ML13339A665

Initial Exam 2013-301 Draft SRO Written Exam

ML13339A665
Person / Time
Site:	Robinson
Issue date:	11/30/2013
From:	NRC/RGN-II
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50-261/13-301
Download: ML13339A665 (180)

Text

F DUKE 1

ENERGY H .B. Robinson ILC-13 NRC Licensing Exam Written Exam Book 2 of 2 AIQN /

NRC Exam

76. 008 AG2.1.7 SRO 001 Given the following plant conditions:

-Plant is at 100% RTP

-PCV-455C, PZR PORV, indicates open

-PT-444, CH I PRZR PRESS, is 2100 psig and lowering

-PT-445, CH II PRZR PRESS, is 2100 psig and lowering

-The OAC manually shuts PCV-455C

-I&C determines that the PCV-455C cant be controlled automatically AOP-019, MALFUNCTION OF RCS PRESSURE CONTROL AOP-025, RTGB INSTRUMENT FAILURE Which ONE (1) of the following completes the statements below?

The crew has entered (1) . lAW ITS, PCV-455C (2) OPERABLE.

A. (1)AOP-019 (2) is NOT B. (1)AOP-025 (2) is NOT C (1)AOP-019 (2) is D. (1) AOP-025 (2)is Tuesday, June 11,20132:10:25 PM 148

NRC Exam The correct answer is C.

A) Incorrect. AOP-019 is the correct procedure because the valve failed open.

PCV-455C is operable because the valve can still be manually operated. Plausible if the student thinks that the valve is inoperable because it had to be manually shut or because automatic control is gone. The valve can still be operable without automatic control.

B) Incorrect. AOP-019 is the correct procedure because the valve failed open.

Plausible because if instrument that inputs into PCV-455C caused it to open, AOP-025 would be the correct procedure to enter. PCV-455C is operable because the valve can still me manually operated. Plausible if the student thinks that the valve is inoperable because it had to be manually shut or because automatic control is gone. The valve can still be operable without automatic control.

C) Correct. AOP-019 is the correct procedure because the valve failed open.

PCV-455C is operable because the valve can still me manually opened and closed.

D) Incorrect, AOP-019 is the correct procedure because the valve failed open.

Plausible because if instrument that inputs into PCV-455C caused it to open, AOP-025 would be the correct procedure to enter. PCV-455C is operable because the valve can still me manually opened and closed.

Question: 76 Tier/Group: 1/1 K/A Importance Rating: RO 4.4 SRO 4.7 K/A: 008 Pressurizer Vapor Space Accident G2.1 .7: Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

References:

Sim/Plant design, AOP-019, LCD 3.4.11 Proposed references to be provided to applicants during the Exam: None Learning Obj: Objective 16 of PZR/PRT, Objective 1 AOP-019 Question Source: New Question History: New Question Cognitive Level: High 10 CFR part 55: (CFR 41.5 /43.5 /45.12 /45.13)

Meets the K/A because the student needs to evaluate plant conditions and determine if the PORV is inoperable. Determining if the PORV is inoperable makes this an SRO question because the SRO must have knowledge of TS bases that is required to analyze TS required actions and terminology.

Thursday, June 13, 2013 10:35:42AM 149

Rev. 18 AOP-019 MALFUNCTION OF RCS PRESSURE CONTROL Page 2 of 20 Purpose and Entry Conditions (Page 1 of 1)

1. PURPOSE This procedure provides instructions in the event RCS pressure is higher OR lower than required for current plant conditions.

This procedure is applicable in Modes 1. 2. and 3.

2. ENTRY CONDITIONS This procedure may be entered when RCS pressure deviates from the desired contro2 band due to a fau]t in pressure control components (AOP 025 covers Instrument Failure)

- END -

Rev. 18 AOP 025 RTCB INSTRUMENT FAILURE Page 2 of 39 Purpose & Entry Conditions (Page 1 of 1)

1. PURPOSE This procedure provides instructions for failure of process variable transmitters which provide input to RTGE controllers.

IF an applicable transmitter fails while the controller is operating in manual OR is being fed from an alternate channel, THEN entry to this procedure is NQ required.

This procedure is applicable in Modes 1, 2, 3, and 4.

2. ENTRY CONDITIONS Failure of any process variable transmitter which affects automatic operation of RTGB controllers with the following exceptions:

• FT-605, RHR Flow

• LT-115, VCT Level

• LT-112, VCT Level

• PR NIS (NI-41, 42, 43, & 44)

• S/C Narrow Range Level

- END -

Pressurizer PORVs B 3.4.11 BASES BACKGROUND the PORVs minimize c:hallenges to the pressurizer safety (continued) valves and also may be used for low temperature overpressure protection (LTOP). See LCO 3.4.12, Low Temperature Overpressure Protection (LTOP) System.

APPLICABLE The PORVs and their respective block valves are provided for SAFETY ANALYSES plant operational flexibility and for limiting the number of challenges to the pressurizer safety valves. Operation of the PORVs is not explicitly considered to be a safety-related function for overpressure protection of the reactor coolant pressure boundary (RCPB) at normal operating temperature and pressure. Plant operators employ the PORVs to depressurize the RCS in response to certain plant transients if normal pressurizer spray is not available.

Operation of the PORVs in MODES 1, 2, and 3 is not classified as a safety-related function (i.e., one on which the results and conclusions of the safety analysis are based and that invokes the highest level of quality and construction). Also, an inadvertent opening of a PORV or a safety valve has been analyzed in the UFSAR (Ref. 1) as an anticipated operational occurrence (AOO) with acceptable consequences. For these reasons, the PORVs are not classified as safety related components.

Generic Letter 90-06 (Ref. 2) provided the NRCs resolution of PORV and block valve reliability concerns (Generic Issue 70). and set forth certain requirements to enhance safety.

The pressurizer PORVs have no safety function and are not assumed to function during any UFSAR design basis accident or transient analysis. However, inclusion of the pressurizer PORVs is consistent with the guidance provided in Generic Letter 90-06. Therefore, they are being retained in Technical Specifications.

LCO The LCO requires the PORVs and their associated block valves to be OPERABLE for manual operation.

An OPERABLE PORV is required to be capable of manually opening and closing and not experiencing excessive seat leakage Automatic control functions are not required for OPERABILITY of the PORV5.

(continued)

HBRSEP Unit No. 2 3.4-54 Revision No. 18

NRC Exam

77. 054 AG2.4.35 SRO 001 Given the following plant conditions:

-The plant is at 98% RTP

-AOP-O1O is entered due a leak in FW Heater4B

-OPS Manager has chosen to repair the leak on line AOP-O1O, MAIN FEEDWATER/CONDENSATE MALFUNCTION OP-407, HEATER DRAINS AND VENTS Which ONE (1) of the following completes the statements below?

The CRS will direct the OAO to use (1) for the specific steps to remove the required FW Heaters from service. The crew (2) have to reduce power for this evolution.

A. (1)OP-407 (2) will NOT B. (1)AOP-O1O (2) will NOT C (1) OP-407 (2) will D. (1)AOP-O1O (2) will Tuesday, June 11,20132:10:25 PM 150

NRC Exam The correct answer is C A) Incorrect. OP407 is correct. AOP-010 will direct you to OP-407 to remove the necessary string of heaters from service, the OAO will perform these actions. Will not have to reduce power is incorrect. Plausible because 98% power is a common power that we reduce to when performing evolutions dealing with AFW flow. The crew will have to reduce power to 659 MW.

B) Incorrect. AOP-010 is incorrect. There are several AOPs that direct performing specific actions either in the main body or in an attachment. This is not one of them.

AOP-010 directs you to OP-407 to remove the heaters. Will not have to reduce power is incorrect. Plausible because 98% power is a common power that we reduce to when performing evolutions dealing with AFW flow. The crew will have to reduce power to 659 MW.

C) Correct. OP-407 is correct. There are several AOPs that direct performing specific actions either in the main body or in an attachment. This is not one of them. AOP-010 directs you to OP-407 to remove the heaters. The crew will have to reduce power to 659 MW.

D) Incorrect. AOP-01 0 is incorrect. There are several AOPs that direct performing specific actions either in the main body or in an attachment. This is not one of them.

AOP-010 directs you to OP-407 to remove the heaters. The crew will have to reduce power to 659 MW is correct.

Question: 77 Tier/Group: 1/1 K/A Importance Rating: RO 3.8 SRO 4.0 K/A: 054 Loss of Main Feedwater G 2.4.35: Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.

Reference(s): AOP-010, OP-407 Proposed References to be provided to applicants during examination: None Learning Objective: Objective 4 of AOP010 lesson plan Question Source: New Question History:

Question Cognitive Level: Low 10 CFR Part 55 Content: (CFR: 41.10 /43.5 /45.13)

Comments:

This question meets the K/A because the candidate has to determine which feedwater heaters need to be removed (operational implication), and what to send the OAO to look at(Local AO tasks). It is SRO level because the candidate must determine which procedure to use to remove the feed water heaters.

Thursday, June 13, 2013 10:44:41 AM 151

Rev. 30 AOP-O1O MAIN FEEDWATER/CONDENSATE MALFUNCTION Page 3 of 27 ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED NOTE Steps 1 is immediate action step.

1. Check Feedwater Regulating Perform the following:

Valves - OPERATING PROPERLY (MANUAL OR AUTO): a. Verify FRV for affected SIG(s) in manual control.

• FCV-478 FRV A

• FCV-488 FRV B b. Attempt to stabilize S/G

• FCV-498, FRV C level using FRV and/or FRV Bypass Valves by matching steam flow with feed flow.

c. Stop any load change in progress.

d. Restore affected S/G level to between 39% and 52%.

e. IF unable to control S/C level, THEN trip the Reactor AND Go To EOP-E-O, REACTOR TRIP or SAFETY INJECTION.

f. Go To Step 37

2. Check Reactor Trip Setpoint - a Reactor Trip Setpoint is BEING APPROACHED approached THEN trip the Reactor and Go To EOP E 0 REACTOR TRIP or SAFETY INJECTION Go To Step 4.

3. Trip The Reactor And Go To EOP-E-O. REACTOR TRIP or SAFETY INJECTION.

4 Make PA Announcement For Procedure Entry

Rev. 30 AOP-O10 MAIN FEEDWATER/CONDENSATE MALFUNCTION Page 4 of 27 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED J__STEP_H

5. Go To The Appropriate Step from The Table Below:

EVENT STEP Main Feed Pump Trip Step 6 Condensate Step 10 Feed Pump Trip Condensate Pump Trip Step 48 Without MFP Trip Heater Drain Pump Trip Step 15 Pipe Break or Leak Step 21 Feedwater Heater Leak HCV-1459 Failed Open Step 35 Other Step 24

6. Check Reactor Power - LESS THAN Trip the Reactor and Go To 70% EOP-E-O, REACTOR TRIP or SAFETY INJECTION.

7. Check Reactor Power - GREATER Go To Step 13.

THAN 60%

Rev. 30 AOP-OlO MAIN FEEDWATER/CONDENSATE MALFUNCTION Page 12 of 27 j STEP H ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE Rapid power reductions may result in the axial flux difference exceeding the operating band values and require a power reduction to less than 50% to comply with ITS 3.2.3 Condition C.

• 21. Reduce Turbine Load At 1%/MIN To 5%/MIN To Match Feedwater Flow AND Steam Flows As Follows:

a. Verify Rods in AUTOMATIC a. Place ROD BANK SELECTOR to M (Manual) AND manually position control rods to maintain Tavg within

-1.5 to +1.5°F of Tref

b. Check Turbine Control Mode - b. Momentarily depress the GV AUTOMATIC (down) button on the EH Control Panel as needed to

1) Depress the IMP IN lower Turbine Load Pushbutton

2) Set The Desired Load In The SETTER

3) Set The Desired Load Rate

4) Depress the GO Pushbutton or the HOLD Pushbutton as Necessary to Reduce Turbine Load

c. Borate Per OP-30l, RCS Boration Quick Checklist, as necessary to maintain AFD within the operating band

d. Check Feed Flow matched with d. Continue with power reduction.

Steam Flow WHEN flows are matched perform Step 2l.e.

Go To Step 22.

e. Stop the power reduction

Rev. 30 AOP-010 MAIN FEEDWATER/CONDENSATE MALFUNCTION Page 13 of 27 STEP_H ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE OP-407 provides instructions for removing Feed Water Heaters from service.

22. Perform the following:

a. Determine location of the leak

• Visual observation of an external leak

• Feed Water Heater level alarms

• Feed Water Heater Normal and Alternate drain valve positions

• Feed Water Heater #1 & #2 Emergency Dump valve positions

b. Isolate the leak. b. Determine appropriate strategy:

1) Consult with Operations Management to choose appropriate strategy

• Repair on line

• Shutdown to repair

• Trip Unit

2) IF the leak will be repaired on line, THEN observe the NQTh prior to Step 38 And Go To Step 38.

3) IF the Unit will be shutdown, THEN Go To GP-006-Ol, Normal Plant Shutdown From Power Operation To Hot Shutdown.

4) IF the Unit will be tripped, H)N trip the Reactor and Go To EOP-E-0, REACTOR TRIP or SAFETY INJECTION.

c. Notify the SM of leak location and method used to isolate leak.

23. Observe The NOTE Prior To Step 38 And Go To Step 38

NRC Exam

78. 055 EA2.06 SRO 001 Given the following plant conditions:

-A Station Blackout has occurred from 100% RTP

-The crew is performing actions of EPP-1, LOSS OF ALL AC POWER

-Immediate actions are complete

-A Charging pump is running

-Offsite power is available to re-energize the SUT

-APP-009-B5, MAIN TRANSF PHASE TRIP, alarm is in OP-603, ELECTRICAL DISTRIBUTION OP-603-3, RESETTING HIGH IMPEDANCE FAULT TRIPS 86P, GENERATOR LOCKOUT RELAY, PRIMARY 86BU, GENERATOR LOCKOUT RELAY, BACKUP Which ONE (1) of the following completes the statement below?

lAW EPP-1, the CRS will direct the use of OP- (1) to restore power. APP-009-B5 is an indication that the (2) lockout needs to be reset.

A (1) 603 (2) 86P B. (1) 603-3 (2) 86P C. (1)603 (2) 86 BU D. (1) 603-3 (2) 86 BU Tuesday, June 11,20132:10:25 PM 152

NRC Exam The correct answer is A A) Correct. OP-603 will provide the guidance to restore power electrical power. 86P will need to be reset based of APP-009-B5. MAIN TRANSF PHASE TRIP feeds into the 86P which in turn, trips the turbine.

B) Incorrect. OP-603-3 is incorrect. Plausible since this procedure does give guidance on restoring power, however, it is restoring power to the DS Bus which has already happened. This is indicated by the A Charging pump running. 86P will need to be reset based of APP-009-B5. MAIN TRANSF PHASE TRIP feeds into the 86P which in turn, trips the turbine.

C) Incorrect. OP-603 will provide the guidance to restore power electrical power. 86BU is incorrect. Plausible since this is one of the two lockouts that need to be reset to restore power, however, this alarm is not associated with 86BU.

D) Incorrect. OP-603-3 is incorrect. Plausible since this procedure does give guidance on restoring power, however, it is restoring power to the DS Bus which has already happened. This is indicated by the A Charging pump running. 86BU is incorrect.

Plausible since this is one of the two lockouts that need to be reset to restore power, however, this alarm is not associated with 86BU.

Question: 78 Tier/Group: 1/1 K/A Importance Rating: RO 3.7 SRO 4.1 K/A: 055 Loss of Offsite and Onsite Power (Station Blackout)

EA2: Ability to determine or interpret the following as they apply to a Station Blackout:

EA2.06: Faults and lockouts that must be cleared prior to re- energizing buses Reference(s): Sim/Plant design, EPP-1, OP-603 Proposed References to be provided to applicants during examination: None Learning Objective: Objective 4 of AOP-010 lesson plan Question Source: RNP Bank Question History: 2009 NRC Exam, Changed format and changed the DS powered pump that was running Question Cognitive Level: Low 10 CFR Part 55 Content: (CFR: 43.5 / 45.13)

Comments:

This question meets the K/A becuase it requires the student to interpret which lockout needs to be reset in order to restore power from offsite. Since the student has to determine which procedure to use in order to restore power, this makes it an SRO question.

Thursday, June 13, 2013 10:47:28 AM 153

çL APP-009-B5 ALARM MAIN TRANSF FAULT TRIP AUTOMATIC ACTIONS

1. Generator Lockout (See Attachment 1) (2/3 logic)

CAUSE

1. Internal electrical fault in Main Transformer A, B, or C (Fault Pressure)

2. APP-045 Annunciator Panel TEST pushbutton.

OBSERVATIONS

1. Exciter Field Breaker Indication

2. Generator Output Breaker Position (52/8 and 52/9)

ACTIONS CK (I

1. IF the reactor has tripped, THEN REFER TO EOP Network.

2. IF the Turbine has tripped while below 40% power, THEN REFER TO AOP-007.

3. IF a fault has occurred on one of the Main Transformers, THEN CONTACT the Load Dispatcher for repairs. (SOER 10-1)

DEVICE/SETPOINTS

1. 63FP / Sensitivity Selection Position Zero (0)

POSSIBLE PLANT EFFECTS

1. Reactor Trip

2. Turbine Trip REFERENCES

1. ITSSR3.8.1.16

2. EOP Network

3. AOP-007, Turbine Trip Below P-8

4. CWD B-I 90628, Sheet 940, Cable P

5. SOER 10-1, Large Power Transformer Reliability, recommendation 6 APP-009 Rev. 54 Page 20

ATTACHMENT 1 LOCKOUT AUTO ACTIONS Generator Lockout ççC\

1. 4KV Fast Bus Transfer

2. Trips Generator Output OCBs 52/8 and 52/9

3. Trips Exciter Field Breaker

4. Actuates Turbine Trip (20/AST and 20/ET)

Startup Transformer Lockout

1. Trips 4KV Breakers 52/12 and 52/17

2. Opens 115KV Motor Operated Disconnect

3. Trips UNIT NO.2 START-UP TRANSF WEST BUS 115 KV OCB

4. Trips UNIT NO.2 START-UP TRANSF EAST BUS 115 KV OCB GENERATOR LOCKOUT RELAY INPUT SIGNALS 86P 86BU BOTH GEN GROUND TRIP GEN NEG SEQ/OCB BU TRIP OCB 52-8 FAILED TO OPEN (APP-009-B4) (APP-009-C2) (APP-009-D5)

MAIN TRANSF FAULT TRIP Exciter Field Breaker Tripped OCB 52-9 FAILED TO OPEN (APP-009-B5) (Loss of field) (APP-009-D6)

AUX TRANSF OVLD/PHASE Unit Differential (87/GT)

A TRIP (APP-009-A6) (APP-009-A5)

AUX TRANSF FAULT TRIP Turbine Trip (Stop valves or (APP-009-B6) Governor valves)

Voltage Control Overcurrent Relay (51-27)

Generator Differential (87/G)

Turbine Trip (63AST)

APP-009 Rev. 54 Page 60 of 60

Rev. 51 EPP-1 LOSS OF ALL AC POWER Page 11 of 66 STEP_H ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

• 16. Perform The Following:

a. Request assistance from Maintenance in restoring AC Power

b. Dispatch an Operator to complete EPP-22, ENERGIZING PLANT EQUIPMENT USING DEDICATED SHUTDOWN DIESEL GENERATOR

c. Contact the Load Dispatcher to determine when Offsite Power will become available

d. Check Off-Site Power (Grid) - d. WHEN Off-Site Power is AVAILABLE available, IIIN perform Step 16.e Observe the NOTE prior to Step 17 and Go To Step 17.

e. Check Startup Transformer - e. j the problem is associated AVAILABLE with the Startup Transformer perform the following:

• personnel are available, THEN restore power using backfeed in EPP-25, ENERGIZING SUPPLEMENTAL PLANT EQUIPMENT USING THE DSDG.

OR

• IF ERO facilities are activated, THEN request a mission to restore power using backfeed in EPP-25, ENERGI Z ING SUPPLEMENTAL PLANT EQUIPMENT USING THE DSDG.

Observe the !Q1 prior to Step 17 and Go To Step 17.

f. Restore normal power using OP-6O3

RESETTING HIGH IMPEDANCE FAULT TRIPS I flCoçr&I OP-603-3 Page 3 of 44 TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE 4 2.0 SCOPE 4 3.0 PRECAUTIONS AND LIMITATIONS 4 4.0 PREREQUISITES 5 5.0 INSTRUCTIONS 6 5.1 Administrative Tasks 6 5.2 480 VAC Emergency Bus E-1 8 5.3 480 VAC Emergency Bus E-2 10 5.4 MCC-5 12 5.5 MCC-6 15 5.6 MCC-9 17 5.7 MCC-10 19 5.8 Instrument Bus I 21 5.9 Instrument Bus 2 24 5.10 Instrument Bus 3 27 5.11 Instrument Bus 4 30 5.12 Instrument Bus 7A& B 33 5.13 Instrument Bus 9A 36 5.14 125 VDC MCC-A and Distribution Panel A 39 5.15 125 VDC MCC-B and Distribution Panel B 41 6.0 RECORDS 43

7.0 REFERENCES

43

CUCCC RESETTING HIGH IMPEDANCE FAULT TRIPS (OP-603-3 Page4of44 1.0 PURPOSE

1. To provide instructions for safely restoring an electrical bus following a high impedance fault trip.

2.0 SCOPE

1. To provide instructions for inspecting busses following a fault trip, determining if it is safe to energize that bus, and remove nonessential loads prior to energizing. Loads required for safe shutdown are loaded after bus has been re-energized. The remaining loads will remain de-energized until full troubleshooting and repair are completed.

3.0 PRECAUTIONS AND LIMITATIONS

1. Resetting of the Generator Lockout Relays 86P and/or 86BU should be done only after all applicable Generator Lockout Signals have been reset and/or removed. Attempting to reset either 86P or 86BU with a Lockout signal present will reactivate the fast bus transfer sequence and may cause a loss of 4KV busses. [Section 7.1.1 Commitment 3]

(CR 390095; OPS-NGGC-1 000, Resetting Protective Devices)

2. Protection devices (breakers, fuses, bistables, overloads, lockouts, and so forth) which have tripped should only be reset, with CRS approval, under the following conditions:

a. The cause of the trip has been identified and corrected.

b. Where no evidence of abnormality is present, it is permissible to restore the protective device ONE TIME (see below on MOV5).

Major control schemes (such as the 86 Generator Lockout scheme) SHALL NOT be reset without understanding the cause of the event AND required corrective action(s), unless it is determined that the action is required to support public health and safety concerns.

3. The SM may approve additional protective device resetting after consultation with engineering. Reenergizing busses without a proper investigation can lead to equipment damage, personnel injury or death.

NRC Exam

79. 058 AG2.1.19 SRO 001 Given the following plant conditions:

-The reactor is at 100% RTP

-B Charging pump is running in Manual

-C Charging pump is running in AUTO

-A reactor trip occurs

-The following indications are seen:

-See next page for references Which ONE (1) of the following completes the statements below?

Based off the indications above, C Charging pump is (1) . The crew will transition from EOP-E-0 to (2) to EPP-27.

A (1) running (2) EOP-ES-0.1 B. (1)tripped (2) EOP-ES-0.1 C. (1) running (2) EPP-7 D. (1) tripped (2) EPP-7 The correct answer is A A) Correct. C Charging pump is still running, however, it has lost control power. The crew will transition from EOP-E-0 to EOP-ES-0.1 to EPP-27.

B) Incorrect. C Charging pump is still running, however, it has lost control power.

Plausible since the indication for C Charging pump goes away during a loss of B DC. The crew will transition from EOP-E-O to EOP-ES-0.1 to EPP-27.

C) Incorrect. C Charging pump is still running, however, it has lost control power.

EPP-7 to EPP-27 is incorrect. Plausible since if the students believe that a loss of DC gives you an SI, you would go from EOP-E-0 to EPP-7 to EPP-27.

D) Incorrect. C Charging pump is still running, however, it has lost control power.

Plausible since the indication for C Charging pump goes away during a loss of B DC. EPP-7 to EPP-27 is incorrect. Plausible since if the students believe that a loss of DC gives you an SI, you would go from EOP-E-0 to EPP-7 to EPP-27.

Tuesday, June 11,20132:10:25 PM 154

NRC Exam CHARGING PUMP C SELECT FIStIC. KEY OR TURII-Oll CODE 1W  : I jJ TEMP1JTJ PLOT LIURARJ GROUP LIBWR REPORT j 35 HOIJR 21 DAYS]

TREIID DTSPLHY DC (Ilili/I1AX) 1:a4 I

øi1 L:27 ijiQo 4:5q t4:

PoinT to DESCRIPTIOI1 TREIID CURREIIT LOU HIGH LIF1 I.PLUE GWIL HLRRFI ALRl1 f5PU355?20 DC 11CC-Il VOET 132.10 132.10 OK 123.00 IIO.OO APU30290 DC [ICC-fl CURREUT -

-6. 6 - -6. 6 OK 0. 0 1000. 0 HPU3023R DC 11CC-B UOLT nan nan ROER 123.00 itO.OO 0.0 F1FAST LEFT F2=LEFT F3-FIEII1J F1TEt1PLHTE CANC F5.RIOHT F.FRST RIGHT CO11OLE.t1ORIIAL PLdIIIT flOUE.1 C

Wednesday, June 12, 2013 6:42:08 PM 81

NRC Exam Question: 79 Tier/Group: 1/1 K/A Importance Rating: RO 3.9 SRO 3.8 K/A: 058 Loss of DC Power AG2.1 .19: Ability to use plant computers to evaluate system or component status.

Reference(s): Sim/Plant design, EPP-27 Proposed References to be provided to applicants during examination: Picture of ERFIS printout and C Charging pump status.

Learning Objective: Objective 4 of AOP-010 lesson plan Question Source: New Question History:

Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 41.10 /45.12)

Comments:

This question meets the K/A because the student must use the ERFIS(plant computer) to determine they have lost DC power. From that, the student will determine the status of C Charging pump and then determine the procedure flow paths. Determining which procedures to use is what makes this question an SRO level question.

Thursday, June 13, 2013 10:56:39 AM 156

CCCQ*

Rev. 4 EOP-ES-O.1 REACTOR TRIP RESPONSE Page 16 of 37 J__STEP_H ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED I

11. Maintain Stable Plant Conditions:

a. PZR pressure - AT 2235 PSIG

b. PZR level - AT 22%

c. S/G narrow range levels -

BETWEEN 8% AND 50%

d. RCS temperature:

• With any RCP running, RCS average temperature -

STABLE AT OR TRENDING TO 5 47° F OR

• With NO RCP running, RCS cold leg temperatures - STABLE AT OR TRENDING TO 547° F

e. Check DC busses A and B - e. Perform the following:

ENERGIZED

• IfDCbusAisNOT energized, perform EPP-26. LOSS OF DC BUS A, while continuing with this procedure.

• ifDCbusBisNOT energized, THEN perform EPP-27, LOSS OF DC BUS B. while continuing with this procedure.

12. Perform Attachment 8. Aligning Balance Of Plant, As Time Permits While Continuing With This Procedure

Rev. 15 EPP-27 LOSS OF DC BUS B Page 17 of 30 INFORMATION USE ATTACHMENT 1 MAJOR EFFECTS / LOAD LIST (Page 2 of 4)

Major DC Loads Lost:

DC Control Power for 4KV Busses 3 & 4 480V Busses 2B 3 & 4 and 480V Emergency Bus E-2 C (c c

7 crc cr& &13 Inverter B ck. Sa Reactor Protection and Safeguards Train B Steam Dump System: Steam Dumps PRV-1324 A-2 & B-3 fail closed.

Steam Driven AFW Pump Control Power 125VDC MCCs B & B-A, Distribution Panels B & Bi, and Power Panel No. 25 PCV-455C, PZR PORV FIC-626, FCV-626 High Flow Closing Controls will 1Q allow the valve to remain open while the valve is energized.

Aux Panel GC Fuse Panel:

Component Function Position Ckt FIC-626 FCV-626 High Flow Controls Closed 01 CVC-303B RCP B Seal Return Open 19 CVC-307 RCP Seal Leakoff Bypass Closed 21 RC-522A PW to RCP A Standpipe Closed 14 RC-519A PW to PRT Closed 24 RC-519B PW to PRT Closed 24 RC-519C PW to PRT Closed 20 RC-523 PRT to RCDT Pump Closed 18 RC-550 PRT N2 Supply Isolation Closed 47 RC-544 RV Flange Leak Detection Open 16 RC-568 RV Head Vent Closed 52 RC-570 PZR Head Vent Closed 52 RC-572 CV Atmos Solenoid Isolation Closed 52

(oC(cC Since an SI does not occur on a loss of DC Bus B, entry to the procedure will be via EOP-ES-0.1 Reactor Trip.

Certain actions will be completed while in EOP-ES-0.1, Attachment 7. These actions are:

a. Transfer of Instrument Bus 3 to MCC-8.

b. Shutdown of Emergency Diesel B.

