQUIRED DC (BD 60) Restoration of power to at least one ESSENTIAL BUSSES REDUCED TO A POWER FOR GREATER THAN 15 MFB within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is NOT likely SINGLE SOURCE FOR GREATER THAN MINUTES (BD 56) 15 MINUTES (BD 58) OPERATING MODE: 1, 2, 3, 4 OR OPERATING MODE: 5, 6 OPERATING MODE: 1, 2, 3, 4 A. Unplanned loss of vital DC power to required Indications of continuing DC busses as indicated by bus voltage less than degradation of core cooling based A. Unplanned loss of vital DC power to required A. AC power capability has been degraded to a 110 VDC on Fission Product Barrier DC busses as indicated by bus voltage less single power source for > 15 minutes due to the monitoring than 110 VDC loss of all but one of the following: AND AND (END)

Unit Normal Transformer (backcharged) Failure to restore power to at least one required Unit SU Transformer DC bus within 15 minutes from the time of loss Failure to restore power to at least one required Another Unit SU Transformer (aligned)

DC bus within 15 minutes from the time of loss CT4 (END)

CT5 (END)

(END)

Loss of Power - Emergency Action Levels (EALs) apply to the ability of electrical energy to perform its intended function, reach its intended equipment. ex. - If both MFBs, are energized but all 4160V switchgear is not available, the electrical energy can not reach the motors intended. The result to the plant is the same as if both MFBs were de-energized. {4}

14

Enclosure 4.6 RP/0/A/1000/001 Fire/Explosions and Security Actions {2} {3} Page 1 of 2 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY

1. FIRES/EXPLOSIONS WITHIN THE 1. FIRE/EXPLOSION AFFECTING (CONTINUE TO NEXT PAGE) (CONTINUE TO NEXT PAGE)

PLANT (BD 65) OPERABILITY OF PLANT SAFETY SYSTEMS REQUIRED TO ESTABLISH/MAINTAIN SAFE OPERATING MODE: All SHUTDOWN (BD 70)

NOTE: Within the plant means:

Turbine Building NOTE: Only one train OPERATING MODE:of a system All needs to Auxiliary Building be affected or damaged in order to satisfy this Reactor Building condition.

Keowee Hydro Transformer Yard B3T B4T Service Air Diesel Compressors A. Fire/explosions Keowee Hydro & associated Transformers AND SSF Affected safety-related system parameter indications show degraded performance OR A. Fire within the plant not extinguished within 15 minutes of Control Room notification or Plant personnel report visible damage to verification of a Control Room alarm permanent structures or equipment required for safe shutdown B. Unanticipated explosion within the plant resulting in visible damage to permanent (Continued) structures/equipment

includes steam line break and FDW line break (Continued) 15

Enclosure 4.6 RP/0/A/1000/001 Fire/Explosions and Security Actions {2} {3} Page 2 of 2 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY

2. CONFIRMED SECURITY CONDITION 2 HOSTILE ACTION WITHIN THE 1. HOSTILE ACTION within the PROTECTED 1. A HOSTILE ACTION RESULTING IN OR THREAT WHICH INDICATES A OWNER CONTROLLED AREA OR AREA (BD 76) LOSS OF PHYSICAL CONTROL OF POTENTIAL DEGRADATION IN THE AIRBORNE ATTACK THREAT. (BD 72) THE FACILITY (BD 79)

LEVEL OF SAFETY OF THE PLANT (BD 67)

A. A HOSTILE ACTION is occurring or has OPERATING MODE: All OPERATING MODE: All occurred within the OWNER CONTROLLED OPERATING MODE: All AREA as reported by the Security Shift A. A HOSTILE ACTION is occurring or has A. A HOSTILE ACTION has occurred such that Supervision. occurred within the PROTECTED AREA as plant personnel are unable to operate A. Security condition that does not involve a reported by the Security Shift Supervision. equipment required to maintain safety HOSTILE ACTION as reported by the B. A validated notification from NRC of an functions Security Shift Supervision. AIRLINER attack threat within 30 minutes of 2. OTHER CONDITIONS EXIST WHICH IN the site. THE JUDGEMENT OF THE EMERGENCY B. A HOSTILE ACTION has caused failure of B. A credible site-specific security threat DIRECTOR WARRANT DECLARATION Spent Fuel Cooling Systems and notification OF A SITE AREA EMERGENCY. (BD 78) IMMINENT fuel damage is likely for a

