Text

302 Final SRO Written Exam
	ML13312A425
Person / Time
Site:	Watts Bar
Issue date:	11/08/2013
From:	; NRC/RGN-II
To:	; Tennessee Valley Authority
References
	50-390/13-302, 50-391/13-302
	Download: ML13312A425 (58)
	v • d • e

Name: ___________________________ 2013-302 SRO Exam Form: 0 Version: 0

76. Given the following:

- The operating crew has entered 1-ES-1 .1, "SI Termination," following an inadvertent Safety Injection.

- As the crew is placing letdown in service, the OAC reports the following:

- 1-LI-68-335A, PZR LEVEL, indicates off scale HIGH.

- 1-LI-68-321A, PZR- COLD CAL LEVEL, indicates off scale HIGH.

- 1-PI-68-334, PZR PRESS, indicates off scale LOW.

- Other Pressurizer pressure indications show pressure is dropping.

- RCS subcooling indicates 58°F and lowering.

Based on the conditions above, which ONE of the following completes the statements below?

Pressurizer level indications 1-LI-68-339 and 1-LI-68-320 will be ( 1)

In accordance with 1-ES-1 .1, the required procedure transition is to GO TO (2)

Note: 1-E-0, "Reactor Trip or Safety Injection" 1-E-1, "Loss of Reactor or Secondary Coolant" ill 0 A. lowering 1-E-0 B. lowering 1-E-1 C. rising 1-E-0 D. rising 1-E-1

2013-302 SRO Exam

77. Given the following:

- Unit 1 is operating at 100% power.

Time 1400 - MFPT 1B is manually tripped due to a leak on the pump discharge.

1410 -A Feed Water Isolation occurs due to erratic level control in one steam generator.

1420 - A loss of offsite power occurs.

1430 - Aux Air compressor A seizes.

Which ONE of the following completes the statements below?

The earliest time that ALL AFW pumps will have automatically started is (1) the loss of offsite power.

In accordance with T/S LCO 3.7.5, Auxiliary Feedwater (AFW) System, (2) of the three AFW Trains remain OPERABLE at 1431 , all other train(s) is(are) INOPERABLE.

ill ~

A. after two B. after one

c. prior to two D. prior to one

2013-302 SRO Exam

78. Given the following:

0000 - Unit 1 reactor trip breakers are opened during a shutdown for a refueling outage.

0740- A Transmission System disturbance results in the WB Hydro Station Switchyard being de-energized.

- DGs 1A-A and 1B-B fail to start.

- DGs 2A-A and 2B-B start and restore power to their respective shutdown boards.

- RCS Average Temperature is currently 252°F and slowly rising.

0756 - The Shift Manager/SED determines the required REP declaration level and declares the event.

Note: Do NOT use SED judgment as a basis for REP declaration.

Based on the conditions above, which ONE of the following completes the statement below?

The correct REP declaration is ( 1) and the declaration {2) require implementation of EPIP-8, "Personnel Accountability and Evacuation."

REFERENCE PROVIDED ill A. an ALERT does B. an ALERT does NOT C. a SITE AREA EMERGENCY does D. a SITE AREA EMERGENCY does NOT

2013-302 SRO Exam

79. Given the following timeline:

0000 The plant is operating at 100% power.

0001 Annunciator window 19-A, 125 DC VITAL CHGRIBATT Ill ABNORMAL, alarms.

0009 Local inspection reveals that the 125 V DC VITAL CHGR Ill output breaker has tripped open and that the vital charger Ill has failed 0011 Vital Battery Board Ill voltage is slightly over 132 volts.

Which ONE of the following completes the statements below?

At 0011, the indicating pointer for 1-EI-57-94, Vital Batt BD Ill AMPS, will be (1) the zero.

Based on the current conditions at 0011, the REQUIRED ACTION of Tech Spec LCO 3.8.4, DC Sources- Operating is (2) to be entered.

ill !£}

A. above required B. above NOT required C. below NOT required D. below required

2013-302 SRO Exam

80. Given the following:

- Unit 1 is operating at 100% power.

- On 10/19/2013 at 0930, the Safety Injection Pump 1A-A is declared INOPERABLE and will remain tagged for maintenance for the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

- On 10/19/2013 at 1200, a 161kv system disturbance results in the degraded voltage (OS) relays operating on 6.9kV Shutdown Board 1B-8 and 30 seconds later 161 kv Line A (Line 1) trips due to line relay operation.

Which ONE of the following completes the statements below?

The 6.9kV Shutdown Board 18-B will be energized from its (1) .

With no further manipulations of the electrical system, the Technical Specification requirement to place Unit 1 in Mode 3 is FIRST required on (2)

REFERENCE PROVIDED (1) (2)

A. Diesel Generator 10/19/2013 B. Alternate power supply 10/19/2013

c. Diesel Generator 10/20/2013 D. Alternate power supply 10/20/2013

2013-302 SRO Exam 81 . Given the following:

- The operating crew entered 1-E-1, "Loss of Reactor or Secondary Coolant."

- A RED path for secondary heat sink is evident and the crew has just entered 1-FR-H.1, "Loss of Secondary Heat Sink," and is about to perform the first step in the procedure to CHECK if secondary heat sink is required.

Currently:

- RCS pressure is 800 psig and stable.

- All S/G pressures are 900 psig and stable.

Which ONE of the following completes the statements below?

Based on the current conditions, 1-FR-H.1 will require (1) and in accordance with Tl-12.04, "User's Guide For Abnormal And Emergency Operating Instructions," the STA is required to monitor status trees (2) ill ~

A. a transition back to 1-E-1 at least once every 10-20 minutes B. continuing in 1-FR-H.1 at least once every 10-20 minutes

c. a transition back to 1-E-1 continuously D. continuing in 1-FR-H.1 continuously

2013-302 SRO Exam

82. Given the following:

- Unit 1 was operating at 100% reactor power when a LOCA occurred due to a pressurizer safety valve failing partially open.

- After initiating an RCS cooldown in accordance with ES-1.2, "Post LOCA Cooldown and Depressurization," the safety valve reclosed and remained closed.

- After ECCS flow has been terminated and normal charging established, 1-HIC-62-93A, CHARGING FLOW PZR LEVEL CONTROL malfunctions.

Subsequently:

0000 The OAC reports the following conditions exist:

Pressurizer level is 93% and rising.

RVLIS is indicating 96% and stable.

0015 A team of instrument mechanics replaces 1-HIC-62-93A and control of 1-FCV-62-93 is regained.

Which ONE of the following completes the statements below?

At 0001 , FR-O, "Status Trees," will indicate entry conditions exist for (1)

At 0016, the OAC, using only 1-HIC-62-93A, (2) be able to reduce charging flow to 0 gpm.

Note:

1-FR-1. 1, "High Pressurizer Level" 1-FR-1.3, "Voids in Reactor Vessel" (1) (2)

A. 1-FR-1.1 will B. 1-FR-1.3 will C. 1-FR-1.1 will NOT D. 1-FR-1.3 will NOT

2013-302 SRO Exam

83. Given the following:

0918 - With Unit 1 operating at 100% power, 1-AOI-27, "Main Control Room Inaccessibility," is entered due to the operating crew becoming irritated by the presence of unidentified fumes in the MCR.

0930 - All MCR actions required by 1-AOI-27 are complete and the crew establishes control in the Auxiliary Control Room.

0930 - Tech Spec 3.0.3 is entered.

1530 - The crew initiates a cooldown.

The following trends are observed:

Time S/G Pressures RCS Hot I~ TemQ 1630 1035 psig 550°F 1730 630 psig 500°F 1830 385 psig 447°F 1930 300 psig 422°F 2030 250 psig 406°F 2130 200 psig 387°F 2230 110 psig 344°F 2330 85 psig 320°F 0030 50 psig 297°F Which ONE of the following completes the statements below?

