ML110980764

Initial Exam 2011-301 Draft SRO Written Exam

ML110980764
Person / Time
Site:	Browns Ferry
Issue date:	04/04/2011
From:	NRC/RGN-II/DRS/OLB
To:	Tennessee Valley Authority
References
50-259/11-301, 50-260/11-301, 50-296/11-301
Download: ML110980764 (203)

Text

{{#Wiki_filter:ES-401 Sample Written Examination Form ES.401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 295001 Partial or Complete Loss of Forced Core Flow Circulation 1 Tier # AA2.02 (IOCFR 55.43.5 SRO Only) Group # 1 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW K/A # 295001AA2.02 CIRCULATION: Neutron monitoring Importance Rating 3.2 Proposed Question: # 76 Unit I was at 100% Reactor Power when Reactor Recirc Pump IA tripped. Total Core Flow indication lowered to 50%. Which ONE of the following completes the statements? Following the trip, APRM Flow Biased Scram set point will be(1)_ Simulated Thermal Power. The APRM Flow Biased Simulated Thermal Power HIGH setpoint is required to be adjusted to Single Loop allowable value within _(2)_ in accordance with T.S. 3.4.1, Recirculation Loops Operating. A. (1)92% (2) 12 hours B. (1)92% (2) 24 hours C. (1)98% (2) 12 hours D. (1)98% (2) 24 hours Proposed Answer: D Explanation A INCORRECT: Part 1 incorrect Plausibility based on Flow biased setpoint (Optional): for Control Rod Block is 0.66(w-Aw) +59%. .66(50-0) +59 = 92% STP. Part 2 incorrect RPS Instrumentation set points for Single Loop Operation must be incorporated within 24 hours of entering SLO perTS 3.4.1. The 12 hour time is recognizable as the time required to place an Inop channel in trip per RPS Instrumentation TS. B INCORRECT: Part 1 incorrect See Explanation C. Part 2 incorrect See Explanation C. C IN CORRECT: Part 1 incorrect See Explanation C. Part 2 correct See Explanation B. D CORRECT: Part 1 correct Flow biased setpoint for reactor scram is 0.66(w-iw) + 65%. .66(50-0) +65 = 98% STP. Part 2 correct RPS - Instrumentation set points for Single Loop Operation must be incorporated within 24 hours of entering SLO perTS 3.4.1.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification: The KA is met because the question tests the candidates ability to determine and interpret APRM flow biased trip signals as they apply to a partial loss of forced core flow as a result of a trip of a Reactor Recirc Pump. SRO Only Justification: This question is SRO Only because it meets the requirements of Clarification Guidance for SRO, Section ll.B Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]. The question involves application of Required Actions (Section 3) in accordance with rules of application requirements (Section 1). See Attached. Candidate must determine time requirement to apply APRM Flow Biased Simulated Thermal Power HIGH setpoint as a result of the Recirc Pump Trip event. Question Cognitive Level: Question rated as C/A because Candidates must use multi-part mental process in recognizing the effects of a Recirc Pump trip and core flow reduction to predict the change to the APRM flow biased set point. Technical Reference(s): 1-AOl-68-1 Rev 3 (Attach if not previously provided) OPL171.148 Rev 12 Ui TS 3.4-1/2 Amm 266 (Including version I revision number) Ui TS B3.4-6 Rev. 45 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPLI 71.074 V.B.2 (As available) Question Source: BFN 0801 #91 (Note changes or attach parent) Question History: Last NRC Exam BFN 0801 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of ever, question.) Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Clarification Guidance for S RO-on ly Questions RevI (0311112010) Figure 1: Screening for SROonly linked to 10 CFR 5&43(b)(2) (Tech Specs> Can question be answered solely by knowing I hour TS/TRM Action? questiori

iNo Can question be answered solely by knowing the LCO/TRM information listed above4he-line? uestion h.

No P. Can question he answered solely by knowing the TS Safety Limits? stion INo Does the question involve one or more of the following for TS, TRM, or 00CM? P.

• Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
• Application of generic LCO requirements (LCO 3.0.1 thru 30.7 and SR 4.0.1 thru 4.04)
• Knowledge of TS bases that is required to analyze TS required actions and terminology No Question might not be linked to 10 CFR 55.43(bX2) for SRO-only Page 5 of 16

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL171.148 Revision 12 Page 23 of 106 INSTRUCTOR NOTES (2> The STP signal is used by the APRM for flow biased rod blocks and scram set points. (3) Flow Biased Scram and Rod Block generation (a) The APRM calculates a flow-biased setpoint by comparing reactor power and reactor recirculation flow (b) At 100% power, both w = flow as recirculation pumps are calculated by the running and the SLO value in APRM instrument. the flow biased calculation is zero (0) tS.w flow 0% for (c) With one recirculation pump 2 loop operation and tripped or secured, a 10% 10% for single loop bias is added to the flow operation. This is the biased calculation to add a conservative bias conservatism to the added to the calculation: calculation during single loop operation. Flow biased setpoint for reactor scram is 0.66(w-Aw) + 65% (e) Flow biased setpoint for Control Rod Block is Obj V.D.7.b 0.66(w-tw) +59% Obj V.D.7.c (f) Examples (i) Given that Neutron Flux indicates 85% .66(40-0) +65 power and Reactor scram setpoint Reclrculatiori Flov 91,4% STP indicates 40% flow, calculate the setpoint .66 (40-0) + 59 = rod block setpoint for the alarm and rod 85.4% STP block. (Assume both reactor recirculation pumps are running.) SLO

                                                     .66(40-1 0)+65 (ii)    How Is this different If 84.8% STP in single loop           .66(40-1 0)+59 operation?               78.8% STP

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Recirculation Loops Operating 83.4.1 BASES (continued) ACTIONS A. 1 With the requirements of the LCO not met, the recircul.ation loops must be restored to operation with matched flows within 24 hours. A recirculation loop is considered riot in operation when the pump in that loop is idle or when the mismatch between total jet pump flows of the two loops is greater than required limits. The loop with the lower flow must be considered not in operation. Should a LOCA occur with one recirculation loop not in operation, the core flow coastdown and resultant core response may not be bounded by the LOCA analyses. Therefore, only a limited time is allowed to restore the inoperable loop to operating status. Alternatively, if the single loop requirements of the LCO are applied to the operating limits and RPS setpoints. operation with only one recirculation loop would satisfy the requirements of the LCO and the initial conditions of the accident sequence. (continuedi BEN-UNIT 1 B 3.4-6 Revision T-45 February 27, 2007

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Recirc Pump TriplCore Flow Decrease 1-AOl-68-IA Unit I OPRMs Operable Rev. 0003 Page 4 of 12 2.0 SYMPTOMS CAUTIONS

1) Operation with one recirc pump out of service and the inservice jet pump loop flow 41 x 106 Ibm/hr (1-l1-68-46 or 1-11-68-48) can result in inaccurate core flow indication. This results from positive jet pump flow in the out of service loop being subtracted instead of added. If operation in this condition is required, contact Reactor Engineers to perform Attachment 2 of i-SR-3.4.i(SLO) to determine actual core flow and to substitute that value into the ICS as necessary.

2) immediately upon the opening of the DRIVE RUNNING contacts, the associated jet pump loop flow is subtracted even though the loop flow is still positive. This results in a severe indicated lowering in core flow, then as the tripped loop flow decays toward zero, the core flow indication will rise toward the actual value. The severity of the indicated core flow perturbation will depend upon the cause of the Recirc pump trip and the speed of the Recirc Drive prior to the trip.

3) [NER!c1. The Natural circulation line on the Power/Flow map only shows the approximate, nominal characteristic for operating with both Recirc loops out of service.

Inaccuracies are evident at low/no-flow conditions. Therefore, indicated core flow in natural circulation operation may NOT fall directly on The natural circulation line as depicted on the Power/Flow map. INRC IFi 96-016, SE SL 516J

4) Per Technical Specifications, the Reactor can be operated indefinitely with one Recirc loop out of service, provided the requirements of T.S. 3.4.1 are implemented within 24 hours of entering single loop operations.

5) Failure to monitor SJAE/OG CNDR CNDS PLOW, 1-Fl-2-42, on Panel 1-9-6 for proper flow may result in SJAE isolation.

6) Changes in Condensate System flow may require adjustment to CNDS SPE BYPASS FLOW CONTROL VALVE, i-FCV-002-0 190, either in the Control Room or locally.

Personnel adjusting this valve locally should be in direct communication with the Control Room NOTE Because a Reactor Recirc Pump seizure provides the same symptoms, the actions described herein cover that condition also. A seizure would most likely NOT be immediately discernible from other pump trips.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Recirculation Loops Operating 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched flows shall be in operation. OR One recirculation loop may be in operation provided the following limits are applied when the associated LCO is applicable:

a. LCO 3.2.1, AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR), single loop operation limits specified in the COLR;

b. LCO 3.2.2, MINIMUM CRITICAL POWER RATIO (MCPR),

single loop operation limits specified in the COLR;

c. LCO 3.3.1.1, Reactor Protection System (RPS)

Instrumentation, Function 2.b (Average Power Range Monitors Flow Biased Simulated Thermal Power High), Allowable Value of Table 3.3.1.1-1 is reset for single loop operation. APPLICABILITY: MODES 1 and 2. BEN-UNIT 1 3.4-1 Amendment No. 236-266 December 29, 2006

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Recircuation Loops Operating 34.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO Ai Satisfy the requirements 24 hours not met. of the LCO B. Required Action and 81 Be in MODE 3. 12 hours associated Completion Time of Condition A not met. OR No recirculation loops in operation. BEN-UNIT I 3.4-2 Amendment No. 236-266 December 29r 2006

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT OPL17t148 Revision 12 Page 23 of 106 INSTRUCTOR NOTES (2) The STP signal is used by the APRM for flow biased rod blocks and scram set points. (3) Flow Biased Scram and Rod Block generation (a) The APRM calculates a flow-biased setpoint by comparing reactor power and reactor recirculation flow (b) At 100% power, both w flow as recirculation pumps are calculated by the running and the SLO value in APRM instrument. the flow biased calculation is zero (0) A flow = 0% for (C) With one recirculation pump 2 loop operation and tripped or secured, a 10% 10% for single loop bias is added to the flow operation. This is the biased calculation to add a conservative bias conservatism to the added to the calculation: calculation during single loop operation. (d) Flow biased setpoini for reactor scram is O.66(w-Aw) + 65% (e) Flow biased setpoint for Control Rod Block is Obj V.D.7,b 0.66(w-w) +59% Obj V.D.7c (f) Examples (I) Given that Neutron Flux indicates 85% .66(40-0) +65 = power and Reactor scram setpoint Recirculatlon Flow 914% STP indicates 40% flow, calcu late the setpoint .66 (40-0) + 59 = rod block setpoint for the alarm and rod 85.4% STP block. (Assume both reactor recirculation pumps are running.) SLO

                                                                        .66(40-1 O)+65 (II)     How is this different if 84.8% STP in single loop           .66(40-1 0)+59 operation?               78.8% STP

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT RPS Instrumentation 3.31.1 3.3 INSTRUMENTATION 3.31.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.1.1-1. ACTIONS

                                                        -NOTE-Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Place channel in trip. 12 hours channels Inoperable. OR A.2 -----NOTE Not applicable for Functions 2.a, lb. 2.c, 2.d, or 2.f. Place associated trip 12 hours system in trip. B. NOTE---------- B.1 Place channel in one trip 6 hours Not applicable for system In trip. Functions 2.a, 2.b, 2.c, 2.d,or2.f. One or more Functions B.2 Place one trip system in 6 hours with one or more required trip. channels inoperable In both trip systems. (continued) BFN-U NIT I 3.3-1 Amendment No. 234, 262, 266 December 29, 2006

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN 0801 #91 Examination Outline Cross-reference: Level 202001 Recirculation Tier # A2.1O (IOCFR 55.43.5 SRO Only) Ability to (a) predict the impacts of the following on the Group # RECIRCULATION SYSTEM; and (b) based on those predictions, K/A # use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

• Recirculation pump seal failure Importance Rating Proposed Question: # 91 Unit I Reactor Recirculation Pump IA was removed from service due to the following indications:
• Number 2 Pump Seal Pressure, 1-Pl-68-63A, is indicating 800 psig and rising slowly
• Recirculation Pump IA Controlled Leakage isI.4 gpm and rising slowly Which ONE of the following completes the statements?

Recirculation Pump 1A parameters indicate _(1)_. The APRM Flow Biased Simulated Thermal Power HIGH setpoint is required to be adjusted to Single Loop allowable value within _(2) in accordance with T.S. 3.4.1, Recirculation Loops Operating. A. (1) a degraded Number 2 Seal (2) 12 hours B. (1) a degraded Number I Seal (2) 24 hours C. (1) a plugged Number 2 Restricting Orifice (2)12 hours D. (1) a plugged Number I Restricting Orifice (2) 24 hours Proposed Answer: B

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 295005 Main Turbine Generator Trip / Tier # 1 G2.1.32 (IOCFR 55A3.2 SRO Only) Ability to explain and apply system limits and precautions. Group # 1 K/A# 295005G2.1.32 Importance Rating 4.0 Proposed Question: # 77 Which ONE of the following completes the statements? In accordance with the Unit I Bases for Tech Spec 3.3.1.1, RPS Instrumentation, an RPS actuation is required as a result of Turbine Stop Valve Closure above a MINIMUM Reactor Power of _(1 )_ to ensure the _(2)_ Safety Limit is not exceeded. A. (1)25% (2) MCPR B. (1)25% (2) RPV Pressure C. (1)30% (2) MCPR D. (1)30% (2) RPV Pressure Proposed Answer: C Explanation A INCORRECT: Part 1 incorrect Plausible in that 25% is a recognizable (Optional): value associated with Main Turbine instrumentation Tech Specs. The feedwater and main turbine high water level trip instrumentation is required to be OPERABLE at 25% RTP. Part 2 correct as detailed in C below. B INCORRECT: Part I incorrect as detailed in A above. Part 2 is incorrect Plausible in that Closure of the TSVs results in the loss of a heat sink that produces reactor pressure. However, ensuring safety limit for RPV Pressure is not exceeded is not the bases for the TSV RPS actuation. C CORRECT: Part 1 correct This Function is required, consistent with analysis assumptions, whenever THERMAL POWER is 30% RTP. This Function is not required when THERMAL POWER is < 30% RTP since the Reactor Vessel Steam Dome Pressure High and the Average Power Range Monitor Fixed Neutron Flux High Functions are adequate to maintain the necessary safety margins. Part 2 correct The Turbine Stop Valve Closure Function is the primary scram signal for the turbine trip event. For this event, the reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the End of Cycle Recirculation Pump Trip (EOC-RPT) System, ensures that the MCPR SL is not exceeded. D INCORRECT: Part 1 correct as detailed in C above. Part 2 incorrect as detailed in B above.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification: The KA is met because the question tests the candidates ability to explain and apply limits associated with Main Turbine Generator Trip by asking the bases of RPS actuation in response to Turbine Control Valve closure and the Reactor Power limit for when the function is required. SRO Only Justification: This question is SRO Only because it meets the requirements of Clarification Guidance for SRO-only Section ll.B Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]. The question involves knowledge of TS bases for Turbine Stop Valve Closure. See attached Figure 1 flow chart. Question Cognitive Level: Question rated as Fundamental Knowledge. Technical Reference(s): ui TS B 3.3-23/24 Rev. 0 (Attach if not previously provided) Proposed references to be provided to applicants during examination: NONE Learning Objective: OPLI 71.028 V.B.9 (As available) Question Source: (Note changes or attach parent) New X Question History: (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.) Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 X Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Clarification Guidance for SRO.only Questions Rev 1(0311112010) Pigure 1: Screening for SRO.only linked to 10 CFR 55.43(b)(2) (Tech Specs) Can question be answered solely by knowing 1 Yes hour TSTRM Action? RO question h jNo 1 Can question be answered solely by knowing the LCOITRM information listed abovethe-line? [NO I I Can question be wered solely by knowing the Yes TS Safety Limits II No Does the question involve one or more of the following for TS, TRM, or 00CM?

• Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1>
• Application of generic LCO requirements (LCO 3.01 thru 3.0.7 and SR 4.0.1 thru 4.0.4) Yes SROonly

       . Knowledge of TS bases that is required to analyze TS                  question required actions and terminology No j

I Question might not be linked to I 10 CFR 5543(b)(2) for SRO-only

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 7a, 7b. Scram Discharge Volume Water Level High - SAFETY ANALYSES, {LS-85-45A, LS-85-45B, LS-85-45C, LS-85-45D, LCO, and LS-85-45E, LS-85-45F, LS-85-45G. and LS-85-45H) APPLICABILITY (continued) Four channels of each type of Scram Discharge Volume Water Level High Function, with two channels of each type in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from these Functions on a valid signal. These Functions are required in MODES 1 and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn. At all other times, this Function may be bypassed.

8. Turbine Stor Valve Closure Closure of the TSVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, a reactor scram is initiated at the start of TSV closure in anticipation of the transients that would result from the closure of these valves.

                    ,. The Turbine Stop Valve Closure Function is the primary scram signal for the turbine trip event analyzed in Reference 7.

J For this event, the reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the End of Cycle Recirculation Pump Tiip (EOC-RPT) System, ensures ThL that the MCPR SL is not exceeded. (continued) BEN-UNIT I B 3.3-23 Revision 0

ES-401 Sample Written Examination Form ES-401-5

    .                  Question Worksheet RPS Instrumentation B 3.3. 1.1 BASES APPLICABLE          8. Turbine Stan Valve Closure (continued)

SAFETY ANALYSES, LCO, and Turbine Stop Valve Closure signals are initiated from position APPLICABILITY switches located on each of the four TSVs. Two independent position switches are associated with each stop valve. One of the two switches provides input to RPS trip system A; the other, to RPS trip system B. Thus, each RPS trip system receives an input from four Turbine Stop Valve Closure channels, each consisting of one position switch. The logic for the Turbine Stop Valve Closure Function is such that three or more TSVs must be closed to produce a scram. This Function must be enabled at THERMAL POWER 30% RTP. This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening the turbine bypass valves may affect this function. The Turbine Stop Valve Closure Allowable Value is selected to be high enough to detect imminent TSV closure, thereby reducing the severity of the subsequent pressure transient. Eight channels of Turbine Stop Valve Closure Function, with four channels in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function if any three TSVs should This Function is required, consistent with analysis assumptions, whenever THERMAL POWER is 30% RTP. This Function is not required when THERMAL POWER is

                        <30% RTP since the Reactor Vessel Steam Dome Pressure            -

High and the Average Power Range Monitor Fixed Neutron Flux

                        -  High Functions are adequate to maintain the necessary safety margins.

close. (continued) BFN-UNIT I B 3.3-24 Revision 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference Level RO SRO 295016 Control Room Abandonment 1 Tier # - G2 1 7 (IOCFR 5543 5- SRO Only) Group # 1 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and KIA # 29501 6G2.1.7 instrument interpretation. Importance Rating 4-3 Proposed Question: # 78 The following occurred on Unit 3:

• Main Control Room has been evacuated due to toxic gas.
• NO Main Control Room actions could be performed.
• The Backup Control Panel is manned five (5) minutes after evacuation of the Main Control Room.
• Required actions from outside the Main Control Room have been performed.
• Twenty five (25) minutes later, the Unit Supervisor is informed that ONE SRV has been continuously open since the Backup Control Panel was manned AND a second SRV has been cycling periodically.

Which ONE of the following completes the statements? Reactor Power is determined to be between (1)_. In accordance with EPIP-1, Emergency Plan Implementing Procedure, the HIGHEST emergency action level classification that is required for these conditions is a (an) _(2). [REFERENCE PROVIDED] A. (1)6%andl4% (2) Alert B. (1) 15% and 23% (2) Alert C. (1)6%andl4% (2) Site Area Emergency. D. (1) 15% and 23% (2) Site Area Emergency Proposed Answer: C Explanation A INCORRECT: Part 1 correct See Explanation C. Part 2 incorrect See (Optional): Explanation B.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet B INCORRECT: Part 1 incorrect With power greater than 15% two SRVs would be open continuously. Part 2 incorrect Although the Backup Control Panel is manned within 5 minutes, the Alert is incorrect due to the inability to establish plant control within 20 minutes which includes controlling reactivity. C CORRECT: Part 1 correct each SRV will pass approximately 6.5% of total steam flow. With one SRV fully open and another cycling reactor power must be between the capacity of one and two relief valves. Part 2 correct - A Site Area Emergency must be declared due the inability to establish plant control within 20 minutes which includes controlling reactivity. D INCORRECT: Part 1 incorrect See Explanation B. Part 2 correct See Explanation C. KA Justification: The KA is met because the question tests the candidates ability to evaluate plant performance and make operational judgments. Based on SRV operation, candidate must conclude Reactor Power and make the required EAL Classification based on this evaluation of plant performance coupled with Control Room Abandonment. SRO Only Justification: This question meets the requirements of Clarification Guidance for SRO-only Questions, Section Il.F Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity. [10 CFR 55.43(b)(6)] (See Attached). This question requires evaluating core conditions based on operating characteristics and determining emergency classifications based on core conditions coupled with Control Room Abandonment. Question Cognitive Level: This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. Technical Reference(s): EPIP-1 Rev. 46 I OPL1 71.009 Rev. 11 (Attach if not previously provided) 3-AOl-i 00-2 Rev. 20 Proposed references to be provided to applicants during examination: EPIP-i EAL Matrix Section 6 Learning Objective: OPL171.075 V.B.2 (As available) Question Source: Bank # Clinton 07 #90 (Note changes or attach parent) Question History: (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.) Question Cognitive Level: Memory or Fundamental Knowledge

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X Comments: Clarification Guidance for SRO-oniy Questions Rev 1 (0311112010) F. Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming. and determination of various Internal and external effects on core reactivity. (10 CFR 5&43(b)(6)J Some examples of SRO exam items for this topic include:

• Evaluating core conditions and emergency classifications based on core r conditions.
• Administrative requirements associated with low power physics testing processes.
• Administrative requirements associated with refueling activities, such as approvals required to amend core loading sheets or administrative controls of potential dilution paths and/or activities.
• Administrative controls associated with the installation of neutron sources.
• Knowledge of TS bases for reactivity controls.