STEP SPECIFIC DESCRIPTION AND RNP DIFFERENCES The following pages will provide the RNP step number and the STEP basis for each step where applicable. This is a Robinson specific EOP, therefore there is no corresponding ERG series of steps. This procedure covers an event that is not covered by the ERGs (Loss of DC). The entire procedure may be categorized as an SSD 10. The steps within this procedure will not interfere with performance of the EOPs since this procedure does not consider any other event in progress other than a loss of DC Bus B. The loss of a DC Bus at RNP is considered beyond design basis and is not analyzed in the UFSAR.

RNP BASIS STEP 1 STEP BASIS This step provides transitional direction for the subsequent step. If the Loss of DC occurred from an at power condition, the main generator output Circuit Breakers, 52/8 and 52/9, will be closed and action will be necessary to trip them.

If the event started from a low power or shutdown condition the subsequent step will not be necessary.

2 STEP BASIS On a loss of DC Bus B the North and South Generator Circuit Breakers, 52/8 and 52/9, will receive a lockout signal. Due to the loss of DC these breakers will not open. This in turn causes backup relaying to open other breakers to isolate the generator. In order to accomplish actions later in the procedure and to allow reclosing the backup Circuit Breakers these breakers must be opened.

There are no local controls that will open the breakers without control power. There is, however, a maintenance control (for testing) at each phase of the Circuit Breakers. This feature will trip the Circuit Breakers one phase at a time. Since this function was not intended to be performed by Site Personnel the Load Dispatcher will be notified to request assistance in opening the breakers.

3 STEP BASIS This continuous action step is provided to initiate efforts to repair the faulted DC Bus. It is placed early in the procedure so that efforts can be made to contact Maintenance personnel.

The high level step provides direction to diagnose the cause and provides transitional guidance. There are three possible failure mechanisms that are the most likely causes:

• Fault on B Battery

• Fault on B Battery Bus

• Fault on MCC-6 The failure, or tripping, of the in-service Battery Charger, is not a likely cause of the loss of DC since warning would be provided via an annunciator with ample time for Operator action to transfer the Chargers.

N4 STEP BASIS The note reminds the Operator that AFW Pump B will not be available due to a loss of Control Power.

4 STEP BASIS This step assures the maintenance of the secondary heat sink by maintaining S/G level at the standard range used throughout the EOP Network. In this case AFW Pump A and the SDAFW pump are specified since AFW Pump B is lost.

EPP-27-BD Rev 15 Page 4 of 9

DISCUSSION This Basis Document provides the step justification for a plant specific EPP. There is no ERG background for this procedure and no Safety Significant Deviation identification number is assigned for the steps since there are no corresponding ERG instructions.

The purpose of this procedure is to provide directions for combating conditions that arise from a loss of DC Bus A.

The procedure is intended to handle situations arising from conditions in which the EPPs are applicable (> 350°F).

This procedure assumes that no other casualty is in progress. Adjustment of the steps may be necessary if other events are in progress. A loss of DC is not an analyzed event at RNP and is not considered a credible event since a passive failure would be required to cause this event.

If the Reactor Trip Breakers are closed, there will always be a Rx Trip from a loss of power to the 52/RTA UV coil.

EDG A will always auto-start, (loss of power to air start solenoids), but without control power, it cannot flash its field or close its output breaker. Since DC Bus A supplies Inverter A, Instrument Bus 2 and 7 will always be lost. However, Loss of DC Bus A has vastly different consequences depending on the initial AC electrical lineup. Initially, if at power, following the reactor and turbine trip, 4 KV and 480V Buses will remain energized as the turbine coasts down. Bus voltage will decrease as the turbine speed decreases. DC Bus A supplies Control Power to Breakers on 4KV Busses I & 2, 480V Busses 1 & 2A, and 480V Emergency Bus E-1. (480V Bus 2B is normally supplied from 4KV Bus 1, 50 it will follow the effects of Bus 1. However, its DC Control Power is supplied by Bus B so it will not lose protective relaying)

• If these busses were initially on the Startup Transformer (SUT) they will remain energized. However, all Busses except for 480V Bus 2B will lose DC Control Power.

• If the busses were initially on the Unit Aux Transformer (UAT), the resultant Rx/Turbine/Generator trip will attempt an auto-transfer, but without DC Control Power, this will not occur. The UAT will be deenergized along with all the busses and components it was supplying. The Loss of E-i results in a loss of Instrument Bus 1. Since Instrument Bus 2 was lost from the Loss of DC Bus A, all bistables in both of these channels will fail and initiate an SI. (This will be a one-train SI since half the plant AC power is lost and the A train Sequencer is failed)

If the Reactor Trip Breakers were closed but all busses were still on the SUT, a Reactor Trip without SI will occur. If the Unit was at power and busses were on UAT, a Reactor Trip with SI will occur. Either way the EOP network will be entered. This EPP will be entered via PATH-i and EPP-7 or EOP-ES-0.1. Certain actions necessary for a Loss of DC Bus A will be completed in EOP-E-0 or EOP-ES-0.i attachments. These are actions that are performed to enable completion of certain steps in PATH-I and steps needed to combat the loss of DC. These steps are:

• Alignment of makeup to the Charging Pump suction by bypassing LCV-i I SB (CVC-358 is opened). LCV-i I 5B fails closed and since Letdown is isolated, a path of water must be aligned to the Charging Pumps.

• Instrument Bus 2 is transferred to MCC-8 in order to regain instrumentation to aid in diagnostics of PATH-i.

• The exciter field breaker is tripped locally to prevent further damage to the Generator and Exciter. As the Generator coasts down, the exciter will attempt to maintain voltage by increasing its output. Normal protection is not available because the control power to trip the Exciter Field Breaker is via DC Bus A.

• A EDG Fuel Racks are tripped to stop the engine. This is the fastest method of stopping the damage to the air start distributor. The EDG is running but can not be loaded because of the lack of control power to the Voltage Regulator.

• Instrument Air is isolated to the EDG to prevent start attempts and conserve air in the starting receiver.

• If MCC-5 is deenergized it is transferred to the DS Bus. If the loss of DC Bus A occurred from an at power condition MCC-S will be lost. Transferring the Bus to the DS bus will regain Instrument Bus i and safety related loads, such as valve operators powered from MCC-5.

This procedure and EPP-27, for DC Bus B have been credited in the evaluation of INPO SOER 81-15, PARTIAL LOSS OF DC POWER, recommendation 2C. No specific steps or sections were identified in the evaluation.

EPP-26-BD Rev. 12 Page3ofli

NRC Exam

80. 077 AA2.07 SRO 001 Given the following plant conditions:

-Plant is at 40% RTP

-The Load Dispatcher reports that grid voltage is degrading

-1 2:00:00 on 1/1/1 3, 480V Bus E2 voltage was reduced to 400V

-12:00:20 on 1/1/1 3, there were no changes to the electrical lineup

-12:30:00 on 1/1/1 3, grid voltage was restored Which ONE (1) of the following completes the statements below?

The B EDG is designed to start under this condition when (1) degraded voltage relays sense their setpoint. If B EDG cannot be restored to service, the latest time the plant can be in MODE 3 is (2) lAW applicable LCO.

(REFERENCE PROVIDED)

A. (1) 1/2 (2) 18:00:20 on 1/7/13 B. (1) 1/2 (2) 18:00:20 on 1/8/13 C. (1)2/3 (2) 18:00:20 on 1/7/13 Dv (1) 2/3 (2) 18:00:20 on 1/8/13 Tuesday, June 11,20132:10:26 PM 157

NRC Exam The correct answer is D A) Incorrect. 1/2 is incorrect. Plausible because this is the relay coincidence required for undervoltage, this is a degraded voltage condition. A degraded grid condition is 430V for 10 seconds. The undervoltage is 328V instantly. Place the plant in MODE 3 by 18:00:20 on 1/7/13 is incorrect. Plausible if they see the 7-day requirement and think that 1/7/13 would be the seventh day. This would cause them to believe that they have missed the time for condition B and would go to condition C which is be in MODE 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

B) Incorrect. 1/2 is incorrect. Plausible because this is the relay coincidence required for undervoltage, this is a degraded voltage condition. A degraded grid condition is 430V for 10 seconds. The undervoltage is 328V instantly. Place the plant in MODE 3 by 18:00:20 on 1/8/13 is correct. This would make the 7 days they had to restore the EDG plus the 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> they have to get the plant in MODE 3.

C) Incorrect. Degraded voltage requires 2/3 relays to start the EDG. A degraded grid condition is 430V for 10 seconds. The undervoltage is 328V instantly. Place the plant in MODE 3 by 18:00:20 on 1/7/13 is incorrect. Plausible if they see the 7-day requirement and think that 1/7/13 would be the seventh day. This would cause them to believe that they have missed the time for condition B and would go to condition C which is be in MODE 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

D) Correct. Degraded voltage requires 2/3 relays to start the EDG. A degraded grid condition is 430V for 10 seconds. The undervoltage is 328V instantly. Place the plant in MODE 3 by 18:00:20 on 1/8/13 is correct. This would make the 7 days they had to restore the EDG plus the 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> they have to get the plant in MODE 3.

Tuesday, June 11,20132:10:26 PM 158

NRC Exam Question: 80 Tier/Group: 1/1 K/A Importance Rating: RO 3.6 SRO 4.0 K/A: 077 Generator Voltage and Electric Grid Disturbances AA2: Ability to determine and interpret the following as they apply to Generator Voltage and Electric Grid Disturbances:

AA2.07: Operational status of engineered safety features Reference(s): Sim/Plant design, LCO 3.8.1 Proposed References to be provided to applicants during examination: LCO 3.8.1 Learning Objective: Objective 15 of EDG lesson plan Question Source: New Question History:

Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 41.5 and 43.5 /45.5, 45.7, and 45.8)

Comments:

This question meets the K/A because based off of electrical grid disturbances, the candidate must know the required number of relays to activate the degraded grid voltage relays to start the EDGs. The candidate must also apply LCO 3.8.3 for the EDG being inoperable. Application of the LCO makes this an SRO level question.

Thursday, June 13, 2013 11:55:18AM 159

AC Sources Operating 3.8.1 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources Operating

[CO 3.8.1 The following AC electrical sources shall be OPERABLE:

a. The qualified circuit between the offsite transmission network and the onsite emergency AC Electrical Power Distribution System: and

b. Two diesel generators (DG5) capable of supplying the onsite emergency power distribution subsystem(s).

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

ACTIONS NOTE

[CO 3.O.4.b is not applicable to DGs.

CONDITION REQUIRED ACTION COMPLETION TIME A. The qualified offsite A.1 Declare required 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from circuit inoperable, feature(s) with no discovery of no offsite power offsite power to available inoperable one train when its redundant concurrent with required feature(s) is inoperabi I ity of inoperable, redundant required feature(s).

AND A.2 Restore offsite 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> circuit to OPERABLE status. AND 8 days from discovery of failure to meet LCO (cont i nued)

HBRSEP Unit No. 2 3.8-1 Amendment No. 203

AC Sources Operating 3.8.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One DC inoperable. B.1 Perform SR 3.8.1.1 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for the offsite circuit. AND Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter AND B.2 Declare required 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from feature(s) supported discovery of by the inoperable DC Condition B inoperable when its concurrent with required redundant inoperability of feature(s) is redundant inoperable. required feature(s)

AND B.3.1 Perform SR 3.8.1.2 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for OPERABLE DC OR B.3.2.1 Determine OPERABLE DG 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is not inoperable due to common cause failure.

AND 8.3.2.2 Perform SR 3.8.1.2 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> for OPERABLE DC.

AND (continued)

HBRSEP Unit No. 2 3.8-2 Amendment No. 176

AC Sources Operating 3.8.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B (continued) B 4 Restore DG to 7 days OPERABLE status.

AND 8 days from discovery of failure to meet LCO C Required Action and C I Be in MODE 3 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A or AND B not met.

C.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />

0. Two or more AC sources NOTE inoperable. Entry into this Required Action may be delayed for no greater than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during performance of Required Action B.3.1 and Required Action 8.3.2.2.

D.1 Enter LCD 3.0.3. Immediately HBRSEP Unit No. 2 3.8-3 Amendment No. 176

( LOP DG Start Instrumentation B 3.3.5 B 3.3 INSTRUMENTATION B 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation BASES BACKGROUND The DGs provide a source of emergency power when offsite power is either unavailable or is insufficiently stable to allow safe unit operation. Undervoltage protection will generate an LOP start if a loss of voltage or degraded voltage condition occurs on the emergency bus. There are two LOP start signals for each 480 V emergency bus.

Undervoltage relays with definite time characteristics are provided on each 480 V emergency bus for detecting a sustained degraded voltage condition or a loss of bus voltage. The Loss of Voltage Function is provided by two relays on each bus. These relays are arranged in a one-out-of-two logic, such that either relay will generate an LOP signal if the voltage is below approximately 68 for a short time (loss of bus voltage). The Degraded Voltage Function is provided by three relays on each bus, which are combined in a two-out-of-three logic to generate an LOP signal if the voltage is below approximately 9OY for a long period of time (degraded voltage). The LOP start actuation is described in UFSAR, Section 8.3 (Ref. 1).

Trip Setpoints and Allowable Values The Trip Setpoints used in the relays are based on the Degraded Grid Voltage Study (Ref. 2). The selection of these Trip Setpoints is such that adequate protection is provided when all sensor and processing time delays are taken into account.

Trip Setpoints and tolerances are specified for each Function in the LCO. If the measured setpoint falls within the tolerance band, the relay is considered OPERABLE.

Operation with a measured setpoint less conservative than the Trip Setpoint, but within the tolerance band, is acceptable provided that operation and testing is consistent with the assumptions of the setpoint calculation. Each Trip Setpoint specified is more conservative than the analytical values determined in Reference 2 in order to account for instrument uncertainties appropriate to the trip function.

(conti nued)

HBRSEP Unit No. 2 B 3.3-115 Revision No. 0

/

I) LOP DG Start Instrumentation

( 8 3.3.5 8 3.3 INSTRUMENTATION B 3.3.5 Loss of Power (LOP) Diesel Generator COG) Start Instrumentation BASES BACKGROUND The DGs provide a source of emergency power when offsite power is either unavailable or is insufficiently stable to allow safe unit operation. Undervoltage protection will generate an LOP start if a loss of voltage or degraded voltage condition occurs on the emergency bus. There are two LOP start signals for each 480 V emergency bus.

Undervoltage relays with definite time characteristics are provided on each 480 V emergency bus for detecting a sustained degraded voltage condition or a loss of bus voltage. The Loss of Voltage Function is provided by two relays on each bus. These relays are arranged in a one-out-of-two logic, such that either relay will generate an LOP signal if the voltage is below approximately 68% for a short time (loss of bus voltage) The Degraded Voltage Function is provided by three relays on each bus, which are combined in a two-out-of-three logic to generate an LOP signal if the voltage is below approximately 90% for a long period of time (degraded voltage). The LOP start actuation is described in UFSAR, Section 8.3 (Ref. 1).

Trin Setpoints and Allowable Values The Trip Setpoints used in the relays are based on the Degraded Grid Voltage Study (Ref. 2). The selection of these Trip Setpoints is such that adequate protection is provided when all sensor and processing time delays are taken into account.

Trip Setpoints and tolerances are specified for each Function in the LCO. If the measured setpoint falls within the tolerance band, the relay is considered OPERABLE.

Operation with a measured setpoint less conservative than the Trip Setpoint, but within the tolerance band, is acceptable provided that operation and testing is consistent with the assumptions of the setpoint calculation. Each Trip Setpoint specified is more conservative than the analytical values determined in Reference 2 in order to account for instrument uncertainties appropriate to the trip function.

(continued)

HBRSEP Unit No. 2 B 33-115 Revision No. 0

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

a. The qualified circuit between the offsite transmission network and the onsite emergency AC Electrical Power Distribution System; and

b. Two diesel generators (DGs) capable of supplying the onsite emergency power distribution subsystem(s).

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

ACTIONS NOTE LCO 3.O.4.b is not applicable to DGs.

CONDITION REQUIRED ACTION COMPLETION TIME A. The qualified offsite A.1 Declare required 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from circuit inoperable, feature(s) with no discovery of no offsite power offsite power to available inoperable one train when its redundant concurrent with required feature(s) is inoperability of inoperable, redundant required feature(s).

AND A2 Restore offsite 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> circuit to OPERABLE status. AND 8 days from discovery of failure to meet LCO (continued)

HBRSEP Unit No. 2 3.8-1 Amendment No. 203

AC Sources Operating 3.8.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One DG inoperable. B.1 Perform SR 3.8.1.1 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for the offsite circuit. AND Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter AND B.2 Declare required 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from feature(s) supported discovery of by the inoperable DG Condition B inoperable when its concurrent with required redundant inoperability of feature(s) is redundant inoperable, required feature(s)

AND B.3.1 Perform SR 3.8.1.2 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for OPERABLE DG OR B.3.2.1 Determine OPERABLE DG 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is not inoperable due to common cause failure.

AND B.3.2.2 Perform SR 3.8.1.2 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> for OPERABLE DG.

AND (continued)

HBRSEP Unit No. 2 3.8-2 Amendment No. 176

qJJ-QR CQ AC Sources Operating 3.8.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.4 Restore DG to 7 days OPERABLE status.

AND 8 days from discovery of failure to meet LCO C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A or B not met.

C.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> D. Two or more AC sources NOTE inoperable. Entry into this Required Action may be delayed for no greater than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during performance of Required Action 8.3.1 and Required Action 8.3.2.2.

D.1 Enter LCO 3.0.3. Immediately HBRSEP Unit No. 2 3.83 Amendment No. 176

NRC Exam

81. W/E 05 EA2.2 SRO 001 Given the following plant conditions:

-The crew is implementing FRP-I-l.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK

-CST is intact, level is 15% and lowering OP-402, AUXILIARY FEEDWATER SYSTEM FRP-H.1, ATTACHMENT 2, SERVICE WATER BACKUP TO MDAFW PUMPS Which ONE (1) of the following completes the statements below?

The correct order in which the crew will attempt to restore Feed Flow to S/Gs is AFW, (1) . If the CST level reaches 10%, the CRS will direct the use of (2) to align Service Water as a backup to the AFW pumps.

A (1) Main Feedwater, Condensate (2) OP-402 B. (1) Main Feedwater, Condensate (2) ATTACHMENT 2 C. (1) Condensate, Main Feedwater (2) OP-402 D. (1) Condensate, Main Feedwater (2) ATTACHMENT 2 The correct answer is A A) Correct. The crew will attempt to restore feed using AFW, Main Feedwater, then Condensate in that order. Because the CST is not damaged, the CRS will direct the use of OP-402 to allign service water to the afw pumps as a backup source.

B) Incorrect. The crew will attempt to restore feed using AFW, Main Feedwater, then Condensate in that order. ATTACHMENT 2 is incorrect. Plausible because FRP-H.1 directs you to attachment 2 if CST level is lowering due to catastrophic failure.

C) Incorrect. AFW, Condensate, Main Feedwater is incorrect. Plausible because FRP-H.1 first has you check to see if the condensate system is available, if so, you first try to start a Main Feed Pump, if you cant, you move on to try and start a Condensate pump. Because the CST is not damaged, the CRS will direct the use of OP-402 to allign service water to the afw pumps as a backup source.

D) Incorrect. AFW, Condensate, Main Feedwater is incorrect. Plausible because FRP-H.1 first has you check to see if the condensate system is available, if so, you first try to start a Main Feed Pump, if you cant, you move on to try and start a Condensate pump. ATTACHMENT 2 is incorrect. Plausible because FRP-H.1 directs you to attachment 2 if CST level is lowering due to catastrophic failure.

Tuesday, June 11,20132:10:26 PM 160

NRC Exam Question: 81 Tier/Group: 2/1 K/A Importance Rating: SRO 4.3 K/A: W/E05 Loss of Secondary Heat Sink EA2: Ability to determine and interpret the following as they apply to the (Loss of Secondary Heat Sink)

EA2.2: Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.

References:

Sim/Plant design, FRP-H.1 and its basis document Proposed references to be provided to applicants during the Exam: None Learning Obj: Objective 5/6 of FRP-H.1 lesson plan Question Source: New Question History:

Question Cognitive Level: High 10 CFR part 55: (CFR 41.10/43.5/45.13)

Meets the K/A because the candidate must determine the correct order for restoring feed flow within FRP-H.1. This is SRO level because the candidate must select a procedure from which to procede given the conditions above.

Thursday, June 13, 2013 12:23:40 PM 161

Rev. 25 FRP-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK Page 4 of 42

- STEP_H ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED

6. Check CST level - GREATER THAN Align SW backup to the AFW Pumps 10% using OP-402. Auxiliary Feedwater System. while continuing with this procedure.

IF the CST is low due to catastrophic failure is inaccessable, align SW backup to the MDAFW Pumps using Attachment 2, SW Backup To MDAFW Pumps.

Go To Step 14.

7. Verify All S/G Blowdown AND Sample Isolation Valves - CLOSED

8. Check AEW Lines - INTACT Isolate break.

IF the break is isolated. HN Go To Step 9.

IF the break can NOT be isolated. THEN Go To Step 14.

Rev. 25 FRP-H.l RESPONSE TO LOSS OF SECONDARY HEAT SINK Page 3 of 42

STEP H ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED I

CAUTION Feed flow is not re-established to any faulted S/G if an intact S/G is available.

1. Check Total Feed Flow - LESS Go To Step 3.

THAN 300 GPM DUE TO OPERATOR ACTION

2. Reset SPDS And Return To Procedure And Step In Effect 3 Determine If Secondary Heat Sink Is Required As Follows:

a. Check RCS pressure - GREATER a. Reset SPDS and return to THAN ANY NON-FAULTED S/C procedure and step in effect.

PRESSURE

b. Check RCS temperature - b. Perform the following:

GREATER THAN 350°F [3100F]

1) Place RER System in service using Supplement I.

2) WHEN adequate cooling with RHR is established, THEN reset SPDS and return to procedure and step in effect.

4 Check Any Two SIG Wide Range any two S/C Wide Range Levels Levels - LESS TH.AN 10% [19Y] lower to less than 10% [19%]

Go To Step 5.

Go To Step 6.

5. Perform The Following:

a. Stop all RCPs

b. Observe CAUTION prior to Step 31 and Go To Step 31

Rev. 25 FRP-I{.l RESPONSE TO LOSS OF SECONDARY HEAT SINK Page 4 of 42 STEP_H ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

6. Check CST level - GREATER THAN Align SW backup to the AFW Pumps 10% using OP-402, Auxiliary Feedwater System, while continuing with this procedure.

IF the CST is low due to catastrophic failure is inaccessable, HN align SW backup to the MDAFW Pumps using Attachment 2, SW Backup To MDAFW Pumps.

Go To Step 14.

7. Verify All SIG Blowdowu MD Sample Isolation Valves - CLOSED

8. Check AW Lines - INTACT Isolate break.

IF the break is isolated, THEN Go To Step 9.

IF the break can NOT be isolated, THEN Go To Step 14.

Rev. 25 FRP-H.l RESPONSE TO LOSS OF SECONDARY HEAT SINK Page 5 of 42 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED H__STEP_H 9 Try To Establish Motor Driven AFW Flow To At Least One SIG As Follows:

a. Check AFW Pump Breakers a. Go To Step 9.c.

TRIPPED

b. Attempt to reclose any tripped breakers as follows:

1) Position the MDAFW Pump Control Switch to the STOP position

2) Reset SI

3) Position the MDAFW Pump Control Switch to the START position

4) Check MDAFW Pump - RUNNING 4) j the tripped breaker will NOT reclose THEN contact I&C to investigate.

Go To Step 10.

c. Verify AFW HDR DISCH Valves -

OPEN:

• V2-16A

• V2-16B

• V2-16C

d. Check AFW flow to S/Cs - d. Co To Step 10.

GREATER THAN 300 GPM

e. Reset SPDS and return to procedure and step in effect

Rev. 25 FRP-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK Page 6 of 42 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

__STEP_H 10 Attempt To Start SDAFW Pump As Follows:

a. Verify STEAM DRIVEN AFW PUMP a. i the steam supply valves STM SHUTOFF Valves - OPEN can NQ be opened, Go To Step 11.

• V1-8A

• V1-8B o V1-8C

b. Verify STEAM DRIVEN AFW PUMP DISCH Valves - OPEN

• V2-14A

• V2-14B

• V2-14C

c. Check AFW flow to S/Gs - c. Go To Step 11.

GREATER THAN 300 GPM

d. Reset SPDS and return to procedure and step in effect 11 Locally Investigate D Attempt To Restore AFW Flow As Follows:

a. Verify AFW Pump suction supply is available

b. Position the MDAFW Pump LOCAL/REMOTE Switch to LOCAL

c. Attempt to start a MDAFW Pump as follows:

1) Depress the MDAFW Pump local STOP Pushbutton

2) Depress the MDAFW Pump local START Pushbutton

3) Check MDAFW Pump - STARTED 3) Place the LOCAL/REMOTE Switch to REMOTE.

Rev. 25 FRP-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK Page 7 of 42 H STEP H ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED I

• 12 Check APW Flow To SIGs - GREATER IF feed flow to at least one S/G T1{AN 300 GPM verified, J]N perform the following:

a. Maintain flow to restore narrow range level to greater than 8% [18%]

b. jN narrow range level is greater than 8% [18%] ,

reset SPDS AND return to procedure and step in effect.

Go To Step 14.

13. Reset SPDS And Return To Procedure And Step In Effect

14. Stop All RCPs

15. Check Condensate System - IN Place the Condensate System in SERVICE service as follows:

a. IF the Condensate System is NOT available, THEN Go To Step 30.

b. Open QCV-10426, COND POL SEC c,

BYP.

c. Close V5-3, COND PUMP DISCH.

d. Momentarily place V5-3 to OPEN.

e. Start one Condensate Pump.

f. WHEN feedwater pressure is greater than 300 psig. THEN verify V5-3 full open.

g. Open HCV-1459, LP HEATERS BYE.

IF at least one Condensate Pump can NOT be started, THEN Go To Step 30.

Rev. 25 FRP-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK Page 8 of 42 RESPONSE RESPONSE NOT OBTAINED H__STEP_jl__ACTION/EXPECTED NOTE The subsequent step will defeat all FW Isolation signals which is necessary to allow starting of a Main Feedwater Pump. Manual Operator action will be required to initiate a FW Isolation.

16 Place ALL The FEEDWATER ISOLATION Key Switches In The OVRD/RESET Position

• STMGENA

• STMGENB

• STMGENC NOTE Local operation of the FRV and B/P valves below is via reverse acting handwheels.

17. Attempt To Establish Feedwater Flow As Follows:

a. Verify the FW HDR SECTION Valves - CLOSED

• V2-6A

• V2-6B

• V2-6C

b. Start one Main FW Pump b. Go To Step 20.

c. Open the FRV Bypass Valves: c. Locally open the FRV Bypass Valve using the Manual

• FCV-479 Handwheel. (Requires small

• FCV-489 Locked Valve Key.)

• FCV-499

d. Check FW Flow - ESTABLISHED d. Go To Step 20.

NRC Exam

82. 001 AA2.05 SRO 002 Given the following plant conditions:

-Unit 2 is raising power from 50% to 100% RTP

-Reactor power is currently at 55% RTP at EOL

-Control Bank D rods are at 171 steps

-30 minutes ago, a 200 gallon dilution to the RCS was performed

-The OAC pulls rods and releases the switch, Tavg and Reactor power steadily rise

-APP-005-E4, DELTA FLUX ALARM, is in

-VCT level has remained stable at 25% for the last 15 minutes Which ONE (1) of the following completes the statements below?

Based on the above conditions, the accident the crew is dealing with is a (1) accident. The basis for the LCO actions to address the Delta Flux Alarm is to (2)

A (1) uncontrolled rod withdraw (2) limit the amount of axial power distribution B. (1) dilution (2) limit the amount of axial power distribution C. (1) uncontrolled rod withdraw (2) limit the gross radial power distribution D. (1) dilution (2) limit the gross radial power distribution Tuesday, June 11,20132:10:26 PM 162

NRC Exam The correct answer is A A) Correct. Tavg and Reactor power rising are indications of a uncontrolled rod withdraw. APP-005-E4 refers you to LCO 3.2.3, Axial Flux Difference, which the basis is to limit the axial power distribution.

B) Incorrect. Dilution is incorrect. Plausible if the student thinks that 200 gallon dilution is too much. A normal dilution at power for temperature control is roughly 20 gallons.

However, 200 gallons would be an appropriate amount for raising power. The candidate should also realize that VCT level is constant, during a dilution, level would be rising. Limit the amount of axial power distribution is correct. APP-005-E4 refers you to LCO 3.2.3, Axial Flux Difference, which the basis is to limit the axial power distribution.

C) Incorrect. Uncontrolled rod withdraw is correct. Limit the gross radial power distribution is incorrect. Plausible if the student does not realize that APP-005-E4 sends you to LCO 3.2.3, not 3.2.4. 3.2.4 is QPTR and that LCO deals with limiting the gross radial power distribution. Both Deal with controlling power distribution, one is radially and the other is axially.

D) Incorrect. Dilution is incorrect. Plausible if the student thinks that 200 gallon dilution is too much. A normal dilution at power for temperature control is roughly 20 gallons.