3. OTHER CONDITIONS EXIST WHICH IN freshly off-loaded reactor core in pool.

C. A validated notification from NRC providing THE JUDGEMENT OF THE information of an aircraft threat EMERGENCY DIRECTOR WARRANT DECLARATION OF AN ALERT (BD 75) OPERATING MODE: All 2. OTHER CONDITIONS EXIST WHICH IN THE JUDGMENT OF THE

3. OTHER CONDITIONS EXIST WHICH A. Other conditions exist which in the judgment of EMERGENCY DIRECTOR WARRANT IN THE JUDGEMENT OF THE the Emergency Director indicate that events are in DECLARATION OF A GENERAL EMERGENCY DIRECTOR WARRANT progress or have occurred which involve actual or EMERGENCY. (BD 81)

DECLARATION OF A NOUE. (BD 69) OPERATING MODE: All likely major failures of plant functions needed for protection of the public or HOSTILE ACTION A. Other conditions exist which in the judgment that results in intentional damage or malicious of the Emergency Director indicate that events acts; (1) toward site personnel or equipment that OPERATING MODE: All are in progress or have occurred which involve could lead to the likely failure of or; (2) that OPERATING MODE: All an actual or potential substantial degradation of prevent effective access to equipment needed for A. Other conditions exist which in the judgment the level of safety of the plant or a security the protection of the public. Any releases are not A. Other conditions exist which in the judgment of the Emergency Director indicate that event that involves probable life threatening expected to result in exposure levels which of the Emergency Director indicate that events are in progress or have occurred which risk to site personnel or damage to site exceed EPA Protective Action Guideline events are in progress or have occurred indicate a potential degradation of the level of equipment because of HOSTILE ACTION.

exposure levels beyond the site boundary. which involve actual or IMMINENT safety of the plant or indicate a security threat Any releases are expected to be limited to small fractions of the EPA Protective Action substantial core degradation or melting with to facility protection has been initiated. No potential for loss of containment integrity or releases of radioactive material requiring off- Guideline exposure levels.

HOSTILE ACTION that results in an actual site response or monitoring are expected (END) loss of physical control of the facility.

unless further degradation of safety systems (END) Releases can be reasonably expected to occurs. exceed EPA Protective Action Guideline exposure levels off-site for more than the (END) immediate site area.

(END) 16

Enclosure 4.7 RP/0/A/1000/001 Natural Disasters, Hazards and Other Conditions Affecting Plant Safety Page 1 of 3 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY

1. NATURAL AND DESTRUCTIVE 1. NATURAL AND DESTRUCTIVE PHENOMENA AFFECTING THE PHENOMENA AFFECTING THE PLANT (CONTINUE TO NEXT PAGE) (CONTINUE TO NEXT PAGE)

PROTECTED AREA (BD 83) VITAL AREA (BD 89)

OPERATING MODE: All OPERATING MODE: All A. Tremor felt and seismic trigger actuates (0.05g)

A. Tremor felt and valid alarm on the strong motion accelerograph NOTE: Only one train of a safety-related system needs to be affected or damaged in B Tornado striking within Protected Area Boundary order to satisfy these conditions.

C. Vehicle crash into plant structures/systems B. Tornado, high winds, missiles resulting from within the Protected Area Boundary turbine failure, vehicle crashes, or other catastrophic event.

D. Turbine failure resulting in casing penetration or damage to turbine or generator seals AND Visible damage to permanent structures or equipment required for (CONTINUED) safe shutdown of the unit.

OR Affected safety system parameter indications show degraded performance.