G. Fuel handling facilities and procedures. [10 CFR 55.43(bX7)j Some examples of SRO exam Items for this topic Include:

• Refuel floor SRO responsibilities.
• Assessment of fuel handling equipment surveillance requirement acceptance criteria.

Prerequisites for vessel disassembly and reassembly.

• Decay heat assessment.
• Assessment of surveillance requirements for the refueling mode.
• Reporting requirements.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QPLI7I .009 Revision 11 Page 15 of 63 (6) The worst over pressure transient is: (a) 3-second closure of all MSIVs neglecting the direct scram (valve position scram). (b) Results in a maximum vessel pressure which, if a neutron flux scram is assumed and 12 valves are operable, results in adequate margin to the code allowable over pressure limit of 1375 psig bottom head pressure. (7) To meet operational design, the analysis of the plant isolation transient (generator load reject without bypass valves) shows that 12 of the 13 valves limit peak pressure to a value well below the limit of 1375 psig.

b. The total safety / relief valve capacity has been established to meet the over pressure protection criteria of the ASME code.

(1) There are 13 Safety/Relief valves. (a) Each SRV has a capacity of 905,000lb/hr © 11 35psig. This gives a total capacity 84,1% (79.5% EPIJ) design steam flow at the reference pressure. (b) Valve leakage is detected by Obj. V.B.6 a temperature element and an Obj. V.C.4 acoustic monitor on each tailpipe. However, only the acoustic monitor will generate an alarm on panel 9-3.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I BROWNS FERRY I I EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX I E Pt PI CONTROL ROOM TURBINE FAILURE EVACUATION uescription Uescraption I I I I 6,3-UI I I I Turbine failure resulting in casing penetration OR C Significant damage to turbine or generator seals during operation, m OPERATING CONDITION: Model,or2 6.2-Al I I I 6,3-Al I I I Control Room Abandonment from entry into Turbine failure resulting in visible structural 1, 2, or 3-AOl-t00-2 or 0-SSll6 for ANY Unit damage to or visible penetration of ANY of the Control Room. following structures from missles:

                                                                 *Reaator Building *Diesel Generator Building          r

• Intake Structure *Control Bay OPERATING CONDITION:

OPERATING CONDITION: Model or 2 ALL 6,2-SI I I I Control Room Abandonment from entry into 1,2, or 3-AOl.l00-2 or 0SS[i6 for ANY Unit Control Room AND m Control of reactor water level, reactor pressure, m and reactor power (for Modes 1, or 2, or 3) or decay heat removal (for Modes 4, or 5) per 1,2, or 3-AOl-100-2 or O-SSI-16 as applicable, can 0 NOT be established within 20 minutes after evacuation is initiated. OPERATING CONDITION: ALL I I I I 0 m z PAGE 55 OF 206 REVISION 46

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Control Room Abandonment 3-AOL-I 00-2 Unit 3 Rev. 0020 Page 8 of 91 4.2 Unit 3 Subsequent Actions [1] IF ALL control rods were NOT fully inserted AND RPS failed to deenergize, TKEN:(Otherwise N/A) DIRECT an operator to Unit 3 Auxiliary Instrument Room to perform Attachment 9. D NOTES

1) The following transfers Reactor Pressure Control to Panel 3-25-32 to allow for pressure control while completing the Panel Checklist.

2) Attachment 7, Alarm Response Procedure Panel 3-25-32, provides for any alarms associated with this instruction.

CAUTIONS

1) Failure to place control switch in desired position prior to transferring to emergency position may result in inadvertent actuation of the component.

2) [NERIC] Operation from Panel 3-25-32 bypasses logic and interlocks normally associated with the components. tGESIL326. Sli

[2] PLACE the following MSRV control switches in CLOSE/AUTO at Panel 3-25-32: Switch No. Description 3-HS-1-22C MAIN STM LINE B RELIEF VALVE D 3-HS-1-5C MAIN STM LINE A RELIEF VALVE D 3-HS-I-41C MAIN STM LINE D RELIEF VALVE D 3-HS-1-34C MAIN STM LINE C RELIEF VALVE C

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Control Room Abandonment 3-AOl-I 00-2 Unit3 Rev. 0020 Page 9 of 91 4.2 Unit 3 Subsequent Actions (continued) [3] PLACE the tollowing MSRV disconnect switches in DISCT at Panel 3-25-32: Switch No Description 3-XS-1-4 MAIN STM LINE A RELIEF VALVE DISCT D 3-XS-1-42 MAIN STM LINE D RELIEF VALVE DISCT D 3-XS-1-23 MAIN STM LINE B RELIEF VALVE DISCT D 3-XS-i-30 MAIN SIM LINE C RELIEF VALVE DISCT D 3-XS-1 -1 80 MAIN STM LINE D RELIEF VALVE DISCT D [4] PLACE the following MSRV transfer switches in EM ERG at Panel 3-25-32: Switch No. Description () 3-XS-i-22 MAIN STM LINE B RELIEF VALVE XFR D 3-XS-1-5 MAIN STM LINE A RELIEF VALVE XFR D 3-XS-1 -41 MAIN STM LINE D RELIEF VALVE XFR D 3-XS-1-34 MAIN STM LINE C RELIEF VALVE XFR C NOTE Use of the following sequence when opening MSRVs should distribute heat evenly in the Suppression Pool. [5] MAINTAIN Reactor Pressure between 800 and 1000 psig using the following sequence at Panel 3-25-32: C A. 3-HS-1-22C, MAIN STM LINE B RELIEF VALVE C B. 3-HS-1-5C, MAIN STM LINE A RELIEF VALVE C C. 3-HS-1-41C, MAIN STM LINED RELIEF VALVE C D. 3-HS-1-34C, MAIN STM LINE C RELIEF VALVE C

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Question 4*: 090 Exam Date: 2007108120 Facility: 481 Reactor Type: BWR-GE6 Exam Level S K/A 295016 AA2.01 QUESTION: The plant was operating at near rated conditions. The Main Control Room has been evacuated due to toxic gas. No Main Control Room actions could be performed. Reactor pressure and water level control have been established at the Remote Shutdown Panel. Required actions from outside the Main Control Room have been performed. Twenty (20) minutes later, the CR5 is informed that one SRV has been continuously open since the Remote Shutdown Panel was manned and a second SRV has been cycling periodically. (1) Reactor Power is determined to be... (2) What Emergency Classification must be declared?

a. (1) Between 6% and 14%

(2) Alert

b. (1) Between 15% and 23%

(2) Alert

c. (1) Between 6% and 14%

(2) Site Area Emergency

d. (1) Between 15% and 23%

(2) Site Area Emergency ANSWER: c.

REFERENCE:

CPS 4003.01, Remote Shutdown R 13c EP-AA-1 003, Radiological Emergency Plan Annex For Clinton Station R 10 HIGHER NEW EXPLANA11ON: a is incorrect each SRV will pass approximately 8.5% of total steam flow. With one SRV fully open and another cycling reactor power must be between the capacity of one and two relief valves. b is incorrect the Alert is incorrect due to the inability to establish plant control. c is correct each SRV will pass approximately 6.5% of total steam flow. With one SRV fully open and another cycling reactor power must be between the capacity of one and two relief valves. A Site Area Emergency must be dedared due the inability to establish plant control within 15 minutes which includes controlling reactivity. d is incorrect with power greater than 15% two SRVs would be open continuously. K/A: Ability to determine and/or interpret the following as they apply to CONTROL ROOM ABANDONMENT: Reactor Power

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference Level R0 SRO 295021 Loss of Shutdown Cooling / Tier # 1 G2 44 (IOCFR 55432- SRO Only) Ability to recognize abnormal indications for system operating Group # 1 parameters that are entry-level conditions for emergency and K/A # 295021 G2.4.4 abnormal operating procedures. Importance Rating - 47 Proposed Question: # 79 Unit 1 is in Mode 4 with RHR IA in Shutdown Cooling. The Drywell Equipment Hatch is open. A leak on RHR Loop I results in the following:

• RHR LOOP I PUMP ROOM FLOOD LEVEL HIGH, (1-9-4C, Window 17), is in alarm
• RHR Loop I is secured AND isolated
• RHR Loop II is placed in service
• Reactor Coolant Temperature is 2150 F and rising Which ONE of the following completes the statements?

Entry into I-EOl-3, Secondary Containment Control, (1)_ required. In accordance with EPIP-1, Emergency Plan Implementing Procedure, _(2)_. [REFERENCE PROVIDED] A. (1)is (2) Emergency Action Level for an Alert is met B. (1)is (2) NO Emergency Action Levels are exceeded C. (1)is NOT (2) Emergency Action Level for an Alert is met D. (1)isNOT (2) NO Emergency Action Levels are exceeded Proposed Answer: A Explanation A CORRECT: Part I correct RHR LOOP I PUMP ROOM FLOOD LEVEL (Optional): HIGH alarm is indicative of Secondary Containment Area Water Level > 2 which is an EOI-3 entry condition. Part 2 correct Reactor moderator temperature can NOT be maintained below 212° F and that with Primary Containment not maintained, Technical Specifications requires Mode 4 conditions, an Alert is required in accordance with EAL 1.5-A. B INCORRECT: Part 1 correct See explanation A. Part 2 incorrect See Explanation D

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet C INCORRECT: Part 1 incorrect See explanation D. Part 2 correct See Explanation A. D INCORRECT: Part 1 incorrect Plausible in that not all alarms associated with degrading conditions occur at the EOl Entry level. Example Drywell Pressure High alarms prior to the EOl entry level. Additionally, EOI-3 Entry is not required in Modes 4 and 5. The candidate must recognize that the event led to change to Mode 3 and therefore, EOl entry is required. Part 2 incorrect Plausible in that if Tech Specs did not require Mode 4 conditions, this would be the correct answer. With the Primary CTMT open, Mode 4 conditions are required. KA Justification: The KA is met because it tests ability to recognize abnormal indications (RHR LOOP I PUMP ROOM FLOOD LEVEL HIGH! Loss of ShUtdown Cooling I Reactor Coolant Temperature 215° F) for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures. Entry levels met for EOl-3 and Loss of Decay Heat Removal EAL. SRO Only Justification: This question meets the requirements of Clarification Guidance for SRO-only Questions, Section ll.F Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity. [10 CFR 55.43(b)(6)] (See Attached). Candidate must evaluate core conditions and determine emergency classifications based on core conditions. They must recognize Reactor moderator temperature can NOT be maintained below 212° F and with Primary Containment not maintained, Technical Specifications requires Mode 4 conditions. This results in declaration of an ALERT. Question Cognitive Level: Question rated as C/A because Candidates must process multiple pieces of data including ECCS Room Flooded, elevated Reactor Coolant Temp, and Loss of S/D Cooling to ascertain EOI and EAL entry requirements. Technical Reference(s): EPIP-1 Rev. 46 / Ui TS 3.6.1.1 Amm 234 (Attach if not previously provided) 1 4C Rev. 18/ OPL1 71.204 Rev. 7 Proposed references to be provided to applicants during examination: EPIP-1 EAL Matrix Section 1 Learning Objective: OPL1 71 .075 V.B.2 (As available) OPL171.204 V.B.2

                                  -rwa Question Source Mod ified Bank #           BEN 1006 #79                   (Note changes or attach parent)

Question History: Last NRC Exam Browns Ferry 2010 (Optional Questions validated at the faculty since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.) Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet 10 CFR Part 55 Content: 55.41 55.43 X Corn ments: Comment made on sample submitted C. (1)is NOT (2) Emergency Action Level for an Alert is met. NRC If secondary containment control is not required (C.(1)), why would an alert be plausible? These two distracters do NOT go together well! The requirement to declare an Alert is not based on the EOl-3 entry but on Reactor Coolant Temp > 212 F and TS requiring Mode 4 conditions. In other words, if RHR Loop I Room level was at 1 inch rather than 2 inches, and all other conditions unchanged, this would be the correct answer and these two distractors would go together. Additionally, this answer was selected in validation.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Clarification Guidance for SRO.only Questions RavI (0311112010) F. Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity. [10 CFR 5543(b)(6)J Some examples of SRO exam items for this topic include:

• Evaluating core conditions and emergency classifications based on core conditions.
• Administrative requirements associated with low power physics testing processes.
• Administrative requirements associated with refueling activities, such as approvals required to amend core loading sheets or administrative controls of potential dilution paths and/or actMties.
• Administrative controls associated with the installation of neutron sources.
• Knowledge of TS bases for reactivity controls.

G. Fuel handling facilities and procedures. [10 CFR 55A3(b)(7)j Some examples of SRO exam Items for this topic include:

• Refuel floor SRO responsibilities.
• Assessment of fuel handling equipment surveillance requirement acceptance criteria.
• Prerequisites for vessel disassembly and reassembly.
• Decay heat assessment.
• Assessment of surveillance requirements for the refiAeling mode.
• Reporting requirements.
• Emergency classifications.

This does not Include Items that the RO may be responsible for at some sites such as fuel handling equipment and refueling related control room instrumentation operability requirements, abnormal operating procedure immediate actions, etc. For example, an RO is required to stop the refueling process when communication is lost between the control room and the refueling floor, therefore, this is a task that is both an RO and SRO responsibility and is not SROonly. Page 9 of 16

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL171.204 Revision 7 Page 5 of 52 A secondary containment floor drain sump water level above maximum normal operating level is an indication that steam or water may be discharging into secondary containment. Maximum normal operating floor drain sump water level is defined to be the highest value of secondary containment floor drain sump water level expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly.

f. Area water levels above 2 inches Secondary containment area water level above maximum normal operating level is an indication that steam or water may be discharging into secondary containment. Maximum normal operating secondary containment water level is defined to be the highest value of secondary containment area water level expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly.

g. Area radiation level above the maximum normal operating value of Table 4.

Secondary containment area radiation level above the maximum normal operating value of Table 4 is an indication that water from a primary system, or from a primary to secondary system leak, may be discharging into secondary containment. Maximum normal operating secondary containment area radiation level is defined to be the highest value of secondary containment areas radiation expected to occur during normal plant operating conditions with all directly associated support and control system functioning properly.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Primary Containment 161.1 3.6 CONTAINMENT SYSTEMS 3.6.1.1 Primary Containment LCO 3.61.1 Primary containment shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TI ME A. Primary containment A.1 Restore primary 1 hour inoperable, containment to OPERABLE status. B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. ANL B.2 Be in MODE 4. 36 hours BEN-UNIT 1 3.6-1 Amendment No. 234

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BROWNS FERRY I EMERGENCY CLASSIFICATION PROCEDURE j EVENT CLASSIFICATION MATRIX EPIP-1 t4-U I I a I I I I Valid MAIN STEAM LINE RADIATION HIGH-HIGH alarm, 1,2, or 3-RA-9O-135C OR Valid OG PRETREATMENT RADIATION HIGH alarm, 1,2, or3-RA-90-157A. OPERATING CONDITION: Modelor2or3 I I I I__ 1,5-Al I Reactor moderator temperature can NOT be I maintained below 212 F whenever Technical Specifications require Mode 4 conditions or during operations in Mode S. 9 OPERATING CONOmON: Mode 4 or 5 I I I I t5-S I CURVE I I I US Suppression Pool temperature, level and RPV pressure can NOT be maintained in the safe area of Curve 1.5-S. m m 0 m OPERATING CONDITION: ModeIor2or3 I I I I I I I 0 rn z rn r rn PAGE 23 OF 206 REVISION 46

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Example of elevated.parameter alarm to support distractors Drywell Pressure Abnormal alarms before EOl Entry of 2.45 psig BFN Panel 9-5 1-ARP-9-5B Unit I i-(A-55-5B Rev. 0016 Page 35 of 42 Sensor/Trip Point: DRYWELL 1-PS-064-0056E 1.65 psig rising 1 -PA-6456 I -PS.064-0056F 0.22 psig towering (Page 1 of 1) Sensor 1-LPNL-925-0005B Location: Elevation 593 Column No. S-R3 Probable A. Drywell S P air compressor failure. Cause: B. Loss of RBCCW. C. Breach of primary containment.

1. Drywell vent valves open or leaking.

2. Drywell vacuum breaker open or leaking.

D. LOCA. E. Sensor malfunction. Automatic None Action: Operator A. VERIFY the alarm using multiple indications. D Action: B. IF RBCCW has been lost, THEN REFER TO 1-AOl-70-1. C. REFER TO 1-AOl-64-1.

References:

1 45E6205-2 1-47E61 0641 1 730E91 5-17

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFNO8IO#79 Proposed Question: # 79 Unit I is shutting down for a refuel outage. The Drywell Equipment Hatch is open.

      . At T=1 2:00, Reactor Temperature is 153 °F Then, a complete loss of Shutdown Cooling occurs.

After 20 minutes, the operators determine that Reactor Coolant Temperature is rising at 16 °F every 10 minutes.

     . At T= 12:20 Reactor Coolant Temperature is 186 °F Which ONE of the following completes the statements?

If the heatup continues at the rate indicated above, a mode change would occur at )_. At T=1 2:45, in accordance with EPIP-1, Emergency Plan Implementing Procedure,. [REFERENCE PROVIDED] A. (1) T=1 2:28 (2) Emergency Action Levels for an Alert is met B. (1)T=12:37 (2) Emergency ActionLevels for an Alert is met C. (1)T=12:28 (2) NO Emergency Action Levels are exceeded D. (1) T=1 2:37 (2) NO Emergency Action Levels are exceeded

ES-401 Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 295024 High Drywefl Pressure Tier # EA2 08 (IOCFR 5543 5 SRO Only) Ability to determine and/or interpret the following as they apply to Group# - HIGH DRYWELL PRESSURE: K/A # 295024EA2.08

• Drywell radiation levels 40 Importance Rating Proposed Question: # 80 Unit 3 was operating at 100% Reactor Power, when a leak in the Drywell resulted in the following conditions:
• Drywell Pressure is 57 psig and rising
• Suppression Chamber Pressure is 54 psig and rising
• Suppression Pool Level is 15 feet
• Drywell Radiation is 2500 RIHr
• Primary Containment Venting will result in release rates above ODCM limits
• Reactor Water Level lowered to (-) 180 inches and is now (-) 170 inches and rising Given these conditions, which ONE of the following identifies the venting requirements in accordance with the EOls?

A. Vent the Drywell in accordance with 3-EOI-APPENDIX-1 3, Emergency Venting Primary Containment. B. Vent the RPV in accordance with 3-EOl-APPENDIX-1 5,RPV Venting for Primary Containment Flooding. C. Vent the Suppression Chamber in accordance with 3-EOl-APPENDIX-1 3,Emergency Venting Primary Containment D. Primary Containment CANNOT be vented because elevated Drywell Radiation levels will result in release rates above ODCM limits. Proposed Answer: C Explanation A INCORRECT: Plausible in that this would be the correct answer if (Optional): Suppression Pool Level was > 20 feet. Drywell Pressure of 57 psig also lends to plausibility. B INCORRECT: With reactor level at (-) 180 inches appendix 15 is plausible but incorrect with level now rising and above -180. Venting per Appendix 15 also requires maintaining offsite radiation levels within table 7 limits. C CORRECT: In accordance with 3-EOI-2, Primary Containment Control, with Suppression Chamber pressure 55 psig and Suppression Pool Level

                                   <20 feet, venting of the Suppression Chamber is required irrespective of offsite release.

ES-401 Written Examination Form ES-401-5 Question Worksheet D INCORRECT: Plausible in that this would be the correct answer if Suppression Chamber pressure of 55 psig was not challenged. KA Justification: The KA is met because the question tests the candidates ability to interpret Drywell Radiation levels as they apply to High Drywell Pressure. Candidate must determine that Venting of the Suppression Chamber is still required with knowledge that Primary Containment Venting will result in release rates above ODCM limits. SRO Only Justification: This question is SRO Only because it meets the requirements of Clarification Guidance for SRO, Section ll.E Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)] See Attached. Candidate must assess plant conditions and then selecting a procedure, 3-EOl-APPENDIX-13, Emergency Venting Primary Containment, due to high Suppression Chamber Pressure to mitigate the event. In making this selection candidate must further recognize that, venting is still required with the knowledge that Primary Containment Venting will result in release rates above ODCM limits. Question Cognitive Level: Question rated as C/A because it tests candidates ability to process multiple pieces of data including Drywell/SC Pressure, Drywell Radiation Levels, Reactor Level and Suppression Pool Level to ascertain Venting requirements. Technical Reference(s): 3-EOI-2 Rev 8 (Attach if not previously provided) Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL171.204V.B.13 (As available) Question Source: Modified Bank # Hatch 09 #97 (Note changes or attach parent) Question History: Last NRC Exam Hatch 2009 (Optional Questions validated at the faculty since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.) Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X Comments:

ES-401 Written Examination Form ES-401-5 Question Worksheet Clarification Guidance for SRO..only Questions Rev 1 (0311112010) Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5) (Assessment and selection of prccedures) Can the question be answered solely by knowing systems knowledge, Le, how the system works, J*fUeStiOn flowpaLh, logic, component location? No P. Can the question be answered solely by knowing immediate operator actions? J Yes 1 question I No Can the question be answered solely by knowing entry conditions far AOPs or plant parameters that require direct entry to major EOPs? No Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? _INoj Does the question require one or more of the following? L . Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps SRO -only
• Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
• Knowledge of administrative procedures that specify hierarchy, implementation, andior coordination of plant normal, abnormal, and emergency procedures No I Question might not be linked to 10 CFR 5543(b)(5) for SRO-only Page 8 of 16

ES-401 Written Examination Form ES-401-5 Question Worksheet b. pcipis L 3-EO-2 PAGE 1 OF I PRIMAFY CONTAINMENT CONTROL UNIT 3 BROWNS FERRY NUCLEAR PLANT RE/: 8

ES-401 Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT

                                            \

MONITOR AND CONTROL PC PRESS BELOW 24 PS) USING ThE VENT SYSTEM (APPX 2)AS NECESS4RY iL I I BFN 1-EOIAPPENDIX-12 UNIT I PRIMARY CONTAINMENT VENTING Rev. 0 Page 1 of 8 LOCATION: Unit 1 Control Room ATTACHMENTS: 1. Vent System Overview /

2. Post-LOCA Release Rate Table ( V )

CAUTION Stack release rates exceeding 1,4 x 10 jiCi!s, or 0-Sl.-4.8.8.1 ,a.1 release fraction above 1.0 will result in 00CM release limits being exceeded.