The Tech Spec required time for placing the unit in Mode 4 (1) met.

In accordance with Tech Specs, the crew (1) violate the limit for RCS cooldown rate.

ill ru A. was did B. was did NOT C. was NOT did D. was NOT did NOT

2013-302 SRO Exam

84. Given the following:

- Unit 1 is operating at 22% power in a nonnal system alignment following restart from a refueling outage in accordance with 1-G0-4, "Normal Power Operation."

- AOI-28, "High Activity In Reactor Coolant," is entered due to a step increase in DOSE EQUIVALENT 1-131 which has risen to and stabilized at 15 JlCi/gm.

- LCO 3.4.16, RCS Specific Activity, REQUIRED ACTION is implemented.

Which ONE of the following completes the statements below?

In order to maintain pressurizer level stable during performance of AOI-28, the charging flow is placed to MANUAL and INCREASED while establishing the required flow through the ( 1) .

The basis for the required action of T/S LCO 3.4.16 is to limit doses at the site boundary in the event of a (2)

A. (1) Cation Bed (2) LOCA in conjunction with the La value of 0.25%/day leakage from containment.

B. (1) Cation Bed (2) Main Steam Line Break in conjunction with an existing SG tube leakage of 150 gpd.

C. (1) Mixed Bed Demin (2) LOCA in conjunction with the La value of 0.25%/day leakage from containment.

D. ill Mixed Bed Demin

@ Main Steam Line Break in conjunction with an existing SG tube leakage of 150 gpd.

2013-302 SRO Exam

85. Given the following:

Time 21:56 - Unit 1 is operating at 100% power with the 1B-B RHR pump out of service.

21 :57 - The plant experienced a seismic event that resulted in a LOCA.

22:13 - The crew implements 1-E-1, "Loss of Reactor or Secondary Coolant."

22:34 - The SRO receives the following reports:

- RHR pump 1A-A just tripped.

- Containment sump level is 85%.

- Containment pressure is now 2.6 psig after peaking at 3.8 psig.

- RWST level is 41% and decreasing.

Based on the given conditions, which ONE of the following identifies the NEXT required procedure transition?

A. Transition to ES-1.3, "Transfer to Containment Sump."

B. Transition to FR-Z.2, "Containment Flooding."

C. Transition to ECA-1 .1, "Loss of RHR Sump Recirculation."

D. Transition to FR-Z.1, "High Containment Pressure."

2013-302 SRO Exam

86. Given the following:

- Unit 1 is in Mode 3.

- Channel calibration of RCP 1 Seal Water Flow Loop 1-LPF-62-1 is in-progress.

- The instrument mechanics are unable to set the "As-Left data to within the "As-Left" tolerances due to the failure of 1-FI-62-1A, RCP 1 SEAL WATER FLOW on 1-M-5.

- The ICS point for RCP #1 Seal Injection Flow (F0131A) which is supplied from the same flow transmitter is providing a good indication of the seal injection flow.

Which ONE of the following completes the statements below?

With NO further actions taken, a requirement to place the plant in MODE 3 will first be required in ( 1) .

NRC approval (2) be required prior to utilizing compensatory instrumentation in the place of 1-FI-62-1A.

REFERENCE PROVIDED A. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months would B. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months would NOT C. 30 days would D. 30 days would NOT

2013-302 SRO Exam

87. Which ONE of the following completes the statement below?

If becomes INOPERABLE, then it must be placed in BYPASS (rather than TRIPPED) in order to satisfy the REQUIRED ACTION of TIS LCO 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation:

A. 1-PT-1-20A, SG 3 PRESS B. 1-LT-63-50, RWST LEVEL C. 1-PT-68-323, PZR PRESS D. 1-LS-3-403-F, SOUTH VALVE VAULT ROOM LEVEL HI

2013-302 SRO Exam

88. Given the following:

0000 Unit 1 is operating at 28% power.

0001 SG #4 main feedwater line ruptures between Turbine Building wall and the South Valve Vault Room.

0002 AOI-38, "Main Steam or Feedwater Line Leak," is entered and the SRO directs a reactor trip.

Which ONE of the following completes the statements below?

Procedural direction to ensure the leak is isolated will be directed (1) the transition from 1-E-0, "Reactor Trip or Safety Injection," to the applicable procedure .

t A+- ooo"3 -in accordance with NPG-SPP-03.5, "Regulatory Reporting Requirements,"

I an 8-hour notification would be required due the actuation of (2)

I REFERENCE PROVIDED A. c.o-MCA.

~"on<JOcf.. A. prior to a single system c.opi~cJ B. onyI after multiple systems 1 to to<A)Orc:.os6 C. only after a single system D. priorto multiple systems

2013-302 SRO Exam

89. Given the following:

- Unit 1 is operating at 100% power.

- The A8 AUO reports:

- The MDAFWP 18-B pump casing is hot to the touch.

TI-3-146, MDAFWP 18-B DISCH LINE TEMP reads 180°F.

- The handheld rounds log requires a notification to be made if the temperature is greater than 11 ooF.

- Last shift 1-TI-3-146 read 89°F.

- Troubleshooting reveals that back leakage is occurring from SG #3.

Which ONE of the following completes the statements below?

Based on the given conditions, the 18-B MDAFWP is to first be ( 1) of SOI-3.02, Auxiliary Feedwater System.

OPDP-8, "Operability Determination Process and Limiting Conditions for Operation Tracking," (2) allow the performance of an Immediate Determination of OPERABILITY (I DO) to be completed by a licensed, on-shift RO.

Note: S0/-3.02 Section 8. 7, "Venting of AFW Pumps" Section 8.1.2, "Manual Startup of AFW Pump B-B" ill ~

A. vented per section 8.7 does B. run per section 8.1.2 does

c. vented per section 8. 7 does NOT D. run per section 8.1.2 does NOT

2013-302 SRO Exam

90. Given the following:

Unit 1 is operating at 100% power

- Q-SI-82-11A, "Monthly Diesel Generator Start and Load Test DG 1A-A",

is in progress

- The 1A-A DG has been paralleled to its shutdown board and has been loaded to 4.2 MW and is providing 1.00 MVARs OUTGOING CSST C Tap changer begins to swing (hunt) uncontrollably Subsequently:

- The US directs that the CSST C Tap changer be taken to manual in accordance with ARI 501-B, CSST C ABNORMAL 1A-A DG remains at a stable loading of 4.2 MW and 1.00 MVARs OUTGOING 1A-A 6.9kV SDBD voltage is stable at 7000V

- ALL minimum voltage requirements for the off-site circuits are met Given the conditions listed, which ONE of the following completes the statements below?

As a result of the current Tap Changer control alignment, the performance of 0-SI-82-2, "8 Hour Diesel Generator AC Power Source Operability Verification" (1) required.

Notification of transmission operator (2) required in accordance with Tl-12.15, "161kV Offsite Power Requirements."

A. (1) IS (2) IS B. (1) IS (2) IS NOT C. (1) IS NOT (2) IS D. (1) IS NOT (2) IS NOT

2013-302 SRO Exam

91. Given the following:

- Unit 1 is operating at 90% and is conducting a downpower to 50%.

- Control Bank D rod H8 CERPI indication is 218 steps

- Control Bank D rod H12 CERPI indication is 214 steps

- All other Bank D rods indicate 201 steps.

- Bank D Demand Counter indicates 203 steps.

- No other equipment is out of service.

Which ONE of the following completes the statement below?

For the given plant conditions, a REQUIRED ACTION of LCO 3.1.5, Rod Group Alignment Limits (1) require Unit to be placed in MODE 3.