However, 200 gallons would be an appropriate amount for raising power. The candidate should also realize that VCT level is constant, during a dilution, level would be rising Limit the gross radial power distribution is incorrect Plausible if the student does not realize that APP-005-E4 sends you to LCO 3.2.3, not 3.2.4. 3.2.4 is QPTR and that LCO deals with limiting the gross radial power distribution. Both Deal with controlling power distribution, one is radially and the other is axially.

Question: 82 Tier/Group: 1/2 K/A Importance Rating: SRO 4.6 K/A: 001 Continuous Rod Withdrawal AA2: Ability to determine and interpret the following as they apply to the Continuous Rod Withdrawal AA2.05: Uncontrolled rod withdrawal, from available indications Reference(s): Simulator/plant design, APP-005-E4, LCO 3.2.3, 3.2.4 Basis Documents Proposed References to be provided to applicants during examination: None Learning Objective:

Question Source: New Question History:

Question Cognitive Level: High 10 CFR Part 55 Content: 43.5 /45.13 Comments:

This question meets the K/A because from the given indications, the student must determine and interpret the casualty. To make it an SRO level question, the student needs to know the basis behind the Axial Flux Difference LCO.

Thursday, June 13, 2013 12:27:08 PM 163

B 3.2.3 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.3 AXIAL FLUX DIFFERENCE (AFD) (PDC-3 Axial Offset Control Methodology)

BASES BACKGROUND The purpose of this LCO is to establish limits on the values of the AFD in order to limit the axial power distribution skewing to either the top or bottom of the core By limiting the amount of power distribution skewing, core peaking factors are consistent with the assumptions used in the safety analyses. Limiting power distribution skewing over time also minimizes the xenon distribution skewing, which is a significant factor in axial power distribution control.

The operating scheme used to control the axial power distribution, PDC-3, involves maintaining the AFD within a tolerance band around a burnup dependent target, known as the target flux difference, to minimize the variation of the axial peaking factor and axial xenon distribution during unit maneuvers.

The target flux difference is determined at equilibrium xenon conditions. The control banks must be positioned within the core in accordance with their insertion limits and Control Bank D should be inserted near its normal position (i.e., 210 steps withdrawn) for steady state operation at high power levels. The power level should be as near RTP as practical. The value of the target flux difference obtained under these conditions divided by the Fraction of RTP is the target flux difference at RTP for the associated core burriup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RTP value by the appropriate fractinal THERMAL POWER level.

Periodic updating of the target flux difference value is necessary to follow the change of the flux difference at steady state conditions with burnup.

The Nuclear Enthalpy Rise Hot Channel Factor (FH) and QPTR LCOs limit the radial component of the peaking factors.

(continued)

HBRSEP Unit No. 2 B 3.2-17 Revision No. 0

QPTR B 3.2.4 8 3.2 POWER DISTRIBUTION LIMITS B 3.2.4 QUADRANT POWER TILT RATIO (QPTR)

ç BASES BACKGROUND The QPTR limit ensures that the gross radial power distribution remains consistent with the design values used in the safety analyses Precise radial power distribution measurements are made during startup testing, after refueling, and periodically during power operation.

The power density at any point in the core must be limited so that the fuel design criteria are maintained. Together, LCO 3.2.3, AXIAL FLUX DIFFERENCE (AFD) (PDC-3 Axial Offset Control Methodology), LCO 3.2.4, and LCO 3.1.6, Control Bank Insertion Limits, provide limits on process variables that characterize and control the three dimensional power distribution of the reactor core. Control of these variables ensures that the core operates within the fuel design criteria and that the power distribution remains within the bounds used in the safety analyses.

APPLICABLE This LCO precludes core power distributions that violate SAFETY ANALYSES the following fuel design criteria:

a. During a large break loss of coolant accident, the peak cladding temperature must not exceed 2200°F (Ref. 1);

b. During a loss of forced reactor coolant flow accident, there must be at least 95 probability at the 95 confidence level (the 95/95 departure from nucleate boiling (DNB) criterion) that the hot fuel rod in the core does not experience a DNB condition (Ref. 2):

c. During an ejected rod accident, the energy deposition to the fuel must not exceed 280 cal/gm (Ref. 3); and

d. The control rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn (Ref. 4).

The LCO limits on the AED, the QPTR, the Heat Flux Hot Channel Factor (FQ(Z)), the Nuclear Enthalpy Rise Hot (conti nued)

HBRSEP Unit No. 2 B 3.2-27 Revision No. 0

NIS & REACTOR CONTROL APP-005 Rev. 36 Page 35 of 45 WINDOW: E4 Page 1 of I CAUSES: 1. Two or more operable Excore Channels outside their Target Bands with Reactor Power above 90% of rated power or 0.9 x APL. A FLUX

2. Two or more operable Excore Channels outside the Operating ALARM Band with Reactor Power above 50% but less than or equal to 90%

of rated power or 0.9 x APL.

3. Accumulation of greater than 60 Penalty Points with Reactor Power above 50% of rated power.

DEVICE: SETPOINT: LOCATION:

1. ERFIS (Refer to FMP-009) Calculated by Core Axial Offset Calculation MUX #4 (CAOC)

OPERATOR ACTIONS

1. MONITOR any of the following parameters:

• Reactor Power

• A Flux

• ERFIS A Flux Program printout

2. NOTIFY Reactor Engineering

3. REFER TO FMP-009, Power Distribution Control, to restore A Flux to acceptable values per Control Room Status Board

4. REFER TO TS 3.2.3, Axial Flux Difference (AFD) (PDC-3 Axial Offset Control Methodology)

REFERENCES

1. B-190628 Sheet 441, Control Wiring Diagram (Cable AV)

2. HBR2-1 1098, Annunciator Window Engraving & Input Tabulation APP-005

TABLE 3.10-1 (Continued)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION Release Pathway/Instrumentation MCO* Corn pensatory Measures Plant Vent (Continued)

f. Plant Vent flow rate 1 With the number of channels operable less than the MCO requirement:

a. Exert best efforts to return the instruments to operable status within 30 days and, if unsuccessful, explain in the next Annual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner in accordance with Technical Specification 5.6.3 and,

b. Effluent releases via this pathway may continue provided that flow rate is estimated once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

2. Containment Vessel via Plant Vent

a. Radionoble gas monitor 1 With the number of channels operable less than the MCO requirement:

(R-12) provides automatic a. Exert best efforts to return the instruments to operable status within 30 days and, termination of Containment if unsuccessful, explain in the next Annual Radioactive Effluent Release Report Vessel releases upon why the inoperability was not corrected in a timely manner in accordance with exceeding alarm/trip Technical Specification 5.6.3 and, Setpoint. b. Effluent releases via this pathway may continue provided that the Plant Vent Radionoble Gas Monitor (RI4C) is operable; otherwise, suspend all releases via this pathway. (note 2)

b. Radioparticulate Monitor i With the number of channels operable less than the MCO requirement:

(R-1 1) provides automatic a. Exert best efforts to return the instruments to operable status within 30 days and, termination of containment if unsuccessful, explain in the next Annual Radioactive Effluent Release Report vessel releases exceeding why the inoperability was not corrected in a timely manner in accordance with alarm/trip setpoints. Technical Specification 5.6.3 and,

b. Effluent releases via this pathway may continue provided that the Plant Vent Radionoble Gas Monitor (RI4C) is operable; otherwise, suspend all releases via this pathway. (note 2)

MCO Minimum Channels Operable HBRODCM 3-84 Rev. 28

TABLE 3.10-1 (Continued)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION Release Pathway/Instrumentation MCO Corn pensatory Measures

6. Radwaste Building Exhaust (Continued)

c. Sampler flow rate gauge 1 With the number of channels operable less than the MCO requirement:

a. Exert best efforts to return the instruments to operable status within 30 days and, if unsuccessful, explain in the next Annual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner in accordance with Technical Specification 5.6.3 and,

b. Effluent releases via this pathway may continue provided the flow rate is estimated once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

7. Deleted.

NA NA MCO - Minimum Channels Operable 12/06 NOTES TO TABLE 3.10-1 Note 1 - No auxiliary sampling is required for periods when normal sampling is off 45 minutes.

Note 2 This MCO is required during Modes 1 2 3 4 and during the movement of recently irradiated fuel assemblies within the containment HBRODCM 3-90 Rev. 28

JSYSTEM OPERATION SYSTEM OPERATION Normal Operation CONTAINMENT HVAC Outdoor Air Makeup and Containment Purge (HVE-IA and HVE-IB)

Prior to activating these systems after shutdown, the containment particulate and gas monitors R-11 and R-12 are used to monitor the airborne activity levels inside the containment as a guide for routine release from the building. One of the following is required for monitoring during CV purge operation: 1) R-11 AND R-12, 2) R-14C (Refer to ODCM Table 3.10-1). In Modes 1,2,3,4 and when moving recently irradiated fuel in the CV, the CV Ventilation Isolation signal from R-11 AND R-12 is required to be OPERABLE (Refer to ITS 3.3.6). If R-11 OR R-12 become inoperable during CV Purge operation when ITS 3.3.6 is applicable, the CV Purge supply and exhaust valves shall be closed IMMEDIATELY. The Outdoor Air Makeup system has an air motor operated damper that opens when the containment purge fan HVE-1A or HVE-1B is started or when the Vacuum Relief system is energized. The air temperature leaving the heating coils is used to control steam to the coils. The Containment Purge system is manually energized from the RTGB.

When the containment purge fan is selected, the outside air louver opens and the butterfly valves open, the fan then starts and the intake damper to that fan opens. Fans are interlocked so that if the running fan motor trips electrically the other fan motor will start. Stopping the fan de-energizes the controls, the dampers, intake louver, and the isolation valves shut.

IF R-11 OR R-12 alarms, THEN the running CV PURGE FAN, HVE-1A or HVE-1B will stop, and the following valves close:

• V12-6, CV PURGE INLET

• V12-7, CV PURGE INLET

• V12-8, CV PURGE OUTLET

• V12-9, CV PURGE OUTLET

• V12-10, CV PRESS RELIEF

• V12-11, CV PRESS RELIEF

• V12-12, CV VAC RELIEF 38

COUQ 3.10 Radioactive Gaseous Effluent Monitoring Instrumentation Applicability Applies to the radioactive gaseous effluent instrumentation system.

Obiective To define the operating requirements for the radioactive gaseous effluent instrumentation system.

Specification CONTROLS 3.10.1 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.10-1 shall be operable with their alarm/trip setpoints set to ensure that the limits of ODCM Specification 3.2.1 are not exceeded. The alarm/trip setpoints of these channels shall be determined in accordance with the ODCM.

ACTIONS 3.10.2 With a radioactive effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above specification, without delay suspend the release of radioactive gaseous effluents, change the setpoint so it is acceptably conservative, or declare the channel not operable.

3.10.3 With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels operable take the action shown in Table 3.10-1.

3.10.4 The provisions of ODCM Specification 8.1 are not applicable.

BASES Radioactive Gaseous Effluent Instrumentation The radioactive gaseous effluent monitoring instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm/trip setpoints for these instruments shall be calculated in accordance with the procedures in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20, Appendix B, Table 2, Column 1. The operability and use of this instrumentation are consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

HBRODCM 3-81 Rev. 28

NRC Exam

83. 060 AG2.2.40 SRO 001 Given the following plant conditions:

-The plant is in MODE 1

-A CV Purge is in progress

-R-14C, PLANT STACK, NOBLE GAS, is OOS

-R-12, CV AIR OR PLANT STACK, NOBLE GAS, fails high Which ONE (1) of the following completes the statements below?

The running CV PURGE EXHAUST UNIT, HVE-IA or HVE-IB will (1) . lAW the ODCM, OFF-SITE DOSE CALCULATION MANUAL, effluent releases (2) continue.

(REFERENCES PROVIDED)

A. (1)trip (2) can B (1)trip (2) can NOT C. (1)NOTtrip (2) can D. (1)NOTtrip (2) can NOT Tuesday, June 11,20132:10:26 PM 164

NRC Exam The correct answer is B.

A) Incorrect. HVE-IA or B, depending on which one is running, will trip from R-12 failing high. R-12 failing high causes a containment isolation signal which in turn, trips the running HVE-1A or B. The effluent release cannot continue. Plausible if the student only goes to the section for R-14C being out of service.

B) Correct. HVE-1A or B, depending on which one is running, will trip from R-12 failing high. R-12 failing high causes a containment isolation signal which in turn, trips the running HVE-1A or B. The effluent release cannot continue on because both R-12 and R-14C are now OOS.

C) Incorrect. HVE-IA or B, depending on which one is running, will trip from R-12 failing high. Plausible if the student does not know what happens on a Containment Isolation signal or thinks that both R-1 I AND R-1 2 must alarm to stop the purge. The effluent release cannot continue. Plausible if the student only goes to the section for R-14C being out of service.

D) Incorrect. HVE-IA or B, depending on which one is running, will trip from R-12 failing high. Plausible if the student does not know what happens on a Containment Isolation signal. The effluent release cannot continue on because both R-12 and R-14C are now OOS.

Question: 83 Tier/Group: 1/2 K/A Importance Rating: SRO 4.7 K/A: 060 Accidental Gaseous Radwaste Release AG 2.2.40: Ability to apply Technical Specifications for a system.

Reference(s): Sim/Plant design, ODCM Proposed References to be provided to applicants during examination: Pgs 130, 131, 133, 139 of the ODCM Learning Objective:

Question Source: New Question History:

Question Cognitive Level: High 10 CFR Part 55 Content: (CFR41.10/43.2/43.5/45.3)

Comments:

Meets the K/A because the student must apply the ODCM(Tech Specs) to the given scenario. Application of the ODCM(Tech Specs) makes this an SRO level question.

Thursday, June 13, 2013 12:33:51 PM 165

3.10 Radioactive Gaseous Effluent Monitoring Instrumentation Applicability Applies to the radioactive gaseous effluent instrumentation system.

Oblective To define the operating requirements for the radioactive gaseous effluent instrumentation system.

Specification CONTROLS 3 10 1 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3 10-1 shall be operable with their alarm/trip setpoints set to ensure that the limits of ODCM Specification 3 2 1 are not exceeded. The alarm/trip setpoints of these channels shall be determined in accordance with the ODCM.

ACTIONS 3.10.2 With a radioactive effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above specification, without delay suspend the release of radioactive gaseous effluents, change the setpoint so it is acceptably conservative, or declare the channel not operable.

3.10.3 With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels operable take the action shown in Table 3.10-1.

3.10.4 The provisions of ODCM Specification 8.1 are not applicable.

BASES Radioactive Gaseous Effluent Instrumentation The radioactive gaseous effluent monitoring instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm/trip setpoints for these instruments shall be calculated in accordance with the procedures in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20, Appendix B, Table 2, Column 1. The operability and use of this instrumentation are consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

HBRODCM 3-81 Rev. 28

TABLE 3.10-1 RADIOACTIVE GASEOUS EFFLU ENT MON ITORING INSTRUMENTATION Release Pathway/Instrumentation MCO Compensatory Measures Plant Vent (R-14) a Radionoble gas monitor 1 With the number of channels operable less than the MCO requirements (R14C) provides a Exert best efforts to return the instruments to operable status within 30 days automatic termination of and if unsuccessful explain in the next Annual Radioactive Effluent Release Waste Gas Decay Tank Report why the inoperability was not corrected in a timely manner in releases upon exceeding accordance with Technical Specification 5 6 3 and alarm/trip setpoint.

b. Effluent releases via this pathway may continue provided that prior to initiating a waste gas decay tank release:

1. Two independent samples are analyzed in accordance with the Surveillance Requirements of ODCM Specification 3.2.1 and;

2. Two members of the facility staff independently verify the release rate calculations and the discharge line valving.

b. Radionoble gas monitor 1 With the number of channels operable less than the MCO requirement:

(R14C) monitors all a. Exert best efforts to return the instruments to operable status within 30 days effluents from Auxiliary and, if unsuccessful, explain in the next Annual Radioactive Effluent Release Building Ventilation Report why the inoperability was not corrected in a timely manner in System without providing accordance with Technical Specification 5.6.3 and, automatic termination of b. Effluent releases via this pathway may continue provided that grab samples release upon exceeding are collected once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and are analyzed for radionoble gases within their respective alarm 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

setpoints.

MCO Minimum Channels Operable HBRODCM 3-82 Rev. 28

Ycooe TABLE 3.10-1 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION Release Pathway/Instrumentation MCO Compensatory Measures Plant Vent (R-14)

a. Radionoble gas monitor I With the number of channels operable less than the MCO requirements:

(RI4C) provides a. Exert best efforts to return the instruments to operable status within 30 days automatic termination of and, if unsuccessful, explain in the next Annual Radioactive Effluent Release Waste Gas Decay Tank Report why the inoperability was not corrected in a timely manner in releases upon exceeding accordance with Technical Specification 5.6.3 and, alarm/trip setpoint.

b. Effluent releases via this pathway may continue provided that prior to initiating a waste gas decay tank release:

1. Two independent samples are analyzed in accordance with the Surveillance Requirements of ODCM Specification 3.2.1 and;

2. Two members of the facility staff independently verify the release rate calculations and the discharge line valving.

b. Radionoble gas monitor I With the number of channels operable less than the MCO requirement:

(RI4C) monitors all a. Exert best efforts to return the instruments to operable status within 30 days effluents from Auxiliary and, if unsuccessful, explain in the next Annual Radioactive Effluent Release Building Ventilation Report why the inoperability was not corrected in a timely manner in System without providing accordance with Technical Specification 5.6.3 and, automatic termination of b. Effluent releases via this pathway may continue provided that grab samples release upon exceeding are collected once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and are analyzed for radionoble gases within their respective alarm 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

setpoints.

MCO Minimum Channels Operable HBRODCM 3-82 Rev. 28

TABLE 3.10-1 (Continued)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION Release Pathway/Instrumentation MCO Compensatory Measures Plant Vent (Continued)

f. Plant Vent flow rate I With the number of channels operable less than the MCO requirement:

a. Exert best efforts to return the instruments to operable status within 30 days and, if unsuccessful, explain in the next Annual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner in accordance with Technical Specification 5.6.3 and,

b. Effluent releases via this pathway may continue provided that flow rate is estimated once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

2. Containment Vessel via Plant Vent

a. Radionoble gas monitor I With the number of channels operable less than the MCO requirement:

(R-12) provides automatic a. Exert best efforts to return the instruments to operable status within 30 days and, termination of Containment if unsuccessful, explain in the next Annual Radioactive Effluent Release Report Vessel releases upon why the inoperability was not corrected in a timely manner in accordance with exceeding alarm/trip Technical Specification 5.6.3 and, Setpoint. b. Effluent releases via this pathway may continue provided that the Plant Vent Radionoble Gas Monitor (R14C) is operable; otherwise, suspend all releases via this pathway. (note 2)

b. Radioparticulate Monitor i With the number of channels operable less than the MCO requirement:

(R-1 1) provides automatic a. Exert best efforts to return the instruments to operable status within 30 days and, termination of containment if unsuccessful, explain in the next Annual Radioactive Effluent Release Report vessel releases exceeding why the inoperability was not corrected in a timely manner in accordance with alarm/trip setpoints. Technical Specification 5.6.3 and,

b. Effluent releases via this pathway may continue provided that the Plant Vent Radionoble Gas Monitor (R14C) is operable; otherwise, suspend all releases via this pathway. (note 2)

MCO Minimum Channels Operable HBRODCM 3-84 Rev. 28

?i rjuc TABLE 3.10-1 (Continued)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION Release Pathway/Instrumentation MCO Compensatory Measures

6. Radwaste Building Exhaust (Continued)

c. Sampler flow rate gauge 1 With the number of channels operable less than the MCO requirement:

a. Exert best efforts to return the instruments to operable status within 30 days and, if unsuccessful, explain in the next Annual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner in accordance with Technical Specification 5.6.3 and,

b. Effluent releases via this pathway may continue provided the flow rate is estimated once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

7. Deleted.

NA NA MCO - Minimum Channels Operable 12/06 NOTES TO TABLE 3.10-1 Note 1 - No auxiliary sampling is required for periods when normal sampling is off 45 minutes.

Note 2- This MCO is required during Modes 1, 2, 3, 4, and during the movement of recently irradiated fuel assemblies within the containment.

HBRODCM 3-90 Rev. 28

NRC Exam

84. 061 AA2.02 SRO 001 Given the following plant conditions:

-The reactor is at 100% RTP

-The BOP reports R-9 is trending up and currently reads 0.6 R/hr OP-920, RADIATION MONITORING SYSTEM OMM-014, RADIATION MONITOR SETPOINTS Which ONE(1) of the following completes the statements below?

The BOP will verify R-9 setpoint lAW (1) . There (2) an EAL declaration that needs to be entered.

(REFERENCE PROVIDED)

A. (1)OMM-014 (2) is NOT B (1) OMM-014 (2) is C. (1) OP-920 (2) is NOT D. (1)OP-920 (2) is The correct answer is B A) Incorrect. OMM-014 will be used to verify R-9 setpoint. There is not an EAL declaration is incorrect. Plausible if the student does not convert R/hr to mr/hr.

B) Correct. OMM-014 will be used to verify R-9 setpoint. There is an EAL declaration that needs to be made. R-9 is currently reading 600 mr/hr. 5U5.1 for R-9 reading

>500mr/hr.

C) Incorrect. OP-920 is incorrect. Plausible since this is the procedure that is used to adjust the alarm setpoint for R-9. There is not an EAL declaration is incorrect.

Plausible if the student does not convert R/hr to mr/hr.

D) Incorrect. OP-920 is incorrect. Plausible since this is the procedure that is used to adjust the alarm setpoint for R-9. There is an EAL declaration that needs to be made.

R-9 is currently reading 600 mr/hr. SU5.1 for R-9 reading >500mr/hr.

Tuesday, June 11,20132:10:26 PM 166

NRC Exam Question: 84 Tier/Group: 1/2 K/A Importance Rating: SRO 3.2 K/A: 061 Area Radiation Monitoring (ARM) System Alarms AA2: Ability to determine and interpret the following as they apply to the Area Radiation Monitoring (ARM) System Alarms:

AA2.02: Normal radiation intensity for each ARM system channel.

Reference(s): Sim/Plant design, OMM-014, Hot Conditions EAL Proposed References to be provided to applicants during examination: Hot Conditions EAL Learning Objective: Self Study procedure Question Source: New Question History:

Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 43.5 /45.13)

Comments:

This question meets the K/A because the candidate must determine if this is a normal reading lAW OMM-014. From this, they must determine if an EAL Declaration is necessary. The EAL decleration is what makes this an SRO level question.

Thursday, June 13, 2013 12:54:19 PM 167

1 0 PURPOSE I I Provide a list of setpoints for the Area Process and Accident Radiation Monitors 1 .2 Provide a discussion of the basis used for calculation of the setpoints.

1 .3 Provide a method of documenting current setpoints and setpoint changes for the listed Area Radiation Monitors, Process Radiation Monitors, Accident Radiation Monitors, and the Radiation Monitoring System (RMS) Recorder.

2.0 REFERENCES

2.1 Improved Technical Specification LCO 3.3.6 and 3.9.3 2.2 UFSAR-Sections 11 and 12 2.3 10CFR 20, Appendix B 2.4 AOP-005, Radiation Monitoring System 2.5 OP-002, Nuclear Instrumentation System 2.6 OP-920, Radiation Monitoring System 2.7 RST-001, Radiation Monitor Source Checks 2.8 Station Curve Book, Section 6 2.9 EMP-013, Operation of R-14 and F-14 2.10 EMP-020, Operation of R-22 and R-38 2.11 EMP-022, Gaseous Waste Release Permits 2.12 EMP-023, Liquid Waste Release and Sampling 2.13 EMP-024, ODCM Surveillance 2.14 EMP-027, Calibration and Operation of GA Monitors R-37 and R-19A, B, and C 2.15 EMP-028, Process Monitor Setpoint Determination 2.16 EMP-034, Operation of R-24 A, B, and C 2.17 CP-014, Primary to Secondary Leak Rate Calculation OMM-014 Rev. 51 Page 4 of 21

1.0 PURPOSE 1.1 Provide instructions for Placing In-Service, Normal Operation, Removing from Service, and Infrequent Operation of the Area Monitors, Process Monitors, the Westronics Series 3000 Recorder, and the Yokogawa VR204 View Recorder.

2.0 REFERENCES

2.1 Improved Technical Specification LCD 3.3.6 and LCD 3.4.15 2.2 ODCM 3.10 and 3.11 2.3 SD-019, Radiation Monitoring System 2.4 DMM-001-12, Minimum Equipment List and Shift Relief 2.5 DMM-014, Radiation Monitor Setpoints 2.6 DP-001, Reactor Control and Protection System 2.7 OP-406, Steam Generator B lowdown/Wet Layup System 2.8 DP-509-1, Condensate Polishing System 2.9 OP-603, Electrical Distribution 2.10 DP-903, Service Water System 2.11 OP-917, Secondary Sampling System 2.12 DST-021, Daily Surveillances 2.13 DST-924-1, Area Radiation Monitoring System 2.14 DST-924-2, Process Radiation Monitoring System 2.15 RST-001, Radiation Monitor Source Checks 2.16 RST-008, Calibration of Radiation Monitor System, Monitors R-1 through R-8 2.17 RST-009, Calibration of Radiation Monitor System, Monitors R-9, R-30, R-31A, B, C and R-33 2.18 RST-010, Calibration of Radiation Monitoring System, Monitor R-11 OP-920 Rev 40 Page 5 of 73

NRC Exam

85. W/E14 EG2.2.38 SRO 001 Given the following plant conditions:

-A Large Break LOCA has occurred

-Neither CV Spray pump started

-CV Pressure has reached 40.5 psig FRP-P.1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK FRP-J.1, RESPONSE TO HIGH CONTAINMENT PRESSURE Which ONE (1) of the following completes the statements below?

The crew will use (1) to mitigate this event. Containment design pressure (2) been exceeded.

A. (1)FRP-P.1 (2) has NOT B (1) FRP-J.1 (2) has NOT C. (1)FRP-P.1 (2) has D. (1) FRP-J.1 (2) has -

Thursday, June 13, 2013 1:05:45 PM 168

NRC Exam The correct answer is B A) Incorrect. FRP-P.1 does not mitigate this casualty since a LBLOCA has occurred.

Plausible because you meet the entry conditions for FRP-P.1, however, since it was a LBLOCA, FRP-P.1 will not mitigate this. Containment design pressure has not been exceeded. Design pressure is 42 psig, this pressure is 40.5 psig.

B) Correct. FRP-J.1 will be used to mitigate the high CV pressure casualty since containment pressure is >lopsig and there are no CV Spray Pumps running.

Containment design pressure has not been exceeded. Design pressure is 42 psig, this pressure is 40.5 psig.

C) Incorrect. FRP-P.1 does not mitigate this casualty since a LBLOCA has occurred.

Plausible because you meet the entry conditions for FRP-P.1, however, since it was a LBLOCA, FRP-P.1 will not mitigate this. Containment design pressure has not been exceeded. Plausible because RNPs old containment pressure value was 40.5 psig.

D) Incorrect. FRP-J.1 will be used to mitigate the high CV pressure casualty since containment pressure is >lopsig and there are no CV Spray Pumps running.

Containment design pressure has not been exceeded. Plausible because RNPs old containment pressure value was 40.5 psig.

Question: 85 Tier/Group: 1/2 K/A Importance Rating: SRO 4.5 K/A: W/E14 High Containment Pressure EG2.2.38: Knowledge of conditions and limitations in the facility license.

Reference(s): Sim/Plant design, FRP-J.1, LCO 3.6.4 Basis Document Proposed References to be provided to applicants during examination: None Learning Objective: Objective lesson plan Question Source: New Question History:

Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 41.7 /41.10 /43.1 /45.13)

Comments:

This question meets the K/A because the student must have knowledge of the basis for LCO 3.6.4, CONTAINMENT PRESSURE, in order to answer this question. This is also what makes this question an SRO level question.

Thursday, June 13, 2013 1:05:45 PM 169

Rev. 9 FRP-J.l RESPONSE TO HIGH CONTAINMENT PRESSURE Page 3 of 8 Purpose and Entry Conditions (Page 1 of 1)

1. PURPOSE This procedure provides actions to respond to high containment pressure.

2. ENTRY CONDITIONS CSF-5, Containment Critical Safety Function Status Tree on a RED, ORANGE or YELLOW condition.

- END -

CSF-5, CONTAINMENT NO NO NO NO GREEN 0

YELLOW ORANGE YELLOW RED CSF-SAT GOTO GOTO GO TO GO TO FRP-J.3 FRP-J.2 FRP-J.1 FRP-J. 1 CSFST Rev. 4 Page 8 of 9

)

CSF-4, RCS INTEGRITY ENTER

+

YELLOW GO TO GO TO GREEN FRP-P.2 FRP-P.2 CSF SAT CSFST Rev.4 Page6of9

Rev. 17 FRP-P.l RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK Page 3 of 25 ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED I

1. Check CST Level - LESS THAN 10% IF CST Level lowers to less than 10%, perform Step 2.

Go To Step 3.

2. Align Service Water to the IF Service Water is unavailable, suction of the AEW Pumps using llN align Fire Water to the CST OP-402. Auxiliary Feedwater using EPP-1, Attachment 3, CST System. Emergency Fill From The Fire System.