ES-401 Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT I-EOl APPENDIX-12 I I BFN UNIT I 12. PRIMARY CONTAINMENT VENTING ADJUST i-FIC-84-19, PATH B VENT FLOW CONT, or l-FIC-84-20, Rev. 0 I Page 5ofaI PATH A VENT FLOW CONT, as applicable, to maintain ALL of the following:

• Stable flow as indicated on controller, AND a 1-PA-84-21, VENT PRESS TO SGT HIGH, alarm light extinguished, AND a Release rates as determined below:

IF ................. PRIMARY CONTAINMENT FLOODING per C-I, Alternate Level Control, is in progress, THEN ........... MAINTAIN release rates below those specified in Attachment 2. ii. IF Severe Accident Management Guidelines are being executed, THEN MAINTAIN release rates below those specified by the TSC SAM Team. iii. IF Venting for ANY other reason than items i or ii above, THEN MAINTAIN release rates below

• Stack release rate of 1.4 x iO pCi/s AND h.
• O-Sl-4.8.B.1 .a.1 release fraction of 1.

ES-401 Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSBIUTY SUPPORT A

                                                                  .wsteu&a MQ L

WHILE EXECUTING STEPS C126 THROUGH C134: L c*4 4flIUE L I WHILE EXECUTING STEPS C148 THROUGH C1-34

                                *7 L

L 3j PAGE 1OF1 ALTERNATE LEVEL CONTROL UNIT 3 BROWNS FERRY NUCLEAR PLANT REV: g

ES-401 Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT (EOI 3-C-i Continued from previous page) OPERATE ALL AVAL4ELECE LA PPLWX 0c IL OPERATE E X7NO r4.JXXE& FAURL$VE PX EJRS os c R.0CUGiONY *40 PQ PGWT4GT ZT#Y lOG, 2O, LGEMRY e. SEY0O0LT 7*O 0I 70 PEG 70 a0Pr3 LEAAU XEAMS tAYLAElE IZ40O I1 AUX R7E 740 4C$T70NLY aL 1440P*0 G1 L thTERLPEIES 40T OONIINUE 1 h L I WHILE EXECUTING THE FOLLOWING STEPS:

                          !!                              ThEN V70R W S A7 4PL                RTOP *fl1NQ TE      (AX )

L 0131 I. MNTA1N RA1EE YIm?AR.E 747R I L I.

ES-401 Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSJBJLITY SUPPORT I. Pc4, 6 pc,p,4s L a-

ES-401 Written Examination Form ES-401-5 Question Worksheet HATCH 2009 97.023-U 001 A major Loss of Coolant Accident (LOCA) has occurred on Unit 2. The following conditions currently exist: o Drywell Pressure 54 psig, slowly increasing o Torus Pressure 52 psig, slowly increasing o Drywell Radiation 2500 R/Hr, slowly increasing o Reactor water level -165 inches, stable o Wide Range Torus Water Level > 300 inches Which ONE of the following identifies the required procedure to vent the Primary Containment and the release rate requirements during the venting process lAW 3 IEO-EOP-012-2, Primaty Containment Control? A. 34S0T48-002-2, Containment Atmospheric Control & Dilution Systems,. Section 7.3.3, Fast Drywell Vent; vent irrespective of offsite release rates. BY 3 1EO-EOP-l 01-2, Emergency Containment Venting; vent irrespective of offsite release rates. C. 34S0-T48-002-2, tontainment Atmospheric Control & Dilution Systems, Section 7.3.3, Fast Drywell Vent.; venting MUST be secured if approaching General Emergency Release Rate Limits. D. 3 1EO-EOP-l 01-2, Emergency Containment Venting; venting MUST be secured if approaching General Emergency Release Rate limits. Friday, May01. 2009 8:37:33 AM 172

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 295028 High Drywell Temperature Tier # 3 EA2.02 (IOCFR 55.43.5) SRO ONLY AbiUty to determine and/or interpret the following as they apply to Group # HIGH DRYWELL TEMPERATURE: F<IA # 295028EA2.02

• Reactor pressure Importance Rating Proposed Question: # 81 Given the following plant conditions on Unit 3:
• A steam line break has occurred inside the Drywell
• ALL Reactor Water Level (RWL) instruments display erratic indication
• Reactor Pressure AND Drywell Temperature are in the Action Required region of RPV Saturation Curve 8 Which ONE of the following completes the statement?

The Unit Supervisor must select EOI flowchart _(1)_ to perform Emergency Depressurization for these conditions. After initiating Emergency Depressurization, the crew must raise injection to establish Reactor Pressure a MINIMUM of(2) above Suppression Chamber Pressure. A. (1) 3-C-4, RPV Flooding (2) 70 psig B. (1) 3-C-2, Emergency Depressurization (2) 70 psig C. (1) 3-C-4, RPV Flooding (2) 90 psig D. (1) 3-C-2, Emergency Depressurization (2) 90 psig Proposed Answer: A Explanation A CORRECT: Part 1 correct 3-EOl-3-C-4 is required because all level (Optional): instruments are unavailable with Reactor Pressure and Drywell Temperature in the unsafe region of Curve 8 and erratic level instrument behavior. All actions associated with flooding and emergency depressurization are in 3-EOl-3-C-4. Part 2 correct In accordance with 3-EOI-3-C-4, after Emergency Depressurizing, the crew must raise injection to establish Reactor Pressure to Minimum RPV Flooding Pressure of 70 psig above SC Pressure but as low as practicable. B INCORRECT: Part 1 incorrect See Explanation D. Part 2 correct See Explanation A.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet o INCORRECT: Part 1 correct See Explanation A. Part 2 incorrect See Explanation D. D INCORRECT: Part 1 incorrect Plausible because 3-EOl-3-C-2 is the normal emergency depressurization flowchart. Part 2 incorrect Plausible because in accordance 1-EOI-1-C-4, the Minimum RPV Flooding Pressure for Unit 1 is 90 psig. Therefore, this would be the correct answer for Unit 1. KA Justification: The KA is met because the question tests ability to interpret Reactor Pressure as it applies to High Drywell Temperature. Candidate must recognize that with Drywell Temp I Reactor Pressure in the unsafe regions of the RPV Saturation curve and erratic level indications that all level indication is lost and then take appropriate actions in response. SRO Only Justification: This question is SRO Only because it meets the requirements of Clarification Guidance for SRO, Section ll.E Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)] See Attached. Candidate must assess plant conditions and to determine that with Drywell Temp I Reactor Pressure in the unsafe regions of the RPV Saturation curve and erratic level indications that 3-0-4, RPV Flooding must be selected to Emergency Depressurize the Reactor to mitigate the event. Question Cognitive Level: The question is high cognitive because; solving it involves a multi-part mental process of assembling, sorting, or integrating the parts to solve a problem. Technical Reference(s): 3-EOI-1 Rev 8 3-EOI-C-4 Rev 8

                                                      ,                                  (Attach if not previously provided)

Proposed references to be provided to applicants during examination: NONE Learning Objective: (As available) Question Source: Bank # Modified Bank # Hatch 09 #79 (Note changes or attach parent) New)i Question History: Last NRC Exam Hatch 2009 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.) Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 x Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Clarification Guidance for SRO-only Questions Rev 1(0311112010) Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b){5) (Assessment and selection of procedures) Can the question be answered so/ely by knowing systems knowledge, Le. how the system works, fIoath, logic, component location? question Ni Can the question be answered solely by knowing 1 immediate operator actions? Yes L RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameter that require direct entry to major EOPs? _ No Can the question be answered solely by knowinq the purpose, overall sequence of events, or es RC) question overall mitigative strategy of a procedure? Noj Does the question require one or more of the following? a Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed a Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps SRU onl\

• Knowledge of diagnostic steps and decision points in the uestio EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
• Knowledge of administrative procedures that specify hierarchy, implementation, andfor coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to I 10 CFR 5543(h)(5) for SRO-only Page 8 of 16

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet CAUTIO4 AM EN Si MAV AFFEG RW WAER LVL INDICATION

                                            ,r 1 L1      OG                                                                                        L
                          +                          MOI4ITiR AND CONTROL RFW WATER LVi.

L CS. LEVELPOWER ROL-I CONTROL VERIFY AS REQIONED:

• PCIS ISOLATIONS GROUPS 1.2W-DIM
• SODS

                                                       *RCIC L

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WT-10001007.COWWA I L 04-23

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT EMERGENCY RPV 3-C2 DEPRESSURIZATION WI EXECUTaN TN PROCEDURE:

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ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT 1-EOI-UC-4 OPENALL ADSiLV L CAN L OFEN ADDITIONAL I.SRWAS NECESSAR TO ESTAELhSS S MSR OPEN CAN ND AT LEAST j CLOSE THE FOLLOWING:

• VSL DRAINS
• RCID STEW iJNEISOLATIONVLVS L

r CAUTION 3 ELEVATED EU IPfI CHIER PRESS CtA TRIP RCIC V( -IPTI OR HOlD SUC7ION TEMP AEO.E Cf L RAISE AND CONTROL LI INTO THE RPV WITH THE FOLOWIN3 INJ SOURCES. IRREEPECTrE OFRUMP NPSH LIMITS AND SIIPPR PL LVL, TO RESTORE AND MAINTAIN:

• ATLEAST4 MSRV OPEN AND
• NW PRESS OT CROPPING AND RPUPRESSALESSTSJPSIAS3VE SUPPR CI-MER PRESS ECASLCN1.4S PRACTIDASLE ELI SOURCE APPX INJ PRESS

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet HATCH 2009 HLT 4 NRC Exam

79. 2j9002G2.4.20 001 Given the followine plant conditions:

o A steam line break inside the Dmvell o All RWL instruments display erratic indication simultaneously The steps to perform Emergency Depressurization for these conditions are contained on hOP flowchart Assuming Orwell temperature is at 215°F, the Wide Range RWL. Instmments be used. to determine RWL after the Minimum Core Flooding Interval has been completed. A. CPA ONLY: can B. CP-2 ONLY: can C. CP-l ONLY: can NOT D CP-2 ONLY; can NOT CPA is required because all level instruments are unavailable (erratic behaviorj. CP-l is the normal emergency depressurization flowchart, All actions associated with flooding and emergency depressunzation are on CP-2. Level instruments can not be used for indication again until DW temp is less than 210 deg F and reference legs refilled. Friday. May 31. 2009 8:37:30 A.1 130

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 295038 High Off-site Release Rate / Tier # 1 G2.4.9 (IOCFR 55435 SRO Only) - Knowledge of low power/shutdown implications in accident (e.g., Group # 1 loss of coolant accident or loss of residual heat removal) mitigation strategies. K/A # 295038G2.4.9 Importance Rating 4.2 Proposed Question: # 82 Unit 1 is at 100% Reactor Power when the following alarms are received o MAIN STEAM LINE RADIATION HIGH-HIGH 1-RA-.90-135C, (1-9-3A, Window 27)

• OG AVG ANNUAL RELEASE RATE EXCEEDED 1-RA-90-1 570, (1-9-40, Window 27)

The Unit Supervisor directs a Core Flow Runback AND Manual Scram to be inserted. Which ONE of the following completes the statement below for this condition? Immediately following the scram, the direction AND criteria to CLOSE MSIVs is contained in _(1)_ AND is based upon a determination that _(2)_. A. (1) 0-EOI-4, Radioactivity Release Control (2) releases are still in excess of Offsite Dose Calculation Manual limits B. (1) Alarm Response Procedure 1-9-3A, Window 27 Section for MAIN STEAM LINE RADIATION HIGH-HIGH (2) releases are still in excess of Offsite Dose Calculation Manual limits C. (1) 0-EOl-4, Radioactivity Release Control (2) the reactor will remain subcritical without boron under all conditions D. (1) Alarm Response Procedure 1-9-3A, Window 27 Section for MAIN STEAM LINE RADIATION HIGH-HIGH (2) the reactor will remain subcritical without boron under all conditions Proposed Answer: D Explanation A INCORRECT: Both parts incorrect as detailed in B,C and D. (Optional): B INCORRECT: First part correct as detailed in D below. Second part incorrect Plausible in that Main Steam Line Radiation High ARP contains action(s) associated with Offgas Radiation and ODCM limits. C INCORRECT: First part incorrect Plausible in that 0-EQ 1-4, does provide direction for isolating primary systems that are discharging into areas outside the primary and secondary containment. However, this step is not applicable under the specified conditions. Second part correct as detailed in D below.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet D CORRECT: The Main Steam Line Rad Hi-Hi alarm, once validated, requires a core flow runback followed by a manual scram. Additionally, ARP specifies that if not in C-5 that MSlVs must be closed. If the reactor is shutdown under all conditions without boron, EOl Contingency C-5 will not be executed. Candidate must understand strategies associated with EOl/Contingency implementation. KA Justification: The KA is met because it tests candidates knowledge of shutdown (ALL RODS IN) implications as they relate to excessive fuel failures inside the reactor core and the resultant high offsite release rates. As the ARP only specifies whether or not you are in C-5,additionally tests the candidates knowledge of strategies associated with EOl and EOl Contingency implementation SRO Only Justification: This question is SRO Only because it meets the requirements of Clarification Guidance for SRO, Section ll.E Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)] See Attached. Candidate must determine whether or not C-5 requires execution for these conditions. The question requires assessing plant conditions to determine if MSIVs should be isolated and selecting the procedure to that provides this guidance to mitigate the event. Question Cognitive Level: This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. Technical Reference(s): 1-ARP-9-4C, Rev. 18 (Attach if not previously provided) 1-ARP-9-3A, Rev. 40 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL171.009 V.B.14.a (As available) Question Source: Bank# Mod ified Bank # BEN 1006 #18 (Note changes or attach parent) r New Question History: Last NRC Exam Browns Ferry 1006 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the in formation will necessitate a detailed review of eveiy question.) Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Clarification Guidance for S RO-only Questions Rev 1(0311112010) Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5) (Assessment and selection of procedures) Can the question be answered so/ely by knowing systems knowledge, Le. how the system works, flowpath. logic, component location? No Can the question be answered solely by knowing immediate operator actions? Yes RO question

                                                           ]

Nol Can the question be answered solely by knowing IYes 1 entR conditions for AQPs or plant parameters i RD ciuestion that require direct entry to major EOPs? NI Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? Does the question require one or more of the following? Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps st RO onl
• Knowledge of diagnostic steps and decision points in the EOP5 that involve transitions to event specific sub-procedures or emergency contingency procedures
• Knowledge of administrative procedures that speci hierarchy, implementation, andlor coordination of plant normal, abnormal. and emergency procedures No Question might not be linked to I 10 CFR 5543(h)(5) for SRQ-only Page 8 of 16

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panel 9-3 1-ARP9-3A Unit I XA55-3A Rev. 0039 Page 40 of 52 Sensor/Trip Point: MAIN STEAM LiNE RADIATION i-RM-90-136 3.0 x normal full power background including i-RM-90-137 N-i 6 contribution and HWC System injection. (Page 1 of 1) Sensor Radiation monitor drawers are on Panel 1-9-10 in the control room. Location: Probable A. Radiation is three times the normal full power background. Cause: B. Sensor malfunctions. C. SI (SR) in progress. Automatic A. Mechanical vacuum pumps trip. Action: B. Vacuum pump sLiction valves 1-FCV-066-0036 and i-FCV-056-0040 close. Operator Action: { A. VERIFY the alarm on 1-RM-90-136 and 1-RM-90-137 on Panel 1-9-10. B. CONFIRM main steam line radiation level on recorder 1 -RR-90-l 35. Panel 1-9-2. C. IF alanii is valid and Reactor Scram has not occLlrred. THEN PERFORM the following:

1. IF core flow is above 60%, THEN LOWER core flow to between 50-60%.

2. MANUALLY SCRAM the Reactor.

C C C C

3. REFER TO i-AOl-lOG-i. C D. IF plant conditions DO NOT require execution of 1-C-S. THEN VERIFY the MSIVs closed. C E. NOTIFY RAD PRO. C F. VERIFY actions of i-ARP-9-3A Window 7 have been completed. C S. IF Technical Specifications limits are exceeded: THEN REFER TO EPIP-1. C

References:

147E610904 730E9159, 10 145E6205

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet CC1BCCR2( CAUTION AMBIENT TEMP MAY AFFECT PPINAER LVL INDICATION L A MO1ItRARDCONTROLRP/WAER LVi. L CC CDNWCL VERIFY AS REOURBD:

                                                          * ?CIStSOLAIONSPLROUPS 12,ANDS
                                                          . BCCS
                                                          . ECIC L

RCL-2 WHILE EXECUTING THE FOLLOWING STEPS: IF

                      *0         rHASH01EENOETEBJINED THAI fiB OTR WILL         LWIN SUBCRITICAL WfifiORON UNDER ALL CONDITIONS EXITRCC ANO ENTER 05. LLVELfPDWERCONTRCN.

EXITRCL AND R1U WAfiR/L CANNOT SE DEVERI.IINED ENTER Cl. WVELOODIND PC WATER LUL LALINi TEE Al NTAINED B BLOW 105 FT STCI NJ INTO fiB RPC FOil SOURCES oc EICSRNAL TO THE PC NOREOUIREO FOR ADEOUA B CORE CODLINO NJ P RN C N MEN P EBBS CAN N MAINTAINED BELOW EN REID RC.-3 N OThS

      /,     THE REAOTDR WILL REMAINSUITICALVNITHOJIEORON
      \I     UNDERALL CONDITIONS WHEN:

• ALL CONTROL RODSARE INSERTED TOOR SETOND POSITION D2
• ALL DONTROL RODS ExOEPrONE ARE INSERTED TO OR SEYOND POSITION o DETERMINED E REACTOR ENDINEERINO ISO STAFF LIA REC OUMEND AN ALTERNATE CURNE FOR STATION SLACKOUT PER O-ADI-S7A

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN Panel 9-4 1-ARP94C Unit I I-XA-554C Rev, 0018 Page 34 of 43 SisoriTri Poin OG AVG ANNUAL RELEASE LIMIT EXCEEDED 1-RE-090-0i57 2.5 R!hr (A[arm from recorder) C2500 mRfhr) 1 -RA-90-1 57C (Paqe 1 of 2) Sensor Elevation 565 Location: Turbine Buildincj Column B-T3 Recorder is on Panel t-J-2. Probable A. Abnormal flow in the off gas system. Cause: B.. Resin trap failure (RWCU or Condensate Demins). C Fuel damage.. Automatic None Action: Operator A.. DETERMINE if the Off Gas Annual Release Rate Limit is exceeded. Action: THEN PERFORM the iollowin9: 1 VERIFY alarm condition on tile following:

a. OFFGAS RADIATION. i-RR-90-266 on Panel 1-0-2.

b. OG PRETREATMENT RADIATION Recorder, 1-RR-90-266, Panel 1-9-2.

c.. OG PRETREATMENT RAD MON RTMR, i-RM-90-157 on Panel 1-9-10. B. NOTIFY Radation Protection. C NOTE High OffGas flow can sweep se:tied pariculates into flow stream and cause momentary rise in monhor reading. Low Off-Gas flow can result in improper dilution and cause monitor reading to hse. C, VERIFY Off-Gas flow normal and oroper samoe flow o the monitor. C Continued on Next Page

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN Panel 9-4 l-ARP-9-4C Unit 1 1-)(A-55-4C Rev. 0018 Page 35 of 43 OS AVG ANNUAL RELEASE LIMIT EXCEEDED, Window 27 (Page 2 of 2) C) perato r Action: (ContnLied) N OTE Load reduction may be recuired to keep Off-Gas within COCM limits.

0. REQUEST Chemistry perform radiochentical analvss to deternine source. D E. WITH OPS MGT acd Shift Managers permisson. PLACE charcoal beds in parallel with another unit. REFER TO 2-01-66. 0 F. IF fuel damaoe is suspected, THEN REFER TO l-SR-3.4.61 for dose equivalent iodine-131 determination. 0 S. REFER TO 0-31-4.8.6.1 .a.1 and SR-3.4.6,1-a for 00CM compliaoce and to determine if power level reduction is required. 0 H. IF directed by Shift rvlanaaer or Unit Supervisor. THEN REDUCE reactor power to maintain off-oas radiation wi:hn 00CM limits. 0

i. REFER TO EPIP-i. 0

References:

GE 729E814 Series 1-47E610-9C-1 00CM 5.5.1 FSAR Secaons 1.6.4.4.6, 7.12.2.2, and 13.6.2 Technca SpeciPcaions 4.6.6.6 and 4.8.6.1 .a.i TechnIcal Requirements Manual 3.3.91, 3.3.5.1 3,7.21

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN Panel 9.3 1.ARP-9-3A Unuit I XA-55-3A Rev. 0040 Page 14 of 52 Sensor/Trip Point: MAIN STEAM LINE RADIATION i-RM-90-136 Channel A [NRCiC] Setpoint is HIGH 1 -RA-90-1 35A i-RM-90-137 Channel C 1.5 X normal full power background including N-IS contribution and HWC System injection[nco 940247001] (Page 1 of 2) Sensor Panel 1 10 Location: Probable A. SI/SR in progress. Cause: B. Air injection from placing standby cond demin in service. C. Resin trap failure (RWCU or Cond Dernin). D Fuel damage. E. Sensor malfunction. F. ROIC in service.