If it will take longer than one hour to correct a rod misalignment, 1-AOI-2 requires that the rod misalignment be corrected with the RBSS in (2)

Note: 1-A0/-2, "Malfunction of Reactor Control System" RBSS, "Rod Bank Selector Switch" REFERENCE PROVIDED A. does manual B. does NOT manual C. does bank select for the affected bank D. does NOT bank select for the affected bank

2013-302 SRO Exam

92. Given the following:

- 30 days after restart from a refueling outage, Unit 1 was operating at 100% when a rapid load decrease was required due to an MSR relief valve inadvertently opening.

- During the down power the MSR relief valve went closed and the unit was stabilized at 72% load.

- Following the load decrease the following annunciators are in ALARM:

83-D, PLANT COMPUTER GENERATED ALARM (SEE ICS) 87-A, ROD INSERTION LIMIT LO

- AFD is more negative than the lower limit on 3 of the 4 NIS channels Which ONE of the following completes the statements below?

To address the current conditions, AOI-34, "Immediate Boration" (1) required to be implemented.

LCO 3.2.3, Axial Flux Difference (AFD), bases state that the limits on the AFD ensure that (2) during either normal operation or in the event of xenon redistribution following power changes.

A. (1)is NOT (2) Heat Flux Hot Channel Factor limit (Fa(Z)) is not exceeded B. (1)is NOT (2)Crud Induced Power Shifts (CIPS) do not occur C. (1)is (2) Heat Flux Hot Channel Factor limit (Fa(Z)) is not exceeded D. {1) is (2) Crud Induced Power Shifts (CIPS) do not occur

2013-302 SRO Exam

93. Given the following:

- The Aux Building Crane is carrying a load of 2040 lbs. on the Auxiliary Hoist.

- No limits are bypassed on the Aux Building Crane.

Based on the above conditions, which ONE of the following completes the statements below?

The Auxiliary Building Crane (1) travel over the spent fuel pit with the conditions stated above.

The basis for the TR 3.9.4, "Crane Travel- Spent Fuel Storage Building,"

states that the weight restriction on the movement of loads over fuel assemblies in the storage pool areas ensures that in the event that the load is dropped, the activity release will be limited to that contained in (2) ill A. can a single fuel assembly B. can NOT a single fuel assembly

c. can two fuel assemblies D. can NOT two fuel assemblies

2013-302 SRO Exam

94. Given the following:

- Unit is operating at 100% power.

- A tornado was sighted over the Watts Bar Training Center, but dissipated without touching down onsite.

- Confirmation was subsequently received that no visible damage had been received to any structures or equipment on site.

- An evaluation of the Radiological Emergency Plan (REP) determines the conditions for an NOUE were initially met, but are now fully resolved.

Which ONE of the following completes the statements below?

In accordance with EPIP-1, "Emergency Plan Classification logic," the condition shall be (1) .

The NRC is required to be notified within a maximum of (2) of the action taken.

ill A. reported to the ODS but the NOUE will NOT be declared 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months B. reported to the ODS but the NOUE will NOT be declared 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months C. declared and tenninated the NOUE at the same clock time 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months D. declared and terminated the NOUE at the same clock time 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

2013-302 SRO Exam

95. Given the following:

- The indicating lights for 1-FCV-30-40, LWR CNTMT PURGE EXH PRESS RLF extinguish on both the 1-M-9 HS and 1-XX-55-6E, CNTMT ISOL STATUS PNL.

- One blown fuse was found in the control circuit for 1-FCV-30-40.

- The valve travelled to its failed position.

- The blown fuse is replaced like-for-like.

- The lost indicating lights are restored.

Which ONE of the following completes the statements below?

In accordance with OPDP-7, "Fuse Control," following fuse replacement, a full stroke of 1-FCV-30-40 ( 1) required to prove operability of the valve.

If 1-FCV-30-40 had not been able to be opened, (2) would be used to maintain containment pressure within the TIS LCO limits.

tu @

A. is EGTS B. is containment purge

c. is NOT EGTS D. is NOT containment purge

2013-302 SRO Exam

96. Given the following:

- Unit 1 operating in MODE 1.

Which ONE of the following identifies a condition where an entry into the Unit 1 narrative log is required for equipment made inoperable, but the completion of an entry into the LCO Tracking Log would NOT be required, in accordance with OPDP-8, "Limiting Conditions for Operations Tracking."

A. The equipment made inoperable will be restored to operable status before the end of the current shift.

B. A Safety Function Determination identifies that no loss of Safety Function exists.

C. The equipment made inoperable is not required unless the unit enters MODE3.

D. The equipment made inoperable is being tracked in accordance with the clearance procedure.

2013-302 SRO Exam

97. Given the following conditions:

- Unit 1 was operating at 100% power.

- A large break LOCA occurred which precipitated entry into the Severe Accident Management Guidelines (SAMGs).

- The TSC and OSC are fully staffed and operational.

Subsequently:

- The TSC RP Manager believes that members of the emergency response organization may be exposed to a Thyroid dose of 11 REM.

In accordance with EPIP-15, "Emergency Exposure Guidelines," which ONE of the following completes the statements below?

The use of Potassium Iodide (KI) (1) be commenced.

The Site Emergency Director (SED) (2) required to approve the distribution of Kl.

ill A. should NOT is B. should is

c. should NOT is NOT D. should is NOT

2013-302 SRO Exam

98. Given the following:

- Unit 1 is at 100% power.

- Waste Gas Decay Tank 'E' needs to be released.

- The minimum time from fill date to release has NOT been met.

In accordance with SOI-77.02, "Waste Gas Disposal System," which ONE of the following correctly identifies whose approval is required to waive the minimum decay time requirement and the minimum required Operations Department approval(s) for this gaseous release permit?

Waive the Decay time Approve the release A. Chemistry Duty Manager Shift Manager or Unit Supervisor B. Chemistry Duty Manager Both Shift Manager and Unit Supervisor

c. Radiation Protection Manager Both Shift Manager and Unit Supervisor D. Radiation Protection Manager Shift Manager or Unit Supervisor

2013-302 SRO Exam

99. Given the following:

- Unit 1 experienced a Reactor Trip and Safety Injection.

- The crew transitioned to 1-FR-Z.1, "High Containment Pressure," from 1-E-1,

- Loss of Reactor or Secondary Coolant."

- While the crew was performing 1-FR-Z.1 step 1, the crew transitioned to 1-FR-P.1, "Pressurized Thermal Shock" and is performing the first step to check RCS pressure greater than 150 psig

- The STA reports the containment pressure has dropped and the containment status tree is GREEN and that no other RED or ORANGE paths exist.

Which ONE of the following completes the statements below in accordance with 1-FR-P.1?

In accordance with the Westinghouse background document for 1-FR-P .1, the basis of the 1-FR-P.1 step checking RCS pressure greater than 150 psig is to (1) .

Based on the given conditions, if a procedure step in 1-FR-P.1 directs the operator to "RETURN TO Instruction in effect," the team is required to transition to (2) .

A. (1) preclude performing 1-FR-P.1 actions, since pressurized thermal shock is not a serious concern for a large-break LOCA (2) 1-E-1 B. (1) preclude performing 1-FR-P.1 actions, since pressurized thermal shock is not a serious concern for a large-break LOCA (2) 1-FR-Z.1 C. (1) determine if the 1-FR-P.1 steps requiring soak periods are required as the procedure is performed (2) 1-E-1 D. (1) determine if the 1-FR-P.1 steps requiring soak periods are required as the procedure is performed (2) 1-FR-Z.1

2013-302 SRO Exam 100. Given the following:

At 0430, the SRO answers the NRC ENS phone and receives the plant status communications check.

Which ONE of the following completes the statement below?

The authentication code which the SRO receives from the NRC is (1)

The authentication code _.....J.::(2:.L.)_ SAFEGUARDS information.