(CONTINUED) 17

Enclosure 4.7 RP/0/A/1000/001 Natural Disasters, Hazards and Other Conditions Affecting Plant Safety Page 2 of 3 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY

2. NATURAL AND DESTRUCTIVE 2. RELEASE OF TOXIC/FLAMMABLE 1. CONTROL ROOM EVACUATION AND PHENOMENA AFFECTING KEOWEE GASES JEOPARDIZING SYSTEMS PLANT CONTROL CANNOT BE HYDRO CONDITION B (BD 85) REQUIRED TO MAINTAIN SAFE ESTABLISHED (BD 96) (CONTINUE TO NEXT PAGE)

================================ OPERATION OR ESTABLISH/ ================================

OPERATING MODE: All MAINTAIN MODE 5 (COLD SHUTDOWN) (BD 91)

A. Reservoir elevation 805.0 feet with all ====================================== OPERATING MODE: All spillway gates open and the lake elevation OPERATING MODE: All continues to rise A. Report/detection of toxic gases in A. Control Room evacuation has been initiated concentrations that will be life-threatening to B. Seepage readings increase or decrease greatly plant personnel AND or seepage water is carrying a significant amount of soil particles B. Report/detection of flammable gases in Control of the plant cannot be established from concentrations that will affect the safe the Aux Shutdown Panel or the SSF within 15 C New area of seepage or wetness, with large operation of the plant:

minutes amounts of seepage water observed on dam,

Reactor Building dam toe, or the abutments

Auxiliary Building

Turbine Building 2. KEOWEE HYDRO DAM FAILURE D. Slide or other movement of the dam or

Control Room (BD 97) abutments which could develop into a failure

3. TURBINE BUILDING FLOOD (BD 93) OPERATING MODE: All E. Developing failure involving the powerhouse or appurtenant structures and the operator believes the safety of the structure is questionable A. Imminent/actual dam failure exists involving OPERATING MODE: All any of the following:

Keowee Hydro Dam

3. NATURAL AND DESTRUCTIVE A. Turbine Building flood requiring use of

Little River Dam PHENOMENA AFFECTING JOCASSEE AP/1,2,3/A/1700/10, (Turbine Building Flood)

Dikes A, B, C, or D HYDRO CONDITION B (BD 86)

Intake Canal Dike

Jocassee Dam - Condition A

====================================== 4. CONTROL ROOM EVACUATION HAS OPERATING MODE: All BEEN INITIATED (BD 94) (CONTINUED)

A. Condition B has been declared for the Jocassee OPERATING MODE: All Dam A. Evacuation of Control Room (CONTINUED)

AND ONE OF THE FOLLOWING:

AND Plant control IS established from the Aux shutdown Panel or the SSF OR Plant control IS BEING established from the Aux Shutdown Panel or SSF (CONTINUED) 18

Enclosure 4.7 RP/0/A/1000/001 Natural Disasters, Hazards and Other Conditions Affecting Plant Safety Page 3 of 3 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4 RELEASE OF TOXIC OR FLAMMABLE 5. OTHER CONDITIONS WARRANT 3. OTHER CONDITIONS WRRANT 1. OTHER CONDITIONS WARRANT GASES DEEMED DETRIMENTAL TO SAFE CLASSIFICATION OF AN ALERT DECLARATION OF SITE AREA DECLARATION OF GENERAL OPERATION OF THE PLANT (BD 87) (BD 95) EMERGENCY (BD 98) EMERGENCY (BD 99)

OPERATING MODE: All OPERATING MODE: All OPERATING MODE: All OPERATING MODE: All A. Report/detection of toxic or flammable gases A. Emergency Coordinator judgment indicates A. Emergency Coordinator/EOF Director A. Emergency Coordinator/EOF Director that could enter within the site area boundary in that: judgment indicates:

judgment amounts that can affect normal operation of the plant Plant safety may be degraded Actual/imminent substantial core (END) degradation with potential for loss of B. Report by local, county, state officials for AND containment potential evacuation of site personnel based on offsite event Increased monitoring of plant functions OR is warranted Potential for uncontrolled (END) radionuclide releases that would

5. OTHER CONDITIONS EXIST WHICH result in a dose projection at the WARRANT DECLARATION OF AN site boundary greater than 1000 mRem UNUSUAL EVENT (BD 88) TEDE or 5000 mRem CDE Adult Thyroid OPERATING MODE: All (END)