3 Determine If RCS Cooldown Is Due To A Large Break LOCA As Follows:

a Check both of the tol]owing a. Go To Step 4.

conditions exist:

. RCS pressure - LESS THAN 275 PSIG [350 PSIG]

AND

. RHR flow on FI-6O5 -

GREATER THAN 1200 GPM

b. Reset SPDS AND return to procedure and step in effect

4. Check RCS Cold Leg Temperature - Go To Step 11.

LOWERING

5. Attempt To Stop RCS Cooldown As Follows:

a. Verify STEAM LINE PORVs -

CLOSED

b. Verify COND DUMPs - CLOSED

c. Check RHR System - ALIGNED c. Go To Step 6.

FOR CORE COOLING

d. Stop cooldown from RHR System

NRC Exam

86. 012 G2.1.20 SRO 001 Given the following plant conditions:

Initial Conditions:

-APP-006-D4, SIG A STM LINE HI FLOW, alarm is in

-APP-006-E4, S!G B STM LINE HI FLOW, alarm is in

-APP-006-F4, SIG C STM LINE HI FLOW, alarm is in

-Steam line pressure is 600 psig Subsequently:

-The crew has transitioned to EPP-16

-AFW flow has been throttled to each S/G

-S/G conditions are now as follows:

A-40% WR level, Pressure is 355 PSIG and rising B-35% WR level, Pressure is 300 PSIG and lowering C-35% WR level, Pressure is 290 PSIG and lowering EOP-E-2, FAULTED STEAM GENERATOR ISOLATION EPP-16, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS Which ONE (1) of the following completes the statements below?

Based off the initial conditions above, the plant (1) received an automatic SI signal. The crew will (2)

A. (1) has NOT (2) remain in EPP-16 B. (1)has (2) remain on in EPP-16 C (1) has (2) transition to EOP-E-2 D. (1) has NOT (2) transition to EOP-E-2 Thursday, June 13, 2013 1:12:32 PM 170

NRC Exam The correct answer is C A) Incorrect. An automatic SI signal has occurred. Plausible if the student does not know the setpoint for the high steam flow with low steamline pressure SI setpoint.

Since the High Steam Line Flow alarms are already in, the only other thing they need is the low steam line pressure < 614 psig, which they have at 600 psig. The crew will transition to EOP-E-2, not continue on with EPP-16. Plausible if the student is unaware that Foldout D is in effect as soon as you enter EPP-16. Since one SIG pressure is rising, the foldout tells you to go to EOP-E-2.

B) Incorrect. An automatic SI signal has occurred. The crew will transition to EOP-.E-2, not continue on with EPP-16. Plausible if the student is unaware that Foldout D is in effect as soon as you enter EPP-16. Since one S/G pressure is rising, the foldout tells you to go to EOP-E-2.

C) Correct. An automatic SI signal has occurred. The crew will transition to EOP-E-2.

EPP-1 6 directs you to Foldout D as soon as you enter the procedure. This foldout has criteria that if S/G pressure is rising, transition to EOP-E-2.

D) Incorrect. An automatic SI signal has occurred. Plausible if the student does not know the setpoint for the high steam flow with low steamline pressure SI setpoint.

Since the High Steam Line Flow alarms are already in, the only other thing they need is the low steam line pressure < 614 psig, which they have at 600 psig. The crew will transition to EOP-E-2. EPP-16 directs you to Foldout D as soon as you enter the procedure. This foldout has criteria that if S/G pressure is rising, transition to EOP-E-2.

Question: 86 Tier/Group: 2/1 K/A Importance Rating: SRO 4.6 K/A: 012 Reactor Protection System G 2.1.20: Ability to interpret and execute procedure steps.

Reference(s): Sim/Plant design, EPP-16, Foldout D, APP-004-E1 Proposed References to be provided to applicants during examination: None Learning Objective: Objective lesson plan Question Source: BANK Question History: Has not been used on an NRC exam Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 41.10/43.5/45.12)

Comments:

This question meets the K/A because the student must determine from the information given if there has been an automatic SI signal (RPS). Then the student must execute procedure steps by determining what the correct transition is based off of Foldout Criteria. Determining which procedure to go with is what makes this an SRO level question.

Thursday, June 13, 2013 1:12:32 PM 171

FIRST OUT REACTOR TRIPS APP-004 Rev. 16 Page 30 of 44 WINDOW: El Page 1 of 2 CAUSES: 1. Steam Line Break downstream of MSIVs and Check Valves. HI STM FLO

2. CV High Pressure (HI-HI). LO TAVGI LO SLP S FG RDITRI P DEVICE: SETPOINT: LOCATION:

1. High Steam Flow:

a. FC-474, FC-475 37.25% 109% (Ramped from 20% to

- Hagan Rack #16, #24 100% Turbine PWR)

b. FC-484, FC-485 37.25% 109% (Ramped from 20% to

- Hagan Rack #16, #25 100% Turbine PWR)

c. FC-494, FC-495 37.25% 109% (Ramped from 20% to

- Hagan Rack #16, #25 100% Turbine PWR)

2. Low TAVG:

a. TC-412E, TC-422E, TC-432E 543°F Hagan Rack #1, #11, #14

3. Low Steam Line Pressure:

a. PC-474A, PC-485A, PC-496A 614 psig Hagan Rack #13, #17, #24

4. CV HI-HI pressure:

a. PC-951A, PC-953A, PC-955A 10 psig Hagan Rack #3, #12, #15

b. PC-950, PC-952, PC-954 10 psig Hagan Rack #3, #12, #15 NOTE:

Any one of the following conditions will cause the associated function(s) to occur:

• High Steam Flow 1/2 flows on 2/3 lines in conjunction with either Low Steam Line Pressure on 2/3 steam lines OR Low TAVG 2/3 channels will cause Steam Line Isolation and Safety Injection)

• CV HI-HI pressure (2/3 channels on 2/2 matrixes will cause a Safety Injection, a Steam Line Isolation and CV Spray Actuation)

• A Safety Injection Signal will cause a Reactor Trip OPERATOR ACTIONS

1. IF Reactor has tripped, THEN GO TO EOP Network

2. IF Reactor has NOT tripped AND either of the following conditions are present:

• Transient operations in progress

• Bistable status panel indicates reactor trip logic has been met THEN:

a. TRIP Reactor

b. GO TO EOP Network

Rev. 21 UNCONTROLLED DEPRESSURIZATION OF ALL STEAM EPP 16 GENERATORS Page 3 of 33 j__STEP_H ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

1. Open Foldout D

2. Perform The Following:

a. Reset SPDS

b. Initiate monitoring of Critical Safety Function Status Trees

Rev. 34 EPP-Foldouts FOLDOUTS Page 15 of 23 CONTINUOUS USE FOLDOUT B (Page 1 of 1)

1. SI REINITIATION CRITERIA IF EITHER condition below occurs, start both SI Pumps:

• RCS Subcooling - LESS THAN 35°F [55°F]

• PZR Level - CAN NOT BE MAINTAINED GREATER THAN 14% [31%]

2. SECONDARY INTEGRITY CRITERIA IX any S/G pressure rises at any time, THEN Go To EOP-E-2, Faulted Steam Generator Isolation, Step 1.

3. COLD LEG RECIRCULATION SWITCHOVER CRITERIA IF RWST level lowers to less than 27%, THEN Go To EPP-9, Transfer To Cold Leg Recirculation.

4. AFW SUPPLY SWITCHOVER CRITERIA IF CST level lowers to less than 10%, II4I switch to backup water supply using OP-402, Auxiliary Feedwater System.

5. RHR PUMP PIT ISOLATION CRITERIA IF ANY condition below occurs, THEN Go To EPP-24, Isolation Of Leakage In The RHR Pump Pit:

• APP-OO1-D4, RHR PIT A HI-HI LEVEL - ILLUMINATED

• APP-OO1-D5, RHR PIT B HI-HI LEVEL - ILLUMINATED

• EITHER RTGB RHR Pit indication - GREATER THAN 24 INCHES

- END -

NRC Exam

87. 061 G2.4.20 SRO 001 Given the following plant conditions:

-The plant is in MODE 3

-The crew is currently in EPP-28, LOSS OF ULTIMATE HEAT SINK

-CST Level is 8%

-Deepwell Pump D is the only available water supply to MDAFW Pump A OP-402, AUXILIARY FEEDWATER SYSTEM ATTACHMENT 6, DEEPWELL COOLING Which ONE(1) of the following completes the statements below?

lAW EPP-28, total AFW flow to the SIGs is limited to (1) gpm. The crew will use (2) to align Deepwell Pump D.

A. (1) 195 (2) ATTACHMENT 6 of EPP-28 B. (1) 140 (2) ATTACHMENT 6 of EPP-28 C (1) 195 (2) OP-402 D. (1) 140 (2) OP-402 The correct answer is C A) Incorrect. 195 gpm is correct per the NOTE in EPP-28. Attachment 6 is incorrect.

Plausible because this attachment does allign deepwell cooling, however, it is to the EDG and not to the MDAFW pumps.

B) Incorrect. 140 gpm is incorrect. Plausible because this is the flow limit for one MDAFW pump alligned to one deepwell pump per OP-402. Attachment 6 is incorrect.

Plausible because this attachment does allign deepwell cooling, however, it is to the EDG and not to the MDAFW pumps.

C) Correct. 195 gpm is correct per the NOTE in EPP-28. EPP-28 tells you to use OP-402 to allign an AFW suction source.

D) Incorrect. 140 gpm is incorrect. Plausible because this is the flow limit for one MDAFW pump alligned to one deepwell pump per OP-402. EPP-28 tells you to use OP-402 to allign an AFW suction source.

Tuesday, June 11, 2013 2:10:26 PM 172

NRC Exam Question: 87 Tier/Group: 2/1 K/A Importance Rating: SRO 4.3 K/A: 061 Auxiliary / Emergency Feedwater (AFW) System G 2.4.20: Knowledge of the operational implications of EOP warnings, cautions, and notes.

Reference(s): Sim/Plant design, EPP-28 and its basis document, OP-402 Proposed References to be provided to applicants during examination: None Learning Objective: Objective 4 of EPP-28 lesson plan Question Source: New Question History:

Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 41.10 / 43.5 /45.13)

Comments:

This question meets the K/A because the candidate must know that the operational implication of the NOTE in EPP-28 is that while using Deepwell Pump D, total AFW flow is limited to 195 gpm. The student must determine which procedure to use to align the Deepwell pump and this is what makes this an SRO level question.

Thursday, June 13, 2013 1:16:40 PM 173

Rev. 13 EPP-28 LOSS OF ULTIMATE HEAT SINK Page 3 of 215

-1f-t ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED CAUTION Subsequent steps may require access to Non-vital Areas. Access to these areas prior to nullification of the threat could cause loss of personnel.

NOTE

• FRPs, EPP Foldouts, EPP-1, AOP-014, AOP-O2O, AND AOP-O22 are NOT applicable for this event. Transition use of these procedures should NOT be made unless otherwise directed in this procedure.

• Any time MCC6 is deenergized. coordination of field activities will require the use of portable radios cell phones. It is recommended that the DS Radios OR Fire Protection radios be used for this communication.

1 Check Reason For Entry - LOSS OF Go to the Section for the INTAKE STRUCTURE current Plant Mode:

MODE SECTION 3 Section E 4 Section F 5 Section C 6 Section H Defueled Section I

Rev. 13 EPP-28 LOSS OF ULTIMATE HEAT SINK Page 4 of 215 H__STEP_H ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED I

2. Go To The Step g Section For The Current Plant Mode:

MODE STEP/SECTION 3 3 4 Section A 5 Section B 6 Section C Defueled Section D

3. Verify The MSIVs D MSIV BYPs Valves - CLOSED

Rev. 13 EPP-28 LOSS OF ULTIMATE HEAT SINK Page 5 of 215 H STEP H ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED I

CAUTION The calculated maximum times the following components may operate without cooling to preclude adverse system effects is as follows:

• EDGs - 40 minutes

• CCW Heat Exchanger - Less than 60 minutes NOTE SW pressure at the CCW Heat Exchanger outlet must be maintained greater than 18 psig.

4 Check Status Of 0ff-Site Power - IF Off Site Power is lost THEN LOST Go To Step 5.

c

çLO (f Perform the following:

a. Dispatch an Operator to perform Attachment 6 Deepwell Cooling for one of the available EDGs.

b. WHEN Attachment 6 is complete, THEN dispatch an Operator to perform Attachment 7, Establishing CCW Cooling, while continuing with this procedure.

c. Verify V6-16C, SW ISOLATION TO TURBINE BUILDING - CLOSED

d. WHEN Attachment 7 is complete, THEN Monitor CCW Temperature maintain Below 125°F By Throttling SW-739 MI1 SW-740 In Equal Increments, while continuing with this procedure.

e. Observe the NOTE prior to Step 21 and Go To Step 21.

Rev. 13 EPP-28 LOSS OF ULTIMATE HEAT SINK Page 17 of 215 H__STEP_H ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED I

46. Transfer AFW To The NDAFW Pump As Follows:

a. Verify ALL V1-8 Valves -

CLOSED

• MS-V1-8A, SC A STM SUPPLY TO STM DRIVEN AFW PUMP

• MS-V1-8B, SC B STM SUPPLY TO STM DRIVEN AFW PUMP

• MS-V1-8C. SC C STM SUPPLY TO STM DRIVEN AFW PUMP

b. Verify ALL V2-14 Valves -

CLOSED

• AFW-V2-14A, STEAM DRIVEN FWP FDWTR DSCHC TO SG A

• AFW-V2-14B, STEAM DRIVEN FWP FDWTR DSCHC TO SC B

• AFW-V2-14C, STEAM DRIVEN FWP FDWTR DSCHC TO SC C

c. Establish AFW flow using the available MDAFW Pump

d. Maintain S/C levels between 39% 50% using the MDAFW Pump

• 47 Check CST Level - LESS THAN 10% WHEN CST Level is less than 10%

III41 observe the prior to Step 48 and perform Step 48.

Observe the NOTE prior to Step 49 and Co To Step 49.

Rev. 13 EPP-28 LOSS OF ULTIMATE HEAT SINK Page 18 of 215

STEP_H ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE When the MDAFW Pumps are being supplied via Deepwell Pump 0 Lhe total AFW flow to the Steam Generators is limited to 195 gpm.

48. Align SW To The AFW Pump Suction Using OP-402 Auxiliary Feedwater System NOTE Attachment 14 should be performed in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or less.

49. Contact Plant Operations Staff To Establish A Mission to Perform Attachment 14, WCCU Cooling

50. Go To Attachment 13, RCS Cooldown To 300°F

51. Contact Plant Operations Staff To Initiate A TSC Repair Mission To Restore At Least 7,500 CPM SW Flow Capability

• 52. Check Ultimate Heat Sink - When the Ultimate Heat Sink is RESTORE]) restored, THEN Go To Step 53.

Go To Step 51.

53. Return To Procedure And Step In Effect

- END -

Section 8.4.2 Page 3 of 7 8.4.2.2 (Continued) INIT VERI NOTE: When Deepwell Water is being supplied to the AFW Pumps, the MDAFW Pumps should have cooling water to the oil coolers and sealing water to the seals supplied from the Firemain VIA direction from the EOP Network. If this is NOT done, lube oil temperature should NOT exceed 140* F (lAW Tech Manual) AND no damage to the seals will occur.

CAUTION Exceeding the Maximum flow rates below could cause the AFW pumps to trip on low discharge pressure.

Number of runnng deepwell pumps 1 2 3 (200 gpm) (400 gpm) (600 gpm)

SDAFW Pump (ONLY) N/A 145 gpm 345 gpm (145 gpm Recirc + 105.2 gpm leakoff) 2 MDAFW Pumps 80 gpm 280 gpm 480 gpm (120 gpm Recirc) NOTE I I MDAFW Pump 140 gpm 325 gpm 325 gpm (60 gpm Recirc) NOTE 1 NOTE 1 NOTE 1: Flowno more than 325 gpm per MDAFW pump to prevent trip on overcu rrent.

d. Based on the above limitations, START AFW Pumps as follows:

1) IF SDAFW Pump is to be used, THEN PERFORM the following:

.a) REMOVE cap AND OPEN AFW-7, SDAFW PUMP SUCTION VENT.

.b) WHEN a solid stream of water issues, THEN CLOSE AFW-7 AND INSTALL the cap.

OP-402 Rev. 82 Page 38 of 108

NRC Exam

88. 063 A2.02 SRO 001 Given the following plant conditions:

-The plant is in MODE 4

-Battery Charger A is in service

-APP-036-F1O, BATT RM A/B HI/LO TEMP, alarms

-It is determined that Battery As representative cells have an average electrolyte temperature of 65°F 3.8.4, DC SOURCES-OPERATING 3.8.5, DC SOURCES-SHUTDOWN 3.8.6, BATTERY CELL PARAMETERS Which ONE (1) of the following completes the statements below?

Per APP-036-F1O, the SRO will direct the OAO to verify the (1) is operating.

The SRO is required to enter LCO(s) (2)

A. (1) alternate heater (2) 3.8.5 only B. (1) air conditioning (2)3.8.5 only C (1) alternate heater (2)3.8.4 and 3.8.6 D. (1) air conditioning (2)3.8.4 and 3.8.6 Tuesday, June 11,20132:10:26 PM 174

NRC Exam The correct answer is C A) Incorrect. Per APP-036-F1 0, you do verify the heater is operating because temperature is below 69°F. LCO 3.8.5 only is incorrect. Plausible if the student is unsure of the MODE requirements for LCO 3.8.5.

B) Incorrect. Verifying the air conditioning is operating is incorrect. Plausible because APP-036-F10 does have you check the air conditioning is operating, however, its only if temperature is high, above 86°F. LCO 3.8.5 only is incorrect. Plausible if the student is unsure of the MODE requirements for LCO 3.8.5.

C) Correct. Per APP-036-F1 0, you do verify the heater is operating because temperature is below 69°F. You are initially in LCO 3.8.6 for the battery cell parameter being <67°F. The battery should be declared inoperable. Based off this, the candidate must determine if they should enter LCO 3.8.4 or 3.8.5.

D) Incorrect. Verifying the air conditioning is operating is incorrect. Plausible because APP-036-F1 0 does have you check the air conditioning is operating, however, its only if temperature is high, above 86°F. You are initially in LCO 3.8.6 for the battery cell parameter being <67°F. The battery should be declared inoperable. Based off this, the candidate must determine if they should enter LCO 3.8.4 or 3.8.5.

Question: 88 Tier/Group: 2/1 K/A Importance Rating: SRO 3.1 K/A 063 DC Electrical Distribution System A2: Ability to (a) predict the impacts of the following malfunctions or operations on the DC electrical systems; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.02: Battery capacity as it is affected by discharge rate Reference(s): Sim/Plant design, LCO 3.8.4, 3.8.6 Proposed References to be provided to applicants during examination: None Learning Objective: Objective 15 of DC lesson plan Question Source: New Question History:

Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 41.5 /43.5/45.3/45.13)

Comments:

This question meets the K/A because the student must determine/predict the status(operable/inoperable)of the batteries due to its parameters. Based off of declaring the batteries inoperable, they will decide which Tech Specs to use to control the consequences of declaring the batteries inoperable.

Thursday, June 13, 2013 1:37:02 PM 175

DC SourcesOperating 3.8.4 3.8 ELECTRICAL POWER SYSTEMS 3 8 4 DC SourcesOperating LCO 3.8.4 The Train A and Train B DC electrical power subsystems shall be OPERABLE.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One DC electrical A.l Restore DC electrical 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> power subsystem power subsystem to inoperable. OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Associ ated Compl eti on Time not met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE_REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify battery terminal voltage is 7 days 125.7 V on float charge.

(continued)

HBRSEP Unit No. 2 3.8-19 Amendment No. 176

Battery Cell Parameters 3.8.6 ACTIONS (conti nued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Declare associated Ininiediately associated Completion battery inoperable.

Time of Condition A not met.

OR One or more batteries with average el ectrol yte temperature of the representative cells

< 67°F.

OR One or more batteries with one or more battery cell parameters not within Category C values.

SURVET LLANCE_REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.6.1 Verify battery cell parameters meet 7 days Table 3.8.6-1 Category A limits.

(continued)

HBRSEP Unit No. 2 3.8-25 Amendment No. 176

APP-036-F1 0 Page 1 of 2 ALARM BATT RM A/B HI/LO TEMP AUTOMATIC ACTIONS

1. None Applicable CAUSE

1. Loss of either Air Conditioning or Heat in the Battery A&B Room OBSERVATIONS

1. None Applicable ACTIONS CK (V)

1. DISPATCH an operator to the Battery A&B Room to check the temperature of the Temperature Indicating Switch TIS-4361.

2. IF the temperature is below 69°F, THEN PERFORM the following:

1) CHECK to see which heater is selected at the HVAC Panel.

2) VERIFY the selected heater is operating.

3) IF the selected heater will NOT operate, THEN SWITCH to the alternate heater.

4) VERIFY the alternate heater operates.

5) CHECK CLOSED breaker MCC-2 (7M), BATTERY ROOM HVAC/HEATER CONTROL PANEL.

6) CHECK CLOSED individual heater breakers (Breaker A and Breaker B located inside the lower control panel).

7) IF neither heater operates, THEN OBTAIN a temporary source of heat.

8) IF the temperature drops to 68°F or lower, THEN INITIATE action to trouble shoot and repair the heaters AND NOTIFY Engineering.

9) IF Station Battery A OR B pilot cell temperature drops below 67°F THEN DECLARE the affected battery inoperable.

3. IF the temperature is above 86°F, THEN VERIFY the air conditioning is operating.

4. For High OR Low temperature, CHECK the temperature in the Battery Room at least every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> AND RECORD the temperature in the Control Operators Log until the alarm resets. (ACR 94-00022)

APP-036 Rev. 83 Page 57 of 98

Battery Cell Parameters 3.8.6 3.8 ELECTRICAL POWER SYSTEMS 3.8.6 Battery Cell Parameters LCO 3.8.6 Battery cell parameters for Train A and Train B batteries shall be within the limits of Table 3.8.6-1 and average electrolyte temperature of representative cells shall be within limit.

APPLICABILITY: When associated DC electrical power subsystems are required to be OPERABLE.

ACTIONS NOTE Separate Condition entry is allowed for each battery.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more batteries A.1 Verify pilot cell 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> with one or more electrolyte level and battery cell float voltage meet parameters not within Table 3.8.6-1 Category A or B Category C limits.

limits.

AND A.2 Verify battery cell 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> parameters meet Table 3.8.6-1 Category C limits.

Once per 7 days thereafter AND A.3 Restore battery cell 31 days parameters to Category A and B limits of Table 3.8.6-1.

(continued)

HBRSEP Unit No. 2 3.824 Amendment No. 176

Battery Cell Parameters 3.8.6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Declare associated Immediately associated Completion battery inoperable.

Time of Condition A not met.

OR One or more batteries with average electrolyte temperature of the representative cell $

< 67°F.

OR One or more batteries with one or more battery cell parameters not within Category C values.

SURVET LLANCE_REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.6.1 Verify battery cell parameters meet 7 days Table 3.8.6-1 Category A limits.

(continued)

HBRSEP Unit No. 2 3.8-25 Amendment No. 176

Battery Cell Parameters 3.8.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.6.2 Verify battery cell parameters meet 92 days Table 3.8.6-1 Category B limits.

AND Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a battery di scharge 110 V AND Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a battery overcharge 150 V SR 3 8 6 3 Verify average electrolyte temperature 92 days of representative cells is 67°F.

HBRSEP Unit No. 2 3.8-26 Amendment No. 176

Definitions 1.1 Table 1.1-1 (page 1 of 1)

MODES REACTIVITY RATED AVERAGE CONDITION THERMAL REACTOR COOLANT MODE TITLE (keff) pOWER(a) TEMPERATURE

(°F) 1 Power Operation 0.99 > 5 NA 2 Startup 0.99 5 NA 3 Hot Standby < 0.99 NA 350 4 Hot Shutdown < 0.99 NA 350 > T,g > 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.

Cc) One or more reactor vessel head closure bolts less than fully tensioned.

HBRSEP Unit No. 2 1.1-6 Amendment No. 176

DC Sources-Shutdown 3.8.5 3.8 ELECTRICAL POWER SYSTEMS 3.8.5 DC Sources-Shutdown LCO 3.8.5 DC electrical power subsystem shall be OPERABLE to support the DC electrical power distribution subsystem(s) required by LCO 3.8.10, Distribution Systems-Shutdown.

APPLICABILITY: MODES 5 and 6, and During movement of irradiated fuel assemblies.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required DC A.1 Declare affected Immediately electrical power required feature(s) subsystems inoperable, inoperable.

OR A.2.1 Suspend CORE Immediately ALTERATIONS.

AND A.2.2 Suspend movement of Immediately irradiated fuel assemblies.

AND (continued)

HBRSEP Unit No. 2 3.8-22 Amendment No. 228

DC Sources Shutdown 3.8.5 ACTTflM (rrint in IPH CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.3 Suspend operations Immediately involving positive reactivity additions that could result in loss of required SDM or boron concentration.

ANfl A.2.4 Initiate action to Immediately restore required DC electrical power subsystems to OPERABLE status.

SURVEI LLANCE_REOUIREMENTS SURVEI LLANCE FREQUENCY SR 3.8.5.1 NOTE The following SRs are not required to be performed: SR 3.8.4.4. SR 3.8.4.5. and SR 3.8.4.6.

For DC sources required to be OPERABLE. the In accordance following SRs are applicable: with applicable SRs SR 3.8.4.1 SR 3.8.4.3 SR 3.8.4.5 SR 3.8.4.2 SR 3.8.4.4 SR 3.8.4.6 HBRSEP Unit No. 2 3.8-23 Amendment No. t7,19O

DC Sources Shutdown 3.8.5 ATTflN (rnni-iniiprfl CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.3 Suspend operations Immediately involving positive reactivity additions that could result in loss of required SDM or boron concentration.

ANfl A.2.4 Initiate action to Immediately restore required DC electrical power subsystems to OPERABLE status.

SURVEI LLANCE REOIJIREMENTS SURVET LLANCE FREQUENCY SR 3.8.5.1 NOTE The following SRs are not required to be performed: SR 3.8.4.4. SR 3.8.4.5. and SR 3.8.4.6.

For DC sources required to be OPERABLE, the In accordance following SRs are applicable: with applicable SRs SR 3.8.4.1 SR 3.8.4.3 SR 3.8.4.5 SR 3.8.4.2 SR 3.8.4.4 SR 3.8.4.6 HBRSEP Unit No. 2 3.8-23 Amendment No. +/-7&,19O

Distribution Systems Shutdown 3.8.10 3.8 ELECTRICAL POWER SYSTEMS 3.8.10 Distribution Systems Shutdown LCO 3.8.10 The necessary portion of AC, DC, and AC instrument bus electrical power distribution subsystems shall be OPERABLE to support equipment required to be OPERABLE.

APPLICABILITY: MODES 5 and 6, and During movement of irradiated fuel assemblies.

ACTI ONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Declare associated Immediately AC, DC, or AC supported required instrument bus feature(s) electrical power inoperable.

distribution subsystems inoperable. Q A.2.1 Suspend CORE Immediately ALTERATIONS.

AND A.2.2 Suspend movement of Immediately irradiated fuel assemblies.

AND (continued)

HBRSEP Unit No. 2 3.8-35 Amendment No. 176

Di stribution Systems Shutdown 3.8.10 ACTTONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.3 Suspend operations Immediately involving positive reactivity additions that could result in loss of required SDM or boron concentration.

ANfl A.2.4 Initiate actions to Immediately restore required AC, DC. and AC instrument bus electrical power distribution subsystems to OPERABLE status.

NU A.2.5 Declare associated Immediately required residual heat removal subsystem(s) inoperable and not in operation.

SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.10.1 NOTE Actual voltage measurement is not required for the AC vital buses supplied from constant voltage transformers.

Verify correct breaker alignments and 7 days voltage to required AC. DC. and AC instrument bus electrical power di stri buti on subsystems.

HBRSEP Unit No. 2 3.8-36 Amendment No. p6.190

NRC Exam

89. 076 A2.01 SRO 001 Given the following plant conditions:

-A plane crashed into the Intake Structure

-The plant was manually tripped

-The SUT tripped

-The CRS has entered EPP-28, EPP-28, LOSS OF ULTIMATE HEAT SINK ATTACHMENT 6, DEEPWELL COOLING ATTACHMENT 7, ESTABLISHING CCW COOLING Which ONE (1) of the following completes the statements below?

The maximum time the crew has to restore cooling to an EDG to preclude adverse effects is (1) minutes. In order to provide cooling to at least one EDG, the CRS will direct ATTACHMENT (2) to be performed.

A. (1)60 (2) 6 B. (1)60 (2) 7 C(1)4O (2) 6 D. (1)40 (2) 7 Tuesday, June 11, 2013 2:10:26 PM 176

NRC Exam The correct answer is C A) Incorrect. 60 minutes is incorrect. Plausible because within the same caution, the caution states that the max time CCW heat exchangers can operate without cooling is 60 minutes. Attachement 6 is correct. This will align deep well cooling to the running EDG.