A. (1) a redundant method by which the identity of a caller may be established as the ENS phone displays caller identification (caller "ID")

(2) IS B. (1) a redundant method by which the identity of a caller may be established as the ENS phone displays caller identification (caller "ID")

(2) IS NOT C. (1) a short, simple means of caller authentication that may eliminate the need to perform a callback (2) IS D. (1) a short, simple means of caller authentication that may eliminate the need to perform a callback (2) IS NOT

ANSWER KEY REPORT for 2013-302 SRO Exam Test Form: 0

ID Points Type Answers I 008 AA2.12 376 1.00 MCS D 2 054 AA2.04 477 1.00 MCS c 3 055 E02.4.41 178 1.00 MCS c 4 058 AA2.01 279 1.00 MCS A 5 077 A02.2.37 580 1.00 MCS c 6 WIEOS E02.1.20 981 1.00 MCS c 7 028 A02.4.21 582 1.00 MCS c 8 068 A02.1.7 183 1.00 MCS B 9 076 AA2.06 584 1.00 MCS D 10 W/E15 EA2.2 185 1.00 MCS B 11 004 02.2.12 286 1.00 MCS D 12 013 02.2.40 587 1.00 MCS B 13 059 A2.05 288 1.00 MCS B 14 061 A2.06 289 1.00 MCS c 15 064 A2.09 590 1.00 MCS c 16 001 02. 1.31 591 1.00 MCS D 17 015 A2.04 92 1.00 MCS A 18 034 A3.0 l 593 1.00 MCS B 19 0 2.1.38 94 1.00 MCS A 20 0 2.2.21 295 1.00 MCS D 21 0 2.2.23 96 1.00 MCS A 22 0 2.3. 14 997 1.00 MCS D 23 0 2.3.6 98 1.00 MCS A 24 0 2.4.18 199 1.00 MCS B 25 0 2.4.43 500 1.00 MCS D SECTION 1 ( 25 items) 25.00 Friday, November01 , 201310:35:41 AM 1

2013-302 WATTS BAR NRC SRO LICENSE EXAM 08/19/2013 REFERENCE PACKAGE For Question 18:

1. ECA-1 .1, Loss of RHR Sump Recirculation, page 29 For Question 22:

1. AOI-11 , Loss of Condenser Vacuum, page 9 (REDACTED)

For Question 31 :

1. 1-SI-74-904-8, Position Indication Verification Residual Heat Removal System (Train 8), pages 19 through 23 (REDACTED)

For Question 78:

1. EPIP-1 , Emergency Plan Classification Logic, pages 26 and 27 For Question 80:

1. LCO 3.8.1 AC Sources- Operating, pages 3.8-1, 3.8-2, 3.8-2a, 3.8-3, 3.8-4 and 3.8-5

2. LCO 3.5.2 ECCS- Operating, page 3.5-4 For Question 86:

1. LCO 3.5.5 - Seal Injection Flow, page 3.5-11

2. Part II - Fire Protection Plan, page 11-59 (REDACTED)

3. 1-SI-68-33, Measurement of Reactor Coolant Pump Seal Injection Flow, page 12 For Question 88:

1. NPG-SPP-03.5, Regulatory Reporting Requirements, pages 19 through 25 (REDACTED)

For Question 91:

1. LCO, 3.1 .5, Rod Group Alignment Limits, pages 3.1-8, 3.1-9 and 3.1-10 (REDACTED)

WBN Loss of RHR Sump Recirculation ECA-1.1 Unit 1 Rev.0012 Figure 1 (Page 1 of 1)

Minimum Sl Flow for Decay Heat vs. Time After Trip 700.0 600.0

"~

500.0 "'\

-E

~

8; 0::

400.0 '\

~

~ Unacceptable Acceptable 1

~ 300.0 Re;Jion Reaion 1 0

u:: "" ~

200.0 ........

l"o 100.0

~

0.0 10 100 1000 10000 Time After Trip (Minutes)

Page 29 of35

§=~

EZ ::;:UJ

o 'A& B' and 'C' CONDENSER VACUUM LO-LO AND LO ALARMS .z

-~

gm......

0--4 6.50 ..--.----..-----,---,.---,---r---r---r---r--..,---.,..--..---,

0~

"OC/l -*

0)>

01-o

--o (1) (1) 8.00 :

ZONE r-()::J 0 oQ; 3 X (") ~

"0 c -*

-m

(/) ,....

0

I a

~ ::J ~ 5.50 c. (")

CD 0 m-

Jo ~

01 BZONES ; :::I c.

<ecn c ..,

CD -., )> CD

~ :;- () :::, < S>> "C

I Cll

<0 (1)

(1)0 (1) ::J w*s.oo S>>

(')

CQ "C CD CD ..,

CD co :J. q- lk: c .:::I a

J 0

<0=

U) c 0 -*

c. ~

3 ... ><

<0 Q)(l) oa.

"G) "'~**~ ~-~---t--~--t-~r--+--~-?~~~-+--~--1-_J -

(") -

<I'

(')

c c

J ...,

0 Q) EL

(/) 3

e-o

-~ lk:

(j)

~Q) w iii "C

(Q::J (1)0. ~ 4.00 I I I I I II I Y ,f 1 1 :::r

~ *~

3 (/)

(1)~ ~

J 0

-c z I

a: 0 3.50 I

(,) I I I '.. .'bY 1, ! I I I I

. - . -I . .. . ..... - --*I *- I I I C)

C)

z ~

0

.... - -LO VAOJUM INC ZONE 0" - LO*LO*VACUUM A&BZONE (1) 3.00 :

3 --- -LO*VAOJUM A&BZONE g_

3i MWe

~ 2.50 :

0 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 LOAO(MWe)

WBN Position Indication Verification 1-SI-74-904-B Unit 1 Residual Heat Removal System Rev.0014 (Train 8) Page 19 of 33 Appendix C (Page 1 of 5) 1-FCV-74-21 Data Sheet Data Package: Page _ __ Date _ __

1.0 1-FCV-74-21 DATA SHEET Handswitch Description RHR PMP B SUCTION Handswitch Location 1-M-6 Timing Direction OPEN TO CLOSED Valve Description RHR PUMP 1B-B SUCTION Valve Location A6V/676 Tracking Number 96-PV-0496 I.Start of Critical Step(s)

NOTES

3) The following is a critical step due to potential to isolate suction to a running RHR pump and cause potential damage.

[1] ENSURE RHR Pump 18-B is NOT in service.

IEnd of Critical Step(s)

[2] RECORD the Stopwatch M&TE ID Number.

M&TE

WBN Position Indication Verification 1-SI-74-904-B Unit 1 Residual Heat Removal System Rev. 0014 (Train B) Page 20 of33 Appendix C (Page 2 of 5)

[3] RECORD as-found position of valves listed in the table below, AND ENSURE valves are in the required position:

AS-FOUND REQUIRED VALVE DESCRIPTION POSITION POSITION PERFORMED RHR PMP 1B-8 TO 1-FCV-63-11 CHG PMP& SIP CLOSED 18-8 SUCT ISOL CONTAINMENT SUMP 1-FCV-63-73 TO RHR PUMP 18-8 CLOSED ISOL RHR SPRAY HDR 8 1-FCV-72-41 CLOSED ISOLATION RHR PUMP 18-B 1-FCV-74-21 OPEN SUCTION

[4] DISPATCH Operator to 1-FCV-74-21 to observe local valve stem movement.

NOTES 2} Abnormal or erratic operation observed during stroke time testing should be recorded, evaluated, and dispositioned as a problem during testing in accordance with SPP-8.1, Conduct of Testing.

[5] PLACE 1-HS-74-21A, RHR PMP A SUCTION, in CLOSE, AND MEASURE stroke time of 1-FCV-74-21 with stopwatch.