0. Placing HWC in service.

Automatic None Action: Operator A. CHECK following radiation recorders on Panel 1-9-2: Action: 1. MAIN STEAM LINE RADIATION momtor, i-RR-90-l35. C

2. OFFGAS PRETREATMENT RADIATION, 1-RR-90-I57. C

3. OFFGAS POST-TREATMENT RADIATION, l-RR-9O-265. C

4. STACK GAS/CONT RM RADIATION, 0-RR-90-l47. C B. NOTIFY RAD PRO. C C. [NRC/C] REQUEST Chemistry to perform radiochemical analysis of primary coolant. [NCO 940247001] C D. IF off-gas PRETREATMENT RAD1ATION, i-RR-90-157, has risen significantly (30% above previous hour average). THEN REQUEST Chemistry to perform analysis of pretreatment off-gas. C E. SHUTDOWN Hydrogen Water Chernistr. REFER TO 1-01-4. C F. REFER TO D-SI-4.8.B. I.A.l for ODCM compliance and to determine if power level reduction is required. C Continued on Next Page

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN Panel 9-3 1-ARP-9-3A Unit I XA-55-3A Rev. 0040 Page 15 of 52 MAIN STEAM LINE RADIATION HIGH, 2-XA-55-135A, Window 7 (Page 2T2J Operator Action: kOontinued) G. [NRC/C] LOWER reactor power to maintain off-gas radiation within ODCM limits as directed by Unit Supervisor. [NCO 940247001] D H. IF ODCM limits are exceeded. THEN REFER TO EPIP-1.

References:

l47E6109Dl 47W00011 l45E6203 1 -729E8 14-i

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSBILLTY SUPPORT iL TABLE 5 OFFSITE RADIOACTIVITY RELEASE CLASSIFICATION LIMITS FOR GENERAL EMERGENCY T22 V2r2GrA:V iV2V7 O3 V V.2 a.as A.2VtOVV<w .V1.A2 N*.- <tV-tCVkV22 2 A

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AbA 5A5

                                                      <255Ab5225A525 trICx*2Vtt2.m.2<                     AAV \ASV 255 255 2) 2)55 A    1VV PAcE 1 OB 1 RADIOACTMTY RELEASE CONTROL UNiT 0 BROWNS FERRY NUCLEAR PLANT REV: 5

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BROWNS FERRY 1006 Examination Outline Cross-reference: 295038 High Off-site Release Rate / 9 G2.4.9 (10CFR 55.41.10) Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies. Proposed Question: # 18 1 Unit 1 has been operating for one week with increasing amounts of fuel bundle leaks. Suppression efforts have been unsuccessful and the trigger point for shutting down the reactor on excessive Stack release rates is rapidly approaching when the following alarms are received

• MAIN STEAM LINE RADIATION HIGH-HIGH 1-RA-90-135C, (1-9-3A, Window 27)
• OG AVG ANNUAL RELEASE RATE EXCEEDED 1-RA-90-157C, (1-9-4C, Window 27)

Which ONE of the following completes the statements for this condition? The requirement to insert a manual scram is directed by a valid _(l) alarm. Immediately following the scram, the ARP-specific IF I THEN directive to CLOSE MSIVs is based upon a determination that (2). A. (1) MAIN STEAM LINE RADIATION HIGH-HIGH 1-RA-90-135C (2) releases are still in excess of Offsite Dose Calculation Manual limits B. (I) OG AVG ANNUAL RELEASE RATE EXCEEDED 1-RA-90-157C (2) the reactor will remain subcritical without boron under all conditions C. (1) MAIN STEAM LINE RADIATION HIGH-HIGH 1-RA-90-135C (2) the reactor will remain subcritical without boron under all conditions D. (I) OG AVG ANNUAL RELEASE RATE EXCEEDED 1-RA-90-157C (2) releases are still in excess of Offsite Dose Calculation Manual limits Proposed Answer: C Explanation A INCORRECT: First part correct as detailed in C below. Second part is (Optional): incorrect in that action(s) will be required if ODCM limits are being exceeded; but closing of the MSIVs is not specified. The ARP for Main Steam Line Rad Hi-Hi is very specific on closing MSIVs if not in Level

                                      /Power Control Contingency C-5 though.

B INCORRECT: First part incorrect (See attached excerpts) The ARP for Offgas Average Annual Release Rate Exceeded is very specific on specifying a power reduction, but does not drive the scram directly; wherein the MS Line Rad Hi-Hi does. Second part correct as detailed in C below.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet c CORRECT: (See attached excerpts) The Main Steam Line Rad Hi-Hi alarm, once validated, requires a core flow runback followed by a manual scram. Additionally, ARP specifies that if not in C-5 that MSIVs must be closed. If the reactor is shutdown under all conditions without boron, EOl Contingency C-5 will not be executed. Candidate must understand strategies associated with EQ I/Contingency implementation. D INCORRECT: Both parts incorrect as detailed in A and B above. RO Level Justification: Tests candidates knowledge of shutdown (ALL RODS IN) implications as they relate to excessive fuel failures inside the reactor core and the resultant high offsite release rates. As the ARP only specifies whether or not you are in C-5,additionally tests the candidates knowledge of strategies associated with EOI and EOI Contingency implementation. Candidate must determine whether or not C-5 requires execution for these conditions. This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 295017 High Off-Site Release Rate Tier # 3 2.2.44 (IOCFR 55.43.5 SRO Only) Ability to interpret control room indications to verify the status and Group # operation of a system, and understand how operator actions and directives affect plant and system conditions. K/A # 29501 7G2.2.44 Importance Rating 4.4 Proposed Question: # 83 UNIT 2 was at 100% Reactor Power when an accident resulted in the following conditions:

• Main Steam Tunnel Temperature in the Turbine Building is 298 °F and rising.
• Main Steam Tunnel Temperature in the Reactor Building is 190 °F and rising.
• Main Steam Line C Inboard AND Outboard MSIVs can NOT be closed.
• Actual Dose Rate at the Site Boundary has been above the General Emergency limit for 16 minutes.
• NO Offsite Emergency Response Facilities are operational.

Which ONE of the following completes the statement? In accordance with the EOls, Emergency Depressurization (1)_ required to be performed for these conditions. The Shift Manager I Site Emergency Director _(2) delegate the determination of Protective Action Recommendation. A. (1) is (2) can B. (1)1sNOT (2) can C. (1)is (2) CANNOT D. (1)isNOT (2) CANNOT Proposed Answer: C Explanation A INCORRECT: Part 1 correct See Explanation C. Part 2 incorrect See (Optional): Explanation B. B INCORRECT: Part 1 incorrect Plausible in that there are not 2 areas above their MAX SAFE limit. If candidate considers only EOI-3 requirements, this would be selected as correct. Part 2 incorrect The Radiation Protection Manager is plausible in that his duties include assessment of site radiological conditions and recommendations for protective actions for onsite personnel.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet C CORRECT: Part 1 correct In accordance with 0-EQ 1-4, Radioactive Release Control, if ED will reduce discharge outside of Primary and Secondary Containment and offsite radiation release is challenging General Emergency limit at the site boundary, ED is required. With failure of MSL C to isolate and temperature in the Turbine Building steam tunnel 298 °F and rising, there is indication of primary system discharging outside Primary and Secondary Containment. Part 2 correct The Site Emergency Director must make any required recommendations (PARS) until the CECC is staffed. This responsibility cannot be delegated until CECC is in operation. Recommendations are required at General Emergency. D INCORRECT: Part 1 incorrect See Explanation B. Part 2 correct See Explanation C. KA Justification: The KA is met because candidate must interpret control room indications for high area temperatures and MSIV position indications along with high offsite release data to verify the status of Primary Containment Isolation to determine correct operator actions and Radiological Emergency Plan actions. SRO Only Justification: This question is SRO Only because it meets the requirements of Clarification Guidance for SRO, Section ll.E Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)] See Attached. Question requires detailed knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures based on interpretation of control room indications to verify that a leak is discharging outside of Primary and Secondary Containment. Also, determination of Protective Action Recommendations is a knowledge / ability unique to the SRO Position. Question Cognitive Level: This question is rated as C/A because it involves the multi-part mental process of assembling, sorting, or integrating the parts to solve the question posed in the stem. Technical Reference(s): 0-EOl-4 Rev 5 I OPL1 71.075 Rev. 25 (Attach if not previously provided) EPIP-5 Rev 39 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL171.075 V.B.7 (As available) Question Source: 8ank# Modified Bank # (Note changes or attach parent) New X Question History: Last NRC Exam (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eveiy question.)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X Comments: SAMPLE SUBMITTAL COMMENTS B. (1)is NOT (2) Radiation Protection Manager NRC Is the RPM an SRO qualified position? NO If not AND the SED is, then the RPM is NP and A and B are NP distracters. FerrResonse Changes made to question to incorporate comments. However, many examples can be found in previous NRC exams that responsibilities were tested and Non SRO positions were used in both the correct answer and the distractors, For Example: BFN 1006 #98 I BFN 0801 #95 I #79 / #72! Crystal River 07 SRO #21 I Dresden 07 #14! #80! Fit 08 #74! Hope Creek 07 #99 1 Nine Mile 08 #71 / Oyster Creek 08 #75 I RBS 08 #95 1 VY 09 #99, I believe it would be a reasonable expectation for the Shift Manager, who has the ultimate responsibility for initial implementation of the Emergency Plan, to be knowledgeable of some key functions and who is responsible. For example, is it a reasonable expectation for the SM to know who initially fills communicator role, starts dose calcs, etc? Even though he does not perform these functions himself, it is important know who does and therefore fair to test on it. The examples from previous NRC Exams listed above appear to support this view. We also have previous validation result in which, given the choice between whether operation or another discipline had a responsibility, the non-operations discipline was selected, C. (1)is (2) Site Emergency Director Is the SAD always an SRO qualified person? No

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Clarification Guidance for SRO-only Questions Rev 1(03/11(2010) Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5) (Assessment and selection of procedures) Can the question be answered solely by knowing systems knowledge, i.e.. how the system works, esRQqestioI flowpath. logic, component location? IN the question be answered solely by knowing immediate operator actions? Yes RO question Noj Can the question be answered solely by knowing ent conditions for AOPs or plant parameters testion that require direct entry to major EOPs? Noj Can the question he answered solely by knowing the purpose, overall sequence of events, or Yes RO question overall mitigative strategy of a procedure? H No Does the question require one or more of the following?

• Assessing plant conditions (normal, abnormal, or emergency and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
• Knowledge of when to implement attachments and appendices, incluclincj how to coordinate these items with procedure steps Yes SRO only Knowledge of diagnostic steps and decision points in the iestioi EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures Knowledge of administrative procedures that specify hierarchy. implementation, and/or coordination of plant normal, abnormal, and emergency procedures No I Question might not be linked to ID CFR 55.43(b)(5) for SRO-only PageS of 16

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet O-EOl-4, Radioactive Release Contro WHILE EXECUTING THIS PROCEDURE: LE T> FLJ JM P$ WVN !4 rLe L IL TAPILE 5 OFFSITE RADIOACT IVITY RELEASE CLASSIrICATION LIMITS FOR GENERAL EMERGENCY 1CRN WEt-CO - LW

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ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BROWNSFERRY GENERAL EMERGENCY EPIP5 3.0 EMERGENCY CLASSIFICATION ACTIONS WHEN... the TSC SED has assumed the responsibilities from the SM SED THEN ... CONTINUE in this procedure at Appendix G. Otherwise continue in this procedure. N OTE

• Procedure steps can be performed concurrently.

      . Procedure Step 3.2.1 CANNOT be delegated. All other procedure steps can be delegated.

• All procedure steps must be completed and remain under the direct oversight of the SED.
• Step 3.2.2 (15 Minutes) and Step 3.5 (60 Minutes) are timed.

CAUTION Ongoing or anticipated security events or severe weather may present a danger to normal staffing and other Emergency Plan implementation processes. Observed all procedural steps carefully during security related events. 3.1 Activation of the Emergency Response Organization (ERO) 3.1.1 NOTIFY.. .a Unit Operator of the General Emergency. Emergency Classification. AND DIRECT... the Unit Operator to implement EPIP-4, Appendix B. Unit Operator Notifications utilizing one of the options listed below: STAGING AREA (If events are on-going or anticipated that may present a danger to normal emergency center staffing such as security related issues.) D DRILL j EMERGENCY PAGE 2 OF 24 REVISION 0059

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BROWNSFERRY GENERAL EMERGENCY EPIP5 3.2 State of Alabama Notification N CT! Notification of the State of Alabama is required to be completed within 15 minutes from the time of emergency classification declaration. 3.2.1 COMPLETE Appendix A (Initial Notification Form) 32.2 DIRECT a member of the Operations staff to COMPLETE Appendix C (State of Alabama Notification) OR COMPLETE Appendix C LI NOTE Confirmation of State of Alabama Notification will be received from the ODS or from a member of the Operations Staff if the ODS could not be contacted. 3.2.3 State of Alabama Notification Confirmation 3.3 Evacuation of Non-Emergency Responders 3.3.1 IF... either of the following conditions exists: 1> A severe weather condition is currently in progress or is projected on-site, such as a tornado. OR

2) An on-site security risk condition exists that may present a danger to site personnel during the Assembly 1 Accountability process as determined by SED/Nuclear Security.

OR

3) Rapid Evacuation of the Protected Area (REPA) has been conducted.

THEN.. .DO NOT initiate the Assembly! Accountability Process AND CONTINUE in this procedure at Step 3.4. Otherwise continue in this procedure. PAGE 3 OF 24 REVISION 0039

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BROWNS FERRY GENERAL EMERGENCY EPIP5 APPENDIX A Page 1 of I GENERAL EMERGENCY INITIAL NOTIFICATION FORM

1. This is a Drill This is an Actual Event - Repeat This is an Actual Event

2. This is Browns Ferry has declared a GENERAL EMERGENCY affecting:

D UNIT I D UNIT2 D UNIT 3 E COMMON

3. EAL Designator: ONLY ONE EAL DESIGNATOR)

4. Brief Description of the Event:

5. Radiological Conditions: (Check one under both Airborne and Liquid column.)

Airborne Releases Offsite Liquid Releases Offsite Minor releases within federally approved limits E Minor releases within federally approved limits Releases above federally approved limits Releases above federally approved limits Release information not known Release information not known ( Tech Specs/ODCM) ( Tech SpecsiODCM)

6. Event Declared: lime: Time) Date:

7. The Meteorological Conditions are: (Use 91 meter data from the Met Tower)

Wind Direction is FROM: Vi,icl Speed: rn.p.h

8. Provide Protective Action Recommendation utilizing Appendix H: (Check either 1 or 2 or 3)

El Recommendation I Recommendation 2

           . EVACUATE LISTED SECTORS

• EVACUATE LISTED SECTORS i2 mile Radius & 10 miles downwind) WIND FROM (2 mile radius 8 5 mile downwnd) 2 DEGREES 2

           . Shelter remainder of 10 mile EPZ.

• SHELTER remainder of 10 m,le EPZ.
• Consider issuance of POTASSIUM Mark nd direction

                                                                                          . Consider issuance or POTASSIUM IODINE in accordance with the State Plan.                  p                   IODIDE in accordance with The State Plan.

A2, 82, F2, 02, E5,E1 0, FE, Fl 0, 05, 010 From 4°- 40° AZ B2, F2, G2, E5, FE, 05 A2, 82. P2, 02, F5, FID. (35, 010, Hi0 730 From 41°- A2, 82, F2, G2, F5, GE A2, B2. F2, 02, 05. 010, H10, 110 From 74°-92° AZ B2.F2, 02,05 A2, 82. P2, G2, AS, 05, HiD, 110. JI0, KID From 93°- 1370 AZ 82. F2, (32, AS, (35 A2. 82. P2, 02. AS, AiD, 110. J TO, KID From 138°- 203° A2, 82, P2, 02, A5 A2. 82 P2, G2, A5, Al 0. 85, BID From 204° 282°

                                                                           -          A2,   82, F2, 02, A5, 35 A2,    92,  F2, 02,  35, 810, ClO, DiD, E5,E10          From 283°-32S°        A2,   82, P2, 02, 85, ES A2. 82, F2. 02,  C10, DiD, E5,E10, F5, FlO           From 327° 30           A2,   82, F2, 02, ES, F5 Recommendation 3

• SHELTER all sectors
• CONSIDER issuance of Potassium Iodide in accordance with the State Plan,

9. Please repeat the information you have received to ensure accuracy.

Action: When completed, fax this appendix as prescribed by procedure. PAGE 7 OF 24 REVISION 0039

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet 0PL171.075 Revision 25 Page 19 of 50 INSTRUCTORS NOTES EPIP-5 contains the directions for activating the Review: EPIP-5 emergency response for the General Emergency Attachment C for and the guidance for making protective action PARs recommendations.

4. The Site Emergency Director must make any Obj. V.8.7 required recommendations until the 08CC is staffed. This responsibility cannot be delegated until CECC is in operation. Recommendations are required at General Emergency.

5. If this is the initial classification, the SM notifies the SM has 5 mm ODS within 5 minutes, and the ODS notifies the local governmental agencmes \vlthmn 1 C) minutes. OD has 15 mm and recommends protective actions. If in a General Emergency and ODS cannot be contacted use phone numbers at bottom of page 2 of EPIP-5 to contact local counties directly and State of Alabama Rad Health Duty Officer.

6. The initiating conditions and emergency action levels which require the General Emergency are explained in the Technical Basis. EPIP-5 directs a Review Appendix continuous mode of evaluation and reevaluation of changing conditions for the event using EPIP.

When those changes are recognized, they are to be communicated to offsite agencies.

7. A plant evacuation of non-emergency responders.

must be conducted in accordance with EPIP-8.

8. Discuss all sections of EPIP-5 and stress Obj. \.B. 9 Protective Action Recommendations (Appendix C).

H. Emergency Organizations EPIP-6 & 7

1. The onsite organization is composed of the Site Ohj. \I.B.1O Emergency Director and technical staff located in the Technical Support Center. the on-shift Operations personnel, and additional support NP REP Plan personnel in the Operations Support Center Appendix A

2. The Technical Support Center (TSC) is staffed EPIP-6 during an ALERT, SITE AREA EMERGENCY, or GENERAL EMERGENCY. TP-1

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet PLAUSIBILITY OF DISTRACTORS SUPPORT L L Sc L 2-EOI-3 PAGE 1 OF 1 SECONDARY CONTALNMENT CONTROL UNIT 2 BROWNS FERRY NUCLEAR PLANT REV: 12

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet PLAUSIBILITY OF DISTRACTORS SUPPORT TABLE 3 SECONDARY CNTMT AREA TEMP RNEL A-S PANEL A-2 WX .IAX ARES ALARM WPIDM TEVP ELEMENT NORMAL SAFE SO (JNLESSNOThO UNLESSNOTEO) VA_UEF VAWEF .SOJRCES RHR SS PUMPS XR-SS-SE-.1 ALSEMED &i FCV-?4-4 1S RHE SVSHPUMPS )LA-SE$E-1 FCV-14-4 745 HPC ROOM XA-SS-SF-iE 73- 5&R ALARMED 210 FCV-73-2, 3, 51, CS SYS I X4--30-10 71-41A ALARMED 103 FCV-?1-2. 3,33 57CC ROOM XA-SS-3D-13 71-1 15. C.D ALARMED 233 F7Ci-?1-2, 3 DPDF TORUS >-.-SS-3F-10 s,:.o ALAR.IED 240 FCV-73-2.251 XA-55-SE-4 74-050. H ALARMED FCV-l.1-4 , IS SSAMUNNE_ (R37 XA-5i,-SD-24 4DA 7PANELA-37 A7_AR,IED 3, F7Oi-s-7, 2, 2 TAL ACCESS )7A-SS-SE-4 14-ASS AARMD 130 FCV-74-4?,45 Z5 E 555W >IA-5!.-SB-32 PANE1.. 0-5) 60-525.5. SC, C ALARMED

                                                                                               -   FC3-7 2 12 (R\SCU PIPE TRENCH:    XA-5545-33 (PANEL A        (AUX INS     0DM)      ALARMED RWOUH.X.ROOM           >35-ES-3D-I?               55-20E,G,H             ALARMED      533  FCV-60-1.2,12 RW3UJMA                >35-SE-3D-Il               60-223                 ALARMED      275  FCV-03-,2, 72 RWOU J               D-17                       60-255                 ALARMED      25   FCV-A3- 2. 12 RE EL 503                  S5-5-.4                74-95C C               *SIAR.IED    105 FCV-74-4?.IR RE EL 62               XSS-3E-.l                 74-AEF                 ALARMED       135 FCM3-13, 14 2-EOl-3                                               PAGE 1         or: j SECONDARY CONTAINMENT CONTROL UNIT 2 BROWNS FERRY NUCLEAR PLANT REVS 12

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BROWNS FERRY ACTIVATION AND OPERATION OF THE TECHNICAL SUPPORT CENTER APPENDIX D Page 2 of 2 RADIATION PROTECTION (RP) MANAGER CHECKLIST Operational Responsibilities

• Direct and/or perform assessment of implant and onsite radiological conditions
• Direct onsite RP activities.
• Coordinate additional RP support with the CECC Radiological Assessment Manager.
• Notity Operations Manager of EOl-4 entry condition Offsite radioactivity release meeting nyEPlP-1, Section 11-4 event classification.) should any EAL be met in EPIP-i, Section 11-4, Radioactivity Release Make recommendations for protective actions for onsite personnel.
• Coordinates assessment of radiological conditions offsite with CECC Radiological Assessment Manager.
• Make final recommendation to SED for entries into radiological hazardous areas.
• Collect and provide plant radiological data to Emergency Centers as applicable.
• Provide assistance to the SED, as needed.
• Provide status update to the SED.
• Provide updates to the OSC RP Manager.
• Ensure accuracy of the RP status maps/boards in the TSC.