A. Emergency Coordinator determines potential degradation of level of safety has occurred (END) 19

Enclosure 4.8 RP/0/A/1000/001 Radiation Monitor Readings for Emergency Classification Page 1 of 1 All RIA values are considered GREATER THAN or EQUAL TO HOURS SINCE RIA 57 R/hr RIA 58 R/hr*

REACTOR TRIPPED Site Area Emergency General Emergency Site Area Emergency General Emergency 0.0 - < 0.5 5.9E+003 5.9E+004 2.6E+003 2.6E+004 0.5 - < 1.0 2.6E+003 2.6E+004 1.1E+003 1.1E+004 1.0 - < 1.5 1.9E+003 1.9E+004 8.6E+002 8.6E+003 1.5 - < 2.0 1.9E+003 1.9E+004 8.5E+002 8.5E+003 2.0 - < 2.5 1.4E+003 1.4E+004 6.3E+002 6.3E+003 2.5 - < 3.0 1.2E+003 1.2E+004 5.7E+002 5.7E+003 3.0 - < 3.5 1.1E+003 1.1E+004 5.2E+002 5.2E+003 3.5 - < 4.0 1.0E+003 1.0E+004 4.8E+002 4.8E+003 4.0 - < 8.0 1.0E+003 1.0E+004 4.4E+002 4.4E+003

RIA 58 is partially shielded 20

Enclosure 4.9 RP/0/A/1000/001 Unexpected/Unplanned Increase In Area Monitor Readings Page 1 of 1 NOTE: This Initiating Condition is not intended to apply to anticipated temporary increases due to planned events (e.g.; incore detector movement, radwaste container movement, depleted resin transfers, etc.).

UNITS 1, 2, 3 MONITOR NUMBER UNUSUAL EVENT 1000x ALERT NORMAL LEVELS mRAD/HR mRAD/HR RIA 7, Hot Machine Shop Elevation 796 150 5000 RIA 8, Hot Chemistry Lab Elevation 796 4200 5000 RIA 10, Primary Sample Hood Elevation 796 830 5000 RIA 11, Change Room Elevation 796 210 5000 RIA 12, Chem Mix Tank Elevation 783 800 5000 RIA 13, Waste Disposal Sink Elevation 771 650 5000 RIA 15, HPI Room Elevation 758 NOTE* 5000 NOTE: RIA 15 normal readings are approximately 9 mRad/hr on a daily basis. Applying 1000x normal readings would put this monitor greater than 5000 mRad/hr just for an Unusual Event. For this reason, an Unusual Event will NOT be declared for a reading less than 5000 mRad/hr.

21

Enclosure 4.10 RP/0/A/1000/001 Definitions/Acronyms Page 1 of 5

1. List of Definitions and Acronyms NOTE: Definitions are italicized throughout procedure for easy recognition.

1.1 ALERT - Events are in process or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

1.2 BOMB - Refers to an explosive device suspected of having sufficient force to damage plant systems or structures.

1.3 COGNIZANT FACILITY STAFF - any member of facility staff, who by virtue of training and experience, is qualified to assess the indications or reports for validity and to compare the same to the EALs in the licensee's emergency classification scheme. (Does not include staff whose positions require they report, rather than assess, abnormal conditions to the facility.)

1.4 CONDITION A - Failure is Imminent or Has Occurred - A failure at the dam has occurred or is about to occur and minutes to days may be allowed to respond dependent upon the proximity to the dam.

1.5 CONDITION B - Potentially Hazardous Situation is Developing - A situation where failure may develop, but preplanned actions taken during certain events (such as major floods, earthquakes, evidence of piping) may prevent or mitigate failure.

1.6 CIVIL DISTURBANCE - A group of persons violently protesting station operations or activities at the site.

1.7 EXPLOSION - A rapid, violent, unconfined combustion, or catastrophic failure of pressurized/energized equipment that imparts energy of sufficient force to potentially damage permanent structures, systems, or components.