WBN Position Indication Verification 1-SI-74-904-B Unit 1 Residual Heat Removal System Rev.0014 (Train B) Page 21 of33 Appendix C (Page 3 of 5)

[6] RECORD the measured stroke time for 1-FCV-74-21 :

MEASURED STROKE STROKE TIME LIMITING VALUE STROKE TIME ACCEPTANCE OF FULL STROKE DIRECTION (sees) CRITERIA TIME (sees)

First Test OtoC 38.88 to 52.60 32.02 to 59.46 Second Test

[7] VERIFY the following conditions exist (Ace Crit):

A. 1-FCV-74-21 locally observed to move from OPEN to CLOSED.

B. 1-HS-74-21A Green light is LIT and Red light is NOT LIT.

[8] IF the MEASURED STROKE TIME for 1-FCV-74-21 is within the LIMITING VALUE OF FULL STROKE TIME, BUT is NOT within the STROKE TIME ACCEPTANCE CRITERIA, THEN A. REPEAT Steps 1.0[3], 1.0[5], and 1.0[6] to obtain a second stroke time.

B. RECORD the date and time of the second test.

Date Time C. IF the second stroke time is within the STROKE TIME ACCEPTANCE CRITERIA, THEN CONTACT the System Engineer to determine the cause of the stroke time deviation, AND DOCUMENT the cause of the stroke time deviation in the test data package.

WBN Position Indication Verification 1-SI-74-904-B Unit 1 Residual Heat Removal System Rev.0014 (Train B) Page 22 of33 Appendix C (Page 4 of 5)

D. IF the second stroke time is within the LIMITING VALUE OF FULL STROKE TIME, but is NOT within the STROKE TIME ACCEPTANCE CRITERIA, THEN CONTACT the System Engineer to analyze the data, AND NOTIFY Operations that the analysis of the valve stroke time data must be completed within 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months of the time recorded in Step 1.0[8]8 above.

[9] VERIFY the MEASURED STROKE TIME for 1-FCV-74-21 is:

0 WITHIN the STROKE TIME ACCEPTANCE CRITERIA range of 38.88 to 52.60 seconds (Acceptance Criteria),

OR 0 WITHIN the LIMITING VALUE OF FULL STROKE TIME range of 32.02 to 59.46 seconds (Acceptance Criteria)

[10] OPEN 1-FCV-74-21 using 1-HS-74-21A.

[11] VERIFY the following conditions exist (Ace Crlt):

A. 1-FCV-74-21 locally observed to move from CLOSED to OPEN.

B. 1-HS-74-21A Red light is LIT and Green light is NOT LIT.

[12] VERIFY 1-FCV-74-21 is OPEN.

IV

WBN Position Indication Verification 1-SI-74*904-B Unit 1 Residual Heat Removal System Rev.0014 (Train B) Page 23 of33 AppendixC (Page 5 of 5)

NOTE The following step or portion of the step may be marked N/A as needed for valves which were found in their required position.

[13] PLACE valves listed in the table below in the as-found position documented in Step 1.0[3], and VERIFY valve position:

VALVE DESCRIPTION PERFORMED IV 1-FCV-63-11 RHRPMP1~BTOCHGPMP&

SIP 1B-B SUCT ISOL 1-FCV-63-73 CONTAINMENT SUMP TO RHR PUMP 1B-B ISOL 1-FCV-72-41 RHR SPRAY HDR B ISOLATION

WBN Emergency Plan Classification Logic EPIP-1 UnitO Rev.0038 Page 26 of 53 Attachment 3 (Page 3 of4) 3 2 Loss of AC (Shutdown)

Mode Initiating/Condition Prolonged loss of Offsite and Onsite AC Nol Applicable power (1 and 2)

1. 1AJWl1B 6.9KV Shutdown Boards de-energized for >15 minutes

2. (a orb)

a. Core Cooling Red m: Orange

b. Restoration of Either 1A 21: 1 B 6 .9KV ShutdQwn Board(s) is not ~kely within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of loss.

Loss of Offslte .l!!!S!. Onslte AC Power Nol Applicable

> 15 minutes

1. 1AJM 18 6.9KV Shutdown Boards de-energized for >15 minutes Loss of Offsite Power for >15 minutes 5,6, or UNPLANNED loss of Offslte Jm1 Onslte AC Defuel power for >15 minutes (1 and 2)

1. 1A and 1 B 6 .9KV Shutdown Boards

1. C ins!. D CSSTs are not available for de-energized for >15 minutes

>15 minutes Also Refer to "Loss of Shutdown Systems" (6. 1)

2. 1A.2! 18 Diesel Generator not available Loss of Offslte Power for >15 minutes 5,6, or UNPLANNED loss of Offslte Power for Defuel >15 minutes (1 and 2)

(1 and 2)

1. C !!!Sl D CSSTs not available for

1. C !rut D CSSTs not available for >1 5 minutes

>15 minutes

2. Either Diesel Generator is supplying power

2. Each Diesel Generator is supplying power to to its respective Shutdown Board its respective Shutdown Board

WBN Emergency Plan Classification Logic EPIP-1 Unit 0 Rev.0038 Page 27 of 53 Attachment 3 (Page 4 of 4)

Refer to "Fission Product Barrier Matrix" and "Loss of Function" (2.2) l 0

s s

Loss of All Vital DC Power fcw >15 minutes 0

1. Vollage <105V DC on 125V DC Vital Battery F Buses 1*1 g 1-lli.!!S! 1*111 g 1*1V for p

>15 minutes 0 Also Refer to 'Fission Product Barrier Matrix", w

'Loss of Function* (2.2). and *toss of E Instrumentation" (2. 1) R u

Also Refer to "Fission Product Barrier Matrix*, 1 "Loss of Function" (2.2), and "Loss of Instrumentation* (2. 1)

UNPLANNED Loss of the Required Train of DC power for >15 minutes (1 or 2)

1. Voltage <105V DC on 125V DC Vital Battery Buses 1*1 ~ 1* 111 for >15 minutes

2. Voltage <105V DC on 125V DC Vital Battery Buses 1-11 !!!!11-IV for >15 minutes

AC Sources - Operating 3.8.1 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources - Operating LCO 3.8.1 The following AC electrical sources shall be OPERABLE:

a. Two qualified circuits between the offsite transmission network and the onsite Class 1E AC Electrical Power Distribution System; and

b. Four diesel generators (DGs) capable of supplying the onsite Class 1E AC Electrical Power Distribution System.

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

The C-S DG may be substituted for any of the required DGs.

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

ACTIONS

*--*----------------NOTE------------------------------------

LCO 3.0.4.b is not applicable to DGs.

Watts Bar-Unit 1 3.8-1 Amendment 55

AC Sources -Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3 Restore offsite circuit to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OPERABLE status.

AND 6 days from discovery of failure to meet LCO B. One required DG B.1 Perform SR 3.8.1.1 for the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable. offsite circuits.

AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND B.2 Declare required feature(s) 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from supported by the inoperable discovery of DG inoperable when its Condition B required redundant feature(s) concurrent with is inoperable. inoperability of redundant required feature(s)

AND B.3.1 Determine OPERABLE DGs 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months are not inoperable due to common cause failure.

QB B.3.2 Perform SR 3.8.1.2 for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months OPERABLE DGs.

AND (continued)

Watts Bar-Unit 1 3.8-2 Amendment 39

AC Sources -Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.4 Restore required DG to 14 days OPERABLE status.

AND 17 days from discovery of failure to meet LCO

c. Two required DGs in Train C.1 Perform SR 3.8.1.1 for the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months A inoperable. offsite circuits.

AND OR Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Two required DGs in Train thereafter B inoperable. AND C.2 Declare required feature(s) 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from supported by the inoperable discovery of DGs Inoperable when its Condition C required redundant feature(s) is concurrent with inoperable. inoperability of redundant required feature(s)

AND C.3.1 Determine OPERABLE DGs are not inoperable due to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months common cause failure.