PAGE 12 OF 49 REVISION 0030

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO. SRO 295029 High Suppression Pool Water Level / Tier # 1 G2.4.47 (IOCFR 55.43.5 SRO Only) Ability to diagnose and recognize trends in an accurate and Groun# 2 timely manner utilizing the appropriate control room reference K/A # 295029G2.4.47 material. Importance Rating 4 Proposed Question: # 84 A leak into Unit 2 Suppression Pool has resulted in the following indications: At 0200 Suppression Pool Level is (-) 3 inch and rising at 1 inch per hour If the current trend continues, which ONE of the following completes the statement? The Unit Supervisor must direct lowering Suppression Pool Level in accordance with _(1). The Tech Spec Limit for 3.6.2.2, Suppression Pool Level, will be reached at (2)_. A. (1) 2-01-71, Reactor Core Isolation Cooling (2) 0315 B. (1) 2-01-74, Residual Heat Removal System (2) 0315 C. (1) 2-01-71, Reactor Core Isolation Cooling (2) 0400 D. (1) 2-01-74, Residual Heat Removal System (2) 0400 Proposed Answer: 0 Explanation A INCORRECT: Part 1 incorrect Plausible in that there is procedural (Optional): guidance for using RCIC in Suppression Pool level control strategy, i.e. can be used to make up with low Suppression Pool Level. Additionally, the physical capability does exist to remove water from the Suppression Pool with RCIC Suction from the Suppression Pool and Test Return to the CST or opening drains with Suction aligned to the Suppression Pool. However, there are no instructions in 2-01-71 for lowering Suppression Pool Level. Part 2 incorrect Plausible in that the Suppression Chamber Water Level Abnormal will be received at this time due to high water level. B INCORRECT: Part I correct See Explanation D. Part 2 incorrect See Explanation A. C INCORRECT: Part 1 incorrect See Explanation A. Part 2 correct See Explanation D. D CORRECT: Parts 1 correct 2-ARP-9-3B, Window 15 Suppression Chamber Water Level Abnormal has crew refer to 2-01-74. Section 8.2 of this 01 provides instructions for removing water from the Suppression Pool. Part 2 correct Tech Spec Limit of (-) 1 inch will be reached at 0400.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification: The KA is met because the question tests candidates ability to diagnose and recognize high Suppression Pool Water Level trend in an accurate and timely manner utilizing the appropriate control room reference material. SRO Only Justification: This question is SRO Only because it meets the requirements of Clarification Guidance for SRO, Section ll.E Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)] See Attached. Question requires assessing plant conditions associated with high Suppression Pool Level due to leakage into the SP and then selecting the appropriate procedure to control Suppression Pool Level. Question Cognitive Level: This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. Technical Reference(s): U2 TS 3.6-29 Am 253 (Attach if not previously provided) 2-ARP-9-3B Rev. 25/ 2-01-74 Rev. 152 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL1 71.016 V.B.1 2 (As available) Question Source: - Bank# - Modified Bank# [ a * Can qestioi be ans ed solely by iowingl_-___ knowledge. i.e, how the system works, Yec I Can the quon be answered solely by knowinci immediate operato r actions? Ye RO ciue1Ion Can the question he answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? Noj Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? No Does the question require one or more of the following? L Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps
• Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
• Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to I ID CFR 55.43(b)5) for SRO-onlv Page 8 of 16

ES-401 Sample Written Examination Form ES-401 -5 Question Worksheet Suppression Pool Water Level 3.6.2.2 36 CONTAINMENT SYSTEMS 3.6.2.2 Suppression Pool Water Level LCO 3.6.2.2 Suppression poo water level shall be -6.25 inches with and -7.25 inches without differential pressure control and -1.0 inches. APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETIoN TIME A. Suppression pool water Al Restore suppression pool 2 hours level not within limits. water level to within limits. B. Required Action and B.[ Be in MODE 3. 12 hours associated Completion Time not met. P B.2 Be in MODE 4. 36 hours BEN-UNIT 2 3.6-29 Amendment No. 253

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN 2-XA-55-3B 2-ARP-9.3 Unit2 Rev. 0025 Page 18 of 38 Sensor/Trip Point: SUPPR CHAMBER LT-64-54 S -5.5 H0 WATER LEVEL - A B N ORMAL 2-LA-64-54A (Page 1 of 1 Sensor RX Bidg, El 519 Location: NW corner room just inside door Probable A. Suppression Chamber water level abnormal. Cause: B. Placing Suppression Pool Cooling in service C. Sensor malfunction. Automatic None Action: Operator A. CHECK Suppression Pool level using multiple indications. Action: B. IF level is low, THEN DISPATCH personnel to check for leaks. D C. IF level is high. THEN CHECK for RCIC. HPCI. Core Spray, or RHR draining to Suppression Pool. and CHECK 2-TR-64-161 and -162. D D. REFER TO 2-01-74, Section 8.0. D E. REFER TO Tech Spec 3.6.2.2 0 F. IF level is above -I or below -6.25, THEN ENTER 2-E0l-2 Flowchart. C

References:

245E6203 247E610641 GE 730E9431 Technical Specifications 3.6.2.2

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Residual Heat Removal System 2..Ol-74 Unit2 Rev.0152 Page 4 of 442 Table of Contents (continued) 6.0 SYSTEM OPERATIONS 46 7.0 SHUTDOWN 47 7.1 Loop 1(11) LPCI Shutdown 47 8.0 INFREQUENT OPERATIONS 50 8.1 RHR System Fill and Vent 50 8.1.1 RHR Loop I Fill and Vent using RHR SYS I CNDS FLUSH & FILL, 2-SHV-074-0699 50 8.1.2 RHR Loop II Fill and Vent via CNDS FLUSH & FILL TO DW SPRAY HDR, 2-SHV-074-0675 55 8.1.3 RHR Loop I and Loop II Fill and Vent using CS SYSTEM I & II FILL FROM CONDENSATE SHUTOFF VALVE, 2-SHV-075-0700 60 8.1.4 Returning a Loop I RHR Pump and Heat Exchanger to Service in an Operable Loop 65 8.1.5 Returning a Loop II RHR Pump and Heat Exchanger to Service in an Operable Loop 70 8.2 Removing Water from the Suppression Pool 76 8.3 Adding Water to the Suppression Pool Through RCIC(HPCI) Minimum Flow Line 80 8.4 Adding Water to the Suppression Pool when Condensate Transfer is Lined up to Core Spray 82 8.5 Initiation of Loop 1(11) Suppression Pool Cooling 83 8.6 Shutdown of Loop 1(11) Suppression Pool Cooling 91 8.7 Loop 1(11) Flush for Shutdown Cooling 94 8.8 InitiationlOperation of Loop 1(11) Shutdown Cooling 110 8.8.1 Initiation / Operation of RHR Loop I in Shutdown Cooling 110 8.8.2 Initiation / Operation of RHR Loop II in Shutdown Cooling 136 8.9 Loop 1(11) Shutdown Cooling Shutdown to Standby Readiness 161 8.10 Initiation of Supplemental Fuel Pool Cooling with RHR Drain PumpA(B) 165 8.11 Shutdown of Supplemental Fuel Pool Cooling with RHR Drain Pump A(B) 171

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet PLAUSIBILITY SUPPORT OPL 171040 Revision 23 Appendix C Paqe6i of 74 RGIC AREA :2 TEWERATURE I, Lk HghStiLe Few ticn Line TP-1: RCIC System Flow Path

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet PLAUSIBILITY SUPPORT BFN SUPPRESSION POOL WATER INVENTORY 1EOfAPPENDlXJT9 UNIT1 REMOVALAND MAKEUP Rev.O Page 3 of 6 4.a Continued)

3) CLOSE and LOCK i-SHV-074-0765A(B), RHR DR PMP NB) DISCH

b. CLOSE l-FCV-74-108, RHR DR PUMP IA/B DISCH HDR VALVE.

c. VERIFY CLOSED I-FCV-74-62. RHR MAIN CNDR FLUSH VALVE.

d. VERIFY CLOSED i-FCV-74-63, RHR RADWASTE SYS FLUSH VALVE.

e. WHEN Suppression Pool level can be maintained between -1 in. and -5.5 in..

THEN EXIT this procedure.

5. IF Directed by SRO to Emergency Makeup to the Suppression Pool from Standby Coolant, THEN CONTINUE in this procedure at Step 9 using RHR Loop II.

6. IF Directed by SRO to add water to suppression pool, THEN MAKEUP water to Suppression Pool as follows:

a. VERIFY OPEN i-FCV-73-4O HPCI OST SIJ OTION VALVE.

b. OPEN I-FCV-73-30, HPCI PUMP MIN FLOW VALVE.

c. IF HPCI is NOT available for Suppression Pool makeup, THEN MAKEUP water to Suppression Pool using RCIC as follows:

1) VERIFY OPEN 1-FCV-71-19, RCIC CST SUCTION VALVE.

2) OPEN i-FCV-71-34, RCIC PUMP MIN FLOW VALVE.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet PLAUSIBILITY SUPPORT BFN 2-XA-55.3B 2-ARP9-3B tJnjt2 Rev.0025 Page 18 of 38 SensopTrip Point: SUPPR CHAMBER LT-64-54 -5.5 H-O WATER LEVEL .. - ABNORMAL -L15 H 0 2 2-LA-64-54A (Page 1 of 1) Sensor RX Bldg. El 519 Location: NW corner room just inside door Probable A. Suppression Chamber water level abnormal. Cause: B. Placing Suppression Pool Cooling in service C. Sensor malfunction. AutomatIc None Action: Operator A CHECK Suppression Pool level using multiple indications. C Action: B. IF level is low. THEN DISPATCH personnel to check for leaks. C C. IF level is high. THEN CHECK for RCIC. HPCI, Core Spray, or RHR draining to Suppression Pool, and CHECK 2-TR-64-161 and -182. C D. REFER TO 2-01-74. Section 8.0. C E. REFER TO Tech Spec 3.6.2.2. C F. IF level is above -i or below -6.25, THEN ENTER 2-EOl-2 Flowchart. C

References:

245E6203 247E6l084i GE 730E9431 Technical Specifications 3.6.2.2

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 295032 High Secondary Containment Area Temperature 1 Tier # EA2.02 (IOCFR 55A3.5 SRO Only) Group # 2 Ability to determine and/or interpret the following as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE K/A # 295032EA2.02 Equipment operability Importance Rating Proposed Question: # 85 Unit 3 was operating at 100% Reactor Power. RHR Pump 3B was tagged out for planned maintenance at 0600 on 1/13/11. At 1000 on 1/14/11, a RCIC steam line leak occurred in the Reactor Building resulting in degraded performance of Loop I Core Spray Room Cooler. Loop I Core Spray Pump area temperature is 1500 F. Based on these conditions, which ONE of the following identifies the EARLIEST time that Unit 3 must be in Mode 3 in accordance with Tech Spec 3.5.1, ECCS-Operating? [REFERENCE PROVIDED] A. 2200on 1/14/11 B. 2300on 1/14/11 C. l8000n 1/20/11 D. 2200on 1/21/11 Proposed Answer: B j Explanation A INCORRECT: Plausible in that this would be the correct answer if TS 3.0.3 (Optional): required Mode 3 in 12 hours. B CORRECT: With Core Spray 3A Pump Room Cooler inoperable, TRM 3.5.3 requires declaring Core Spray Loop I inoperable immediately. With Loop I CS and RHR Pump 3B INOP, TS 3.5.1 Condition H requires TS 3.0.3 Immediately. TS 3.0.3 requires Mode 3 in 13 hours. C INCORRECT: Plausible in that this would be the correct answer if loss of the Core Spray Room Cooler did not require declaring the associated ECCS Pump inoperable. D INCORRECT: Plausible in that this would be the correct answer if candidate believed that one inoperable RHR Pump does not result in the loop being considered inoperable and therefore not entering Condition A until the subsequent Core Spray 3A inoperability.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification: The KA is met because the question tests ability to determine and/or interpr et Equipment operability (Operability of ECCS Room Cooler and its impact on operability of CS System) as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE (A steam leak in RCIC Room and CS Pump Room Temp 1500 F). SRO Only Justification: This question is SRO Only because it meets the requirements of Clarification Guidance for SRO, Section lI.B Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]. The question involves application of Required Actions (Section 3) in accordance with rules of application requirements (Section 1). See Attached. Candidate must determ ine that the CS Room Cooler is inoperable since it cannot maintain area temperature < 148° F and then determine that CS Loop I must also be declared inoperable. Then, they must apply the requirements of TS 3.5.1 and TS 3.0.3. Question Cognitive Level: This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question and use reference to solve a problem. Technical Reference(s): U3 TS 3.5-1 to la Amm 244 (Attach if not previously provided) U3 TS 3.5-3 Amm 229 TRM 3.5.3 Rev. 0 U3 TS 3.5.1 Bases Rev. 0 U3 TS 3.0.3 Amm 226 Proposed references to be provided to applicants during examination: TS 3.5.1 No Bases Learning Objective: OPL171.045 V.B.6 (As available) Question Source: Bank # Modified Bank# (Note changes or attach parent) New X Question History: Last NRC Exam - (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the in formation will necessitate a detailed review of every question.) Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Clarification Guidance for SRO-only Questions Rev 1(0311112010) Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2) (Tech Specs) Can question be answered solely by knowing 1 Yes hour TSFTRM Action? R() question No Can question be answered solely by knowing the LCQ!TRM information listed above-the-lineT Qri.estior INo Can question be answered solely by knowing the TS Safety Limits? Oquestior LNO Does the question involve one or more of the following for TS. TRM, or ODCM?

• Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section i)
• Application of generic LCD requirements (LCD 3.0.1 thru 3.0.7 and SR 4.0.1 him 4.0.4> Yes SRO-only
• Knowledge of TS bases that is required to analyze TS question required actions and terminology No Question might not he linked to 10 CFR 55A;3(b)( ) for SRO-only Page 5 of 16

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Equipment Area Coolers TR 3.5.3 TR 3.5 EMERGENCY CORE COOLING SYSTEMS TR 3.5.3 Equipment Area Coolers LCO 3.5.3 The equipment area cooler associated with each RHR pump and the equipment area cooler associated with each set of Core Spray pumps (A and C or B and 0) must he OPERABLE at all times when the pump or pumps served by that specific cooler is considered to be OPERABLE. APPLICABILITY: Whenever the associated subsystem is required to be OPERABLE ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.i Declare the pump(s) Immediately Equipment Area served by that cooler Cooler inoperable, inoperable. (Refer to applicable TS and TRM LCOs) TECHN ICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY TSR 3.5.3.1 Verity each Equipment Area Cooler 92 days automatically starts when the associated Core Spray or RHR pump is started. BEN-UNIT 3 3.5-5 TRM Revision 0

ES-401 Sample Written Examination Form ES-401 -5 Question Worksheet ECCS Operatino 3.5.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TI ME G. Two or more ADS valves (3.1 Be in MODE 3. 12 hours inoperable. AND OR (3.2 Reduce reactor steam 33 hours Required Action and dome pressure to associated Completion 150 psig. Time ot Condition C, D. E. or F not met. H. Two or more low pressure H.i Enter LCO 3.0.3. Immediately ECCS injection/spray subsystems inoperable for reasons other than Condition A. OR HPCI System and one or more ADS valves inoperable. BFN-IJNIT 3 3.5-3 Amendment No. 2-i2 229 March 12. 2001

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet LCO Applicability 3.0 3.0 LIMITING CONDITION EOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCOs shall he met duhng the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2 arid LCO 3.0.7. LCO 3.0.2 Upon discovery of a tailLire to meet an LCO, the Required Actions of the associated Conditions shall he met, except as provided in LCO 3.0.5 and LCO 3.0.6. If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required, unless otherwise stated. LCO 3.0.3 When an LCO is riot met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, he unit shall he placed in a MODE or other specified conchtion in which the LCO is not applicable. Action shall be initiated within 1 hour to place the unit, as applicable, in:

a. MODE 2 within 1Cr hours;

b. MODE 3 within 13 hours; and

c. MO[:E 4 within 37 hours.

Exceptions to this Specification are stated in the individual Specifications. Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LCO 3.0.3 is not required. LCO 3.0.3 is only applicable in MODES 1, 2, and 3. (continued) BEN-UNIT 3 3.0- I Amendment No. 21-27 226 November 21. 2000

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet EOOS Operating B 3.5.1 BASES BACKGROUND P05 is still pressurized. If HPCI fails, it is hacked up by ADS in (continued) combination with LPOI and OS. In this event, either the vessel would be manually depressurized or the ADS timed sequence would be allowed to time out and open the selected safety/relief valves (S/RVs) depressurizing the P05, thus allowing the LPOI and OS to overcome ROS pressure and inject coolant into the vessel. If the break is large, ROS pressure initially drops rapidly and the LPOI and CS cool the core. Water from the break returns to the suppression pool where it is used again and again. Water in the suppression pool may be circulated through a heat exchanger cooled by the RHR Service Water System. Depending on the location and size of the break, portions of the EGOS may he ineffective: however, the overall design is effective in cooling the core regardless of the size or location of the piping break. All EGOS subsystems are designed to ensure that no single active component failure will prevent automatic initiation and successful operation of the minimum required EGOS equipment. The OS System (Ref. 1; is composed of two independent subsysiems. Each subsystem consists of two 50% capacity motor driven pumps, a spray sparger above the core, and piping and valves to transfer water from the suppression pool to the sparger. The LOCA analysis (Ref. 13) requires both pumps in a subsystem (loop) to be OPERABLE for the subsystem to be OPERABLE. Failure of one OS pump results in the loss of the associated OS loop for LOCA mitigation. The OS System is designed to provide cooling to the reactor core when reactor pressure is low. Upon receipt of an initiation signal, the CS pumps in both subsystems are automatically started (A pump fcontinued) BFN-UNIT 3 B 3.5-2 Revision 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT ECCS Operating 3.5. 1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - Operating LCO 3.5. 1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of six safety/relief valves shall be OPERABLE. APPLICABILITY: MODE i, MODES 2 and 3, except high pressure coolant injection (H PCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig. ACTIONS NOTE-- LCO 3.0.4.b is not applicable to HPCI. CONDITION REQUIRED ACTION COMPLETION TI ME A. One low pressure ECCS Al Restore low pressure 7 days injection/spray subsystem ECCS injection/spray inoperable, subsystem(s) to OPERABLE status. OR One low pressure coolant injection (LPCI) pump in both LPCI subsystems inoperable. (continued) BEN-UNIT 3 3.5-1 Amendment No. 212. 220, 244 December I 2003

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT ECCS OperazirQ ACTIONS (contmuecl) CONDITION REQUIRED ACTION COMPLETION TI ME B. Required Action and B.i Be in MODE 3. 12 hours associated Completion Time of Condition A not AN met. - B.2 Be m MODE 4. :3b hours (continued) BEN-UNIT 3 Amendment No. 244 December 1. 2003

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level 203000 RHRJLPCI: Injection Mode Tier # G2.2.4 (IOCFR 55.43.5 SRO Only) (multi-unit license) Ability to explain the variations in control Group # board/control room layouts, systems, instrumentation, and K/A # 2.2.4 procedural actions between units at a facility. Importance Rating 3.6 Proposed Question: # 86 Unit I has experienced a Loss of Offsite Power concurrent with a LOCA. Multiple equipment failures have resulted in need for RHR Crosstie to be lined up for injection into the reactor. Which ONE of the following completes the statements below? Unit I RHR can be crosstied to Unit 2 RHR (1)_ in accordance with 1-EOl Appendix 7C, Alternate RPV Injection System Lineup RHR Crosstie. The Unit 2 RHR Pump Suction Valve interlocks must be defeated in accordance with _(2)_. A. (1)Loopl (2) 2-01-74, Residual Heat Removal System B. (1)Loopl (2) I-EOI Appendix 7G. Alternate RPV Injection System Lineup RHR Crosstie C. (1)LoopII (2) 2-01-74, Residual Heat Removal System D. (1) Loop II (2) 1-EOl Appendix 7C, Alternate RPV Injection System Lineup RHR Crosstie Proposed Answer: B Explanation A INCORRECT: Part 1 correct See Explanation B. Part 2 incorrect See (Optional): Explanation C. B CORRECT: Part 1 correct In accordance with 1-E0I Appendix 7C, Unit 1 RHR can be crosstied to Loop I Unit 2 RHR ONLY. Part 2 correct RHR Pump Suction interlocks must be defeated to complete the crosstie and the instructions to defeat the interlocks is contained in 1-EOI Appendix 7C C INCORRECT: Part 1 incorrect Plausible in that this would be the correct answer if Unit 3 RHR is crosstied to Unit 2 RHR. Part 2 incorrect Plausible in that defeating interlocks is sometimes directed in the associated Operating Instruction rather than the Appendix being performed. For example, when injecting CS per Appendix 6E with a loss of associated ECCS ATU Panel, defeating the reactor low pressure interlock would be performed in accordance with 01-75. Also, 2-01-74 contains instructions for defeating various interlocks such as: Defeating the Rx Low Pressure Interlock on the RHR Loop 1/2 Injection and Inhibiting RHR Pump Auto Start and Auto Injection Logic D INCORRECT: Part 1 incorrect See Explanation C. Part 2 correct See Explanation B.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification: The KA is met because the question tests ability to explain variations in systems and procedures between units associated with RHR / LPCI Crosstie capabilities AND differences between Unit I and Unit 2 EOI Appendix 7G. SRO Only Justification: This question is SRO Only because it meets the requirements of Clarification Guidance for SRO, Section ll.E Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)} See Attached. Unit Supervisor is required to analyze plant conditions and select the correct procedures to complete the required hardware modifications and to support the crosstie of Unit 1 and Unit 2 RHR. Question Cognitive Level: This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. Technical Reference(s): 1-EOl Appendix 7C Rev. 1 (Attach if not previously provided) OPL1 71 .044 Rev. 17 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL1 71.044 V.B.3 (As available) Question Source: (Note changes or attach parent) Question History: (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.) Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet COMMENTS FROM SAMPLE SUBMITTED A. (1)Loopl (2) 2-01-74, Residual Heat Removal System NRC Are any interlocks mentioned or defeated in 2-01-74? If not then 2-01-74 may not be plausible. BFN Response Yes, for example, 2-01-74 contains instructions for Defeating the Rx Low Pressure Interlock on the RHR Loop 1/2 Injection (Sections &38 / &39) and Inhibiting RHR Pump Auto Start and Auto Injection Logic(Section &40). See Attached. Also, this 01 was selected in validation.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Clarification Guidance for SRO-only Questions Revl (0311112010) Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5) (Assessment and selection of procedures) Can the question be answered so/ely by knowing systems knowIedge, Le., how the system works, flowpath, logic, component location? Lquestion

lNo Can the question be answered solely by knowing immediate operator actions?