1.7 EXTORTION - An attempt to cause an action at the station by threat of force.

1.8 FIRE - Combustion characterized by heat and light. Sources of smoke, such as slipping drive belts or overheated electrical equipment, do NOT constitute fires. Observation of flames is preferred but is NOT required if large quantities of smoke and heat are observed.

1.9 FRESHLY OFF-LOADED CORE - The complete removal and relocation of all fuel assemblies from the reactor core and placed in the spent fuel pool. (Typical of a "No Mode" operation during a refuel outage that allows safety system maintenance to occur and results in maximum decay heat load in the spent fuel pool system).

22

Enclosure 4.10 RP/0/A/1000/001 Definitions/Acronyms Page 2 of 5 1.10 GENERAL EMERGENCY - Events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.

Releases can be reasonably expected to exceed EPA Protective Action Guidelines exposure levels offsite for more than the immediate area.

1.11 HOSTAGE - A person(s) held as leverage against the station to ensure demands will be met by the station.

1.12 HOSTILE ACTION - An act toward an NPP or its personnel that includes the use of violent force to destroy equipment, takes HOSTAGES, and/or intimidates the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities, (e.g., violent acts between individuals in the owner controlled area.)

1.13 HOSTILE FORCE - One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

1.14 IMMINENT - Mitigation actions have been ineffective, additional actions are not expected to be successful, and trended information indicates that the event or condition will occur. Where IMMINENT timeframes are specified, they shall apply.

1.15 INTRUSION - A person(s) present in a specified area without authorization. Discovery of a BOMB in a specified area is indication of INTRUSION into that area by a HOSTILE FORCE.

1.16 INABILITY TO DIRECTLY MONITOR - Operational Aid Computer data points are unavailable or gauges/panel indications are NOT readily available to the operator.

1.17 LOSS OF POWER - Emergency Action Levels (EALs) apply to the ability of electrical energy to perform its intended function, reach its intended equipment. Ex. - If both MFBs, are energized but all 4160v switchgear is not available, the electrical energy can not reach the motors intended. The result to the plant is the same as if both MFBs were de-energized.

1.18 PROJECTILE - An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.

1.19 PROTECTED AREA - Typically the site specific area which normally encompasses all controlled areas within the security PROTECTED AREA fence.

23

Enclosure 4.10 RP/0/A/1000/001 Definitions/Acronyms Page 3 of 5 1.20 REACTOR COOLANT SYSTEM (RCS) LEAKAGE - RCS Operational Leakage as defined in the Technical Specification Basis B 3.4.13:

RCS leakage includes leakage from connected systems up to and including the second normally closed valve for systems which do not penetrate containment and the outermost isolation valve for systems which penetrate containment.

A. Identified LEAKAGE LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE).

LEAKAGE, such as that from pump seals, gaskets, or valve packing (except RCP seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank; LEAKAGE through a steam generator (SG) to the Secondary System (primary to secondary LEAKAGE): Primary to secondary LEAKAGE must be included in the total calculated for identified LEAKAGE.

B. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE.

C. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall or vessel wall.

1.21 RUPTURED (As relates to Steam Generator) - Existence of Primary to Secondary leakage of a magnitude sufficient to require or cause a reactor trip and safety injection.

1.22 SABOTAGE - Deliberate damage, mis-alignment, or mis-operation of plant equipment with the intent to render the equipment inoperable. Equipment found tampered with or damaged due to malicious mischief may not meet the definition of SABOTAGE until this determination is made by security supervision.

1.23 SECURITY CONDITION - Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION.

1.24 SAFETY-RELATED SYSTEMS AREA - Any area within the Protected area which contains equipment, systems, components, or material, the failure, destruction, or release of which could directly or indirectly endanger the public health and safety by exposure to radiation.

24

Enclosure 4.10 RP/0/A/1000/001 Definitions/Acronyms Page 4 of 5 1.25 SELECTED LICENSEE COMMITMENT (SLC) -Chapter 16 of the FSAR 1.26 SIGNIFICANT PLANT TRANSIENT - An unplanned event involving one or more of the following:

(1) Automatic turbine runback>25% thermal reactor power (2) Electrical load rejection >25% full electrical load (3) Reactor Trip (4) Safety Injection System Activation 1.27 SITE AREA EMERGENCY - Events are in process or have occurred which involve actual or likely major failures of plant functions needed for the protection of the public. or HOSTILE ACTION that results in intentional damage or malicious act; (1) toward site personnel or equipment that could lead to the likely failure of or; (2) that prevents effective access to equipment needed for the protection of the public. Any releases are NOT expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the Site Boundary.