B) Incorrect. 60 minutes is incorrect. Plausible because within the same caution, the caution states that the max time CCW heat exchangers can operate without cooling is 60 minutes. Attachement 7 is incorrect. Plausible because you will utilize attachernent 7 in epp-28, however, it is to establish cooling to the CCW heat exchangers and not the EDG. While performing attachment 7, there are multiple valves that are aligned in the EDG room.

C) Correct. 40 minutes is the maximum time the EDG may operate without cooling to preclude adverse system effects. The CRS will direct attachment 6 of epp-28 to restore cooling to the EDG using the D deepwell pump.

D) Incorrect. 40 minutes is the maximum time the EDG may operate without cooling to preclude adverse system effects. Attachement 7 is incorrect. Plausible because you will utilize attachement 7 in epp-28, however, it is to establish cooling to the CCW heat exchangers and not the EDG. While performing attachment 7, there are multiple valves that are aligned in the EDG room.

Question: 89 Tier/Group: 2/1 K/A Importance Rating: SRO 3.7 K/A: 076 Service Water System (SWS)

A2: Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.01: Loss of SWS Reference(s): Sim/Plant design, EPP-28 Proposed References to be provided to applicants during examination: None Learning Objective: Objective 5 of EPP-28 lesson plan Question Source: New Question History:

Question Cognitive Level: Low 10 CFR Part 55 Content: (CFR: 41.5 / 43.5 / 45/3 / 45/13)

Corn ments:

This question meets the K/A because the student must predict the impact a total loss of service water has on the EDG. The student must then determine which attachment to use to restore cooling to the EDG. Determining which attachment to use is what makes this an SRO level question.

Thursday, June 13, 2013 1:45:22 PM 177

Rev. 13 EPP28 LOSS OF ULTINATE HEAT SINK Page 5 of 215 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED j__STEP_H CAUTION The calculated maximum times the following components may operate without cooling to preclude adverse system effects is as follows

• EDGs - 40 minutes

• CCW Heat Exchanger - Less than 60 minutes NOTE SW pressure at the CCW Heat Exchanger outlet must be maintained greater than 18 psig.

4. Check Status Of Off-Site Power - IF Off-Site Power is lost, THEN LOST Go To Step 5.

Perform the following:

a. Dispatch an Operator to perform Attachment 6.

Deepwell Cooling for one of the available EDGs.

b. WHEN Attachment 6 is complete, THEN dispatch an Operator to perform Attachment 7, Establishing CCW Cooling, while continuing with this procedure.

c. Verify V6-l6C, SW ISOLATION TO TURBINE BUILDING - CLOSED

d. WHEN Attachment 7 is complete. THEN Monitor CCW Temperature maintain Below 125°F By Throttling SW-739 AIffl SW-74O In Equal Increments, while continuing with this procedure.

e. Observe the NOTE prior to Step 21 and Go To Step 21.

Rev. 13 EPP-28 LOSS OF ULTIMATE HEAT SINK Page 6 of 215

__STEP_H ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED I

5. Establish Cooling To ONE EDG Using Deepwell Pump D As Follows:

a. Check EDG B - AVAILABLE a. IF A EDO is available, THEN dispatch an Operator to perform Attachment 6, Deepwell Cooling, for A EDO A!i2 Go To Step 6 IF neither EDG is available.

THEN Go To Section J, SBO With No Service Water.

b. Dispatch an Operator to perform Attachment 6, Deepwell Cooling, for B EDO

6. On the EDG To Remain Running Verify The Following Components:

a. 1 CCW Pump - RUNNING

b. 1 Charging Pump - RUNNING

c. MDAFW Pump - STOPPED

d. BOTH SW Pumps - STOPPED

e. SWBP - STOPPED

f. BOTH HVH Units - STOPPED NOTE If a valve was open when power was lost on the bus it may be assumed to remain open. There are no spurious valve operations for this event.

7. Verify The Following CCW Valves

- OPEN:

• CC-716A, CCW TO RCP

• CC-716B, CCW TO RCP

• CC-735, THERM BAR OUT ISO

. FCV-626, THERM BAR FLOW CONT

Rev. 13 EPP-28 LOSS OF ULTIMATE HEAT SINK Page 7 of 215

STEP_H ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED

8. Check Attachment 6 - COMPLETE WI Attachment 6 is complete, THEN Go To Step 9.

9. Check MCC-5 - ENERGIZED Co To Step 11.

10. Locally Verify The Following Breakers At MCC-5 - CLOSED:

• BATTERY CHARGER A-i (CMPT-5B)

• BATTERY CHARGER A (CMPT-11CR) ii. Check MCC-6 - ENERGIZED Go To Step 13.

12. Locally verify the following breakers at MCC-6 - CLOSED:

• BATTERY CHARGER B-i (CMPT-14M)

• BATTERY CHARGER B (CMPT-1SFR) 13 Dispatch An Operator To Perform LP Attachment 7 Establishing CCW Cooling

\

14. Verify V6-16C, SW ISOLATION TO TURBINE BUILDING - CLOSED

• 15. Check Attachment 7 - COMPLETED WHEN Attachment 7 has been completed, THEN observe the NOTE prior to Step 16 and perform Step 16.

Observe the NOTE prior to Step 17 and Go To Step 17.

NOTE SW pressure at the CCW Heat Exchanger outlet must be maintained greater than 18 psig.

16. Monitor CCW Temperature Maintain Below 125° F By Throttling SW-739 SW-740 In Equal Increments

NRC Exam

90. 103 A2.03 SRO 002 Given the following plant conditions:

-A main steam line break inside CV results in a Safety Injection and CV Spray actuation

-EOP-E-0, REACTOR TRIP OR SAFETY INJECTION, has been implemented

-CV pressure is currently 14 psig

-CC-735, THERM BAR OUT ISO, is open SUPPLEMENT A, SAFETY INJECTION COMPONENT ALIGNMENT SUPPLEMENT B, PHASE B AND CV SPRAY COMPONENT ALIGNMENT Which ONE (1) of the following completes the statement below?

Based on current plant conditions, CC-735 (1) in the correct position and the CRS may direct the use of (2) to verify its position.

A. (1)is (2) SUPPLEMENT A B. (1)isNOT (2) SUPPLEMENT A C. (1)is (2) SUPPLEMENT B D (1) is NOT (2) SUPPLEMENT B Tuesday, June 11,20132:10:26 PM 178

NRC Exam The correct answer is D A) Incorrect. CC-735 is in the correct position is incorrect. Plausible if the student does not know that CC-735 is shut on a Phase B signal. Also plausible if they student does not know that at this CV pressure, a Phase B signal is automatically generated.

Supplement A is incorrect. Plausible if the student believes that CC-735 gets shut by a phase A signal.

B) Incorrect. CC-735 is not in its correct position is correct. It should be shut since the plant received a Phase B signal based of CV pressure being> 10 psig. Supplement A is incorrect. Plausible if the student believes that CC-735 gets shut by a phase A signal.

C) Incorrect. CC-735 is in the correct position is incorrect. Plausible if the student does not know that CC-735 is shut on a Phase B signal. Also plausible if they student does not know that at this CV pressure, a Phase B signal is automatically generated.

Supplement B is correct. This supplement gives you specific valve positions for all Phase B valves.

D) Correct. CC-735 is not in its correct position is correct. It should be shut since the plant received a Phase B signal based of CV pressure being> 10 psig. Supplement B is correct. This supplement gives you specific valve positions for all Phase B valves.

Question: 90 Tier/Group: 2/1 K/A Importance Rating: SRO 3.8 K/A: 103 Containment System A2: Ability to (a) predict the impacts of the following malfunctions or operations on the containment system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations A2.03: Phase A and B isolation Reference(s): Sim/Plant design, EOP-E-0, Supplement B Proposed References to be provided to applicants during examination: None Learning Objective: Objective 2 Question Source: New Question History:

Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 41.5 / 43.5 / 45.3/45.13)

Comments:

This question meets the K/A because the student must determine which isolation signal (phase A or B) did not function properly based off the valve position in the stem. Then, must determine which procedure/portion of a procedure with which to use to verify the specific phase B valves are shut. Determining which procedure or section of a procedure is what makes this an SRO level question.

Thursday, June 13, 2013 1:48:04 PM 179

Rev. 45 EPP Supplements

- SUPPLEMENTS Page 19 of 103 CONTINUOUS USE Supplement B Phase B And CV Spray Component Alignment (Page 1 of 1)

To establish Phase B Containment Isolation verify the following valves - CLOSED

a. RCP Cooling e CVC-38l, SEAL WTR RTRN ISO

• FCV-626, THERM BAR FLOW CONT

• CC-735, THERM BAR OUT ISO

• CC-716A, CCW TO RCP ISO

• CC 7163 CCW TO RCP ISO

• CC-73O, ERG OUTLET ISO

b. MSIV AND MSIV BYEs - CLOSED

2. To establish CV Spray, perform the following:

a. Verify valves positioned as follows:

SI-844A, PUMP A INLET - OPEN

• SI-844B, PUMP B INLET - OPEN

• SI-845A, SAT DISCH - OPEN

• SI-845B, SAT DISCH - OPEN

• SI-845C, SAT THROTTLING - THROTTLED TO APPROXIMATELY 12 GPM

• SI-880A, PUMP A DISCH - OPEN

• S1-88O3, PUMP A DISCH - OPEN

• SI-88OC, PUMP B DISCH - OPEN

• SI-88OD, PUMP B DISCH - OPEN

b. Return to procedure and step in effect.

- END -

Lo(cQ Rev. 45 EPP- Supplements SUPPLEMENTS Page 5 of 103 CONTINUOUS USE Supplement A Safety Injection Component Alignment (Page 1 of 14)

1. Go to the appropriate step from the table below:

FUNCTION STEP To Establish Cold Leg Injection Step 2 To Establish Hot Leg Injection Step 4 To Establish Phase A Containment Isolation Step 6 To Establish Containment Ventilation Isolation Step 8 To Establish Motor Driven AFW Pump alignment Step 10 To Establish Feedwater Isolation Step 12 Shift Control Room Ventilation to Emergency Pressurization Mode Step 14

Rev. 45 EPP-Supplements SUPPLEMENTS Page 11 of 103 CONTINUOUS USE Supplement A Safety Injection Component Alignment (Page 7 of 14)

NOTE Fuse pullers are located with AOP/EOP/DSP Tool Kits.

6. To establish Phase A Containment Isolation, perform the following verification:

a. Verify the following letdown valves - CLOSED

• CVC-200A, LTDN ORIFICE

• CVC-200B, LTDN ORIFICE

• CVC-200C, LTDN ORIFICE

• CVC-204A, LTDN LINE ISO

• CVC-204B, LTDN LINE ISO

b. Check IA PCV-l716, INSTRUMENT AIR ISO TO CV - CLOSED

c. IF PCV-1716 does NOT indicate closed, RI obtain fuse pullers AND fail PCV-1716 closed by removing the following fuses on Auxliary Panel GC:

• Aux. Panel GC, circuit 32, fuse 67

• Aux. Panel CC, circuit 32, fuse 68 (CONTINUED NEXT PAGE)

Rev. 45 EPP-Supplements SUPPLEMENTS Page 12 of 103 CONTINUOUS USE Supplement A Safety Injection Component Alignment (Page 8 of 14)

6. (CONTINUED)

d. Verify the following Sample Line valves - CLOSED

• PS-956A, PZR STEAM SPACE SAMPLE

• PS-956B, PZR STEAM SPACE SAMPLE

• PS-956C, PZR LIQUID SPACE SAMPLE

• PS-956D, PZR LIQUID SPACE SAMPLE

• PS-956E, RCS HOT LEG SAMPLE

• PS-956F, RCS HOT LEG SAMPLE

• PS-956G, ACCUM SAMPLE LINE AIR OPERATED ISOLATION

• PS-956H, ACCUM SAMPLE LINE AIR OPERATED ISOLATION

a. Verify the following PRT valves - CLOSED

• RC-516, PRT TO GAS ANALYZER

• RC-553, PRT TO GAS ANALYZER

• RC-550, PRT NITROGEN SUPPLY

• RC-5l9A AND B, PRIMARY WATER TO PRESSURIZER RELIEF TANK

f. Verify CC-739, EXCESS LTDN HX OUTLET - CLOSED NOTE Local operation of the valve below is via a reverse acting handwheel.

g. Verify S1-855, ACCUMULATOR NITROGEN SUPPLY - CLOSED (CONTINUED NEXT PAGE)

\c(j Rev. 45 EPP-Supplements SUPPLEMENTS Page 13 of 103 CONTINUOUS USE Supplement A Safety Injection Component Alignment (Page 9 of 14)

6. (CONTINUED)

h. Verify the following Fire Protection System valves - CLOSED

• FP-248, ELECT. PENETRATION CV ISOLATION

• FP-249, ELECT. PENETRATION CV ISOLATION

• FP-256. RCP SPRINKLER ISOLATION VALVE

• FP-258, RCP SPRINKLER ISOLATION VALVE

i. Verify the following Waste Disposal System valves - CLOSED

• WD-l721, RCDT PUMP DISCHARGE LINE AUTO ISOLATION

• WD-l722, RCDT PUMP DISCHARGE LINE AUTO ISOLATION

• WD-l723, CONTAINMENT SUMP PUMP DISCHARGE AUTO ISOLATION

• WD-1728, CONTAINMENT SUMP PUMP DISCHARGE AUTO ISOLATION

• WD-l786, RCDT VENT

• WD-1787, RCDT VENT

• WD-1789, RCDT SAMPLE LINE TO GAS ANALYZER

• WD-1794, RCDT SAMPLE LINE TO GAS ANALYZER (CONTINUED NEXT PAGE)

Rev. 45 EPP -Supplements SUPPLEMENTS Page 14 of 103 CONTINUOUS USE Supplement A Safety Injection Component Alignment (Page 10 of 14)

6. (CONTINUED)

j. Check the following S/G blowdown and sample valves CLOSED

. SGB-FCV-1930A, STEAM GENERATOR A BLOWDOWN LINE

. SGB-FCV-1930B, STEAM GENERATOR A BLOWDOWN LINE

. SGB-FCV-1931A, STEAM GENERATOR B BLOWDOWN LINE

. SGBFCV-1931B, STEAM GENERATOR B BLOWDOWN LINE

. SGBFCV-1932A, STEAM GENERATOR C BLOWDOWN LINE

. SGB-FCV-1932B, STEAM GENERATOR C BLOWDOWN LINE

. SGB-FCV-1933A, STEAM GENERATOR A SAMPLE LINE

. SGB-FCV-l933B. STEAM GENERATOR A SAMPLE LINE

. SGBFCV-1934A, STEAM GENERATOR B SAMPLE LINE

. SGB-FCV-1934B, STEAM GENERATOR B SAMPLE LINE

. SGBFCV-1935A. STEAM GENERATOR C SAMPLE LINE

. SGB-FCV-1935B, STEAM GENERATOR C SAMPLE LINE

k. j any S/G blowdown and sample valves do indicate closed, jI deenergize the R-19 skid (A/B/C) for the affected valves.

1. Verify the following Radiation Monitoring System valves CLOSED

• RMS-l, CONTAINMENT OUTLET TO R-ll AND R-12

• RMS-2, R-ll AND R-12 INLET ISOLATION

• RNS-3, CONTAINMENT INLET FROM R-ll AND R-12

• RMS-4, R-ll AND R-12 OUTLET ISOLATION (CONTINUED NEXT PAGE)

\( 11uC Rev. 45 EPP-Supplements SUPPLEMENTS Page 15 of 103 CONTINUOUS USE Supplement A Safety Injection Component Alignment (Page 11 of 14) 6 (CONTINUED)

m. Check the IVSW Sytem automatic header isolation valves OPEN.

• PCV-1922A. AUTOMATIC HEADER PRESSURE CONTROL VALVE

• PCV-1922b, AUTOMATIC HEADER PRESSURE CONTROL VALVE

n. IF PCV-l922A OR PCV1922B is closed, THEN fail air to the affected valve(s) by performing the following:

1) Isolate the affected valve(s) air supply isolation valve.

2) Open the affected valve(s) air regulator petcock to bleed air from the valve.

7. Return to procedure and step in effect.

8. To establish Containment Ventilation Isolation, verify the following valves - CLOSED

• V12-6. CONT PURGE VALVES

• Vl2-7, CONT PURGE VALVES

• V12-8, CONT PURGE VALVES

• V12-9, CONT PURGE VALVES

• V12-lO. CONTAINMENT PRESSURE RELIEF

• V12-ll, CONTAINMENT PRESSURE RELIEF

• V12-12. CONTAINMENT VACUUM RELIEF

• V12-13, CONTAINMENT VACUUM RELIEF

9. Return to procedure and step in effect.

NRC Exam

91. 001 A2.13 SRO 001 Given the following plant conditions:

-APP-004-E4, CV HI PRESS SFGRDJTRIP, flashes and is confirmed valid

-The reactor is at 100% RTP

-CETs are 1225°F and rising FRP-S.1, RESPONSE TO NUCLEAR POWER GENERATION/ATWS FRP-C.1, RESPONSE TO INADEQUATE CORE COOLING SACRM-1, SEVERE ACCIDENT CONTROL ROOM MANAGEMENT INITIAL

RESPONSE

Which ONE (1) of the following completes the statements below?

lAW FRP-S. 1, the crew unsuccessfully tried tripping the reactor by opening the (1) . The CRS is required to transition to (2)

A (1) Generator A & B Circuit Breakers (2) SACRM-1 B. (1) Generator A & B Circuit Breakers (2) FRP-C.1 C. (1) feeder breaker to 480V busses 2B and 3 (2) SACRM-1 D. (1) feeder breaker to 480V busses 2B and 3 (2) FRP-C.1 Tuesday, June 11,20132:10:26 PM 180

NRC Exam The correct answer is A A) Correct. The immediate actions of FRP-S.1 has you trip the Generator A & B Circuit Breakers. This is an effort to de-energize the control rods and let them fall into the core. The CRS will transition to SACRM-1 because CETs are >1200°F.

B) Incorrect. The immediate actions of FRP-S.1 has you trip the Generator A & B Circuit Breakers. This is an effort to de-energize the control rods and let them fall into the core. FRP-C.1 is incorrect. Plausible because you would have a red on CSFST for Core Cooling because CETs are >1200°F which sends you to FRP-C.1. You would not go there because Criticality is a higher priority red ball on CSFSTs.

C) Incorrect. Opening the feeder breaker to 480V busses 2B and 3 is incorrect.

Plausible because this would de-energize the Rod Drive Motor Generators. However, FRP-S.1 has you trip the feeder breaker FROM 480V 2b and 3 to the Rod Drive Motor Generators. The CRS will transition to SACRM-1 because CETs are >1200°F.

D) Incorrect. Opening the feeder breaker to 480V busses 2B and 3 is incorrect.

Plausible because this would de-energize the Rod Drive Motor Generators. However, FRP-S.1 has you trip the feeder breaker FROM 480V 2b and 3 to the Rod Drive Motor Generators. FRP-C.1 is incorrect. Plausible because you would have a red on CSFST for Core Cooling because CETs are >1200°F which sends you to FRP-C.1. You would not go there because Criticality is a higher priority red ball on CSFSTs.

Question: 91 Tier/Group: 2/2 K/A Importance Rating: SRO 4.6 K/A: 001 Control Rod Drive System A2: Ability to (a) predict the impacts of the following malfunction or operations on the CRDS- and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.13: ATWS Reference(s): Simulator/plant design, FRP-S.1 Proposed References to be provided to applicants during examination: None Learning Objective: Objective 5 of FRP-S.1 Lesson Plan Question Source: New Question History:

Question Cognitive Level: High 10 CFR Part 55 Content: 41.5/43.5/45.3/45.13 Comments:

This question meets the K/A because the stduent must know that the crew will try to de-energize the control rods because of the ATWS event in progress. Because they couldnt trip the rods, the CRS must know the procedure that the crew will use to mitigate this event.

Thursday, June 13, 2013 1:53:43 PM 181

Rev. 19 FRP-S.1 RESPONSE TO NUCLEAR POWER GENERATION/ATWS Page 4 of 21

STEP 1{ ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED CAUTION RCPs should NQ be tripped with Reactor Power GREATER THAN 5%.

NOTE Steps 1 ANP. 2 are Immediate Action steps.

1. Check REACTOR TRIP As Follows: Perform the following:

• REACTOR TRIP MAIN AN BYP a. Depress both Reactor Trip BKRs - OPEN Pushbuttons.

• Rod Position indication - b. jf Reactor Trip Breakers will ZERO open, ¶I perform the following

• Rod Bottom lights -

ILLUMINATED 1) Insert Control Rods.

• Neutron Flux - LOWERING 2) Dispatch an operator to the MG SET Room to trip the following breakers:

• REACTOR TRIP BREAKER A

• REACTOR TRIP BREAKER B

• GENERATOR A CIRCUIT BREAKER

• GENERATOR B CIRCUIT BREAKER

3) Dispatch an operator to 480V Busses 2B and 3 to trip the following breakers:

• ROD DRIVE MOTOR GENERATOR SET A

• ROD DRIVE MOTOR GENERATOR SET B

Rev. 19 FRP-S.1 RESPONSE TO NUCLEAR POWER GENERATION/ATWS Page 13 of 21 H STEP H ACTION/EXPECTED RESPONSE Hf RESPONSE NOT OBTAINED I

CAUTION A loss of DC power may occur if the DC busses are at maximum load and the battery chargers are not restarted within 60 minutes of a loss of all AC power.

20. Check Battery Chargers - IN SERVICE:

APP-036-Di, BATT CHARGER

• Locally Check In Service A/Al TROUBLE - EXTINGUISHED Battery Charger A A-i -

OPERATING In Service Battery Charger is operating, THEN restart tripped BATTERY CHARGER A A-i, using OP-60l DC Supply System.

while continuing with this procedure.

AND APP-036-D2, BATT CHARGER

• Locally Check In Service B/Bl TROUBLE - EXTINGUISHED Battery Charger B . B-i -

OPERATING.

In Service Battery Charger is IIQ operating, THEN restart tripped BATTERY CHARGER B B-i, using OP-60i DC Supply System.

while continuing with this procedure.

21. Check Core Exit TICs - LESS THAN Core Exit TICs are greater 1200°F than 1200°F rising, THEN Go To SACRM-i, Severe Accident Control Room Management -

Initial Response Core Exit T/Cs are greater than 1200°F lowering. II1I Go To Step 22.

CSF-2, CORE COOLING Ef\(FER Lo1 GREEN YELLOW ORANGE YELLOW ORANGE RED CSF-SAT GOTO GOTO GOTO GO TO GOTO FRP-C.3 FRP-C.2 FRP-C.3 FRP-C.2 FRP-C.1 CSFST Rev. 4 Page 4 of 9

NRC Exam

92. 068 A2.04 SRO 001 Given the following plant conditions:

-An approved radioactive liquid waste release is in progress

-Subsequently, R-18, LIQUID EFFLUENT WASTE DISPOSAL, is reading above its alarm setpoint

-The release is still in progress Which ONE(1) of the following completes the statements below?

R-1 8 reading above the alarm setpoint (1) have terminated the release. Per the ODCM, the release may continue provided (2)

(REFERENCES PROVIDED)

A (1) should (2) two independent samples are analyzed and two facility staff independently verify the release rate calculations and discharge line valving B. (1) should NOT (2) two independent samples are analyzed and two facility staff independently verify the release rate calculations and discharge line valving C. (1) should (2) the flow rate is estimated at lease once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases D. (1) should NOT (2) the flow rate is estimated at lease once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases Thursday, June 13, 2013 2:17:44 PM 182

NRC Exam The correct answer is A.

A. Correct. Since the monitor reading has exceeded its alarm setpoint, RCV-018 should have shut automatically. Per ODCM Table 2.6-1, two independent samples are analyzed and two facility staff independently verify the release rate calculations and discharge line valving.

B. Incorrect. R-18 should NOT have terminated the release is incorrect. Plausible if the student does not know that R-18 alarming automatically shuts RCV-018. Per ODCM Table 2.6-1, two independent samples are analyzed and two facility staff independently verify the release rate calculations and discharge line valving.

C. Incorrect. Since the monitor reading has exceeded its alarm setpoint, RCV-018 should have shut automatically. The flow rate being estimated every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is incorrect. Plausible if the student uses the wrong portion of the table from Table 2.6-1. This part is if the flow meter is not in service.

D. Incorrect. R-18 should NOT have terminated the release is incorrect. Plausible if the student does not know that R-18 alarming automatically shuts RCV-018. The flow rate being estimated every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is incorrect. Plausible if the student uses the wrong portion of the table from Table 2.6-1. This part is if the flow meter is not in service.

Question: 92 Tier/Group: 2/2 K/A Importance Rating: SRO 3.4 K/A: 068 Liquid Radwaste System (LRS)

A2: Ability to (a) predict the impacts of the following malfunctions or operations on the Liquid Radwaste System ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.04: Effluent release Reference(s): SIM/Plant Design, ODCM Table 2.6-1 Proposed References to be provided to applicants during examination: ODCM Table 2.6-1 Learning Objective: Objective 9 from RMS lesson plan Question Source: RNP Bank Question History:

Question Cognitive Level: High 10 CFR Part 55 Content: 41.5 / 43.5 / 45.3 / 45.13 Comments:

This question meets the K/A because the candidate must predict what should have happened to the release based off of R-1 8 being above the alarm setpoint, and from that, use the ODCM to control the release.

Thursday, June 13, 2013 2:17:44 PM 183

TAt3LE 2.6-I RADIOACTIVE LIQUID 5 Et F LUENT MONITORING INSTRUMENTATION Release Pathway? Instrumentation MCO Compensatory Measures Liquid Raclwante Effluent Discharge Line 1 With the nunibercf channels operable less than the MCO requirements:

a. Exert best efforts to return the inslruinerrts to operable status within

a. Monitor (R.18) 30 days and, if unsuccessfuL explain in the next Annual Radioacbve Effbent Release Report why the inopersbility was not corrected in a timely manner is accordance with Technical Specification 5,6.3 and,

b. Effluent releases via this pathway may conthrue prodded that prior to ir?tiating a release:

/

(J)c1/C I. Two independent samples are analyzed in accordance with the Surveillance Requirements of 00CM Specif cation 2,2,1 and;

2. Two members of the fadhitystaff Independendy verify the release rate calculabons and the discharge line valving.

b. Flow rule measurement deelce 1 With the number of channels operable less than the MCO requirement

a. Exert best efforts to return the instruments to operable status within 30 days and, if unsuccesalul, explain in the next Annual Radioaetr,e Effkient Release Report why the Inoperability was not corrected In a çQ,(1 timely manner kr accordance with Technical Specitcaticn 6.6.3 and,

b. Efflient relaasen xis this pathway may be continued, provided that the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump performance curves generated in situ and tank volumes may be used to eslinrale flow.

MCO Minhaum Channels Operable HI3RODCM 2-26 Rev. 28

ICOMPONENT DESCRIPTION R-18, Liquid Waste Disposal Effluent

1. Uses a MD-51 Gamma Scintillation Detector.

2. This channel continuously monitors all Waste Disposal System liquid releases from the plant.

3. The alarm setpoint is determined for each liquid waste release. Upon alarm, RCV 018 will automatically close.

4. The detector is mounted in a shielded assembly.

Figure 17 R-18; Liquid Waste Disposal Effluent Monitor 41

TABLE 2.61 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Release Pathway/ Instrumentation MCO Compensatory Measures Liquid Radwaste Effluent Discharge Line I With the number of channels operable less than the MOO requirements:

a. Exert best efforts to return the instruments to operable status within 30 days and if unsuccessful, explain in the next Annual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner in accordance with Technical Speciiication 6.6.3 and,

b. Effluent releases via this pathway may continue provided that prior to initiating a release:

1. Two independent samples are analyzed in accordance with the Surveillance Requirements of ODOM Specification 2.2.1 and:

2. Two members of the facility staff independently verify the release rate calculations and the discharge line valving.

b. Flow rate measLirement device 1 With the number of channels operable less than the MOO requirement:

a. Exert best efforts to return the instruments to operable status within 30 days and, if unsuccessful, explain in the next Annual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner in accordance with Technical Specification 5.6.3 and,

b. Effluent releases via this pathway may be continued, provided that the flow mte is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump performance curves generated in situ and tank volumes may be used to estimate flow.