                                                           ]

Yes I RO question

p. INOI Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?

h. I

p. I No Can the question be answered solely by knowing the purpose, overall sequence of events, or question overall mitigative strategy of a procedure?

p. [No Does the question require one or more of the following? Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps RO-only
• Knowledge of diagnostic steps and decision points in the uestion EOPs that involve transitions to event specific subS procedures or emergency contingency procedures
• Knowledge of administrative procedures that specify hierarchy, implementation, andlor coordination of plant normal, abnormal, and emergency procedures No I Question might not be linked to I 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN ALTERNATE RPV INJECTION SYSTEM 1EOI APPENDIX-7C UNIT I LINEUP RHR CROSSTIE j Rev. I Page 1 of 5 LOCATION: Unit I Control Room ATTACHMENTS: 1. Tools and Equipment

2. Unit 2, Panel 2-9-33 Relay Arrangement (____

NOTIFY Unit 2 Operator that Unit 2 RHR will be crosstied to Unit I as directed by the EQIs.

2. DISPATCH personnel to Unit 2 Auxiliary Instrument Room to perform the following:

a. REFER to Attachment I and OBTAIN tools and equipment from EO1 Equipment Storage Box.

b. REFER to Attachment 2 and BOOT the following relay contacts on Unit 2, Panel 2-9-32, Front:

• 2-RLY-074-1OA-K1 9A contact 1-2
• 2-RLY.-074-IOA-K22A, contact 1-2.

c. NOTIFY Unit I and Unit 2 Operators that RHR Pump 2A and 2C Suction Valve interlocks have been defeated.

3. DISPATCH personnel to CLOSE the following breakers:

• 1-8KR-074-0098, RHR PUMP 16 SUCT XTIE VLV, (480V RMOV Board 18, Compartment IC)
• 1-BKR-074-O1Ol, UNITS 1-2 DISCHARGE CROSSTIE (480V RMOV Board 18, Compartment 19C)
• 2-BKR-074-O100, RHR HEAT EXCHANGER CROSSTIE VALVE FCV-74-100 (MO1O-171)(480V RMOV Board 18, Compartment 19A)
• 2-BKR-074-0096, RHR PUMP 2A SUCTION CROSSTIE VALVE FCV-74-96 (480V RMOV Board 18, Compartment 19E)
• 1-BKR-074-0099, RHR PUMP 10 SUCTION CROSSTIE VALVE FCV-74-99, (480V RMOV Board 26, Compartment 17E)
• 2-BKR-074-0097, RHR PUMP C SUCTION CROSSTIE VALVE, 2-FCV-74-97 (480V RMOV Board 28, Compartment 19C).

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN ALTERNATE RPV INJECTION SYSTEM 1-EOI APPENDIX-7C UNIT I LINEUP RHR CROSSTIE Rev. I Pane 3 of 5

5. (Cont) k.. SLOWLY THROTTLE 2-FCV-23-34(40). RHR HX 2A(2C)

RHRSW OUTLET VLV, to obtain between 1350 and 4500 gpm flow through the desired RHR exchanger. I. NOTIFY Unit 1 Operator when complete.

6. PLACE i-HS-74-149, RHR SYSTEM II MIN FLOW INHIBIT, switch in INHIBIT on Unit 1. Panel l-9-3 T VERIFY CLOSED 1-FCV-74-30, RHR SYS II MIN FLOW VLV.

8. VERIFY CLOSED 1-FCV-74-67, RHR SYS II LPCI INBD INJECT VALVE.

9. CLOSE 1-FCV-74-66, RHR SYS II LPCI OUTBD INJECT VALVE.

10. DISPATCH personnel to RACK OUT the following Unit 1 RHR Pump breakers:

• 1-BKR-074-0028, RHR PUMP lB (4KV Shutdown Board 1C, Compartment 17)
• 1-BKR-074-0039, RHR PUMP ID (4KV Shutdown Board ID, Compartment 16).

11. OPEN the following valves on Unit 1, Panel 1-9-3:

• 1-FCV-74-98, RHR PUMP lB SUCT U-2 XTIE
• i-FCV-74-99. RHR PUMP ID SUCT U-2 XTIE
• 1-FCV-74-i01, UNITS 1-2 DISCHARGE CROSSTIE.

12. CHECK 1-Pl-74-65, RHR SYS II DISCH PRESS, indicates above 45 psig.

13. NOTIFY Unit 2 Operator to START the RHR Pump (2A or 2C) for the RHR heat exchanger aligned in Step 5.k.

14. OPEN 1-FCV-74-67, RHR SYS Ii LPCI INBD INJECT VALVE.

15. THROTTLE OPEN 1-FCV-74-66, RHR SYS II LPCI OUTBD INJECT VALVE, to control injection below 5000 gpm.

16. NOTIFY Chemistry that RHRSW has been aligned to in-service RHR Heat Exchangers.

END OF TEXT

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI 71.044 Revision 17 Page 22 of 146 INSTRUCTOR NOTES (2) Discharge is through a motor-operated flow only use the B control valve to the FPC pump return line to RHR drain pump. the Spent Fuel Pool. (3) Discharge MOVs can only be throttled locally ADHR performs the supplemental (4) Flow rate is monitored using flow instrument cooling function F 1-74-76 (0-1 500 gpm) on Panel 9-3 now such that this mode is unlikely to be needed. (5) CAUTION: Initiation of flow in this mode must SER 03-05 be performed very slowly to allow flow Converting SFSP velocity in the pool drains to increase and level head to handle the added flow otherwise the SFSP drain line velocity ventilation will be flooded, head takes time (or more level head).

f. RHR Crosstie TP-1 and 2 Unit 1 Loop 2 to/from Unit 2 Loop 1 g Unit 2 Loop 2 to/from Unit 3 Loop I Obj. VE:2 (1) i.e. Unit I can crosstie loop 2; Flow through the Unit 3 can crosstie loop 1; crosstie is limited to Unit 2 can crosstie either loop. The unit 5000 gpm.

needing to crosstie can have the other unit operate its pumps and heat exchangers to EFFECTIVE perform the desired function. COMMUNICAT1ON (2) Can be used for LPCI, Containment Cooling Requires booting (Drywell Spray, Suppression Pool Spray, contacts or lifting Suppression Pool Cooling) or Shutdown leads to run Cooling pumps without their interlocked suction path. (3) LPCI and Containment Cooling Suction from the affected unit Suppression Pool, through the crosstie line, to another units pumps and heat exchanger and back to affected unit (4) Shutdown Cooling Suction from the affected unit SDC cooling supply, through the crosstie to another loops pumps and heat exchanger and back to affected unit.

g. Miscellaneous

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT OPLI 71.044 Revision 17 Page 22 of 146 INSTRUCTOR NOTES (2) Discharge is through a motor-operated flow only use the B control valve to the FPC pump return line to RHR drain pump. the Spent Fuel Pool. (3) Discharge MOVs can only be throttled locally ADHR performs the supplemental (4) Flow rate is monitored using flaw instrument cooling function F 1-74-76 (0-1 500 gpm) on Panel 9-3 now such that this mode is unlikely to be needed. (5) CAUTION: Initiation of flow in this mode must SER 03-05 be performed very slowly to allow flow Converting SFSP velocity in the pool drains to increase and level head to handle The added flow otherwise the SFSP drain line velocity ventilation will be flooded, head takes time (or more level head).

f. RHR Crosstie TP-1 and 2 Unit 1 Loop 2 to/from Unit 2 Loop 1 1 Unit 2 Loop 2 to/from Unit 3 Loop I Obj. V.E.2 r

(1) i.e. Unit 1 can crosstie loop 2; Flow through the Unit 3 can crosstie loop 1; crosstie is limited to Unit 2 can crosstie either loop. The unit 5000 gpm. needing to crossüe can have the other unit operate its pumps and heat exchangers to EFFECTIVE perform the desired function. COMMUNICATION (2) Can be used for LPCI, Containment Cooling Requires booting (Dryweti Spray, Suppression Pool Spray, contacts or lifting Suppression Pool Cooling) or Shutdown leads to run Cooling pumps without their interlocked suction path. (3) LPCI and Containment Cooling Suction from the affected unit Suppression Pool, through the crosstie tine, to another units pumps and heat exchanger and back to affected unit (4) Shutdown Cooling Suction from the affected unit SDC cooling supply, through the crosstie to another loops pumps and heat exchanger and back to affected unit.

g. Miscellaneous

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BPN INJECTION SUBSYSTEMS LINEUP 1-EOl APPENDIX-6E UNIT I CORE SPRAY SYSTEM Page 1 LOCATION: Unit 1 Control Room ATTACHMENTS: 1. NPSH Monitoring ( I) 1, VERIFY OPEN the following valves:

• l-FCV75-30, CORE SPRAY PUMP lB SUPPR POOL SUCT VLV
• 1-FCV-75-39, CORE SPRAY PUMP ID SUPPR POOL SUCT VLV
• 1-FCV-75-51, CORE SPRAY SYS II OUTBO INJECT VALVE.

2. VERIFY CLOSED 1-FCV-75-50, CORE SPRAY SYS II TEST VALVE.

3, VERIFY CS Pump 1 B and/or ID running.

4. WHEN RPV pressure Is below 450 psig, THEN THROTTLE 1-FCV-75-53, CORE SPRAY SYS II INBD INJECT VALVE, as necessary to control injection at or below 4000 gpm per pump.

CAUTION Continuous operation with inadequate NPSH may result In pump damage or pump inoperability.

5. MONITOR Core Spray Pump NPSH using Attachment 1.

END OF TEXT

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT A BFN Core Spray System 1-01-75 Unit I Rev, 0020 Page 97 of 108 8.14 DefeatIng the Rx Low Pressure Interlock on the Core Spray Injection Valves. NOTE This section will defeat the 450 psig interlock art the Care Spray injection valves 1-FCV-75-25(53). This allows the valves to open regardless of reactor pressure. This section is intended to be used when ECCS inverter(s) are deenergized or any failure which prevents energizing the 450 psig relays, 14A-KSA(SB) and 14A-K23A(23B). [1] VERIFY Reactor pressure less than 450 psig. 1st [2] IF It is desired to defeat the 450 psig interlock for 1-FCV-75-25, THEN INSTALL a Jumper between the banana jacks on terminal EE-27 and terminal EE-28 located in 1-PNLA-009-0032, bay 2. 1St 2nd [3] IF it is desired to defeat the 450 psig interlock for 1-FCV-75-53, THEN INSTALL a jumper between the banana jacks on terminal EE-27 and terminal EE-28 located in 1-PNLA-009-0033, bay 2. 1St 2nd [4] PLACE the signed copy of Section 8.14 in the daily configuration log.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BPN Residual Heat Removal System 2.01-74 Unit2 Rev.0152 Page 6 of 442 Table of Contents (continued) 8.35 Initiation of Standby Coolant using RHR Loop 1(11) 326 8.36 Reactor Vessel Draindown Assist/Level Control to the Main Condenser or Radwaste using RHR Drain System 329 8.37 Reactor Vessel Draindown/Level Control Assist to the Suppression Pool using RHR Drain System 334 8.38 Defeating the Rx Low Pressure Interlock on the RHR Loop 1 Injection Valves 338 8.39 Defeating the Rx Low Pressure Interlock on the RHR Loop 2 Injection Valves 340 8,40 Inhibiting RHR Pump Auto Start and Auto Injection Logic 342 8.41 Restoring RHR Pump Auto Start and Auto Injection Logic 344 8.42 Removing RHR Seal Heat Exchanger 345 8.43 Returning RHR Seal Heat Exchanger To Service 347 8.44 Warming an Idle Recirc Loop Using Shutdown Cooling 349 8.45 Alternate Method for Removing Water from the Suppression Pool by Flushing the RHR Pump A Suction Piping with Suppression Pool Water using RHR Drain Pump A 352 8.46 Alternate Method for Removing Water from the Suppression Pool by Flushing the RHR Pump C Suction Piping with Suppression Pool Water using RHR Drain Pump A 357 8.47 Alternate Method for Removing Water from the Suppression Pool by Flushing the RHR Pump B Suction Piping with Suppression Pool Water using RHR Drain Pump B 362 8.48 Alternate Method for Removing Water from the Suppression Pool by Flushing the RHR Pump D Suction Piping with Suppression Pool Water using RHR Drain Pump B 367 8.49 Draining Inboard and Outboard Drywell Spray Valve Piping 372 8.50 RHR LOOP 1(11> HEAT EXCHANGER/DISCHARGE PIPING DRAIN USING RHR DRAIN SYSTEM 374 8.50.1 Draining RHR LOOP I Heat Exchangers A(C) and associated discharge piping 374 8.50.2 Draining RHR LOOP II Heat Exchangers B(D) and associated discharge piping 380

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 262001 AC Electrical Distribution Tier # 2 24.41 (IOCFR 5543.5 SRO ONLY) Knowledge of the emergency action level thresholds and Group # classifications. K/A # 1 G2.4.41 Importance Rating 46 Proposed Question: # 87 The following conditions exist on Unit 3:

• Reactor Power is 100%
• Emergency Diesel Generator 3EA is tagged out of service The following sequence of events occur:
• 1130 ALL Offsite power is lost
• 1130 EDGs 3EB AND 3EC start AND tie to their associated Board
• 1130 EDG 3ED trips on differential overcurrent
• 1135 EDG 3EC trips on low lube oil pressure
• 1135 EDG 3EB trips for unknown reason
• 1155 EDG 3EB is restarted AND tied to its associated Board Which ONE of the following identifies the HIGHEST emergency classification required AND the MAXIMUM amount of time allowed to make the initial notification to the State of Alabama once a formal declaration of the event is made?

[REFERENCE PROVIDED] A. Alert; 15 minutes B. Alert; 30 minutes C. Site Area Emergency; 15 minutes D. Site Area Emergency; 30 minutes Proposed Answer: C Explanation A INCORRECT: Part 1 incorrect See Explanation B. Part 2 correct See (Optional): Explanation C.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet B INCORRECT: Part 1 incorrect Plausible in that this would be the correct answer for Modes 4 or 5. Part 2 incorrect Assessment of an EVENT commences when recognition is made that one or more of the conditions associated with the event exists. Implicit in this definition is the need for timely assessment, i.e. within 15 minutes. This combined with requirement to contact the State of Alabama within 15 minutes could be added to incorrectly conclude 30 minutes is allowed make notifications. C CORRECT: Part 1 correct In accordance with EPIP-1, EAL 5.1-S, Loss of voltage to ALL unit specific 4KV shutdown boards from Table 5.1 for greater than 15 minutes in Modes 1 ,2,or 3 requires declaration of a Site Area Emergency. Part 2 correct The State of Alabama shall be contacted within 15 minutes of the emergency classification. D INCORRECT: Part 1 correct See Explanation C. Part 2 incorrect See Explanation B. KA Justification: The KA is met because the question tests Emergency Action Level threshold and classification associated with AC Electrical Distribution with the loss of offsite power and subsequent Emergency Diesel Generator failures. SRO Only Justification: This question meets the requirements of Clarification Guidance for SRO-only Questions, Section Ill. (See Attached). Classification of Emergencies is a knowledge I ability unique to the SRO position. Candidate must evaluate AC Electrical Distribution status and determine emergency classifications. This results in declaration of a Site Area Emergency. Question Cognitive Level: To solve the question the examinee must use a multi part mental process to assemble, sort, and integrate the parts of the plant conditions. Technical Reference(s): EPIP-1, Rev 46 (Attach if not previously provided) OPL1 71.075 Rev 25 Proposed references to be provided to applicants during examination: EPIP-1, Rev 46 Section 5 Learning Objective: V.B.2 (As available) Question Source: Bank # Brunswick 08 #82 (Note changes or attach parent) Question History: Last NRC Exam Brunswick 2008 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eveiy question.) Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Comments: The question has been modified from the original Brunswick 2008 #82 to be valid for Browns Ferry. However, it does not meet the requirements of a Significantly Modified Question so it is identified as a Bank Question. Original is attached. Clarification Guidance for SRO-only Questions RevI (0311112010) Ill. Justification for Plant Specific Exemptions The 25 SRO-only questions shall evaluate the additional knowledge and abilities required for the higher license level in accordance with 10 CFR 55.43(b). INUREG 1021, Section ES.401D2.d] The fact that a facility licensee trains its ROs to master certain 10 CFR 55.43 knowledge, skills, and abilities does NOT mean that they can no longer be used as a basis for SRO-only questions. Operator Licensing Feedback Web page Item 401.36 http:llwwwnrc.gov/reactorsIoperator-TicensinqIopIicensinq-flles/oi-feedback.pdf] The SRO-only test item is required to be tied to one of the 10 CFR 55.43(b) items. However, if a licensee desires to evaluate a knowledge/abilIty that is not tied to one of the 10 CFR 55.43(b) items, then the licensee can classify the knowledge/ability as unique to the SRO positior, provided that there is documented evidence that ties the knowledge/ability to the licensees SRO job position duties in accordance with the systematic approach to training (SAT). Justification: A question that is tied to one of the 10 CFR 55.43(b) Items can still be classified as SRO-only provided the licensee has documented evidence to prove that the knowledge/ability is unique to the SRO position at the site. An example of documented evidence includes: o The question is linked to a learning objective that is specifically labeled in the lesson plan as being SRO-only (e.g., some licensee lesson plans have columns in the margin that differentiate AO, RO, and SRO learning objectives) [NUREG 1021, ES-401, Section D2.d] AND/OR o A questIon Is linked to a task that Is labeled as an SRO-only task, and the task is NOT listed in the RO task list. Page 10 of 16

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX NOTES 5iU Loss of normal and alternate supply voltage implies inability to restore voltage from any qualified source to normal or alternate feeder for at least one of the unit specific boards within 15 minutes, At least two boards must be energized from Diesel power to meet this classificatiort If only one board can be energized and that board has only one source of power then refer to EAL 5,1-Al or 5,1-A2. 5,141 Only one source of power (Diesel or Offsite) is available to any one of the listed unit specific 4KV Shutdown Boards. No power is available to the three remaining boards. 5.142 Loss of voltage to all unit specific 4KV Shutdown Boards applies to those boards which normally supply emergency AC power to the affected unit only. Determination of the event classification depends on the affected unit operating mode. For units in operation 5,1-S would apply. 51-S Lass of voltage to all unit specific 4KV Shutdown Boards applies to those boards which normally supply emergency AC power to the affected unit only. Determination of the event classification depends on the affected unit operating mode. For units in Shutdown or Refuel 5142 would apply. 5.1-6 Loss of voltage to all unit specific 4KV Shutdown Boards applies to those boards which normally supply emergency AC power to the affected unit only. CURVES/TABLES: Table ii UNIT 4KV SHUTDOWN BOARD APPLICABILITY APPLICABLE UNIT APPLICABLE 4KV SHUTDOWN BOARDS UNIT 1 A,BC,andD UNIT2 A B, C. andO UNIT 3 3A, 3B, 3C, and 3D

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BROWNS FERRY I EMERGENCY CLASSIFICATION PROCEDURE I I I EVENT CLASSIFICATION MATRIX E P1 P-I Loss ot normal and ALL unit specific 4KV shutdown boards from Table 5.1 z for greater than 15 minutes AND At least two Diesel Generators supplying power to unit specific 4KV shutdown boards listing in Table 5.1. m OPERATING CONDITION: m ALL 5.1-Al I I NOTE I TABLE I US 5.1-A2 I I NOTE I TABLE I Us - Loss of voltage to ANY THREE unit specific 4KV Loss of voltage to ALL unit specific 4KV shutdown shutdown boards from Table & I for greater than boards from Table 5.1 for greater than 15 minutes. 15 minutes r Only ONE source of power available to the m remaining board, OPERATING CONDITION: OPERATING CONDITION: Mode 1 or 2 or 3 Mode 4 or 5 or Defueled 5.1-S I I NOTE ITABLEI US I I I I Loss of voltage to ALL unit specific 4KV shutdown boards from Table 5.1 for greater than 15 minutes. In In I OPERATING CONDITION: -< Mode I or 2 or 3 5.1-G I I NOTE I TABLE I US I I I Loss of voltage to ALL unit specific 4KV shutdown boards from Table 5.1 AND m Either of the following conditions exists;

          . Restoration of at least one 4KV shutdown board is NOT likely within three hours.                                                                     rn
          . Adequate core cooling can NOT be assured.