OR C.3.2 Perform SR 3.8.1.2 for OPERABLE DGs.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND (continued)

Watts Bar-Unit 1 3.8-2a Amendment 39

AC Sources -Operating 3.8.1 CONDITION REQUIRED ACTION COMPLETION TIME

c. (continued) C.4 Restore at least one required 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months DG to OPERABLE status.

AND 6 days from discovery of failure to meet LCO D. Two offsite circuits D.1 Declare required feature(s) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months from inoperable. inoperable when its redundant discovery of required feature(s) is Condition D inoperable. concurrent with inoperability of redundant required features AND D.2 Restore one offsite circuit to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE status.

(continued)

Watts Bar-Unit 1 3.8-3 Amendment 30, 39

AC Sources -Operating 3.8.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. One offsite circuit -------- ------NOTE--------*-----

inoperable. Enter applicable Conditions and Required Actions of LCO 3.8.9, AND "Distribution Systems - Operating,"

when Condition E is entered with no AC One or more required power source to any train.

DG(s) in Train A ----------------------------------- --

inoperable.

E.1 Restore offsite circuit to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months QB OPERABLE status.

One or more required Q.B DG(s) in Train B inoperable. E.2 Restore required DG(s) to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months OPERABLE status.

F. One or more required F.1 Restore required DGs in Train 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months DG(s) in Train A A to OPERABLE status.

inoperable.

Q.B AND F.2 Restore required DGs in Train One or more required B to OPERABLE status 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months DG(s) in Train 8 inOPerable.

G. Required Action and G.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Associated Completion Time of Condition A, B, C, AND D, E, or F not met.

G.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued)

Watts Bar-Unit 1 3.8-4 Amendment 39

AC Sources -Operating 3.8.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME H. Two offsite circuits H.1 Enter LCO 3.0.3. Immediately inoperable.

AND One or more required DG(s) in Train A inoperable.

OR One or more required DG(s) in Train B inoperable.

I. One offsite circuit 1.1 Enter LCO 3.0.3. Immediately inoperable.

8t!lQ One or more required DG(s) in Train A inoperable.

AND One or more required DG(s) in Train B inoperable.

Watts Bar-Unit 1 3.8-5 Amendment 39

ECCS - Operating 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS- Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE.

NOTES-------------------------- --

1. In MODE 3, both safety injection (SI) pump flow paths may be isolated by closing the isolation valves for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to perform pressure isolation valve testing per SR 3.4.14.1.

2. In MODE 3, the safety injection pumps and charging pumps may be made incapable of injecting to support transition into or from the Applicability of the LCO 3.4.12, Cold Overpressure Mitigation System (COMS) for up to four hours or until the temperature of all the RCS cold legs exceeds 375°F, whichever occurs first.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more trains A.1 Restore train(s) to OPERABLE 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable. status.

AND At least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available.

B. Required Action and 8 .1 BeinMODE3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not IT!et. AND 6.2 Be in MODE4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Watts Bar-Unit 1 3.5-4 Amendment 55

Seal Inj ection Flow 3 . 5 .5 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

3. 5.5 Seal Injection Flow LCO 3.5 . 5 Reactor coolant pump seal injecti on f low s hall be ~ 40 gpm wi th chargi ng pump dis charge header pressure ~ 2430 psig and the pressurizer l evel control valve full open.

APPLICABILI TY: MODES 1, 2, and 3 .

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Seal injection flo w A.l Adjust manual seal 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months not wi t hin limit. i njection t hrottle valves to give a f low within limit with charging pump discharge header pre ssure

~ 2 430 psig and the pressurizer level control valve full open.

B. Required Action and B.1 Be in MODE 3 . 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Ti me no t AND met.

B. 2 Be in MODE 4 . 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Watts Bar- Unit 1 3 . 5 - 11

PLAN Rev. 22 14.9 Emergency Battery Lighting Units Emergency battery lighting units provided for FSSD shall be Operable whenever the illuminated associated fire safe shutdown equipment is required.

14.9.1 With any of the emergency battery lighting units provided for FSSD inoperable, restore the inoperable units to Operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months -OR- ensure alternate lighting is available.

14.9.2 Restore the inoperable emergency battery lighting unit to Operable status within 14 days. If not restored within 14 days, continue the compensatory actions AND perform 10CFR50.72 and/or 10CFR50.73 reviews per site administrative procedures.

14.10 Fire Safe Shutdown Equipment The equipment listed on Table 14.10 is required for Fire Safe Shutdown(FSSD) and shall be Operable (or in its FSSD condition) when the unit is in modes 1, 2, and 3. The non-System 26 valves noted on the plants mechanical flow diagrams as being administratively locked in the open, closed, or throttled position (with breaker open) for Appendix R shall be maintained in that condition when the unit is in Modes 1, 2 and 3.

14.10.1 With one or more required equipment in Table 14.10 inoperable (or not in its FSSD condition), restore to operable status (or its FSSD condition) within 30 days.

14.10.2 With one or more of the breakers and/or valves specified in design output documents not in the noted position or condition, return the breakers and/or valve to the required position within 30 days.

II- 59

WBN Measurement Of 1-SI-68-33 Unit 1 Reactor Coolant Pump Rev.0013 Seal lnj_ection Flow Page 12 of20 Data Package: Page __ of _ _ Date _ __

6.2 Determination of Seal Leakage (continued)

NOTE Charging Pump discharge header pressure ~ 2440 psig with 1-FCV-62-93 full open are the test conditions for this instruction. 1-HIC-62-89A, CHRG HDR-RCP SEALS FLOW CONTROL, may be adjusted as required to obtain proper seal injection flow of s 37.6 gpm total , as long as charging header pressure indicates ~ 2440 psig AND RCS pressure is between 2215 and 2255 psig.

[5] ADJUST 1-HIC-62-89A, CHRG HDR-RCP SEALS FLOW CONTROL, to maintain a pressure reading of ~ .2440 psig on test gauge installed at 1-PT-62-92A.

[6] ALLOW system to stabilize, AND RECORD the following pressure readings:

Test gauge at 1-PT-62-92A (Ace Crft: ~ 2440 psig)

- - --- psig Computer Point P0142A (information only) _ _ _ _ _ psig NOTE Only one flow rate per pump is required for Step 6.2[7]. Data not taken may be marked NA.

[7] RECORD the seal injection flowrate to each of the four Reactor Coolant Pumps (RCPs) using the corresponding computer point or flow indicator on 1-M-5:

A. RCP #1 F0131A: (COMPLIANCE) _ _ __ gpm 1-FI-62-1A, RCP 1 SEAL WATER FLOW: (COMPLIANCE) _ _ _ _ gpm B. RCP#2 F0129A: (COMPLIANCE) _ _ __ gpm 1-FI-62-14A, RCP 2 SEAL WATER FLOW: (COMPLIANCE) _ _ __ gpm

NPG Standard Regulatory Reporting Requirements NPG-SPP-03.5 Programs and Rev.0008 Processes Page 19 of99 Appendix A (Page 1 of 15)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 1.0 PURPOSE This Appendix identifies reporting requirements; a.nd instructions for determining reportability, preparation, and transmittal of LERs; and notification to NRC for events occurring at TVA's licensed nuclear plants.

2.0 SCOPE TVA is required by §50.72 and §50.73 to promptly report various types of conditions or events and provide written follow-up reports, as appropriate. This appendix provides reporting guidance applicable to licensed power reactors.