MOO Minimum Channels Operable HBRODCM 2-25 Rev. 28

TABLE 2.6-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Release Pathway / Instrumentation MCO Compensatory Measures

2. Steam Generator Blomclown Effluent I per With the number of channels operable less than the MCO requirement:

Line S/G a. Exert best efforts to return the instruments to operable status within 30 days and, if unsuccessful, explain in the next Annual Radioactive Effluent

a. Monitor (R-19A.B. and C) Release Report why the i operability was not corrected in a timely manner in accordance vrth Technical Specification 5.6.3 and,

b. Effluent releases via this pathway may continue provided that grab samples are analyzed for gross radioactivity (beta or gamma> with a lower limit of detection of at least 1 .OE-O7pCiml or are analyzed for principle gamma emitters consistent with Table 2.8-1

1. Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the spedfic activity of the secondary coolant is 50.01 pCi/mI Dose Ecluivalent 1.131, or:

2. Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> aMen the spedftcactivity of the secondary coolant is >0 .01 pCi/mI dose Equivalent 1-131.

b. Flow rate measurement devices I per With the number of channels operable less than the MCO requirement due to inoperable

- each Steam Generator has its SIG equipment:

own blowclown Il ow rate a. Exert best efforts to relutn the instruments to operable status within 30 days measuring device. These and, if unsuccessful. explain in the next Annual Radioactive Effluent Release devices only measure flow Report why the inoperabilitywas not corrected in a timely manner in directed through the heat accordance with Technical Specification 5.6.3 recovery system, and will not measure flow which bypasses With the number of channels operable less than the MCO requirement due to inoperable fits heat recovery system. equipment, if the steam generator blowdown system is aligned such that any flow bypasses the flow measurement device(s) (i.e. heat recovery is not in service):

b. Effluent releases via this pathway may continue provided that the flow rate for the affected blowdown line(s) is estimated at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

MCO Minimum Channels Operable HBRQUCM 2-26 Rev. 28

TABLE 2.6-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Release Pathway/Instrumentation MCO* Compensatory Measures

2. Steam Generator Blowdown Effluent 1 per With the number of channels operable less than the MCO requirement due to Line (continued) S/G inoperable equipment:

c. R-1 9A, B and C flow a. Exert best efforts to return the instruments to operable status within measurement device each 30 days and, if unsuccessful, explain in the next Annual Radioactive monitor has its own flow rate Effluent Release Report why the inoperability was not corrected in a measurement device timely manner in accordance with Technical Specification 5,6,3 and,

b. Effluent releases via this pathway may continue provided that the flow rate for the affected monitor line(s) is estimated at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3. Discharge Canal Flow Note 1 With the number of channels operable less than the MCO requirement suspend effluent release via this pathway.

4. Tank Level Indicating Devices With the number of channels operable less than the MCO requirement:

a. Exert best efforts to return the instruments to operable status within 30 days and, if unsuccessful, explain in the next Annual

a. Refueling Water Storage 1 Radioactive Effluent Release Report why the inoperability was not Tank corrected in a timely manner in accordance with Technical Specification 56.3 and,

b. Monitor Tanks Tank A 1 b. Liqind additions to the affected tank(s) may continue provided that Tank B 1 the liquid level for the affected tanks is estimated during all liquid additions to the affected tank(s),

c. Waste Condensate Tanks TankC 1 TankD 1 TankE 1

d. Outside Temporary Tanks 1 per Tank (Note 2)

MCO Minimum Channels Operable HBRODCM 2-27 Rev. 28

TABLE 2.6-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Release Pathway/I nstrumeritation MCO Compensatory Measures

5. Containment Fan Cooling Water 1 With the number of channels operable less than the MCO requirement:

Monitor a. Exert best efforts to return the instruments to operable status (Service Water Effluent Line) within 30 days and, if unsuccessful, explain in the next Annual Radioactive Effluent Release Report why the inoperability was not

a. Monitor (R-16) does not provide corrected in a timely manner in accordance with Technical automatic termination of release Specification 5.6.3 and, upon exceeding alarm setpoint.

b. Effluent releases via this pathway may continue provided that, once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, grab samples are collected and analyzed for gross radioactivity (beta or gamma) with a lower limit of detection of at least 1 .OE-07 pCi/mI or are analyzed for principal gamma emitters consistent with Table 2.81.

6. Composite Samplerfor Settling Ponds 1 With the number of channels operable less than the MCO requirement:

a. Exert best efforts to return the instruments to operable status within 30 days and, if unsuccessful, explain in the next Annual Radioactive Effluent Release Report whythe inoperability was not corrected in a timely manner in accordance with Technical Specification 5.6.3 and,

b. Effluent releases via this pathway may continue provided that, grab samples are collected and composited three times per week and analyzed in accordance with Table 2.8-1.

02109 MCO Minimum Channels Operable

TABLE 2.6-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Release Pathway/Instrumentation MCO Compensatory Measures

7. Condensate Polisher Liquid Waste 1 With the number of channels operable less then tile MCO requirement:

Monitor a. Exert best efforts to return the instruments to operable status within 30 days and, if unsuccessful, explain in the next Annual

a. Monitor (R-37) provides Radioactive Effluent Release Report why the inoperability was not automatic termination of release corrected in a timely manner in accordance with Technical upon exceeding alarm/trip Spedfication 5.6.3 and, setpoint

b. Effluent releases via this pathway may continue provided that.

once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, grab samples are collected and analyzed for gross radioactivity (beta or gamma) with a lower limit of deteion of at least 1 .OE-07 pCi/mI or are analyzed for principal gamma emitters consistent with Table 2.8-1.

MCO Minimum Channels Operable HBRODCM 2-29 Rev. 28

NRC Exam

93. 071 G2.4.46 SRO 001 Given the foflowing plant conditions:

-The crew has started a gaseous waste release from WGDT C

-ATTACHMENT 10.3, GASEOUS WASTE RELEASE PERMIT-WASTE GAS DECAY TANK, is complete (SEE REFERENCE ON NEXT PAGE)

-During the release, R-14C, PLANT STACK, NOBLE GAS, read 12K cpm Which ONE(1) of the following completes the statements below?

The (1) had to approve the release before it could begin. Based off of ATTACHMENT 10.3, R-14C (2) alarm during the release.

A. (1) E&C Supervisor (2) did NOT B. (1) E&C Supervisor (2) did C. (1)SM (2) did NOT D (1) SM (2) did Thursday, June 13, 2013 2:21:20 PM 184

ATTACHMENT 10.3 Page 1 of 2 GASEOUS WASTE RELEASE PERMIT WASTE GAS DECAY TANK RELEASE NUMBER: I3O3 SSN: DATE: T This revisior is the latest revision available by:

M Name (Print) Initial Signature Dte PART I: PRE-RELEASE INFORMATION (E&C)

A I B i) 0 Estimated Release Start ic- -

Waste Gas Decay Tark Date Time (Circle Appropriate Letter) Estimated Release Sthp IDz31 Date Time Monitor Setpoint Basis (Circle One) CV Purge (Circle One) (NCR 410785)

R-14C - CPM EC In Service Not In Service Maximum WDT Flow Rate: ICX CFM I

PART II: RADIATION MONITOR INFORMATION (OPS arid E&C)

READING 2 R-14C PRIOR (Channel Check) CPM SOURCE CHECK 2 E&CINI.

SETPOINTVERF.AT 1c-j - CPM 4

UPDATESTATUSBOARD OPSINI.

DURING RELEASE CPM AFTER RELEASE CPM SETPOINTRETURNEDTO CPM I STATUS BOARDUPOATED 4 OPSINI. --

(N

NRC Exam The correct answer is D A) Incorrect. E&C Supervisor is incorrect. Plausible since they can approve certain releases, however, this is not one they can approve. R-14C did not alarm is incorrect.

Plausible if the student misuses attachment 10.3 and thinks that the setpoint in PART 1 for R-14C gets reset to 1.OOE+6 cpm. Since R-14C is already set at a more conservative value, 1.01 E+4 cpm, R-14Cs setpoint would not have changed for the release.

B) Incorrect. E&C Supervisor is incorrect. Plausible since they can approve certain releases, however, this is not one they can approve. R-14C did alarm during the release. R-14Cs setpoint for the release was 1.01 E÷4. The highest value it reached was 12K cpm, which exceeds its setpoint.

C) Incorrect. SM is the correct person who will approve this release. SM is the correct person who will approve this release. R-14C did not alarm is incorrect. Plausible if the student misuses attachment 10.3 and thinks that the setpoint in PART 1 for R-14C gets reset to 1 .OOE+6 cpm. Since R-14C is already set at a more conservative value, 1.OIE+4 cpm, R-14Cs setpoint would not have changed for the release.

D) Correct. SM is the correct person who will approve this release. R-14C did alarm during the release. R-14Cs setpoint for the release was 1.01 E+4. The highest value it reached was 12K cpm, which exceeds its setpoint.

Print out Attachment 10.3 from Z:\ILC-13 NRC Written Exam\SRO References\071 G2.4.46 SRO, titled release. They will need this sheet to answer the question.

Question: 93 Tier/Group: 2/2 K/A Importance Rating: SRO 4.2 K/A: 071 Waste Gas Disposal System (WGDS) 2.4.46 Ability to verify that the alarms are consistent with the plant conditions.

Reference(s): EMP-022 Proposed References to be provided to applicants during examination: Attachment 10.3 of EMP-022 Learning Objective:

Question Source: New Question History:

Question Cognitive Level: High 10 CFR Part 55 Content: 41.10 / 43.5 /45.3 /45.12 Comments:

This question meets the K/A because the student must determine if R-14C alarmed during a release based off of setpoints that could have changed on Attachment 10.3.

Asking who approves the release is an SRO level task and that is what makes this an SRO level question.

Thursday, June 13, 2013 2:32:28 PM 185

ATTACHMENT 10.3 Page 1 of 2 GASEOUS WASTE RELEASE PERMIT WASTE GAS DECAY TANK RELEASE NUMBER: SSN: DATE:

This revision is the latest revision available as verified by:

Name (Print) Initial Signature Date PART I: PRE-RELEASE INFORMATION (E&C)

A I B I C I D Estimated Release Start Waste Gas Decay Tank Date Time (Circle Appropriate Letter) Estimated Release Stop Date Time Monitor Setpoint Basis (Circle One) CV Purge (Circle One) (NCR 410785)

R-14C CPM EC Activity In Service Not In Service Maximum WGDT Flow Rate: CFM This release can be made within the limits of IOCFR2O and IOCFR5O using the setpoints and restrictions stated herein.

Prepared By: Peer Review:

Verified By Only required if performing a WGDT gaseous waste release with R-14C out of service.

E&C Supervisor:

PART II: RADIATION MONITOR INFORMATION (OPS and E&C)

READING 2 R-14C 1 (Channel Check)

PRIOR CPM SOURCE CHECK 2 E&C INI.

SETPOINT VERF. AT 3 CPM UPDATE STATUS BOARD4 OPS INI.

DURING RELEASE CPM AFTER RELEASE CPM SETPOINT RETURNED TO 5 3 CPM STATUS BOARD UPDATED 4 OPS INI.

Approved for Release: (CR 97-00059)

Shift Manager EMP-022 Rev. 54 Page 52 of 60

NRC Exam

94. G2.1.38 SRO 001 Which ONE (1) of the following completes the statement below?

The CRS is responsible for conducting a(n) every 45 60 minutes during long lasting events.

A. Crew Update B. Alignment Brief C. Crew Shift Brief Dv Plant Status Brief The correct answer is D.

A. Incorrect. Plausible because it is one of the listed types of briefs in OPS-NGGC-1314 and the title of the brief does not suggest when it is performed.

B. Incorrect. Per OPS-NGGC-1314, Alignment briefs should be performed when transitioning between event procedures.

C. Incorrect. Plausible because it is one of the listed types of briefs in OPS-NGGC-1314 and the title of the brief does not suggest when it is performed.

D. Correct. A Plant status brief is conducted lAW OPS-NGGC-1314 ever 45-60 minutes during long lasting events.

Question: 94 Tier/Group: 3 (SRO)

K/A Importance Rating: RO 3.7 SRO 3.8 K/A: 2.1 Conduct of Operations 2.1.38 Knowledge of the stations requirements for verbal communication when implementing procedures.

Reference(s): OPS-NGGC-1 314 Proposed References to be provided to applicants during examination: None Learning Objective:

Question Source: New Question History:

Question Cognitive Level: Low 10 CFR Part 55 Content: 43.5 Comments:

This question meets the K/A because the candidate needs to know the station expectations for communicating information during implementation of AOP and EOPs.

Thursday, June 13, 2013 2:36:52 PM 186

9.4.2 Middle-of-Shift Brief

. RLO(1

1. A middle-of-shift brief of the CRS, BOP and AOs shall be conducted. The purpose of this meeting is to discuss remaining I ongoing activities for the remainder of the shift, identify roadblocks, and assure a successful error-free outcome. The middle-of-shift brief may be waived at the discretion of the CRSISM.

9.4.3 Crew Update

1. A Crew Update is to be used when a time-critical, status-critical or a key piece of information would be beneficial for the crew to know.

2. When an update is used, it should contain the following elements:

a. Announcement of Crew Update

b. Ensure all affected personnel are attentive signified by each raising a hand.

c. Concise statement of the critical information.

d. Announcement End of Update.

3. Any member of the control room staff may use a Crew Update to communicate significant changes or discoveries to the entire control room staff. For example:

Crew update. Maintenance has completed setting the head on the vessel. End of update.

4. The following are examples of when a Crew Update should be considered:

a. Announcement of Emergency Action Level classification

b. Significant changes in plant status (adverse containment values in effect, off site power restored, etc.).

c. Trip of a critical component.

d. Significant milestones in an evolution OPS-NGGC-1 314 Rev. 1 Page 18 of 45

/

9.4.5 Alignment Briefs

a. Alignment Briefs are shorter in duration and less formal than Plant Status Briefs. Alignment Briefs should last for less than 30 seconds when possible.

The purpose of an Alignment Brief is to ensure that crew members are aligned when time does not permit the performance of a formal Plant Status Brief, or when a formal Plant Status Brief is not warranted.

b. Alignment Briefs are held at the discretion of the CRS/SM. Alignment Briefs can be requested by any crew member.

c. Alignment Briefs should be used at the following times:

1) When entering or transitioning between event procedures.

2) When any crew member has a question related to mitigation strategy.

3) After an event in which plant response is not understood.

4) Prior to taking actions that will have a significant impact on plant operations.

d. When an Alignment Brief is used, it should contain the following elements:

1) CRS states Attention in the Control Room for an Alignment Brief

2) Ensure all affected personnel are attentive, signified by each raising a hand.

3) Discuss information as required for alignment. The discussion may include input/questions from crew members, provided that the discussion remains brief. Crew member input is not required.

4) The CRS announces End of brieV.

EXAMPLE Alignment Briefs Example 1: Attention in the Control Room for an Alignment Brief. We are entering AP 510, Rapid Power Reduction, to lower power to 70% due to elevated vibrations on Reactor Coolant Pump A. Ted- you are going to manually maintain level in the A Steam Generator between 40% and 60% due to the feed reg valve failure. Any questions or comments? End of brief.

Example 2: Attention in the Control Room for an Alignment Brief. We will be tripping the Reactor due to a malfunctioning Main Feedwater Pump. OAC, you will trip the Reactor and perform immediate actions. BOP, you will initiate Emergency Feedwater and trip Main Feedwater Pumps. Any questions or comments? End of brief.

Example 3: Attention in the Control Room for an Alignment Brief. Do we understand why Reactor Coolant pressure is lowering? ... [Brief discussion with crew members] ... End of brief.

OPS-NGGC-1314 Rev. I Page 21 of 45

9.4 Briefs 941 CrewShiftBrief Standards a A Shift Brief is conducted at the beginning of each operating shift following the watch station turnovers This briefing is expected to be free of distractions in order to facilitate the effective communication of plant conditions.

b. Personnel of the on-coming shift attend and participate in this briefing. It is the Shift Managers responsibility to make sure that the meeting has a clearly defined beginning and end, and that all participate in the information exchange.

c. Communication during the briefing focuses on recent or pending status changes.

d. The shift briefing is led by the duty Shift Manager, who establishes an atmosphere of open communication that is free from unnecessary distractions (side conversations, phone calls, etc.); during the brief, Operations standards for formality and professionalism are maintained.

2. Expectations

a. The Shift Brief is lead by the Shift Manager or designee and shall be normally conducted outside the Control Room.

b. The CRS will ensure the telephone access is restricted in the Control Room prior to the brief and will be restored at the conclusion as required.

c. The OAC will participate in the brief to the extent that it does not distract from the primary responsibility to monitor the reactor.

d. All participants must speak loudly enough to be heard over any background noise.

e. At the discretion of the SM/CRS, the crew briefing may be stopped if unusual conditions exist which demand the prompt attention of the operating crew. The SM/CRS resumes the brief when the situation is again conducive to an effective exchange of information.

f. Attachment 1, Shift Brief Checklist should be used as a guide for the shift brief to ensure all pertinent information is discussed.

g. SM should review the crew composition for the shift. This review considers the crew makeup for individuals who do not routinely work or train together as a crew. Based on the results of this assessment compensatory measures are taken as appropriate.

OPS-NGGC-1314 Rev. I Page 17 of 45

9.4.4 Plant Status Briefs

1. Plant Status Briefs are conducted for the purpose of bringing Control Room personnel to the same level of understanding of the present plant status, overall mitigation plan, and emergency plan status.

2. Plant Status Briefs are held at the discretion of the CRS and are normally held following plant stability during emergency procedures. Additionally, Plant Status Briefs should be held:

a. After mitigation of a major symptom.

b. At the request of Control Room personnel.

c. When exiting event procedures.

d. Prior to performing complex/significant evolutions. [R4]

e. Every 45 60 minutes during long lasting events.

3. Plant Status Briefs shall not interfere with the performance of time critical actions or immediate actions of event procedures.

4. ROs shall continue to maintain awareness of plant parameters during the brief.

5. The CRS should notify crew members of the intent to conduct a Plant Status Brief I 2 minutes before the brief occurs.

6. A Plant Status Brief is performed by the CRS utilizing the following format:

a. B BEGIN:

• Announce Attention in the Control Room for a Plant Status Brief

• Ensure all affected personnel are attentive, signified by each raising a hand

b. R RECAP:

• Provide overview of sequence of events and current plant status

• Identify major equipment failures

• Identify procedures that are in effect

• Discuss status of actions performed outside the MCR OPS-NGGC-1 314 Rev. I Page 19 of 45

NRC Exam

95. G2.2.13 SRO 001 Given the following plant conditions:

- The letdown line has been removed from service and cleared for maintenance to install several new vent valves.

- An approved test procedure has been provided with the work order package to perform the line hydrostatic test following installation.

- A clearance boundary change must be implemented to initiate the system hydrostatic test.

Which ONE (I) of the following completes the statements below?

lAW OPS-NGGC-1 301, Equipment Clearance, each boundary change shall be authorized by (1) . The (2) will provide concurrence for Maintenance personnel to introduce fluids into the clearance boundary for the system hydrostatic test.

A. (1) anSRO (2) WCC SRO B. (1) the CRS only (2) CRS C(1) anSRO (2) CRS D. (1) the CRS only (2) WCC SRO The correct answer is C.

A. Incorrect. SRO is correct for boundary change but WCC SRO is incorrect for the concurrence.

B. Incorrect. Maintenance Supervisor is incorrect for the boundary change.

C. Correct. Reference OPS-NGGC-1 301 Section 9.1.21 and 9.2.4.1.5. SRO is required for the boundary change and CRSS is required for the concurrence to introduce the fluids into the system.

D. Incorrect. Maintenance Supervisor is incorrect for the boundary change and WCC SRO is incorrect for the concurrence.

Tuesday, June 11,20132:10:26 PM 187

NRC Exam Question: 95 Tier/Group: 3 (SRO)

K/A Importance Rating: RO 4.1 SRO 4.3 K/A: 2.2 Equipment Control 2.2.13 Knowledge of tagging and clearance procedures.

Reference(s): OPS-NGGC-1 301 Proposed References to be provided to applicants during examination: None Learning Objective:

Question Source: RNP Bank Question History:

Question Cognitive Level: Fundamental Knowledge 10 CFR Part 55 Content: 43.3 Comments:

This question meets the K/A because it tests the candidate on SRO responsibilities contained in the Equipment Clearance procedure.

Thursday, June 13, 2013 2:39:08 PM 188

Subsection 9.1, General Administration (Contd)

41. If an Emergent Equipment Clearance is needed and the clearance computer program is unavailable, the appropriate attachments in the back of this procedure may be used to manually prepare the clearance.

A Clearance Log Sheet (Attachment 1) will be filled out for tracking purposes starting with clearance number YY-XXXX. (YY current year and XXXX will be sequential numbered starting at 0001). Once the computer system has been restored, all information should be entered into the computer.

Any tags that are still hanging should be replaced with the computer generated tags.

The manually generated clearance forms should be saved until the computer generated clearance has been closed and all pertinent information captured.

42. Specific Plant Addendums should be referenced for all employees working under this procedure.

43. Attachment 10, Clearance Process Checklist, is required to be used for all clearances generated in conjunction with this procedure.

44 The introduction of fluids either gas or liquid, into a system within a clearance boundary may be performed by work groups other than Operations provided it is according to an approved procedure and with the concurrence of the CRS or SM The location at which the fluid is to be introduced into the system shall be specified in the procedure.

45. Any approved clearance older than 6 months, must be evaluated to ensure that it meets the intent of the current revision to this procedure. If not, it must be re-written using the current revision.

46. When equipment that operates at temperatures above ambient is removed from service, it should be allowed to cool prior to starting work or precautionary measures, such as wearing the appropriate PPE, shall be taken to eliminate the potential burn hazard.

OPS-NGGC-1301 Rev.30 Page27of 113

9.5 Clearance Development: Lift Checklist

1. During Lift Checklist development, a Clearance Preparer shall determine:

• Restoration positions. The restoration position should be based on current plant status / conditions. This may include reviewing applicable procedure steps and/or valve lineups to determine plant status.

• Specify the proper restoration sequence, paying particular attention to restoring in an order to prevent personnel injury or equipment damage.

• Independent Verification requirements using OPSNGGC-1 303

2. If relying on a system alignment for restoration, the procedure and section or valve lineup being relied on should be specified in the Special Instructions. [R18]

NOTES: 1. Boundary Change Checklists may be prepared prior to receiving Attachment 6, Boundary Change Form.

2. Steps 3.0 and 4.0 of Attachment 6 may be performed concurrently.

3. Hang and lifts shall be on separate checklists unless there are no holders.

9.6 Clearance Development: Boundary Change

1. Use Attachment 6, Boundary Change Form, for all boundary changes.

2. A Clearance Preparer shall develop a boundary change in accordance with Attachment 10 and identifies all affected clearance orders.

3. A second Clearance Preparer who is a Licensed Operator shall verify a boundary change is adequate.

4. Each boundary change shall be authorized by an SRO. This authorization is indicated by the Checklist Status being Distributed and indicates the SRO has verified the following:

• Plant conditions are correct for the boundary change 0 The boundary change will not adversely impact plant operation

• Applicable compensatory actions have been initiated

• The Control Room has been notified, as necessary

5. The boundary change is then assigned to a Tag Hanger for implementation. The Tag Hanger shall initial each step and print their Passport short name at least once with their initials for each Checklist.

6. Workers and holders shall perform Zero Energy Checks after any change to the clearance.

OPS-NGGC-1 301 Rev. 30 Page 36 of 113

NRC Exam

96. G2.2.I8SR0001 Given the following plant conditions:

- Two trains of RHR are OPERABLE

- RCS is at 135°F and has been drained to -30 inches standpipe level for RCP seal replacement

- Maintenance has requested that both CV Personnel Hatch doors be opened to allow cables and hoses to be routed through the doors to support maintenance activities Which ONE (1) of the following completes the statements below?

lAW OMM-033, IMPLEMENTATION OF CV CLOSURE, CV Personnel Hatch doors can be opened provided CV closure can be implemented within (1) hours. The hoses and cables can be routed through the hatch provided (2) for removal.

A (1) 0.5 (2) the hoses and cables have quick disconnects B. (1)4 (2) the hoses and cables have quick disconnects C. (1)0.5 (2) dedicated personnel are stationed inside the CV with tools needed D. (1)4 (2) dedicated personnel are stationed inside the CV with tools needed Tuesday, June 11,20132:10:26 PM 189

NRC Exam The correct answer is A.

A. Correct. Per OMM-033, CV closure time for all penetrations is 30 minutes with the conditions provided in the stem. Hoses and cables blocking the door must be capable of immediate removal using quick disconnects, clamps, etc..

B. Incorrect. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is the maximum time of isolation for the Equipment Hatch.

Plausible because the Personnel Hatch is a smaller penetration and it could be assumed that because it has a smaller release area that it could be assigned the longer duration for isolation.

C. Incorrect. Per OMM-033, CV closure time for all penetrations is 30 minutes with the conditions provided in the stem. Penetrations must be capable of being isolated from outside containment so personnel if required to be stationed would not be stationed inside containment. Plausible because there are requirements in some cases to have personnel stationed at the penetration and a lot penetrations can be isolated from either side.

D. Incorrect. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is the maximum time of isolation for the Equipment Hatch.

Plausible because the Personnel Hatch is a smaller penetration and it could be assumed that because it has a smaller release area that it could be assigned the longer duration for isolation. Penetrations must be capable of being isolated from outside containment so personnel if required to be stationed would not be stationed inside containment. Plausible because there are requirements in some cases to have personnel stationed at the penetration and a lot penetrations can be isolated from either side.

Question: 96 Tier/Group: 3 (SRO)

K/A Importance Rating: RO 2.6 SRO 3.8 K/A: Equipment Control G2.2.1 8: Knowledge of the process for managing maintenance activities during shutdown operations, such as risk assessments, work prioritization, etc..

Reference(s): OMM-033 Proposed References to be provided to applicants during examination: None Learning Objective: RCSOI2 Question Source: RNP Bank Question History: 09 NRC EXAM Question Cognitive Level: Comprehension 10 CFR Part 55 Content: 43.5 Comments: This question meets the K/A because the candidate must know the requirements for allowing maintenance to perform work and the restrictions required when shutdown.

Thursday, June 13, 2013 2:47:10 PM 190

UOS \S on c-QUESTIONS REPORT for ILC-09 NRC Written Exam Questions

1. 02.2.18 002 Given the following:

- Unit in Mode 5 for RCP seal replacement.

- RCS is at 135°F and has been drained to -30 inches standpipe level.

- Reactor has been shutdown for 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br />.

- Maintenance has requested that both CV Personnel Hatch doors be opened to allow cables and hoses to be routed through the doors to support maintenance activities.

Which ONE (1) of the following describes the requirements that must be met for the CV Personnel Hatch lAW OMM-033, Implementation of CV Closure?

CV Personnel Hatch doors A. cannot be opened until 67 hours7.75463e-4 days <br />0.0186 hours <br />1.107804e-4 weeks <br />2.54935e-5 months <br /> 5 minutes have elapsed since reactor shutdown.

B can be opened provided CV closure can be implemented within 30 minutes and the hoses I cables are provided with quick disconnects for removal. Closure must be implemented from outside of the CV.

C. can be opened provided CV closure can be implemented within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and the hoses I cables are provided with quick disconnects for removal. Closure must be implemented from outside of the CV.

D. can be opened provided CV closure can be implemented within 30 minutes.

However, the hoses I cables that impede the closing of either door cannot be allowed.

Thursday, June 13, 2013 2:45:31 PM

3.5 It is the responsibility of the Work Group Supervisor to [SOER 09-1, Recommendation 11]:

3.5.1 Prepare the CV Closure Exception Permit.

3.5.2 Provide knowledgeable personnel on a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> basis, available on site AND properly briefed, to secure all penetrations held open under this procedure that are his direct responsibility.

NOTE: Power or air operated tools are acceptable, however closure shall be capable of being performed entirely from outside the CV (except for the Fuel Transfer Tube lAW Section 8.2) with only manual hand tools in the event that air or electrical power is NOT available.

3.5.3 Ensure that the following are available at the site of the open penetration:

A device providing a positive form of closure for the open penetration, capable of withstanding at least 19 psia CV pressure, and being installed in a timely manner.

All tools required to install the closure device.

3.5.4 Ensure that the tools and materials required to isolate each open penetration are inspected each shift and discrepancies corrected.

3.5.5 Ensure that the personnel designated to secure penetrations, are familiar with the installation method for all types of closure devices that fall under their responsibility.

3.5.6 Ensure that personnel who are responsible for closure of penetrations are familiar with this procedure.

3.5.7 Ensure the lead person designated to secure the penetration on each shift signs the Attachment 10.3, Penetration Closure Responsibility Log, prior to OR during turnover before the off-going designated person leaves the site. These actions should be completed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of watch relief.

3.6 The work leader for the particular activity is responsible for having the tools, materials, and manpower required to isolate the penetration at the work site and for notifying the WCC of any turnover of responsibility.

3.7 Individuals designated as responsible for verifying penetrations are closed are responsible for notifying the WCC SRO to identify another individual if job responsibilities or temporary absences prevent performing the function.

OMM-033 Rev. 30 Page 7 of 41

5.15 In the event that the Equipment Hatch has been removed AND the CV Purge is lost due to reasons NOT involving a loss of a safety function (See OMM-046) OR a fuel handling accident is NOT in progress, THEN deployment of the Equipment Hatch Membrane is NOT required.

5.16 An aggregate impact assessment shall be performed by the WCC SRO using Attachment 10.5 when a new CV Closure Exception Permit is initiated OR a change in status of an existing CV Closure Exception Permit occurs.

5.17 Pre-outage scheduling and risk reviews should have considered aspects of open penetrations aggregate assessment. However, emergent issues or delayed activities may impact the aggregate assessment, and the number of penetrations open at any one time may be limited by the resources or methods available to effect closure.

5.18 A CV purge will remain in operation during any core alteration or movement of irradiated fuel assemblies while the equipment hatch is removed, even if time since shutdown is greater than 4 days.

5.19 Plant Aligned for Natural Circulation Table Plant Aligned for Time After Time to Boil Equipment Hatch Natural circulation Shutdown Yes 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> N/A May be removed and the runway installed.