G) rn z OPERATING CONDITION: Mode 1 or 2 or 3 PAGE 47 OF 206 REVISION 46

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE j EVENT CLASSIFICATION MATRIX EPIP-1 LOSS OF AC POWER v ueseripnon 5.1-U I NOTE TABLE I US I I I Loss of normal and alternate supply voltage to ALL unit specific 4KV shutdown boards from Table 51 z for greater than 15 minutes AND At least two Diesel Generators supplying power to unit specific 4KV shutdown boards listing in Table 5.1. OPERATING CONDITION: m ALL 5.1-Al j NOTE I TABLE I US 5.1-A2 I NOTE TABLE I US Loss of voltage to ANY THREE unit specific 4KV Loss of voltage to ALL unit specific 4KV shutdown shutdown boards from Table 5.1 for greater than boards from Table 5.1 for greater than 15 minutes. 15 minutes A 1 Only ONE source of power avadable to the m remaining board. OPERATING CONDITION: OPERATING CONDITION: Mode I or 2 or 3 Mode 4 or S or Defueled 5.1-S I INOTEITABLEI US I I I I Loss of voltage to ALL unit specifIc 4KV shutdown boards from Table 5.1 for greater than 15 minutes. m m M C) m z C) OPERATING CONDITION: Mode I or 2 or 3 5.l-G I j NOTE I TABLE I US I I I Loss of voltage to ALL unit specific 4KV shutdown 0 boards from Table 5.1 AND m Either of the following conditions exists

           . Restoration of at least one 4KV shutdown board is NOT likely within three hours.                                                                      m
           . Adequate core cooling can NOT be assured.

nl OPERATING CONDITION: Mode I or 2013 PAGE 47 OF 206 REVISION 46

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet

82. The following sequence of events have occurred:

1130 Offsite power is lost to both units. 1130 All EDGs start and tie to their associated E-Buses. 1135 DG3 trips on low lube oil pressure 1145 DG4 trips on differential overcurrent 1200 El is crosstied to E3 1205 Current Time (reference provided) Which one of the following identifies the highest emergency classification reached during the event AND the maximum amount of time allowed to make initial notification to State and local governments once formal declaration of the event is made? A. Alert; 15 minutes B, Alert; 30 minutes Cv Site Area Emergency; 15 minutes D. Site Area Emergency; 30 minutes

REFERENCE:

EALs to be prtwided to the exarninee only. PEP-2.1 Initial Ernergenci Actions, 6,0 Electrical and Power Failures EXPLANATION: The inability to power wither 4KV bus from off-site power AND loss of all on-site AC power capability indicated by failure of diesel generators to start or synchronize AND lasting more than 15 minutes Site Area Emergency. After declaration of event 15 minutes is the requirement to notify State and local govt. CHOICE GA .. Incorrect CHOICE B Incorrect CHOICE C Correct Answer CHOICE D - Incorrect Page 114 of 147

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level 261000 Standby Gas Treatment System Tier # A2.12 (IOCFR 55.43.5 SRO Only) Ability to (a) predict the impacts of the following on the STANDBY Group # GAS TREATMENT SYSTEM and (b) based on those predictions, K/A # 261000A2.12 use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High fuel pool ventilation radiation: Plant-Specific 3.4 Importance Rating Proposed Question: # 88 Unit 3 is at 100% Reactor Power. Standby Gas Treatment System (SGTS) A was tagged out of service on 1/16/Il at 0600. SGTS B has been manually started. At 1000 on 1/16/11, a container is removed from the Unit 3 Spent Fuel Pool (SFP) resulting in the following Refuel Zone Radiation Monitor indications:

• 3-RM-90-140 Detector A is reading 73 mr/hr
• 3-RM-90-140 Detector B is reading 72 mr/hr
• 3-RM-90-141 Detector A is reading 71 mr/hr
• 3-RM-90-1 41 Detector B is reading 71 mr/hr SGTS C did NOT start. The container was placed back in the SFP AND Refuel Zone Radiation Monitor indications returned to normal.

Which ONE of the following completes the statements below? A Tech Spec required shutdown condition must be entered at_(1)_ in accordance with Tech Spec 3.6.4.3, Standby Gas Treatment System. The initiation of plant a shutdown required by plant Tech Specs _(2)_ a 4 hour report to the NRC in accordance with SPP-3.5, Regulatory Reporting Requirements. [REFERENCE PROVIDED] A. (1)l000on 1/16/11 (2) requires B. (1)0600 on 1/23/11 (2) requires C. (1)l0000nl/16/11 (2) does NOT require D. (1)0600 on 1/23/11 (2) does NOT require Proposed Answer: A

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Explanation A CORRECT: Part 1 correct With Refuel Zone Radiation Monitor Channels (Optional): A and D above the set point for automatic initiation of SGTS and the failure of SGTS C to start, SGTS C must be declared inoperable. With SGTS A and C inoperable, TS 3.6.4.3 Condition D requires immediate entry into TS 3.0.3. Part 2 correct In accordance with SPP-3.5, Regulatory Reporting Requirements, the initiation of any nuclear plant shutdown required by the plants Technical Specifications requires a 4 hour NRC notification. B INCORRECT: Part 1 incorrect See Explanation D. Part 2 correct See Explanation A. C INCORRECT: Part 1 correct See Explanation A. Part 2 incorrect See Explanation D. D INCORRECT: Part 1 incorrect Plausible in that if the right combination of channels for Automatic Start of SGTS did not exceed the set point, this would be the correct answer. SGTS C would still be operable so a shutdown condition would not be entered until SGTS A was tagged out for 7 days in accordance with TS 3.6.4.3 Conditions A and B. Part 2 incorrect Plausible in that candidate may believe that reportability requirement is 1 hour or 8 hours. KA Justification: The KA is met because the question tests the candidates ability to predict the impact of High fuel pool ventilation radiation on SGTS and with one train all ready out of service. Then, utilize Tech Specs and OPDP-8,Limiting Conditions for Operation Tracking, to control the consequences of this abnormal condition. SRO Only Justification: This question is SRO Only because it meets the requirements of Clarification Guidance for SRO, Section II.B Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)J. The question involves application of Required Actions (Section 3) in accordance with rules of application requirements (Section 1). See Attached. Candidate must determine that SGTS C is inoperable because it failed to start when the required number of channels reached the initiation set point. Then, they must determine when a TS shutdown condition is entered and reportability requirements. Determination of reportability requirements is also a function unique to the SRO position. Question Cognitive Level: This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): U3 TS 3.6-51/52 Amm 249 (Attach if not previously provided) U3 TS 3.6-54 Amm 215 U3 TS 3.0-1 Amm 226 OPL171.033 Rev. 13/ SPP-3.5 Rev. 0 Proposed references to be provided to applicants during examination: U3 TS 3.6.4.3 Learning Objective: OPL171.033 V.B.5 (As available) Question Source: (Note changes or attach parent) New X Question History: (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.) Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Clarification Guidance for SRO-only Questions RevI (03/1112010) Figure 1: ScreenIng for SRO.oniy linked to 10 CFR 55.43(b)(2) (Tech Specs) Can question be answered solely by knowing 1 hour TS/TRM Action? Oqueion No Can question be answered solely by knowing the Yes LCO/TRM information listed above-the-line? RO question I No Can question be answered solely by knowing the Yes - TS Safety Limits? RO question No Does the question involve one or more of the following for TS, TRM, or ODCM?

• Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section I)
• Application of generic LCO requirements (LCO 3.01 thru 3.0.7 and SR 4.0.1 thru 4.0.4) Yes I SRO-onlyl
• Knowledge of TS bases that is required to analyze TS required actions and terminology j question No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 1$

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet SGT System 3.6.4.3 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and Ci Place two OPERABLE Immediately associated Completion SGT subsystems in Time of Condition A not operation. met during OPDRVs. OR C.2 Initiate action to suspend Immediately OPDRVs. D. Two or three SGT 0.1 Enter LCO 3,0.3. Immediately subsystems inoperable in MODE 1, 2, or 3. (continued) BFN-UNIT 3 3.6-52 Amendment No. 242 249 September 27, 2004

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet SGT System 3.6.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.3.1 Operate each SGT subsystem for 10 31 days continuous flours with heaters operating. SR 3.6.4.3.2 Perform required SOT filter testing in In accordance accordance with the Ventilation Filter Testing with the VETP Program (VETP). SR 3.6.4.3.3 Verify each SOT subsystem actuates on an 24 months actual or simulated initiation signal. SR 3.6.4.3.4 \erity the SOT decay heat discharge 12 months dampers are in the correct position. BEN-UNIT 3 3.6-54 Amendment No. 215 November30, 1998

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet LCD Applicability 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2 and LCO 3.01. LCO 3.0.2 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6. If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required, unless otherwise stated. LCO 3.0.3 When an LCD is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour to place the unit, as applicable, in:

a. MODE 2 within 10 hours;

b. MODE 3 within 13 hours; and

c. MODE 4 within 37 hours.

Exceptions to this Specification are stated in the individual Specifications. Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LCD 3.0.3 is not required LCO 3.0.3 is only applicable in MODES I, 2, and 3. (continued) BFN-UNIT 3 3.0-1 Amendment No. 21-2 226 November 21, 2000

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL 171.033 Revision 13 Page 28 of 75 INSTRUCTOR NOTES

b. Four gamma sensitive GM instrumentation Obj. V.B.2 channels monitor the radiation from the Obj. V.C.2 reactor zone exhaust and four identical Ob]. V.B.4.a channels monitor the radiation from the Obj. V.0.31 refueling zone Obj, V.0.9 (1) These are physically located on the side of the ventilation ducts on the refuel floor (2) Refuel Zone monitors RM-90-140(A & Obj. V.B.3.f B) and Reactor Zone monitors Obj. V.C.3.f 90-142(A & B) are powered from RPS N

(3) Refuel Zone monitors RM-90-141(A & B) and Reactor Zone monitors 90-143(A & B) are powered from RPS B (4) The instrumentation channels are Lesson OPL 171.034 similar to an area radiation monitoring Area Rad. Mon. system channel. c, Alarms Obj. V.B.5 Obj. V.C.5 (1) REACTOR ZONE EXHAUST RADIATION HIGH (55-3A-21)Alami setpoint is 72 mrlhr (a) Reactor zone and refueling zone monitors work independently of each other for trip actuation (b) High radiation trip setpoint is 72 Rad-monitor auto mrfhr for the refueling and resets when alarm is reactor zones clear

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI71 .033 Revision 13 Page 29 of 75 INSTRUCTOR NOTES (c) Trip logic for the refueling and the reactor zones is identical, and the following combinations will generate a trip: Two high level trips in the same Two-out-of-two, once F channel, (division)

                             -DR One downscale trip in each          One-out-of-two, twice channel (division)
                             -DR One monitor INOP in each            One-out-of-two, twice channel (division)                  Obj. V.B.3.f Obj. V.0.31
                             -OR-Loss of RPS power to either channel (2)    Automatic actions                         Obj. V.B,1 ,3.e Obj. V.C.1,3.e (a)   Refuel Zone Trip                    Obj V.D.6 (i)      Isolate Refuel Zone S             (ii)     Starts Standby Gas Treatment System (iii)    P015 Group 6 isolation      Obj. V.B.1 ,3.f, 3.g Obj. V.C.1,3.t 3.g (iv)     Starts CREVs (b)  Reactor Zone Trip (i)      Isolate Control Room, Reactor Zone, and Refueling Zone ventilation (ii)     Starts Standby Gas Treatment System (iii)    Start OREVs (iv)     P018 Group 6 isolation

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet T NPG Standard Programs and Regulatory Reporting Requirements NPG-SPP-03.5 Rev. 0000 Processes Page 17 of 71 Appendix A (Page 3 of 11) Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 31 immediate Notification NRC (continued) NOTE If it is discovered that a condition existed which met the Emergency Plan criteria but no emergency was declared and the basis for (he emergency class no longer exists at the time of discovery, an ENS notification (and notification of the Operations Duty Specialist), within one hour of discovery of the undeclared (or misclassified) event, shall be made. However, actual declaration of the emergency class is not necessary in these circumstances.

3. §50.72(b).(1)) Any deviation from the plants Technical Specifications authorized pursuant to §50.54(x).

C. The following criteria require 4-hour notification:

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

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

3. §50.72(b)(2)(iv)(8) Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

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

D. The following criteria require 8-hour notification: NOTE The non-emergency events specified below are only reportable if they occurred within three years of the date of discovery.

1. §50.72(b)(3)(ii)(A) Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT SGT System 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 Three SGT subsystems shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3, During operations with a potential for draining the reactor vessel (OPDRVs). ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SGT subsystem A.1 Restore SOT subsystem 7 days inoperable, to OPERABLE status. B. Required Action and B. 1 Be in MODE 3. 12 hours associated Completion Time of Condition A not NP meUn MODE 1, 2, or 3. B.2 Be in MODE 4. 36 hours (continued) BEN-UNIT 3 3.6-51 Amendment No. 242, 249 September 27, 2004

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT NPG Standard Regulatory Reporting Requirements NPG-SPP-03.6 Programs and Rev. 0000 Processes Page 16 of 71 Appendix A (Page 2 of 11) Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 3.0 REQUIREMENTS NOTES

1) Internal management notification requirements for plant events are found in Appendix 0. The Operations Shift Manager is responsible for notifying Site Operations Management and the Duty Plant Manager. The Duty Plant Manager is responsible for making the remaining internal management notifications.

2) NRC NUREG-I 022, Supplements and subsequent revisions should be used as guidance for determining reportability of plant events pursuant to §50.72 and §50.73.

3.1 Immedlate Notification NRC - WA is required by §50.72 to notify NRC immediately if certain types of events occur. This appendix contains the types of events and the allotted time in which NRC must be notified. (Refer to Form NPG-SPP-03.5-1 or NRC Form 361). Operations is responsible for making the reportability determinations for §50.72 and §50.73 reports. Operations is responsible for making the immediate notification to NRC in accordance with §50.72. Notification is via the Emergency Notification System. If the Emergency Notification System is not operative, use a telephone, telegraph, mailgram, or facsimile. NOTE The NRC Event Notification Worksheet may be used in preparing for notifying the NRC. This Worksheet may be obtained directly from the NRC website (www.nrc.gov) under Form 361, 0rTVA NPG Form NPG-SPP-03.5-i may be used. A. The Immediate Notification Cnteria of §50.72 is divided into i-hour, 4-hour, and 8- hour phone calls, Notify the NRC Operations Center within the applicable time limit for any item which is identified in the Immediate Notification Criteria. B. The following cnteria require 1-hour notification:

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

2. §50.72 (a)(l )(i) The declaration of any of the Emergency classes specified in the licensees approved Emergency Plan.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT NP Standard Regulatory Reporting Requirements NPG-SPP-035 Programs and Rev. 0000 Processes Page 17 of 71 Appendix A (Page 3 of 11) Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 31 ImmedIate Notification NRC (continued> NOTE If it is discovered that a condition existed which met the Emergency Plan criteria but no emergency was declared and the basis for the emergency class no longer exists at the time of discovery, an ENS notification (and notification of the Operations Duty Specialist), within one hour of discovery of the undeclared (or misclassified) event, shall be made. However, actual declaration of the emergency class is not necessary in these circumstances.

3. §50.72(b).(1)) Any deviation from the plants Technical Specifications authorized pursuant to §5(154(X).

C. The following criteria require 4-hour notification:

1. §5tL72(b)(2)(i) The initiation of any nuclear plant shutdown required by the plants Technical Specifications.

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

3. §50.72(b)(2)(iv)(8) Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

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

D. The following criteria require 8-hour notification: NOTE The non-emergency events specified below are only reportable if they occurred within three years of the date of discovery.

1. §50.72(b)(3)(ii)(A) Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level 264000 Emergency Generators (Diesel/Jet) Tier # A2.09 (IOCFR 55.41.5 SRO ONLY) Ability to (a) predict the impacts of the following on the Group # EMERGENCY GENERATORS (DIESEL/JET); and (b) based on KA # those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of A.C. power Importance Rating Proposed Question: # 89 You are the UNIT I Unit Supervisor. Unit I is operating at 100% power, in a normal electrical lineup, when the following alarms and indications occur:

• 4kV UNIT BDs 1C, 2C, 3C, 3A, 3B AUTO TRANSFER
• 4kV RECIRC PUMP DRIVEs AUTO TRANSFER
• 4kV COMMON BDs A, B AUTO TRANSFER
• The reactor scrams.
• ALL MSIVs close.
• Generator MWe lowers to zero.

Based on the above conditions, which ONE of the following responses completes the statement? The Unit 1/2 Diesel Generators will (1) AND you should direct entry into 0-AOl (2) A. (1) start ONLY (2) IC, Loss of 161 KV B. (1) startAND load (2) 1C, Loss of 161KV C. (1) start ONLY (2) 1 B, Loss of 500KV D. (1) startAND load (2) 1 B, Loss of 500KV Proposed Answer: D Explanation A INCORRECT: (1) incorrect (2) Incorrect. Plausible because candidate may (Optional): have misconception that 161kV is the normal supply to the Unit Bds, Recirc Bds, and Common Bds, 500 kV is. Also that fast transfer would occur to 500kV that would supply Unit Bds, Recirc Bds, and Common Bds. DGs start in -1 .5 sec after loss of voltage and Unit Bds re-energize (fast Transfer), alternate feeder breaker fast transfer for SD Bds is delayed, would not occur after DG bkr closes.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet B INCORRECT: (1) Correct (2) Incorrect. DG will start and load. DG will start and DG Output breaker closes. (2) Plausible because candidate may believe 161kV is the normal supply to the Unit Bds, Recirc Bds, and Common Bds. C INCORRECT: (1) incorrect. DGs start in 1.5 sec after loss of voltage and misconception that alternate feeder breaker fast transfer occurs in > 3 sec and 4kV SD Bd voltage would be restored from 161 KV before DG output breaker closed. (2) Loss of 500 KV gives these indications. D CORRECT: (1) Correct (2) Correct. 500kV is the normal supply to the Unit Bds, Recirc Bds, and Common Bds. When under voltage is sensed on the 4kV SD Bd the DG starts in 1.5 sec. The Unit Bds fast transfer but the DG ties to the SD Bds before power can be supplied by the Unit Bds to the SD Bds. KA Justification: The K/A is matched because; the candidate must assess the plant conditions, determine which AC power is lost, predict the DG response, and select the correct procedure to direct to mitigate, recover and proceed. SRO Only Justification: This question is SRO Only because it meets the requirements of Clarification Guidance for SRO, Section ll.E Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)J See Attached. Candidate must assess plant conditions to determine that 0-AOl-57-1 B, Loss of 500 kV, must be selected. Question Cognitive Level: To resolve the issues presented in the question the examinee must utilize a multi-part mental process to assemble, sort, and integrate the information/facts. Technical Reference(s): 0-AOI-57-1B Rev 14 (Attach if not previously provided) OPL171.036 Rev 12 Proposed references to be provided to applicants during examination: NONE Learning Objective: V.B.6 (As available) Question Source: Bank # (Note changes or attach parent) New X Question History: Last NRC Exam (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.) Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 x Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Clarification Guidance for SRO.only Questions Rev 1(0311112010) FIgure 2: Screening for SROony linked to 10 CFR 55.43(b)(5) (Assessment and selection of procedures) Can the question be answered solely by knowing systems knowledge, Le, how the system works, flowpath, logic, component location? Can the question be answered solely by knowing immediate operator actions? Yes I RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? No Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? ZLfijueon No Does the question require one or more of the following?

• Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps
• Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
• Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 5543(b)(5) far SROonly Page 8 of 16

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Loss of 500KV O-AOl-57-1 B UnitO Rev. 0014 Page 4 of 10 1.0 PURPOSE This abnormal operating instruction provides symptoms, automatic actions, and operator actions for a loss of 500kV distribution. 2.0 SYMPTOMS A. The following annunciators on Panel 9-8 are in alarm: N OTE 4kV Unit Boards 1A(2A), I B(2B) will NOT auto transfer solely on loss of voltage.

1. 4KV UNIT BD 1C (2C, 3C)AUTO XFR, XA-57-10 (XA-55-88, window 7)

2. 4KV UNIT BD 3AAUTO XFR, XA-57-4 (XA-55-8B, window 10)

3. 4KV UNIT BD 38 AUTO XFR, XA-57-7 (XA-55-88, window 12)

4. 4KV RECIRC PMP DRIVE AUTO XFR, XA-57-13 (3-XA-55-8B, window 15)

5. 4KV RECIRC PMP DRIVE AUTO XFR, XA-57-13 (2-XA-55-88, window 15)

6. 4KV COMMON BD A AUTO XFR, XA-57-91 (XA-55-8C, window 1)

7. 4KV COMMON BD B AUTO XFR, XA-57-92 (XA-55-8C, window 2)

B. If the 4kV Shutdown Bds supplying the 480V Shutdown Bds are being powered from a 500kV source, the following will occur:

1. Reactor scram

2. MSIV isolation C. All diesel generators (DIGs) may start with associated annunciation. (D/Gs start 1.5 +/- 0.1 seconds after loss of voltage and alternate feeder transfers take 3 or more seconds.)

D. If reactor power is greater than 30% (first stage turbine pressure greater than 147 psig) a reactor scram will occur. E. If a 40% mismatch (cross under pressure vs stator output amps) occurs a Turbine-Generator load reject (control valve fast closure) will occur. F. Generator MWe lowers to zero.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL1 71.036 Revision 12 Page 26 of 60 (a) It receives a trip signal anytime the associated Unit 3 diesel output breaker is not shut. (b) It can not receive a shut command unless the load side of the breaker is de-energized (no voltage present from Unit 1/2). (2) The 4kV Bus Tie Board cable to CT Swgr is removed lifted and the board can be used as an emergency feed to U1/2 SD boards from U3.