1.28 SITE BOUNDARY - That area, including the Protected Area, in which DPC has the authority to control all activities including exclusion or removal of personnel and property (1 mile radius from the center of Unit 2).\

1.29 TOXIC GAS - A gas that is dangerous to life or health by reason of inhalation or skin contact (e.g.; Chlorine).

1.30 UNCONTROLLED - Event is not the result of planned actions by the plant staff.

1.31 UNPLANNED - An event or action is UNPLANNED if it is not the expected result of normal operations, testing, or maintenance. Events that result in corrective or mitigative actions being taken in accordance with abnormal or emergency procedures are UNPLANNED.

1.32 UNUSUAL EVENT - Events are in process or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.

1.33 VALID - An indication or report or condition is considered to be VALID when it is conclusively verified by: (1) an instrument channel check; or, (2) indications on related or redundant instrumentation; or, (3) by direct observation by plant personnel such that doubt related to the instruments operability, the conditions existence, or the reports accuracy is removed. Implicit with this definition is the need for timely assessment.

25

Enclosure 4.10 RP/0/A/1000/001 Definitions/Acronyms Page 5 of 5 1.34 VIOLENT - Force has been used in an attempt to injure site personnel or damage plant property.

1.35 VISIBLE DAMAGE - Damage to equipment or structure that is readily observable without measurements, testing, or analyses. Damage is sufficient to cause concern regarding the continued operability or reliability of affected safety structure, system, or component.

Example damage: deformation due to heat or impact, denting, penetration, rupture.

1.36 VITAL AREA - An area within the protected area where an individual is required to badge in to gain access to the area and that houses equipment important for nuclear safety. The failure or destruction of this equipment could directly or indirectly endanger the public health and safety by exposure to radiation.

26

Enclosure 4.11 RP/0/A/1000/001 Operating Modes Defined In Improved Page 1 of 1 Technical Specifications MODES REACTIVITY % RATED AVERAGE CONDITION THERMAL REACTOR COOLANT MODE TITLE POWER (a) TEMPERATURE (Keff) (°F) 1 Power Operation >0.99 >5 NA 2 Startup >0.99 <5 NA 3 Hot Standby <0.99 NA >250 4 Hot Shutdown (b) < 0.99 NA 250 > T > 200 5 Cold Shutdown (b) < 0.99 NA < 200 6 Refueling (c) NA NA NA (a) Excluding decay heat.

(b) All reactor vessel head closure bolts fully tensioned.

(c) One or more reactor vessel head closure bolts less than fully tensioned 27

Enclosure 4.12 RP/0/A/1000/001 Instructions For Using Enclosure 4.1 Page 1 of 2

1. Instructions For Using Enclosure 4.1 - Fission Product Barrier Matrix 1.1 If the unit was at Hot S/D or above, (Modes 1, 2, 3, or 4) and one or more fission product barriers have been affected, refer to Enclosure 4.1, (Fission Product Barrier Matrix) and review the criteria listed to determine if the event should be classified.

1.1.1 For each Fission Product Barrier, review the associated EALs to determine if there is a Loss or Potential Loss of that barrier.

NOTE: An event with multiple events could occur which would result in the conclusion that exceeding the loss or potential loss thresholds is imminent (i.e. within 1-3 hours). In this situation, use judgement and classify as if the thresholds are exceeded.

1.2 Three possible outcomes exist for each barrier. No challenge, potential loss, or loss.

Use the worst case for each barrier and the classification table at the bottom of the page to determine appropriate classification.