4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Permit. 16 bolts must be installed and torque to 397 ft-lbs in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. (CV Closure Exception Permit for CV Equipment Hatch with Outside Runway)

No <4 days N/A Must be installed.

No 4 days < 41 minutes Must be installed.

No 4 days 41 minutes May be removed but must be installed by Time AND < 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to Boil. 4 bolts must be installed and torque to 397 ft-lbs in Time to Boil. (CV Closure Exception Permit for CV Equipment Hatch)

No 4 days 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> May be removed and the runway installed.

4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> permit. 4 bolts must be installed and torque to 397 ft-lbs in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. (CV Closure Exception Permit for CV Equipment Hatch with Outside Runway)

OMM-033 Rev. 30 Page 10 of 41

Section 8.1 Page 1 of I REFERENCE USE 8.0 INSTRUCTIONS 8.1 Determining Penetration Closure Times NOTE: CV Closure time is NOT applicable with the core fully off loaded to the Spent Fuel Building.

8.1.1 IF the following conditions are satisfied, THEN allowed CV Closure Time is 30 minutes for all penetrations except the CV Equipment Hatch:

1. Two Trains of RHR are OPERABLE AND Reactor Coolant System average temperature is less than or equal to 200°F Reactor Coolant System level is above -36 inches OR

2. One Train of RHR OPERABLE with refueling cavity level between 16 and 29 inches as indicated on the Refueling Cavity Level Indicator AND Reactor Coolant System average temperature is less than or equal to 200°F.

3. IF NEITHER condition above is met, THEN CV Closure Time shall be determined from Plant Curve 3.5, Time to CV Closure, in the Station Curve Book.

8.1 .2 IF the estimated time to close the open penetration exceeds 30 minutes, THEN an evaluation should be performed for the open penetration AND Operations Manager approval on Attachment 10.1 will be required.

8.1.3 CV Personnel Hatch

1. WHEN opening the CV Personnel Hatch, THEN at least one of the doors will be capable of being closed.

2. Any equipment impeding the closing of one of the doors shall be located in such a way that it can be immediately removed from the opening through the use of quick disconnects, clamps, etc.

8.1 .4 IF the equipment hatch will be removed, THEN refer to Precaution and Limitation 5.18 for determination of required closure time OMM-033 Rev.30 Page 12of41

NRC Exam

97. G2.3.13 SRO 001 Given the following plant conditions:

- A Site Area Emergency has been declared due to a LOCA outside containment

- All Emergency Response Organization facilities are staffed and functional

- The TSC has determined that the leak can be isolated

- Expected exposure to isolate the leak is> 5 Rem TEDE

- An operator has been briefed and is awaiting approval for entry into the Auxiliary Build ing to isolate the leak Which ONE (1) of the following completes the statement below?

Authorization of the exposure (1) be delegated to the Radiological Control Director. If authorized, the operator is allowed to receive up to (2) REM for this entry.

A. (1) can (2) 10 B. (1) can (2) 25 C (1) can NOT (2) 10 D. (1) can NOT (2) 25 The correct answer is C A) Incorrect. This is the wrong authorizer. Plausible since the person can perform many of the same functions with concern to radiological controls as the Site Emergency Coordinator lAW EPOSC-04. However, this person cannot authorize individuals to exceed 5 REM TEDE in a year. 5 REM is incorrect. Plausible since this is the maximum limit workers can receive for normal activities.

B) Incorrect. This is the wrong authorizer. Plausible since the person can perform many of the same functions with concern to radiological controls as the Site Emergency Coordinator lAW EPOSC-04. However, this person cannot authorize individuals to exceed 5 REM TEDE in a year. 10 REM is correct for repair efforts during a casualty.

C) Incorrect. This is the correct authorizer. 5 REM is incorrect. Plausible since this is the maximum limit workers can receive for normal activities.

D) Correct. The SEC can authorize this entry and approve the individual exceeding the 5 EM TEDE in a year.

Tuesday, June 11,20132:10:26 PM 191

NRC Exam Question: 97 Tier/Group: 3 K/A Importance Rating: RO 3.2 SRO 3.7 K/A: G2.3.4: Knowledge of radiation exposure limits under normal or emergency conditions.

References:

EPOSC-04, EPCLA-1 Proposed references to be provided to applicants during the Exam: None Learning Obj:

Question Source: RNP bank (Modified G2.3.4 002)

Question History: No NRC Exams Question Cognitive Level: Low 10 CFR part 55: 43.4 Meets the K/A because the student must know radiation exposure limits under emergency conditions and requirements for approval.

Thursday, June 13, 2013 2:49:44 PM 192

2.21 RNP-RA105-0082, Response to NRC Bulletin 2005-02, Emergency Preparedness and Response Actions for Security-Based Events 2.22 NUREG-0654, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants 2.23 EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents, U.S. Environmental Protection Agency, Washington, D.C., May 1992 2.24 NRC Regulatory Issue Summary 2005-08, Range of Protective Actions For Nuclear Power Plant Incidents 2.25 NRC Interim Compensatory Measures Order, Section B.5.b 2.26 NRC Regulatory Issue Summary 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events 2.27 NEI 99-01 Revision 4, Methodology for Development of Emergency Action Levels 2.28 CAPR 258209, GE Declaration during Graded Exercise 2.29 EMG-NGGC-0005, Activation Of The Emergency Response Organization Notification System 2.30 EPEOF-lO, Recovery Manager And Recovery Operations 2.31 NSIR/DPR-ISG-01, Interim Staff Guidance 2.32 EPRAD-03, Dose Projections 2.33 NRC RIS Issue Summary 2008-26, Clarified Requirements of Title IOCFR5O.54(y)

When Implementing IOCFR5O.54(x) to Depart From a License Condition or Technical Specification 3.0 RESPONSIBILITIES 3.1 The Site Emergency Coordinator (SEC) has immediate and unilateral authority to implement this procedure.

3.2 The SEC may not delegate:

3.2.1 The decision to notify offsite authorities; 3.2.2 Making offsite Protective Action Recommendations (PAR);

3.2.3 Classifying or terminating the emergency; 3.2.4 Authorizing exposures in excess of 10 CFR 20 limits during a declared emergency.

EPCLA-01 Rev. 35 Page 10 of 74

8.6.6 IF the monitored dose rates OR stay times encountered during the entry exceed the limits set forth for the mission, THEN the team will immediately communicate with OSC and seek further direction.

8.6.7 Once the mission has been completed, the team will follow established monitoring AND personnel decontamination procedures OR take other actions specified by the ROD.

8.7 Emergency Worker Dose Limits 8.7.1 Regulatory limits shall be observed for planned radiation exposures to emergency workers UNLESS the SEC AUTHORIZES the individual to exceed 5 Rem TEDE in a year.

8.7.2 The table shown below identifies the Emergency worker dose limits. In addition to the categories listed in the table doses should be limited as follows:

1. The lens of the eye should be limited to three (3) times the stated TEDE value.

2. Any other organ (including skin and body extremities) should be limited to ten (10) times the stated TEDE value.

Dose Limit Activity Condition 5REM All 10 REM Repair and reentry Lower dose not efforts practicable 25 REM Lifesaving or protection Lower dose not of large populations practicable

>25 REM Lifesaving or protection Only on a voluntary of large populations basis to persons fully aware of the risks involved 8.7.3 There may be situations where saving a life is not the issue, AND it is necessary to enter a hazardous area under repair/reentry efforts, to protect valuable installations, OR to make the facility more secure against events which could lead to radioactivity releases (e.g.,

assessment actions, entry of damage repair parties who are to repair valve leaks, OR add iodine-fixing chemicals to spilled liquids).

EPOSC-04 Rev. 10 Page 9 of 22

NRC Exam

98. Given the following plant conditions:

-A Large Break LOCA has occurred

-Containment pressure is 32 psig and rising

-HVH-1 and HVH-2 have tripped

-Containment Sump level is rising

-R-32A, CV High Range, has the current indication:

Which ONE (1) of the following completes the statement below?

Based on the information provided a (1) must be declared due to (2)

(References Provided)

A. (1) Site Area Emergency (2) Loss of the Reactor Coolant AND Fuel Cladding Barriers B. (1) General Emergency (2) Loss of the Reactor Coolant, Fuel Cladding AND Containment Barriers C. (1) Site Area Em&gency (2) Loss of the Reactor Coolant Barrier AND a Potential Loss of the Containment Barrier D. (1) General Emergency (2) Loss of the Reactor Coolant AND Fuel Cladding Barrier AND Potential Loss of the Containment Barrier Thursday, June 13, 2013 5:13:02 PM 98

NRC Exam The correct answer is A A) Correct. A SAE will be declared. R-32A indicates approximately 300 R/hr. This correlates to a loss of Fuel Cladding Barrier due to item #3 and a loss of Reactor Coolant System Barrier due to item #1. This gives you a loss of any two barriers per FS1.1.

B) Incorrect. A GE is incorrect. Plausible because R-32A indicates approximately 300 R/hr. This correlates to a loss of Fuel Cladding Barrier due to item #3 and a loss of Reactor Coolant System Barrier due to item #1. If the student misinterprets sump level rising for response not consistent with LOCA conditions (#2 of Containment Barrier),

this would get them to a GE.

C) Incorrect. A SAE will be declared. R-32A indicates approximately 300 R/hr. This correlates to a loss of Fuel Cladding Barrier due to item #3 and a loss of Reactor Coolant System Barrier due to item #1. The Potential loss of containment barrier is plausible if the student mistakes HVH-1 and HVH-2 tripping for < one full train of depressurization equipment operating. CV spray is still operating and so is the B train of HVH.

D) Incorrect. A GE is incorrect. Plausible because R-32A indicates approximately 300 R/hr. This correlates to a loss of Fuel Cladding Barrier due to item #3 and a loss of Reactor Coolant System Barrier due to item #1. The Potential loss of containment barrier is plausible if the student mistakes HVH-land HVH-2 tripping for <one full train of depressurization equipment operating. CV spray is still operating and so is the B train of HVH.

Question: 98 Tier/Group: 3 K/A Importance Rating: SRO 2.9 K/A: Radiation Control G2.3.5: Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Reference(s): Sim/Plant design, Hot Conditions EAL Matrix Proposed References to be provided to applicants during examination: Hot conditions EAL Matrix Learning Objective: Objective EAL-004 Question Source: RNP Bank Question History: ILC 11-1 NRC exam. This question was modified so a previously wrong answer is now correct. Removed the loss of off-site power from the stem.

Removed the loss of B EDG. Added HVH-2 tripping as well as HVH-1. Also added containment sump level status to aid in the distractors.

Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 41.11 / 41.12 / 43.4 / 45.9)

Comments:

This question meets the K/A because the student must use the rad monitoring system (R-32A) to aid in making an EAL Classification which is an SRO task. The EAL Classification is what makes this an SRO question.

Thursday, June 13, 2013 2:54:30 PM 194

NRC Exam

99. G2.4.23 SRO 001 Given the following plant conditions:

-The crew is in EPP-28, LOSS OF ULTIMATE HEAT SINK, for a total loss of the intake structure

-Subsequently a loss of all AC power occurs Which ONE (1) of the following completes the statement below?

The CRS (1) transition to EPP-1, LOSS OF ALL AC POWER, because (2)

Av (1) will NOT (2) EPP-28 contains actions to deal with a loss of all AC power B. (1) will (2) EPP-1 contains actions to deal with a loss of all AC power C. (1) will (2) EPP-1 has priority over all other EPP procedures D. (1)will NOT (2) EPP-28 has priority over all other EPP procedures The correct answer is A A) Correct. The CRS will not transition to EPP-1. EPP-28 has a note in the beginning of it saying that it has higher priority over EPP-1. The basis for this is that EPP-28 has actions to take care of a loss of all ac.

B) Incorrect. The CRS will transition to EPP-1 is correct. Plausible because EPP-1 is a higher priority EPP than all other EPPs except EPP-28. EPP-1 does not have priority over all procedures. Plausible because EPP-1 does contain the actions to deal with a loss of all AC power.

C) Incorrect. The CRS will transition to EPP-1 is incorrect. Plausible because EPP-1 is a higher priority EPP than all other EPPs except EPP-28. EPP-1 does not have priority over only the EPP procedures. Plausible because in most procedures, you cannot follow their mitigation strategy without power to the components. EPP-28 is the one exception that takes into account a loss of all ac power.

D) Incorrect. The CRS will not transition to EPP-1 is correct. EPP-28 does not have priority over all the EPP procedures, plausible because there are two procedures that it doesnt have priority over, EPP-21/25. EPP-28 sends you to EPP-21 to energize your pressurizer heaters and EPP-25 to energize supplemental plant equipment.

Tuesday, June 11,20132:10:27 PM 195

NRC Exam Question: 99 Tier/Group: 3 K/A Importance Rating: SRO 4.4 K/A: Emergency Procedures I Plan G2.4.23: Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations.

Reference(s): Sim/Plant design,EPP-28 and its basis document Proposed References to be provided to applicants during examination: None Learning Objective: Objective 4 of EPP-28 Lesson Plan Question Source: New Question History:

Question Cognitive Level: Low 10 CFR Part 55 Content: (CFR: 41.10 /43.5 /45.13)

Comments:

This question meets the K/A because the student must prioritize which EPP to use and know the reason why(basis) to use that procedure. This question is SRO level because they must decide which procedure to use to mitigate this event.

Thursday, June 13, 2013 3:10:00 PM 196

Rev. 13 EPP-28 LOSS OF ULTIMATE HEAT SINK Page 3 of 215 STEP_H ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED I

CAUTION Subsequent steps may require access to Non-vital Areas. Access to these areas prior to nullification of the threat could cause loss of personnel.

NOTE

• FRPs EPP Foldouts EPP 1 AOP-014 AOP-020 AND AOP-022 are NOT applicable for this event Transition use of these procedures should be made unless otherwise directed in this procedure.

• Any time MCC-6 is deenergized, coordination of field activities will require the use of portable radios cell phones. It is recommended that the DS Radios OR Fire Protection radios be used for this communication.

1. Check Reason For Entry - LOSS OF Go to the Section for the INTAKE STRUCTURE current Plant Mode:

MODE SECTION 3 Section E 4 Section F 5 Section G 6 Section H Defueled Section I

RNP BASIS STEP Main Body LOIS MODES I THROUGH 3 CI STEP BASIS Subsequent steps throughout the procedure contain local manual operator actions outside the Control Room. Prior to performing these steps the threat should be nullified. Previous steps in AOP-034 have directed the Control Room to have personnel seek shelter. An exact time can not be determined for when the threat will be nullified. The intent of the caution is to remind the Operator that prior to sending an operator to perform an action; he should have a reasonable expectation that the Operator can reach the area and accomplish the task.

INI STEP BASIS The note explains that FRBs, EPP Foldouts, EPP-1, AOP-014, AOP-020, and AOP-022 are not applicable for this event and that these procedures should not be implemented except as directed by this procedure. These procedures assume that the UHS is available and will not provide any meaningful strategies for this event. I AOP-014, AOP-020, and AOP-022 have actions or individual attachments which will provi e assistance, this procedure will direct their implementati Any individual steps that need rn lemeritation have been included in this procedure. T i ote s applicable for the entire procedure; therefore it has been placed at the beginning of the procedure.

2N1 STEP BASIS Most of the events that this procedure is intended to mitigate could involve a loss of 480V Bus E-2 and MCC-6. The Plant PA is powered from a bus fed by MCC-6, therefore under these conditions the plant PA will not be available. The note recommends that the Fire Protection or DS Radios be used for coordination of field activities. The plant cell phone repeaters are powered from non vital power and are not expected to last for longer than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> following a loss of power. The DS/Fire Protection radios will provide a more reliable form of communications. The DS radios and fire protection radios both operate on the same frequency.

1-2 STEP BASIS These are decision steps to control the applicable portion of the procedure to be performed. Since the time frames for the two components of interest are so divergent, the procedure has been divided into different sections (see general). If both the intake and the dam have been lost the step is arranged such that the intake section is performed since this is much more limiting. Step one checks for a loss of Intake or Dam. The RNO (LOTG) sends the operator to the appropriate Section for LOTG dependent on the Mode at the time of the event. Step 2 sends the operator to the appropriate section for a loss of the intake dependent on Mode.

NOTE: The remainder of the steps in the Main Body or for a LOIS event for Modes I through 3.

3 STEP BASIS This step ensures the MSIV and Bypass valves are closed to limit the amount of cooldown. Based on simulator observations, when the turbine is tripped from greater than approximately 25% power, auxiliary feedwater is initiated from SIG low level resulting in additional cooling and a lower PZR level. In this circumstance when direction is given to stop RCPs later in the procedure, an SI is initiated due to steam line delta pressure. Isolating the steam lines early and establishing control of auxiliary feedwater flow prior to stopping RCPs helps to minimize the heat loss and reduce the likelihood for an SI initiation.

C4 STEP BASIS Based on calculation RNP-M/MECH-1 769, Operations needs to provide well water to the running EDG within 40 minutes after it starts to assure continued operation for the EDG or shut it down, and valve in well water to the CCW heat exchanger within one hour into the event to prevent CCW temperature from exceeding 125F. The 125°F CCW temperature is based on ensuring the integrity of RCP seals. This caution provides the times available for establishing the alternate cooling from well water for operator awareness.

EPP-28-BD Rev 13 Page 9 of 50

NRC Exam 100. G2.4.30 SRO 001 Given the following plant conditions:

- Today at 0100 LCO 3.4.13, RCS OPERATIONAL LEAKAGE, was entered due to an 8 gpm leak from a crack on the spray line penetration weld at the PZR

- The crew commenced a shutdown at 0200

- The plant reached MODE 3 at 0800 Which ONE (1) of the following completes the statement below?

Notifications must be made to the NRC by (1) and (2)

(REFERENCES PROVIDED)

A. (1) 0300 (2) 0800 B (1) 0600 (2) 0800 C. (1) 0300 (2) 1200 D. (1) 0600 (2) 1200 Thursday, June 13, 2013 3:13:58 PM 197

NRC Exam The correct answer is B.

A. Incorrect. 0300 is incorrect because the notification requirement is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> vice 1.

Plausible because RCS leakage is also an E-Plan classification, however the value is 10 gpm for pressure boundary leakage.

B. Correct. Commencing a Technical Specification required shutdown requires a notification to the NRC within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> lAW AP-030 based on 10CFR requirements.

The shutdown commenced at 0200 therefore the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> notification must be made by 0600. Another notification is required to be made to the NRC within one 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of exceeding the LCO shutdown time requirements per the E-Plan under EAL Matrix SU3.1. The LCO was entered at 0100 with a 6 hr requirement to be in Mode 3 which expired at 0700. This requires an Unusual Event classification. The NRC is required to be notified as soon as possible not to exceed one hour which would be 0800.

C. Incorrect. 0300 is incorrect because the notification requirement is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> vice 1.

Plausible because RCS leakage is also an E-Plan classification, however the value is 10 gpm for pressure boundary leakage. 1200 is incorrect because the E-Plan classification requires a one hour notification. 1200 is plausible because a notification is required lAW AP-030 for completion of a TS required shutdown, however it is a 60 day LER.

D. Incorrect. 1200 is incorrect because the E-Plan classification requires a one hour notification. 1200 is plausible because a notification is required lAW AP-030 for completion of a TS required shutdown, however it is a 60 day LER.

Thursday, June 13, 2013 3:13:58 PM 198

NRC Exam Question: 100 Tier/Group: 3 (SRO)

K/A Importance Rating: RO 2.7 SRO 4.1 K/A: Emergency Procedures/Plan G2.4.30: Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

Reference(s): AP-030, EAL Matrix, TS 3.4.13 Proposed References to be provided to applicants during examination: EAL Matrix, AP-030 ,Attachment 11.1-11.3, LCO 3.4.13 Learning Objective:

Question Source: New Question History:

Question Cognitive Level: Analysis 10 CFR Part 55 Content: 43.2, 5 Comments:

This question meets the K/A because the candidate must determine the required notifications to an external agency (NRC) based on conditions given in the stem.

You have completed the test!

Thursday, June 13, 2013 3:19:33 PM 199

ATTACHMENT 11.1 Page 1 of 9 IMMEDIATE (ONE HOUR) NOTIFICATIONS TO THE NRC IMMEDIATE (ONE HOUR) NOTIFICATIONS TO THE NRC 10 CFR 50.72 states that immediate reports shall be made to the NRC Operations Center of these Emergency Events via the NRC Emergency Telecommunications System (ETS) as specified in the Emergency Plan. 10 CFR 50.72 additionally identifies Non-Emergency Events which are to be reported within one-hour, four-hours, or eight hours to the NRC. ETS Telephones, which are identified, are located in the Control Room, the TSC, and the EOF. In the event that the ETS is not available, 10 CFR 50.72(a)(2) permits the use of commercial telephone.

EVENT KEYWORDS REQUIREMENT I EXAMPLES NOTE: 10 CFR 50.72 recognizes the Emergency Plan and its four Emergency Classes of Unusual Event, Alert, Site Area Emergency and General Emergency.

EMERGENCIES Emergency HBRSEP shall notify the NRC of the Declaration of an Unusual Event, Alert, Site Unusual Event declaration of any of the Emergency Classes Area Emergency, or General Emergency.

Alert specified in the Emergency Plan.

Site Area Discovery of an event that should have Emergency (See EPNOT-01) resulted in an Emergency Classification, but General no emergency was declared.

10 CFR 50.72(a)(i) Emergency Discovery that a declared emergency 10 CFR 30.32(i)(3)(viii) ISFSI exceeded the Emergency Action Levels for a 10 CFR 40.31 (i)(3)(viii) higher emergency declaration, but the higher 10 CFR 72.75(a) classification was not declared.

EROS ACTIVATION ERDS HBRSEP shall activate the ERDS as soon as An Alert, Site Area Emergency, or General Emergency possible but not later than one hour after Emergency is declared.

declaring an Alert, Site Area Emergency, or General Emergency.

10 CFR 50.72(a)(4)

DEVIATION FROM TS (10 CFR Deviation Any deviation from the TS authorized Intentional deviation from an approved plant 50.54(X)) Departure pursuant to 10 CFR 50.54(x). procedure in order to preserve plant safety License 10 CFR 50.54(x).

10 CFR 50.72(b)(1) Condition AP-030 Rev. 48 Page 15 of 65

ATTACHMENT 11.2 Page 1 of 3 FOUR HOUR NOTIFICATIONS TO THE NRC FOUR HOUR NOTIFICATIONS TO THE NRC If not reported under paragraphs (a) or (b)(1) of 10 CFR 50.72, HORSEP shall notify the NRC Operations Center via ETS as soon as practical and in all cases, within four hours of the occurrence of any of the followino:

EVENT KEY WORDS REQUIREMENT EXAMPLES SHUTDOWN REQUIRED BY TS Shutdown The initiation of any shutdown required by Reactor is in MODEs 1 or 2 and the Control TS Shutdown the TS. Room takes action to reduce power (i.e.,

Power negative reactivity insertion) in order to Reduction comply with a Required Action to be in MODE 3 within a Completion Time. Reduction in power for some other purpose than compliance with the shutdown requirement is not reportable. MODE changes required by TS when reactor is in MODEs 3, 4, or other non-power conditions, are not reportable.

If allowed outage time plus required shutdown time to MODE 3 is less than the expected restoration time of the LCO and power is 10 CFR 50.72(b)(2)(i) reduced in anticipation of the required shutdown, the shutdown is reportable.

ECCS DISCHARGE INTO RCS ECCS Any event that results or should have Manual or automatic Safety Injection System Actuation resulted in emergency core cooling actuation in response to a valid signal that Safety system (ECCS) discharge into the reactor resulted in or should have resulted in Injection coolant system as a result of a valid signal discharge into the reactor coolant system.

except when the actuation results from and is part of a pre-planned sequence during 10 CFR 50.72(b)(2)(iv)(A) testing or reactor operation.

RPS INITIATION RPS Actuation Any event or condition that results in - Manual or automatic reactor trip from critical (MANUAUAUTOMATIC) Reactor actuation of the reactor protection system through RTP of 100%. Trips which occur as DURING OPERATION Protection (RPS) when the reactor is critical except part of planned evolutions in accordance with System when the actuation results from and is part procedures are not reportable.

RPS of a pre-planned sequence during testing 10 CFR 50.72(b)(2)(iv)(B) Reactor Trip or reactor operation.

AP-030 Rev. 48 Page 24 of 65

RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE;

b. 1 gprn unidentified LEAKAGE;

c. 10 gpm identified LEAKAGE; and

d. 75 gal Ions per day primary to secondary LEAKAGE through any one steam generator (SG).

APPLICABILITY: MODES 1, 2, 3, and 4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RCS operational A.1 Reduce LEAKAGE to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> LEAKAGE not within within limits.

I imits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.

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

B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR Pressure boundary LEAKAGE exists.

OR Primary to secondary LEAKAGE not within limit.

HBRSEP Unit No. 2 3.4-35 Amendment No. 212

Co°Q RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE;

b. 1 gpm unidentified LEAKAGE;

c. 10 gpm identified LEAKAGE; and

d. 75 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

APPLICABILITY: MODES 1, 2, 3, and 4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RCS operational A.1 Reduce LEAKAGE to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> LEAKAGE not within within limits.

limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.

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

B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR Pressure boundary LEAKAGE exists.

OR Primary to secondary LEAKAGE not within I imit.

HBRSEP Unit No. 2 3.4-35 Amendment No. 212

?C ATTACHMENT 11.1 Page 1 of 9 IMMEDIATE (ONE HOUR) NOTIFICATIONS TO THE NRC IMMEDIATE (ONE HOUR) NO11FICATIONS TO THE NRC 10 CFR 50.72 states that immediate reports shall be made to the NRC Operations Center of these Emergency Events via the NRC Emergency Telecommunications System (ETS) as specified in the Emergency Plan. 10 CFR 50.72 additionally identifies Non-Emergency Events which are to be reported within one-hour, four-hours, or eight hours to the NRC. ETS Telephones, which are identified, are located in the Control Room, the TSC, and the EOF. In the event that the ETS is not available, 10 CFR 50.72(a)(2) permits the use of commercial teleohone.

EVENT KEY WORDS REQUIREMENT I EXAMPLES NOTE: 10 CFR 50.72 recognizes the Emergency Plan and its four Emergency Classes of Unusual Event, Alert, Site Area Emergency and General Emergency.

EMERGENCIES Emergency I-IBRSEP shall notify the NRC of the Declaration of an Unusual Event, Alert, Site Unusual Event declaration of any of the Emergency Classes Area Emergency, or General Emergency.

Alert specified in the Emergency Plan.

Site Area Discovery of an event that should have Emergency (See EPNOT-01) resulted in an Emergency Classification, but General no emergency was declared.

10 CFR 50.72(a)(i) Emergency Discovery that a declared emergency 10 CFR 30.32(i)(3)Ctii) ISFSI exceeded the Emergency Action Levels for a 10 CFR 40.31 (i)(3)(viii) higher emergency declaration, but the higher 10 CFR 72.75(a) classification was not declared.

ERDS ACTIVATION ERDS HBRSEP shall activate the ERDS as soon as An Alert, Site Area Emergency, or General Emergency possible but not later than one hour after Emergency is declared.

declaring an Alert, Site Area Emergency, or General Emergency.

10 CFR 50.72(a)(4)

DEVIATION FROM TS (10 CFR Deviation Any deviation from the TS authorized Intentional deviation from an approved plant 50.54(X)) Departure pursuant to 10 CFR 50.54(x). procedure in order to preserve plant safety License 10 CFR 50.54(x).

10 CFR 50.72(b)(1) Condition AP-030 Rev. 48 Page 15 of 65

ATTACHMENT 11.1 Page 2 of 9 IMMEDIATE (ONE HOUR) NOTIFICATIONS TO THE NRC IMMEDIATE (ONE HOUR) NOTIFICATIONS TO THE NRC HBRSEP shall immediately notify the NRC Operations Center via ETS as soon as practical and in all cases within one hour of the occurrence of any of the following:

EVENT KEY WORDS REQUIREMENT EXAMPLES SAFETY LIMIT, LIMITING SAFETY Safety Limit If any safety limit is exceeded, shut down The limits of TS Figure 2.1.1-1 are exceeded.

SYSTEM SETTiNG EXCEEDED Limiting Safety the reactor. HBRSEP shall notify the NRC System Setting [within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> via ETS per 10 CFR 50.72(a)(1), See Emergency Plan Procedures]. Operation must not be resumed until authorized by the NRC.

10 CFR 50.36(c)(1)(i)(A) NRC Region II must also be notified within 1 UFSAR Section 17.3A, Paragraph hour and the Vice President Robinson 3.1.a Nuclear Plant within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SAFETY SYSTEM DOES NOT ESF HBRSEP shall notify the NRC if the A failure mechanism is discovered that FUNCTION AS REQUIRED RPS automatic safety system [to correct an indicates that the RPS will not function to trip Limiting Safety abnormal situation before a safety limit is the reactor under certain required conditions.

System Setting exceeded] has been determined not to 10 CFR 50.36(c)(1)(iiXA) function as required.