NOTES

1) Appendix B provides additional reporting criteria found in §Part 20, 30, 40, and 70 that may be applicable to events involving byproduct, source or special nuclear material possessed by the licensed nuclear plant. Site Licensing and Site RadCon are responsible for making the reportability determinations for §Part 20, 30, 40, or 70 events associated with their site. Corporate Licensing and Corporate RadChem are responsible for making the reportability determinations for §Part 20, 30, 40, or 70 events associated with all other TVA licensed activities. Licensing is responsible for developing (with input from affected organizations) and submitting the immediate notification and written reports to NRC in accordance with §Part 20, 30, 40, or 70 requirements. Reporting requirements for personnel exposure required by §Part 20 are contained in RCTP-105, Personnellnprocessing and Dosimetry Administrative Processes.

2) Appendix C contains the criteria for reporting if events or conditions affecting ISFSI.

TVA, as the general licensee of the ISFSI, is required by §72.216 to make initial and written reports in accordance with §72.74 and §72.75. Operations is responsible for making the reportability determinations for §72.74 and §72.75 reports. For any event, condition, or issue having the potential for being reportable, contact Site Licensing for consultation and concurrence on the reportability determination. In no event shall the lack of licensing concurrence result in a failure to meet specified reporting timeframes.

Operations is responsible for making the immediate notifiCation to NRC in accordance with §72.74. Operations is responsible for making the immediate, 4-hour, and 24-hour notifications to NRC in accordance with §72.75. Licensing is responsible for developing (with input from affected organizations) and submitting the written reports required by §72.75.

3) Reporting requirements for events or conditions affecting the physical protection of the licensed nuclear plant specified in §73. 71 are contained in NSDP-1, Safeguards Event Reporting Guidelines. Responsibilities for reportability determinations and immediate notification requirements are assigned to Site Nuclear Security and Corporate Nuclear Security. Licensing is responsible for developing (with input from affected organizations) and submitting the written reports required by §73.71.

NPG Standard Regulatory Reporting Requirements NPG-SPP-03.5 Programs and Rev.0008 Processes Page 20 of99 Appendix A (Page 2 of 15)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 3.0 REQUIREMENTS NOTES

1) Internal management notification requirements for plant events are found in Appendix D. The Operations Shift Manager is responsible for notifying Site Operations Management and the Duty Plant Manager. The Duty Plant Manager is responsible for making the remaining internal management notifications.

2) NRC NUREG-1022, Revision 3 and subsequent revisions should be used as guidance for determining reportability of plant events pursuant to §50. 72 and §50. 73.

A text searchable copy of NUREG-1022 is maintained on the TVA NPG Nuclear licensing Webpage.

3.1 Immediate Notification - NRC TVA is required by §50.72 to notify NRC immediately if certain types of events occur. This appendix contains the types of events and the allotted time in which NRC must be notifted.

(Refer to Form NPG-SPP-03.5-1 or NRC Form 361 ). Operations is responsible for making the reportability determinations for §50.72 and §50.73 reports. For any event, condition, or issue having the potential for being reportable, contact Site Licensing for consultation and concurrence on the reportability determination. In no event shall the lack of licensing concurrence result in a failure to meet specified reporting timeframes. Operations is responsible for making the immediate notification to NRC in accordance with §50.72.

Notification is via the Emergency Notification System. If the Emergency Notification System is not operative, use a telephone, telegraph, mailgram, or facsimile.

NOTE The NRC Event NotifiCation Worksheet may be used in preparing for notifying the NRC. This Worksheet may be obtained directly from the NRC website (www.nrc.gov) under Form 361 , or TVA NPG Form NPG-SPP-03.5-1 may be used.

A. The Immediate Notification Criteria of §50.72 is divided into 1-hour, 4-hour, and 8- hour phone calls. Notify the NRC Operations Center within the applicable time limit for any item which is identified in the Immediate Notification Criteria.

B. The following criteria require 1-hour notification:

1. (Technlcal Specifications) - Safety Limits as defined by the Technical Specifications which have been violated.

NPG Standard Regulatory Reporting Requirements NPG-SPP-03.5 Programs and Rev.0008 Processes Page 21 of99 Appendix A (Page 3 of 15)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 3.1 Immediate Notification - NRC (continued)

2. §50.72 (a)(1 )(i)-The declaration of any of the Emergency classes specified in the licensee's approved Emergency Plan.

3. §50.72(b).(1))- Any deviation from the plant's Technical Specifications authorized pursuant to §50.54(x).

4. 10 CFR 73, Appendix G, paragraph 1- Safeguards Events. The requirements of

§73.71, Reporting of Safeguard Events, are also applicable. Refer to NSDP-1 ,

PSafeguards Event Reporting Guidelines," for additional information.

a. Any event in which there is reason to believe that a person has committed or caused, or attempted to commit or cause, or has made a credible threat to commit or cause:

(1) A theft or unlawful diversion of special nuclear material; or (2) Significant physical damage to a power reactor or any facility possessing SSNM or its equipment or carrier equipment transporting nuclear fuel or spent nuclear fuel, or to the nuclear fuel or spent' nuclear fuel a facility or carrier possesses; or (3) Interruption of normal operation of a licensed nuclear power reactor through the unauthorized use of or tampering with its machinery, components, or controls inCluding the security system. [Note: a Confirmed Cyber Attack at any NPG site is reported to the NRC in accordance with the requirements of 10 CFR 73, Appendix G. Review the 'Incident Categorization' section in NPG-SPP-12.8.8.]

b. An actual entry of an unauthorized person into a protected area, material access area, controlled access area, vital area, or transport.

c. Any failure, degradation, or the discovered vulnerability in a safeguard system that could allow unauthorized or undetected access to a protected area, material access area, controlled access area, vital area, or transport for which compensatory measures have not been employed.

NPG Standard Regulatory Reporting Requirements NPG*SPP-03.5 Programs and Rev.0008 Processes Page22 of99 Appendix A (Page 4 of 15)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 3.1 Immediate Notification -NRC (continued)

d. The actual or attempted introduction of contraband into a protected area, material access area, vital area, or transport.

C. The following criteria require 4-hour notiftcation:

1. §50.72(b)(2)(i)- The initiation of any nuclear plant shutdown required by the plant's Technical Specifications.

2. §50.72(b)(2)(iv)(A)- Any event that results or should have resulted in Emergency Core Cooling System (ECCS) discharge into the reactor coolant system as a result of a valid signal except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

3. §50.72(b)(2)(iv)(B)- Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

NOTES

1) NPG-SPP-05.14 provides additional instructions regarding addressing and informally communicating events to outside agencies involving radiological spills and leaks.

2) Routine or day-to-day communications between TVA organizations and state agencies typically do not constitute a formal notification to other government agencies that would require a report in accordance with §50.72(b)(2)(xi).

4. §50.72(b)(2)(xi)- Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Such an event may include an onsite fatality or inadvertent release of radioactive contaminated materials.

D. The following criteria require 8-hour notification:

NOTE With the exception of "Events or Conditions that Could Have Prevented Fulfillment of a Safety Function," ENS notifications are required for any event that occurred within three years of discovery, even if the event was not ongoing at the time of discovery.

NPG Standard Regulatory Reporting Requirements NPG*SPP-03.5 Programs and Rev.OOOS Processes Page 23 of99 Appendix A (Page 5 of 15)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 3.1 Immediate Notification *NRC (continued)

1. §50.72(bX3}(iiXA)- Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.

2. §50. 72(b)(3Xii)(B)- Any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.

3. §50.72(bX3)(iv)(A)- Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) [see list below], except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

a. The systems to which the requirements of paragraph §50.72(b)(3XivXA) apply are:

NOTE Actuation of the RPS when the reactor is critical is also reportable under §50.72(b)(2)(iv)(B) above.

(1) Reactor protection system (RPS) including: reactor scram or reactor trip.

(2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs).

(3) Emergency core cooling systems (ECCS) for pressurized water reactors (PWRs) including: high-head, intermediate-head, and low-head injection systems and the low pressure injection function of residual (decay) heat removal systems.