1.3 The numbers in parentheses out beside the label for each column can be used to assist in determining the classification. If no EAL is met for a given barrier, that barrier will have 0 points. The points for the columns are as follows:

Barrier Failure Points RCS Potential Loss 4 Loss 5 Fuel Clad Potential Loss 4 Loss 5 Containment Potential Loss 1 Loss 3 1.3.1 To determine the classification, add the highest point value for each barrier to determine a total for all barriers. Compare this total point value with the numbers in parentheses beside each classification to see which one applies.

1.3.2 Finally as a verification of your decision, look below the Emergency Classification you selected. The loss and/or potential loss EALs selected for each barrier should be described by one of the bullet statements.

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Enclosure 4.12 RP/0/A/1000/001 Instructions For Using Enclosure 4.1 Page 2 of 2 EXAMPLE: Failure to properly isolate a 'B' MS Line Rupture outside containment, results in extremely severe overcooling.

PTS entry conditions were satisfied.

Stresses on the 'B' S/G resulted in failure of multiple S/G tubes.

RCS leakage through the S/G exceeds available makeup capacity as indicated by loss of subcooling margin.

Barrier EAL Failure Points RCS SGTR > Makeup capacity of one HPI pump in Potential Loss 4 normal makeup mode with letdown isolated Entry into PTS operating range Potential Loss 4 RCS leak rate > available makeup capacity as Loss 5 indicated by a loss of subcooling Fuel Clad No EALs met and no justification for No 0 classification on judgment Challenge Containment Failure of secondary side of SG results in a Loss 3 direct opening to the environment RCS 5 + Fuel 0 + Containment 3 = Total 8 A. Even though two Potential Loss EALs and one Loss EAL are met for the RCS barrier, credit is only taken for the worst case (highest point value) EAL, so the points from this barrier equal 5.

B. No EAL is satisfied for the Fuel Clad Barrier so the points for this barrier equal 0.

C. One Loss EAL is met for the Containment Barrier so the points for this barrier equal 3.

D. When the total points are calculated the result is 8, therefore the classification would be a Site Area Emergency.

E. Look in the box below "Site Area Emergency". You have identified a loss of two barriers. This agrees with one of the bullet statements.

The classification is correct.

29

Enclosure 4.13 RP/0/A/1000/001 References Page 1 of 1 1

References:

1. PIP O-05-02980

2. PIP O-05-4697

3. PIP O-06-0404

4. PIP O-06-03347

5. PIP O-09-00234

6. PIP O-10-1055

7. PIP O-10-01750

8. PIP O-11-02811

9. PIP O-12-1590

10. PIP O-10-7809

11. PIP O-12-9201

12. PIP O-12-9198

13. PIP O-12-11227

Examination KEY for: ILT44 ONS SRO NRC Examin Question Answer Number 1 B 2 B 3 C 4 C 5 B 6 A 7 D 8 B 9 A 10 A 11 A 12 C 13 D 14 C 15 B 16 C 17 D 18 D 19 B 20 C 21 D 22 A 23 B 24 B 25 D Printed 12/17/2013 8:20:05 AM Page 1 of 4

Examination KEY for: ILT44 ONS SRO NRC Examin Question Answer Number 26 A 27 C 28 A 29 C 30 D 31 D 32 D 33 C 34 A 35 A 36 C 37 B 38 C 39 C 40 A 41 B 42 A 43 C 44 A 45 D 46 A 47 C 48 C 49 D 50 D Printed 12/17/2013 8:20:06 AM Page 2 of 4

Examination KEY for: ILT44 ONS SRO NRC Examin Question Answer Number 51 B 52 B 53 D 54 D 55 B 56 C 57 C 58 A 59 B 60 D 61 C 62 A 63 A 64 B 65 B 66 B 67 B 68 B 69 D 70 A 71 B 72 D 73 A 74 A 75 B Printed 12/17/2013 8:20:06 AM Page 3 of 4

Examination KEY for: ILT44 ONS SRO NRC Examin Question Answer Number 76 D 77 C 78 D 79 A 80 B 81 B 82 D 83 A 84 A 85 A 86 D 87 B 88 D 89 A 90 C 91 A 92 A 93 C 94 B 95 C 96 B 97 D 98 A 99 B 100 A Printed 12/17/2013 8:20:06 AM Page 4 of 4

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