AP-030 Rev. 48 Page 16 of 65

ATTACHMENT 11.1 Page 3 of 9 IMMEDIATE (ONE HOUR) NOTIFICATIONS TO THE NRC IMMEDIATE (ONE HOUR) NO11FICATIONS TO THE NRC SECURITY SAFEGUARDS EVENTS HBRSEP shall notify the NRC Operations Center via the ETS within one hour after discovery of the safeguards events described as follows (10 CFR 73.71(bl(1ll:

EVENT KEY WORDS REQUIREMENT EXAMPLES THEFT/UNLAWFUL DIVERSION OF SNM Any discovery of the loss of any shipment of Shipment Emergency Event SNM OR SPENT FUEL SHIPMENT Spent Fuel SNM or spent fuel, and within one hour after Security recovery of or accounting for such lost Safeguards shipment 10 CFR 73.71(a)(1)

THEFT/UNLAWFUL DIVERSION OF Theft of SNM Any event in which there is reason to Shipment Emergency Event SNM Diversion believe that a person has committed or Security caused, or attempted to commit or cause, or Safeguards has made a credible threat to commit or 10 CFR 73.71(b)(1) cause:

10 CFR 73, Appendix G, l(a)(1) (1) A theft or unlawful diversion of SNM SABOTAGE OF PLANT Sabotage [Any event in which there is reason to Shipment Emergency Event EQUIPMENT Damage to Plant believe that a person has committed or Security Event (SEC-NGGC-2147)

SNM caused, or attempted to commit or cause, or Spent Fuel has made a credible threat to commit or Security cause:]

Safeguards (2) Significant physical damage to a power reactor...or its equipment or carrier equipment transporting nuclear fuel or spent 10 CFR 73.71 (b)(1) nuclear fuel, or to the nuclear fuel or spent 10 CFR 73, Appendix G, l(a)(2) fuel a facility or carrier possesses.

In response to NRC Bulletin 2005-02, RNP committed to make an accelerated call to the NRC within approximately 1S minutes following discovery of an imminent threat or attack against the station. The primary purpose is to allow for the NRC to timely notify other licensees of a potential common threat. The accelerated call should not be allowed to interfere with plant or personnel safety, physical security response, or notification of local law enforcement agencies.

The information provided in the accelerated call can be limited to:

• Site name

• Emergency Classification if already determined do not delay call for the purpose of classifying

• Nature of the threat briefly described, if known, including the type of attack (e.g., armed assault by land, water or aircraft) and the attack status (e.g.,

imminent, in progress, or repelled)

AP-030 Rev. 48 Page 17 of 65

ATTACHMENT 11.1 Page 4 of 9 IMMEDIATE (ONE HOUR) NOTIFICATIONS TO THE NRC IMMEDIATE (ONE HOUR) NOTIFICATIONS TO THE NRC SECURIT, SAFEGUARDS EVENTS HBRSEP shall notify the NRC Operations Center via the ETS within one hour after discovery of the safeguards events described as follows (10 CFR 73.71(b)(lfl:

EVENT KEY WORDS REQUIREMENT EXAMPLES UNAUTHORIZED TAMPERING Unauthorized [Any event in which there is reason to Security Event (SEC-NGGC-2147)

WITH PLANT EQUIPMENT Use believe that a person has committed or Tampering caused, or attempted to commit or cause, or Security System has made a credible threat to commit or Safeguards cause:]

(3) Interruption of normal operation of HBRSEP through the unauthorized use of or tampering with its machinery, components, or controls including the 10 CFR 73, Appendix G, l(a)(3) security system.

ENTRY OF UNAUTHORIZED Unauthorized An actual entry of an unauthorized person Security Event (SEC-NGGC-2147)

PERSON INTO PROTECTED OR Entry into a protected area, material access area, VITAL AREA Security controlled access area, vital area, or 10 CFR 73, Appendix G, 1(b) Safeguards transport.

FAILURE, DEGRADATION, OR Degradation Any failure, degradation, or the discovered - Procedure SEC-NGGC-2147 DISCOVERED VULNERABILITY OF Vulnerability vulnerability in a safeguard system that SAFEGUARD SYSTEM Safeguards could allow unauthorized or undetected Unauthorized access to a protected area, material access Undetected area, controlled access area, vital area or Access transport for which compensatory measures 10 CFR 73, Appendix G, 1(c) Security have not been employed.

INTRODUCTION OF Contraband The actual or attempted introduction of - Contraband applies to items that could be CONTRABAND INTO VITAL OR Unauthorized contraband into a protected area, material used to commit radiological sabotage as PROTECTED AREA Security process area, vital area, or transport. defined in 10 CFR 73.2.

10 CFR 73, Appendix G, 1(d) Safeguards See footnote on the previous page regarding a goal for a 15 minute call to the NRC in regard to an imminent security threat or attack.

AP-030 Rev. 48 Page 18 of 65

ATTACHMENT 11.1 Page 5 of 9 IMMEDIATE (ONE HOUR) NOTIFICATIONS TO THE NRC IMMEDIATE (ONE HOUR) NOTIFICATIONS TO THE NRC SOURCE, BYPRODUCT AND SNM HBRSEP shall immediately notify the NRC Operations Center via ETS, when:

EVENT KEY WORDS REQUIREMENT EXAMPLES EXTERNAL EXPOSURE FROM Byproduct Notwithstanding any other requirements for BYPRODUCT, SOURCE, OR SNM Source notification, immediately notify the NRC of any (5X ANNUAL LIMIT) SNM event involving byproduct, source, or SNM Exposure possessed by HBRSEP that may have caused or Dose threatens to cause any of the following Release conditions:

Occupational 1. An individual to receive:

(i) A total effective dose equivalent of 25 rems or more; or (ii) An eye dose equivalent of 75 rems or more; or (iii) A shallow dose equivalent to the skin or extremities of 250 rads or more; or

2. The release of radioactive material, inside or outside the restricted area, so that, had an individual been present for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the 10 CFR 20.2202(a)(1) individual could have received an intake five times the occupational annual limit on intake.

AP-030 Rev. 48 Page 19 of 65

ATTACHMENT 11.1 Page 6 of 9 IMMEDIATE (ONE HOUR) NOTIFICATIONS TO THE NRC IMMEDIATE (ONE HOUR) NOTIFICATIONS TO THE NRC SOURCE, BYPRODUCT AND SNM HBRSEP shall immediately notify the NRC Operations Center via ETS, when:

EVENT KEY WORDS REQUIREMENT EXAMPLES INTERNAL EXPOSURE FROM Intake The release of radioactive material, inside or BYPRODUCT, SOURCE, SNM (>5X Ingestion outside the restricted area, so that, had an OCCUPATIONAL LIMIT) Release individual been present for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the Source individual could have received an intake five times Byproduct the occupational annual limit on intake.

10 CFR 20.2202(a)(2) SNM IMMEDIATE (ONE HOUR) NOTIFICATIONS TO THE NRC ISFSI -

HBRSEP shall immediately notify the NRC Operations Center via ETS, when:

EVENT KEY WORDS REQUIREMENT EXAMPLES ISFSI ACCIDENTAL CRI11CALITY

- ISFSI The licensee shall notify the NRC Operations Unusually high radiation readings OR LOSS OF SNM Criticality Center via ETS within one hour of discovery of discovered in the vicinity of the ISFSI that 10 CFR 7274 SNM accidental criticality or any loss of SNM. could indicate possibility of a criticality

. Loss event IMMEDIATE (ONE HOUR) NO11FICATIONS TO THE NRC SNM SHIPMENTS -

HBRSEP shall immediately notify the NRC Operations Cen er via ETS, when:

EVENT KEY WORDS REQUIREMENT EXAMPLES LOST OR UNACCOUNTED Shipment HBRSEP shall notify the NRC Operations Center Shipment Emergency Event SHIPMENT OF SNM Loss via the ETS within one hour after discovery of any Security Event (SEC-NGGC-2147)

SNM loss of any shipment of SNM or spent fuel or any Spent Fuel incident in which an attempt has been made, or is Diversion believed to have been made, to commit a theft or 10 CFR 70.52(b) Safeguards unlawful diversion of SNM.

10 CFR 73.71(a)(1) Security LOST OR UNACCOUNTED Recovery HBRSEP shall notify the NRC Operations Center SHIPMENT OF SNM RECOVERY

- Accounting via the ETS within one hour after recovery of, or Shipment accounting for, any lost shipment of SNM.

SNM Security 10 CFR 73.71(a)(1) Safeguards AP-030 Rev. 48 Page 20 of 65

00 1 Qr o ATTACHMENT 11.1 Page 7 of 9 IMMEDIATE (ONE HOUR) NOTIFICATIONS TO THE NRC IMMEDIATE (ONE HOUR) NOTIFICATIONS TO THE NRC FOLLOW-UP -

With respect to the telephone notifications made under paragraphs (a) and (b) of 10 CFR 50.72 or paragraphs (a), (b), (c), or (d) of 10 CFR 72.75, in addition to making the required initial notification, HBRSEP shall during the course of the event immediately report:

EVENT KEY WORDS REQUIREMENT EXAMPLES FOLLOW-UP NOTIFICATION Degradation (i) any further degradation in the level of Refer to EPNOT-01 Emergency Class safety of the plant or ISFSI or other Change worsening conditions, including those that Update require the declaration of any of the Termination Emergency Classes, if such a declaration ISFSI has not been previously made, or (ii) any change from one Emergency Class 10 CFR 50.72(c)(1) to another, or (iii) a termination of the 10 CFR 72.75(f)(1) Emergency Class.

FOLLOW-UP NOTIFICATION Result (i) the results of ensuing evaluations or Evaluation assessments of plant or ISFSI conditions, Effectiveness (ii) the effectiveness of response or Unknown protective messures taken, and 10 CFR 50.72(c)(2) ISFSI (Hi) information related to plant or ISFSI 10 CFR 72.75(fl(2) behavior that is not understood.

FOLLOW-UP NOTIFICATION Open Maintain an open, continuous Refer to EPNOT-01 Continuous communication channel with the NRC 10 CFR 50.72(c)(3) Communication Operations Center upon request by the 10 CFR 50.72.75(0(3) ISFSI NRC.

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ATTACHMENT 11.1 Page 8of9 IMMEDIATE (ONE HOUR) NOTIFICATIONS TO THE NRC IMMEDIATE (ONE HOUR) NOTIFICATIONS NRC REGION II OFFICE HBRSEP shall immediately notify the final delivery carrier and, by telephone and telegram, mailqram, or facsimile, the NRC Reqion II Office when:

EVENT KEY WORDS REQUIREMENT EXAMPLES THEFTIUNLAWFUL DIVERSION OF Incident Any incident in which an attempt has been 10 Curies of tritium discovered missing from TRIT1UM Theft made or is believed to have been made to the Chemistry Laboratory, and reason exists Tritium commit a theft of more than 10 curies of to suspect that the tritium was stolen Attempt tritium (outside of spent fuel) at any one Security time or more than 100 curies of tritium in 10 CFR 30.55(c) Safeguards one calendar year THEFTIUNLAWFUL DIVERSION OF Incident Any incident in which an attempt has been A source assembly is discovered missing SOURCE MATERIAL Attempt made or is believed to have been made to from a new fuel shipment.

Theft commit a theft or unlawful diversion of more Diversion than 15 pounds of Source Material at any Source one time or 150 pounds of Source Material 10 CFR 40.64(c) Security in any one calendar year Safeguards SHIPPING PACKAGE Contamination Removable radioactive surface New or Spent Fuel Shipment Cask arrives RADIOACTIVELY CONTAMINATED Shipment contamination exceeds the limits of with surface contamination in excess of limits.

10 CFR 20.1906(d)(1) 10 CFR 71.87 SHIPPING PACKAGE EXCEEDING Radiation External radiation levels exceeds of the New or Spent Fuel Shipment Cask arrives EXTERNAL DOSE RATE LIMITS Dose Rate limits of 10 CFR 71.47 with external radiation levels in excess of 10 CFR 20.1906(d)(2) Shipment limits.

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ATTACHMENT 11.1 Page 9 of 9 IMMEDIATE (ONE HOUR) NOTIFICATIONS TO THE NRC IMMEDIATE (ONE HOUR) NOTIFICATIONS TO THE NRC FFD -

The NRC Region II Administrator must be notified immediately by telephone of the following:

EVENT KEY WORDS REQUIREMENT EXAMPLES NRC EMPLOYEE NOT FIT FOR Alcohol If HBRSEP has a reasonable belief that an DUTY Influence NRC employee or NRC contractor may be Substance under the influence of any substance, or is NRC employee otherwise unfit for duty, the licensee or FFD other entity may not deny access but shall 10 CFR 26.77(c) Fitness for Duty escort the individual. In any such instance, the licensee or other entity shall immediately notify the Region II Administrator by telephone, followed by written notification (e.g., e-mail or fax) to document the oral notification. If the Region II Administrator cannot be reached, the licensee or other entity shall notify the NRC Operations Center.

IMMEDIATE (ONE HOUR) NOTIFICATIONS TO THE NRC IAEA -

The NRC Director, NRR or Director, NMSS must be notified immediately by telephone of the following:

SURPRISE VISIT OF IAEA IAEA HORSEP shall immediately communicate If the IAEA representatives credentials have OFFICIAL International by telephone, within one hour with respect not been confirmed by the NRC, the licensee Atomic to the credentials of any person who claims shall not admit the person until the NRC has Energy to be an IAEA representative and shall confirmed the persons credentials. The Agency accept written or electronic confirmation of licensee, shall notify the Commission 10 CFR 75.8(c) Credential the credentials from the NRC. promptly, by telephone, whenever an IAEA representative arrives at a facility or location without advance notification.

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ATTACHMENT 11.2 Page 1 of 3 FOUR HOUR NOTIFICATIONS TO THE NRC FOUR HOUR NOTIFICATIONS TO THE NRC If not reported under paragraphs (a) or (b)(1) of 10 CFR 50.72, HBRSEP shall notify the NRC Operations Center via ETS as soon as practical and in all cases, within four hours of the occurrence of any of the following; EVENT KEY WORDS REQUIREMENT EXAMPLES SHUTDOWN REQUIRED BY TS Shutdown The initiation of any shutdown required by Reactor is in MODEs 1 or 2 and the Control TS Shutdown the TS. Room takes action to reduce power (i.e.,

Power negative reactivity insertion) in order to Reduction comply with a Required Action to be in MODE 3 within a Completion Time. Reduction in power for some other purpose than compliance with the shutdown requirement is not reportable. MODE changes required by TS when reactor is in MODEs 3, 4, or other non-power conditions, are not reportable.

If allowed outage time plus required shutdown time to MODE 3 is less than the expected restoration time of the LCO and power is 10 CFR so.72(b)(2)c) reduced in anticipation of the required shutdown, the shutdown is reportable.

ECCS DISCHARGE INTO RCS ECCS Any event that results or should have Manual or automatic Safety Injection System Actuation resulted in emergency core cooling actuation in response to a valid signal that Safety system (ECCS) discharge into the reactor resulted in or should have resulted in Injection coolant system as a result of a valid signal discharge into the reactor coolant system.

except when the actuation results from and is part of a pre-planried sequence during 10 CFR 50.72(b)(2)(iv)(A) testing or reactor operation.

RPS INITIATION RPS Actuation Any event or condition that results in - Manual or automatic reactor trip from critical (MANUALIAUTOMATIC) Reactor actuation of the reactor protection system through RTP of 100%. Trips which occur as DURING OPERATION Protection (RPS) when the reactor is critical except part of planned evolutions in accordance with System when the actuation results from and is part procedures are not reportable.

RPS of a pre-planned sequence during testing 10 CFR 50.72(b)(2)(iv)(B) Reactor Trip or reactor operation.

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Uft Qo ATTACHMENT 11.2 Page 2 of 3 FOUR HOUR NOTIFICATIONS TO THE NRC FOUR HOUR NOTIFICATiONS TO THE NRC If not reported under paragraphs (a) or (b)(1) of 10 CFR 50.72, HBRSEP shall notify the NRC Operations Center via ETS as soon as practical and in all cases, within four hours of the occurrence of any of the following:

EVENT KEY WORDS REQUIREMENT EXAMPLES PRESS RELEASES AND News Release Any event or situation, related to the health Any News release concerning GOVERNMENT NOTIFICATIONS Press and safety of the public or on-site - A fatality, Radio personnel, or protection of the - Inadvertent release of radioactively Television environment, for which a news release is contaminated materials to public areas Fatality planned or notification to other government - unusual or abnormal releases of radioactive Environment agencies has been or will be made. Such effluents, or Public an event may include an on-site fatality or - Information associated with an Emergency Health and Safety inadvertent release of radioactively Event except when the ERO is activated Release contaminated materials. (EPNOT-01).

ISFSI Notification to other government agencies concerning:

- A fatality on site,

- Health and safety of the public or site personnel,

- Inadvertent release of radioactively 10 CFR 50.72(b)(2)(xi) contaminated materials to public areas, 10 CFR 72.75(b)(2) - Discovered endangered species kill.

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0 Qc ATTACHMENT 11.2 Page 3 of 3 FOUR HOUR NOTIFICATIONS TO THE NRC FOUR HOUR NOTIFICA11ONS TO THE NRC HBRSEP shall notify the NRC Operations Center via ETS as soon as possible but not later than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the discovery of any of the following events or conditions involvinci sources or spent fuel.

EVENT KEY WORDS REQUIREMENT EXAMPLES LOSS OR THEFT OF LICENSED Loss Immediately notify the NRC, after its A radiography source is discovered missing. The MATERIAL (>I000X 10 CFR 20 Theft occurrence becomes known, any lost, stolen, source is licensed to the radiography contractor.

LIMITS) Missing or missing licensed material in an aggregate If the contractor does not make the required Licensed quantity equal to or greater than 1,000 times notification, HERSEP should notify the NRC Radioactive the quantity specified in [10 CFR 20] Operations Center via ETS.

Material Appendix C under such circumstances that it Recovery appears to HBRSEP that an exposure could 10 CFR 20 2201 result to persons in unrestricted areas.

. Follow-up written report required within subsequent 30 days.

Note If the lost, stolen, or missing source exceeds a Quantity of Concern as specified in HPP-018, then the NRC desires to be notified within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of any subsequent recovery of the source.

ISFSI DEPARTURE FROM

- ISFSI An action taken in an emergency that Action taken in an emergency that departs from LICENSE CONDITION Emergency departs from a condition or a technical procedure that is deemed necessary to prevent Departure specification contained in a license or releases or radiation doses to the public in Deviation certificate of compliance issued under excess of 10 CFR 20 limits (See Health and 10 CFR 72 when the action is immediately PRO-NGGC-0200).

Safety needed to protect the public health and License safety and no action consistent with license Condition conditions or technical specifications that can 10 CFR 72.75(b)(1) provide adequate or equivalent protection is immediately apparent.

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Qco_, i ATTACHMENT 11.3 Page 1 of 5 EIGHT HOUR NOTIFICATIONS TO THE NRC EIGHT HOUR NOTIFICA11ONS TO THE NRC If not reported as a declaration of an Emergency Class under parsgraph (a) of 10 CFR 50.72, HBRSEP shall notify the NRC Operations Center via ETS as soon as practical and in all cases within eight hours of the occurrence of any of the following:

EVENT KEY WORDS REQUIREMENT EXAMPLES PRINCIPAL SAFETY BARRIERS Degraded Any event or condition that results Fuel cladding failures in the reactor, or in the storage pool, that SERIOUSLY DEGRADED Safety in the condition of the nuclear exceed expected values, or that are unique or widespread, or Barriers power plant, including its principal that are caused by unexpected factors, and would involve a Fission safety barriers, being seriously release of significant quantities of fission products Product degraded; Cracks and breaks in the piping or reactor vessel, or major Barrier components in the reactor coolant system that have Safety relevance (steam generators, reactor coolant pumps, valves, etc.)

Significant welding or material defects in the RCS Low temperature overpressure transients where the pressure temperature limits are violated Loss of relief and/or safety valve functions during operation Loss of Containment function or integrity

- Complete loss of containment integrity function including (1) containment leakage rate greater than allowed value per SR 3.6.1.1 (i.e., entry into LCD 3.6.1 Condition A), (2) loss of containment penetration isolation functional capability (i.e., both 10 CFR 50 72(b)(3)(ii)(A) barriers), or loss of containment spray capability UNANALYZED PLANT Safety Any event or condition that results OTeJ setpoints are declared inoperable due to summator CONDITION Function in the nuclear power plant being in module lag constants. The channel response time exceeded the Unanalyzed an unanalyzed condition that value assumed in the accident analysis (analytical limits).

Condition significantly degrades plant safety. Accumulation of voids in systems designed to remove heat from the reactor that could inhibit the ability to adequately remove heat from the core, particularly under natural circulation conditions.

Any power level excursion above 2346 MWt should be evaluated to determine if the condition posed an unanalyzed condition that significantly degrades nuclear plant safety. Operation slightly in excess of 2346 MWt for short periods are not expected to trigger the 10 CFR 50.72(b)(3)(ii)(B) criterion.

10 CFR 50.72(b)(3)(ii)(B)

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9c i ATTACHMENT 11.3 Page 2 of 5 EIGHT HOUR NOTIFICATIONS TO THE NRC EIGHT HOUR NOTIFICATIONS TO THE NRC If not reported as a declaration of an Emergency Class under paragraph (a) of 10 CFR 50.72, HBRSEP shall notify the NRC Operations Center via ETS as soon as practical and in all cases within eight hours of the occurrence of any of the following:

EVENT KEY WORDS REQUIREMENT I EXAMPLES LOSS OF EMERGENCY Selective Any event that results in a major Loss of 15 or more of 59 Public Warning Sirens as indicated ASSESSMENT, OFF-SITE Signaling loss of emergency assessment on the siren activation system for a period of at least 30 RESPONSE, OR System capability, off-site response minutes at any one time COMMUNICATIONS CAPABILITY capability, or communications Loss of greater than 50% of communications capability (i.e.,

Sirens capability (e.g. significant portion ofisite communications systems which include the Selective of control room indication, ETS, or Signaling System, the Essex System and the Local ETS off-site notification System) Government Radio System>

ERFIS Loss of Emergency Assessment Capability. This may include EROS planned or unplanned losses of an Emergency Response Facility (ERF). Typically, this would be the TSC but may include the EOF. (1)

ETS communications function unavailable. This does not apply to minor interruptions in site or corporate telecommunications systems. It is intended to apply to serious conditions during which the telecommunication system can no longer fulfill the requirements of the Emergency Plan or provide ETS functionality. (1)

Loss of commercial telephone system to the extent that required communications could not be made to official offsite locations (e.g., EOC5, Warning Points)

Inoperability of ERFIS and ERDS is not capable of being restored within one hour. (1) 10 CFR 50.72(b)(3)(xiii>

(1) See Attachment 11.16, Guidance On Reporting Loss Of Emergency Assessment Or Communications Capability, for additional guidance.

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ATTACHMENT 11.3 Page 3of5 EIGHT HOUR NOTIFICATIONS TO THE NRC EIGHT HOUR NOTIFICA11ONS TO THE NRC If not reported ass declaration of an Emergency Class under paragraph (a) of 10 CFR 50.72, HBRSEP shall notify the NRC Operations Center via ETS as soon as practical and in all cases within eight hours of the occurrence of any of the followina:

EVENT KEY REQUIREMENT EXAMPLES WORDS RPS1SAFETY SYSTEM Manual Any event or condition that results in valid Auxtiary Feedwater initiation/actuation INITIATION Automatic actuation of any of the systems listed below Reactor Trip (Manual/Automatic) while subcritical (MANUALIAUTOMATIC) Actuation except when the actuation results from and Reactor Trip while critical is reportable per Attachment 11.2 Engineered is part of a pre-planned sequence during EDG start due to an undervoltage trip signal on emergency bus El or E2 Safety testing or reactor operation. A single train of Containment Isolation actuates Feature The systems to which the requirements of A valid signal for Containment Ventilation Isolation occurs ESF this paragraph apply are:

Valid (1) Reactor protection system (RPS) Valid actuations are those actuations that result from valid signals or Clearance including: reactor scram and reactor from intentional manual initiation, unless it is part of a preplanned test.

RPS trip. Valid signals are those signals that are initiated in response to actual Actuation (2) General containment isolation signals plant conditions or parameters satisfying the requirements for the Reactor affecting containment isolation valves initiation of the safety function of the system. They do not include Protection in more than one system or multiple actuations which are the result of other signals. (NUREG 1022)

System main steam isolation valves (MSIV5).

(3) Emergency core cooling systems Invalid actuations are, by definition, those that do not meet the criteria for RPS being valid. Thus invalid actuations include actuations that are not the Reactor Trip (ECCS) for pressurized water reactors (PWR5) including: high-head, result of valid signals and are not intentional manual actuations.

intermediate-head, and low-head Except for actuations of the Reactor Protection System (RPS) when the injection systems and the low pressure reactor is critical or in MODE 1, invalid actuations are not reportable by injection function of residual (decay) telephone under 10 CFR 50.72. In addition, invalid actuations are not heat removal systems.

reportable under 10 CFR 50.73 in any of the following:

(4) PWR auxiliary or emergency feedwater system. - The invalid actuation occurred when the system is already properly (5) Containment heat removal and removed from service. This means all requirements of plant depressurization systems, including procedures for removing equipment from service have been met. It containment spray and fan cooler includes required clearance documentation, equipment and control systems. board tagging, and properly positioned valves and power supply (6) Emergency AC electrical power breakers.

systems, including: emergency diesel generators (EDG5) - The invalid actuation occurs after the safety function has already 10 CFR 50.72(b)(3)(iv)(A)(B) been completed. An example would be RPS actuation after the control rods have already been inserted into the core.

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ATTACHMENT 11.3 Page 4 of 5 EIGHT HOUR NOTIFICATIONS TO THE NRC EIGHT HOUR NOTIFICA11ONS TO THE NRC If not reported as a declaration of an Emergency Class under paragraph (a) of 10 CFR 50.72, HBRSEP shall notify the NRC Operations Center via ETS as soon as practical and in all cases within eiciht hours of the occurrence of any of the followinq:

EVENT KEY WORDS REQUIREMENT EXAMPLES CONDITION THAT COULD Loss of Safety Any event or condition that at the Loss (inoperability) of both Trains, e.g., ECCS, Low PREVENT FULFILLMENT OF Function time of discovery could have Temperature Overpressure Protection System, or Lake SAFETY FUNCTIONS Residual Heat prevented the fulfillment of the Robinson water level below LCO 3.7.8 limit Mitigation safety function of structures or systems that are needed to: Loss of one Train of required equipment, and the cause of the Shutdown -

failure could fail the other train, and there is a reasonable Generic (A) Shut down the reactor and maintain it in a safe expectation that the other train would not fulfill its safety Setpoint Drift shutdown condition; function if required Engineering Evaluation (B) Remove residual heat; 1) Contaminated lubrication fluid degrades SI Pump Operability (C) Control the release of operation (a single condition could prevent fulfillment of a Determination radioactive material, or safety function if both trains could be reasonably Common Mode (D) Mitigate the consequences expected to be inoperable).

Failure of an accident. 2) EDG Air Start Solenoids (if it demonstrates a design, procedural, or equipment deficiency that could prevent Events covered in this section may the fulfillment of a safety function, i.e., if both diesels are include one or more procedural susceptible to same problem) errors, equipment failures, and/or discovery of design, analysis, Multiple Control Rod failures (if failure prevented the fulfillment fabrication, construction, and/or of a safety function) procedural inadequacies. Operator action to inhibit the RPS (actions would prevent However, individual component fulfillment of a safety function) failures need not be reported in accordance with this paragraph if redundant equipment in the same system was operable and 10 CFR 50.72(b)(3)(v) available to perform the required safety function.

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ATTACHMENT 11.3 Page 5 of 5 EIGHT HOUR NOTIFICATIONS TO THE NRC EIGHT HOUR NOTIFICATIONS TO THE NRC If not reported as a declaration of an Emergency Class under paragraph (a) of 10 CFR 50.72, HBRSEP shall notify the NRC Operations Center via ETS as soon as practical and in all cases within eight hours of the occurrence of any of the following:

EVENT KEY WORDS REQUIREMENT EXAMPLES ISFSI DEFECT IMPORTANT TO

- ISFSI A defect in any spent fuel storage A defect discovered in the design or construction of ISFSI SAFETY Defect structure, system, or component that units that could result in releases or radiation doses to the Safety is important to safety, public in excess of 10 CFR 20 limits.

10 CFR 72.75(c)(1)

ISFSI REDUCTION IN

- ISFSI A significant reduction in the Wear or degradation of ISFSI units that could result in EFFECTIVENESS Confinement effectiveness of any spent fuel releases or radiation doses to the public in excess of 10 Reduction storage cask confinement system CFR 20 limits.

10 CFR 72.75(c)(2) Effectiveness during use.

TRANSPORT OF CONTAMINATED Contaminate Any event requidng the transport of a Any event requiring the transport of a radioactively INJURED PATIENT Injured radioactively contaminated person to contaminated or potentially contaminated (NUREG 1022)

Person an off-site medical facility for person to an off-site medical facility for treatment.

Medical treatment.

Transport Rescue 10 CFR 50.72(b)(3)(xii) Hospital 10 CFR 72.75(c)(3) ISFSI j AP-030 Rev. 48 Page 31 of 65