(4) ECCS for boiling water reactors (BWRs) including: core spray systems; high-pressure coolant injection system; low pressure injection function of the residual heat removal system.

(5) BWR reactor core isolation cooling system.

(6) PWR auxiliary or emergency feedwater system.

(7) Containment heat removal and depressurization systems, inclUding containment spray and fan cooler systems.

NPG Standard Regulatory Reporting Requirements NPG-SPP-03.5 Programs and Rev.0008 Processes P~ge 24 of99 Appendix A (Page 6 of 15)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 3.1 Immediate Notification -NRC (continued)

(8) Emergency ac electrical power systems, including: Emergency diesel generators (EOGs).

NOTE For systems within scope, the inadvertent TS inoperability of a system in a required mode of applicability constitutes an event or condition for which there is no longer reasonable expectation that equipment can fulfill its safety function. Therefore, such events or conditions are reportable as an "Event or Condition that Could Have Prevented Fulfillment of a Safety Function."

4. §50.72(b)(3)(v)- Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to:

(A) Shut down the reactor and maintain it in a safe shutdown condition; (B) Remove residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident.

NOTE According to §50.72 (b)(3)(vi) events covered by §50.72(b)(3)(v) may include one or more procedural errors, equipment failures, and/or discovery of design, analysis, fabrication, construction, and/or procedural inadequacies. However, individual component failures need not be reported pursuant this paragraph if redundant equipment in the same system was operable and available to perform the required safety function.

5. §50.72(b)(3)(xii) - Any event requiring the transport of a radioactively contaminated person to an offsite medical facility for treatment.

6. §50.72(b)(3)(xiii)- Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of control room indication, emergency notification system, or offsite notification system).

NPG Standard Regulatory Reporting Requirements NPG.SPP-03.5 Programs and Rev.0008 Processes Page25 of99 Appendix A (Page 7 of 15)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 3.1 Immediate Notification - NRC (continued)

E. Follow-up Notification (§50.72(c))

With respect to the telephone notifications made under paragraphs (a) and (b) [§50.72 (a) and §50.72 (b), respectively] of this section [§50.72], in addition to making the required initial notification, during the course of the event:

1. Immediately report:

(i) Any further degradation in the level of safety of the plant or other worsening plant conditions including those that require the declaration of the Emergency Classes, if such a declaration has not been previously made; or (ii) Any change from one Emergency Class to another, or (iii) A termination of the Emergency Class.

(1) Immediately report:

(i) The results of ensuing evaluations or assessments of plant conditions, (ii) The effectiveness of response or protective measures taken, and (iii) Information related to plant behavior that is not understood.

(2) Maintain an open, continuous communication channel with the NRC Operations Center upon request by the NRC.

3.2 Twenty-Four Hour Notification - NRC Any violation of the requirement contained in specific operating license conditions, shall be reported to NRC in accordance with the license condition.

3.3 Two-Day Notification - NRC

§50.9(b) - The NRC shall be notified of incomplete or inaccurate information which contains significant implications for the public health and safety or common defense and security.

Notification shall be provided to the administrator of the appropriate regional office within two working days of identifying the information. licensing is responsible for determining reportability (with input from affected organizations) and notifying NRC in accordance with

§50.9.

Rod Group Alignment Limits 3 . 1. 5 3.1 REACTIVITY CONTROL SYSTEMS 3.1 . 5 Rod Group Alignment Limits APP LICABILITY : MODES 1 and 2 .

ACTIONS CONDITION REQUIRED ACT I ON COMPLETI ON TIME A. One or mo re rod(s) A. 1. 1 Verify SDM is ~ 1. 6% 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months untrippable. -k/k.

OR A. 1 . 2 Initiate boration t o 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore SDM to within limit.

AND A. 2 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months B. One r od not wi thin B. l Restore rod t o within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months alignment limits . alignment l imits .

B. 2.1.1 Verify SOM is 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

~ 1.6% -k/k .

OR (con tinued)

Wat ts Bar-Unit 1 3.1 - 8

Rod Group Alignment Limits l 3~1 . 5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) 8.2.1 . 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 8 . 2.2 Reduce THERMAL POWER 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to < 75 % RTP.

AND 8 .2 . 3 Verify SDM is Once per

~ 1. 6 % - k/k 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND 8 . 2. 4 Perform SR 3.2 . 1.1. 72 hou r s 8.2.5 Perform SR 3.2 . 2.1. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months AND 8 . 2. 6 Re-evaluate safety 5 days analyses and confirm results remain valid for duration of operation under these conditions .

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition B not met .

(continued)

Watts Bar-Unit 1 3 . 1- 9

Rod Group Alignment Limits

' 3 .1. 5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. More than one rod not D.1.1 Verify SDM is 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months within alignment ~ 1.6% -k/k.

limit.

OR D.1.2 Initiate boration to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months resto re required SDM to within limi t.

AND D.2 Be in MODE 3 . 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Watts Bar-Unit 1 3 . 1 - 10

Questions Asked During Administration ofWBN NRC Written Exam 10 .. 30-2013 Question # )S~

Question Asked:

TH-e- P...cw~(OfJ 0~ (He fPj ,~ A t/!JrJI. jl./oTlftl!AfloiV Voc *ro (1/r- f1-4:vv1 3c~NZ. C~tTJCf>.c . WH(:'N SrcctFtck<-Y Pro"(..r~ ~---- ~ (ft.~

No-nhtkllor-J .fv~ fA£-;- (;') c{ Qu6'flc.,v-J r ~ Tl+c 'KPS Pts&(JN cso To (fNr Bcw RoRntAJfJ /+;. ~ 1 /lou( MvfiCA77or/ As f ~\J' ~ ?

Applicant Name: <T, u~17"' G~<-~~td- 1_~\~~~--tt_---+-=::::::::=:=------ Time: 0 9o 5"'

Printed ~

Response Provided:

+L-

~

I*

CfUe$-l'tO'f\

+o ...,..,..

_j

"',....ell .*

,,_A+ 0<:):)3 ir~ c:a.ecd'dor>>ce I

'---:=z>~--------

Proctor Signature: _ ___....0::.......::-:

Applicant S i g n a t u r e : - - - - - - - - - - - - - - Time: _ __

. Questions Asked During Administration of WBN NRC Written Exam 10-30-2013 Question# g (.

Question Asked:

r:;r ~ q u,-en plal7+ ccncli>o/ls, Can we ass~ -Me-+ we.

av-e. +a 1:::, ~:! +he "il"' re ~ a .,_-t.,.., s IF ~ a f'P'""P<'"--k CcX\chllC-\\ .3<\d '"'~\1-._ 0o+ ?n>qve s=sed -+c:, CiL SUo.SE?<ft)42.(\+

ec-nJmvn. ~ree of ~e ~vr ~dch()(\_s. h6v-e_ a ~t{Dut~--0 fc\\CIN -to p\o. U2- ~ en \1-- '\..) t'J\ t> <k_ '3 Applicant Name: \-\en~s-~

Printed

~~~ Signature Time: 1 I 50 Response Provided:

.f)...,o qlte.st>OI'\ a~ w r t' i -}C#) <j' ve-.. 4~

An.s o.)(3 r prov) dod. ~n ~r~ ~on (do no-+ GtSSWVt-o

-t>>,of no-f e >< p Ii c...i t ly s-fJ-oel Qt"l'f-t ""-*" ~ IS 4ho q~s-tc/\ ).

j1\

Proctor Signature:

~~------------

/;:) #

Applicant Signature: -~......:.

{. . u. .;. . :. . ._ -=-.:. ._" : .= =-: - . :. .__ _ _ _ Time: JZ2Z

ML13312A425

Text

Navigation menu

Search