2. The following conditions are required for the Obj. V.CA.

DIG breaker to auto close Obj. V.B.16

a. DG up to speed (above 870 rpm) Situational Awareness

b. All Shutdown Board feeders open

c. No lockout on Shutdown Board

d. No lockout on diesel generator

e. No lockout on normal or alternate feeder breakers

f. Under voltage on Board

3. The following conditions will trip the DIG Questioning breaker. Attitude

a. ESTR Engine stop relay (1) Stop signal (2) 86 Generator

b. 86SX Lockout on S/D Board

c. Overspeed

d. 86GX Lockout on DIG

e. Manual trip

f. RIA relay (stop signal) Obj. V.B.12

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 300000 Instrument Air System (lAS) Tier # 2 G2.2.36 (IOCFR 55.43.2 SRO Only) Ability to analyze the effect of maintenance activities, such as Group # 1 degraded power sources, on the status of limiting conditions K/A # 300000G2.2.36 for operations. Importance Rating Proposed Question: # 90 Unit 3 is at 100% Reactor Power. Plant Control Air has been aligned to Drywell Control Air to allow maintenance on the Nitrogen Storage Tanks. Which ONE of the following completes the statement? Technical Requirements Manual Section 3.6.3, Drywell Control Air System, requires Reactor Thermal Power be reduced to less than or equal to _(1) power within _(2)_ if Plant Control Air is being used to supply the pneumatic control system inside primary containment. A. (1)15% (2) 12 hours B. (1)15% (2) 24 hours C. (1)25% (2)12 hours D. (1)25% (2) 24 hours Proposed Answer: B Explanation A INCORRECT: Part 1 correct See Explanation B. Part 2 incorrect See (Optional): Explanation C. B CORRECT: Part 1 and 2 correct Technical Requirements Manual Section 3.6.3 requires reactor thermal power be reduced to less than or equal to 15% power within 24 hours if plant control air is being used to supply the pneumatic control system inside primary containment. C INCORRECT: Part 1 and 2 incorrect Plausible in that 25% Reactor Power and 12 hours are common power level I time requirements associated with Tech Spec Applicability and Surveillance Requirements. Example: SR 3.3.1.1.2 Not required to be performed until 12 hours after THERMAL POWER> 25% RTP. D INCORRECT: Part 1 incorrect See Explanation C. Part 2 correct See Explanation B.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification: The KA is met because the question tests the candidates ability to analyze the effect of maintenance activities on the status of limiting conditions for operations associated with the Control Air Systems. SRO Only Justification: This question is SRO Only because it meets the requirements of Clarification Guidance for SRO, Section Il.B Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]. The question involves application of Required Actions (Section 3) in accordance with rules of application requirements (Section 1). See Attached. Candidate must determine power limitations and allowed time to achieve with Plant Control Air aligned to Drywell Control Air aligned to allow maintenance. Question Cognitive Level: Question rated as Fundamental Knowledge. Technical Reference(s): TRM 3.6-5 Rev. 55 (Attach if not previously provided) 3-Ol-32A Rev. 25 / OPL1 71.054 Rev. 15 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL1 71 .054 V.B.8 (As available) Question Source: (Note changes or attach parent) Question History: (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.) Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 X Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Clarification Guidance for SRO-only Questions Rev 1(0311112010) Figure 1: ScreenIng for SRO-only linked to 10 CFR 55.43(b)(2) (Tech Specs) Can question be answered solely by knowing I hour TS/TRM Action? oquesn No Can question be answered solely by knowing the Yes - LCO/TRM information listed above-the-line? RO question No Can question be answered solely by knowing the Yes TS Safety Limits? RO question No Does the question involve one or more of the following for TS, TRM, or ODCM?

      ,. Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)

• Application of generic LCO requirements (LCO 3.01 thru 3.0.7 and SR 4.0.1 thru 4.0.4)
• Knowledge of TS bases that is required to analyze TS Yes j SRO-only question required actions and terminology No Question might not be linked to 10 CFR 55.43(bX2) for SRO-only Page 5 of 16

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Drywell Control Air System 3-OI-32A Unit 3 Rev. 0025 Page 6 of 11 3.0 PRECAUTIONS AND LIMITATIONS A. While shutdown, plant control air is normally aligned to the Drywell Control Air System. During power operation, the Containment Inerting System is normally aligned to Drywell Control Air. B. Technical Requirements Manual Section 3.6.3 requires reactor thermal power be reduced to less than or equal to 15% power within 24 hours if plant control air is being used to supply the pneumatic control system inside primary containment. C. The CAD to Drywell Control Air (DWCA) crosstie provides long term MSRV accumulator gas supply in order to fulfill Appendix F fire requirements, It can also be used during short periods as a backup supply to the DWCA. DCN W17937A ensure that CAD is operable per Tech Spec 3.6.3.1 by replacing the CAD to DWCA crosstie piping with Seismic Class 1 components and verifying that the ability of CAD to perform its primary safety function is NOT altered when crosstied to DWCA. DCN Wi 7937A also ensures that CAD meets Primary Containment Integrity/Isolation per Tech Spec 3.6. D. Regulators PREG-32-49A & -49B have a relatively low volume flow rate to accommodate normal DWCA usage. As such, regulator bypass valve BYV-32-141 should be used to fill empty or very low pressure DWCA receiver tanks A & B to save wear on the regulators. For example when these tanks have been emptied for maintenance or other activities.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Drywall Control Air System TR 3.6.3 TR 3.6 CONTAINMENT SYSTEMS TR 3.6.3 Drywell Control Air System LCO 3(13 The pneumatic control system inside primary containment shaH be supplied from the Drywall Control Air system or the Containment Atmosphere Dilution system. APPLICABILITY: When primary containment inerting is required ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Plant control air is A.1 Reduce THERMAL 24 hours being used to supply POWER to 15% RTP. the pneumatic control system inside primary containment. TECHNICAL SURVEILLANCE REQUIREMENTS TSR 3.6.3.1 The plant control air supply valve located Prior to outside primary containment for the completing pneumatic control system inside primary primary containment shall be verified closed, containment inerting during reactor startup AND Every 31 days thereafter BFN-UNIT 3 3.6-5 TRM Revision 55 March 09, 2006

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI71 .054 Revision 15 Page 37 of 69

b. If Oxygen concentration is outside limit, restore to within limit within 24 hours. If limit is not restored within 24 hours, reduce Thermal Power 15% within 8 hours.

3. TRM Sect. 3.6.3 Drywell Control Air System.

a. Units 1, 2, and 3 require that Thermal Power be reduced to 15% within 24 hours if plant control air is being used to supply the pneumatic control system inside the primary containment.

F. Procedural Requirements Obj. V.B.8 Obj. V.C.8 t Review the procedure requirements for the control air Obj.V.D.i1 system, the service air system, and the drywell control Obj. V.E.13 air system.

a. 0-01-32, SER 03-05

b. 0-01-33 Emphasize indications

c. 1-0l-32A, 2-Ol-32A and 3-Ol-32A and component response of loss of air

d. 0-AOl-32-1, 1 -AOl-32-2, 2-AOl-32-2 and 3-A0l-32-2 events.

e. 1-AOl-32A-1, 2-AOl-32A-i and 3-AOI-32A-1

f. 1,2,3 -E0l-Appendix Ii A; Alternate pressure control Unit differences MSRVs permits the initiation of EOI appendix 8G.

On Units 1 ,2 CAD A supplies 7 MSRVs and CAD B See panel 9-3 label next supplies 6. to Ll-64-159A; Supp Pool Level OR Label Unit 3 CAD A supplies 6 MSRVs and CAD B behind 2-XS supplies 7 MSRVs 161)1 62; Supp pool selector

g. EOl Appendix 8G; Crosstie CAD to DW Control Air MSIVs (Inboard) A&B CAD A MSIVs (inboard) C&D CAD B

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS

                       -*                            NOTES

1. RefertoTable 3.31.1-i to determine which SRs applyfor each RPS Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains RPS trip capability.

SURVEILLANCE FREQUENCY SR 3.31.11 Perform CHANNEL CHECK. 24 hours SR 3.3.1.1.2 Not required to be performed until 12 hours after THERMAL POWER 25% RTP. Verify the absolute difference between the 7 days average power range monitor (APRM) channels and the calculated power is 2% RTP while operating at 25% RTP. SR 3.3.1.1.3 NOTE ------ Not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2. Perform CHANNEL FUNCTIONAL TEST. 7 days (continued) BFN-UNIT 3 3.3-4 Amendment No. 213 September 03, 1998

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RD SRO 202001 Recirculaton Tier # 2 A2.13 (IOCFR 55.43.5 SRO Only) Ability to (a) predict the impacts of the following on the Group # 2 RECIRCULATION SYSTEM; and (b) based on those predictions, K/A # 202001A2.13 use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: 2 8 Carryunder Importance Rating Proposed Question: # 91 Which ONE of the following completes the statements below in accordance with Tech Spec 3.3.1.1, Reactor Protection System (RPS) Instrumentation AND its associated Bases? ONE of the bases of Reactor Vessel Water Level - Low, Level 3 RPS function is to prevent significant carryunder to protect _(1)_. If this function is lost due to TWO inoperable channel in a trip system, RPS trip capability must be restored with a completion time of(2)_. A. (1) the accuracy of Reactor Level Instrumentation (2) Immediately B. (1) the accuracy of Reactor Level Instrumentation (2) 1 hour C. (1) available Reactor Recirc Pump Net Positive Suction Head (2) Immediately D. (1) available Reactor Recirc Pump Net Positive Suction Head (2) 1 hour Proposed Answer: D Explanation A INCORRECT: Part 1 incorrect Plausible in that Reactor Level (Optional): instrumentation has taps in the Downcomer region where the dynamics are altered as a result of significant carryunder. Part 2 incorrect Plausible in that Immediate is a common completion time in Tech Specs. B INCORRECT: Part 1 incorrect See Explanation A. Part 2 correct See Explanation D. C INCORRECT: Part 1 correct See Explanation D. Part 2 incorrect See Explanation A. D CORRECT: Part 1 correctIn accordance with TS 3.3.1.1 Bases, The Reactor Vessel Water Level Low, Level 3 Allowable Value is selected to ensure that during normal operation the steam dryer skirt is not uncovered (this protects available recirculation pump net positive suction head (NPSH) from significant carryunder). Part 2 correct In accordance with TS 3.3.1.1, Condition C, with one or more Functions with RPS trip capability not maintained, restore trip capability within 1 hour.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification: The KA is met. To answer the question, the candidate must predict the impact of carryunder on the Recirculation System. Then, utilize Tech Specs and associated implementing procedures to mitigate the consequences loss of Level 3 RPS Channel designed to protect the Recirculation System from the impact of carryunder. SRO Only Justification: This question is SRO Only because it meets the requirements of Clarification Guidance for SRO-only Section ll.B Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)j. The question involves knowledge of TS bases for Level 3 RPS. Question Cognitive Level: Question rated as Fundamental Knowledge. Justification: Question requires knowledge of Tech Spec bases and is therefore, SRO-Only. Technical Reference(s): Ui TS 3.3.1-2 Amm 262 (Attach if not previously provided) Ui TS B 3.3.-18 Rev. 0 (Including version I revision number) Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL1 71.028 V.B. 14 (As available) Question Source: Barik# Modified Bank# (Note changes or attach parent) New X Question History Last NRC Exam (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.) Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 X Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Clarification Guidance for SRO-only Questions RevI (0311112010) igure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2) (Tech Specs) Con question he answered solely by knowing . 1 Yes hour TS/TRM Action? RO question Can question be answered solely by knowing the Yes - LCOITRM information listed above-the-lineT RO question No r Con question be answered solely by knowing the Yes - TS Safety Limits? PC question r Does the question involve one or more of the following for TS, TRM, or ODCM?

• Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
• Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.04) Yes SRO-oi Knowledge of TS hoses that is required to analyze TS question required actions and terminology No j

Question might not he linked to 10 CFR 55.43(h)(2) for SRQ-only Page 5 of 16

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet RPS Instrumentation 31.1.1 ACTIONS (cofltiflLled) CONDITION REQUIRED ACTION COMPLETION TI ME C. One or more Functions Ci Restore RPS trip 1 hour with RPS trip capability capability. not maintained. D. Required Action and D.i Enter the Condition Immediately associated Completion referenced in Time of Condition A. B. or Table 33.11- I for the C not met. channel. E. As required by Required E. 1 Reduce THERMAL 4 hours Action D. I and POWER to < 30% RTP. referenced in Table 3.3.1.1-i. F. As required by Required F.i Be in MODE 2. 6 hours Action D. 1 and referenced in Table 3.3:1.1-i. G. As required by Required G,i Be in MODE 3. 12 hours Action D. 1 and referenced in Table 3.3.1.1-i. H. As required by Required H.i Initiate action to fully Immediately Action D.i and insert all insertable referenced in control rods in core cells Table 3.3.1.1-I. containincl one or more fuel assemblies. BEN-UNIT 1 Amendment No. 234262 September 27. 2006

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet RPS Instrumentation B 331.1 BASES APPLICABLE 4. Reactor Vessel Water Level Low, Level 3 SAFETY ANALYSES. (LIS-3-203A. LIS-3-203B. LIS-3-203C. and LlS-3-203D) LCO, and APPLICABILITY Low RPV water level indicates the capability to cool the fuel (continued) may be threatened. Should RPV water level decrease too far. fuel damage could result. Therefore, a reactor scram is initiated at Level 3 to substantially reduce the heat generated in the fuel from fission. The Reactor Vessel Water Level Low, - Level 3 Function is assumed in the analysis of the recirculation line break (Ref 6). The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the Emergency Core Cooling Systems (ECCS). ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. Reactor Vessel Water Level Low, Level 3 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor \!essel Water Level Low, Level 3 Function, with two channels in each trip system arranged in a oneoutoftwo logic, are required to he OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. r The Reactor \/essel Water Level Low. Level 3 Allowable Jalue is selected to ensure that (a) during normal operation the steam dryer skirt is not uncovered (this protects available j recirculation pump net positive suction head (NPSH) from significant carryund.er), and (b) for transients involving loss of all normal feeciwater flow, initiation of the low pressure ECCS subsystems at Reactor \/essel Water Low Low Low, Level 1 will not be required. (continued) BFN-UNIT 1 B 3.3-18 Revision 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT RPS Instrumentation 3.3 Li ACTIONS (continuedt CONDITION REQUIRED ACTION COMPLETION TI ME C. One or more Eunc!ions C.i Restore RPS trip 1 hour with RPS trip capability capability. not maintained. D. Required Action and D.i Enter the Condition Immediately associated Completion referenced in Time of Condition A, B. or Table 3.3.1 .1-I for the C not met. channel. E. As required by Required E. 1 Reduce THERMAL 4 hours Action D.i and POWER to < 30% RTP. referenced in Table 3.3.1.1-i. F. As required by Required F.l Be in MODE 2. 6 hoLirs Action D. l and referenced in Table 3.3.1.1-i. G. As required by Required 0.1 Be in MODE 3. 12 hours Action D. I and referenced in Table 3.3.1.1-I. H As required by Required H.i Initiate action to fully Immediately Action D. 1 and insert all insertable referenced in control rods in coie cells Table 3.3.1.1-i. containinq one or more fuel assemblies. BFN-IJNIT 3.3-2 Amendment No. 234,262 September 27, 2006

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT OPLI71.003 Revision 19 Appendix C Page 63 o 66 CTUA LEVELS

                             ,/

I ACCDSNT u nnru H u PJ ULU II LR332 TP-3 VESSEL LEVEL INSTRUMENT RANGES

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 216000 Nuclear Boiler instrumentation Tier # 2 G2.4.45 (IOCFR 55.43.5) Group# 2 Ability to prioritize and interpret the significance of each annunciator oralarm. KIA# 216000G2.4.45 Importance Rating 4.3 Proposed Question: # 92 The following alarms AND indications exist on Unit 3:

• DRYWELL PRESS HIGH, (3-9-3B, Window 23), is in alarm
• REACTOR VESSEL WTR LVL CH A LOW-LOW-LOW (3-9-5B, Window 4), is in alarm
• REACTOR VESSEL WTR LVL CH B LOW-LOW-LOW (3-9-5B, Window 5), is in alarm
• DRYWELL EQPT DR SUMP PUMP EXCESSIVE OPRN, (3-9-4B, Window 11), is in alarm
• Drywell Floor Drain Leakage is calculated at 100 gpm
• Group I PCIS Logic A Success light is NOT illuminated
• ALL other PCIS Logic Success lights are illuminated Which ONE of the following completes the statement below?

These alarms AND indications establish that A. a loss of the Fuel Clad Barrier ONLY exists B. a loss of the Reactor Coolant System Barrier ONLY exists C. a loss of the Reactor Coolant System Barrier AND Fuel Clad Barrier ONLY exists D. a loss of the Containment Barrier AND Reactor Coolant System Barrier ONLY exists Proposed Answer: B Explanation A INCORRECT: The threshold for fission product barrier loss a Reactor (Optional): coolant sample that yields a result of 300 pCi/gm lodine-131 equivalent is indicative of cladding failure. There is no indication of elevated coolant samples. Plausible in that with indications of a Loss of Coolant Accident and very low Reactor Water, candidate may conclude that fuel damage has occurred. B CORRECT: The threshold for Reactor Coolant System fission product barrier loss is considered to be consistent with Reactor coolant leakage of at least 50 GPM from the primary system. C INCORRECT: The threshold for fission product barrier loss a Reactor coolant sample that yields a result of 300 pCi/gm Iodine-i 31 equivalent is indicative of cladding failure. Plausible in that with indications of a Loss of Coolant Accident and very low Reactor Water, candidate may conclude that fuel damage has occurred. The first part is correct.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet D INCORRECT: The threshold for Primary Containment fission product barrier loss is considered to be consistent with the following: Refer to 2.5-U. Unexplained Loss Of Containment Pressure / Exceeding SI-4.7.A.2.a Limits (Excessive N2 Makeup) / Inability To Isolate Any Line Exiting Containment When Isolation Is Required I Venting Irrespective Of Offsite Release Rates Per EOIs / SAMGs. Plausible in that REACTOR VESSEL WTR LVL CH A/B LOW-LOW-LOW alarms establish that MSIV isolation is required. Although the Group 1 PCIS Logic A Success light not illuminated indicates failure of the logic channel, one channel would meet the requirement to isolate the Main Steam Lines. The second part is correct. KA Justification: The KA is met because the question requires the candidate to interpret alarms and indications associated with the Nuclear Boiler System to determine Barrier losses in accordance with EPIP Bases. SRO Only Justification: This question meets the requirements of Clarification Guidance for SRO-only Questions, Section ll.F Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity. [10 CFR 55.43(b)(6)] (See Attached). This question requires evaluating core conditions, Reactor Coolant System Barrier and Containment Barrier in accordance with the Emergency Classification Procedure Technical Bases Question Cognitive Level: This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): EPIP-1 Rev. 46 (Attach if not previously provided) Proposed references to be provided to applicants during examination: NONE Learning Objective: (As available) Question Source: Bank # Modified Bank# BFN 1006 #100 (Note changes or attach parent) PERRY 07 SRO #10 New Question History: Last NRC Exam Browns Ferry 1006 Perry 2007 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.) Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Clarification Guidance for SRO-only Questions Rev 1(03111/2010) F. Procedures and limitations involved in initial core loading, alterations in core configuration. control rod programming, and determination of various internal and external effects on core reactivity. 110 CFR S543(by6)1 Some examples of SRO exam items for this topic include:

      .      Evaluating core conditions and emergency classifications based on core conditions.

• Administrative requirements associated with low power physics testing processes.
• Administrative requirements associated with refueling activities, such as approvals required to amend core loading sheets or administrative controls of potential dilution paths and/or activities.
• Administrative controls associated with the installation of neutron sources.
• Knowledge of TS bases for reactivity controls.

G. Fuel handling facilities and procedures. [10 CFR 55.43(b)(7)1 Some examples of SRQ exam items for this topic include:

• Refuel floor SRO responsibilities.
• Assessment of fuel handling equipment surveillance requirement acceptance criteria.

a Prerequisites for vessel disassembly and reassembly.

• Decay heat assessment.
• Assessment of surveillance requirements for the refueling mode.
• Reporting requirements.

Emergency classifications. This does not include items that the RO may be responsible for at sonic sites such as fuel handling equipment and refueling related control room instrumentation operability requirements, abnornial operating procedure immediate actions. etc. For example. an RO is required to stop the refueling process vhen communication is lost between the control room and the refueling floor, therefore, this is a task that is both an RD and SRO responsibility and is not SRO-only. Page S of 16

ES-401 Sample Written Examination Form ES-40I-5 Question Worksheet BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE TECHNICAL BASIS E PIPI GENERAL EMERGENCY EAL: Events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential for toss of containment ntegrity or HOST1LE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area. OR Loss of any two barriers and potential loss of third harrier. OPERATING CONDmON: ALL BASIS: This event classification is intended to address unanticipated conditions not explicitly addressed elsewhere but that warrant declaration ot an emergency because conditions exist which are believed by the Site Emergency Director to tall under the General Emergency classification. BFN EALs were developed primarily utilizing the symptom based grouping methodology. This approach is consistent with the BFN EOI methodology. It is important to note here that the consideration & fission product barriers has been incorporated within this symptom based approach. Barrier-based EALs refer to the level of challenge to principal barriers useo to assure containment of radioactive material. For radioactive materials that are contained within the reactor core, these barriers are: fuel cladding, reactor coolant system pressure boundary, and containment. The level of challenge to these harriers encompasses the extent of damage (loss or potential loss) and the number of barriers currently under challenge. Site Emergency Directors should be continuously aware of all challenges to these barriers and the number of barriers loss or potentially loss. Also Site Emergency Directors should consider that when the loss or potential loss thresholds is imminent (i.e., I to 3 hours) use judgment and classify as if the thresholds are exceeded. Loss or potential loss of all fission product harriers must be considered along with inability to monitor fission product barriers during extreme conditions. The threshoid for fission product barrier loss is considered to he consistent wth the following: Fuel clad A Reactor coolant sample that yields a result of 300 pCi/gm lodine-i31 equivalent is indicative of cladding failure (Refer to 1.3-A). RCS barrier Reactor coolant leakage of at least 50 GPM from the primary system (Refer to 2.4-A). Prima Containment barrier Refer to 2.5-U,

REFERENCES:

Reg Guide 1.101 Rev. 3, (NUMARC HG2, FG) NRC Bulletin 2005-02, Juiy 18, 2005 Attachment 2 (Emergency Classification Level changes) NEI White Paper, Enhancements to Emergency Preparedness Programs for Hostile Action, May 2005 (Revised November 18. 2005) PAGE 204 OF 206 REVISION 46

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ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BROWNS FERRY 1006 Examination Outline Cross-reference: Level RO SRO G2.4.45 (IOCFR55.43.5SROOnIy) Tier# 3 Ability to prioritize and interpret the significance of each Grou p # annunciator or alarm. K/A # G2.4.45 Importance Rating 4.3 Proposed Question: # 100 The following alarms AND indications exist on Unit 3:

• DRYWELL PRESS HIGH, (3-9-3B, Window 23), is in alarm o Reactor Level is (-) 130 inches and lowering slowly
• DRYWELL EQPT DR SUMP PUMP EXCESSIVE OPRN, (3-9-4B, Window 11), is in alarm
• Drywell Floor Drain Leakage is calculated at 100 gpm
• Reactor Coolant Sample yields a result of 310 pCi/gm Iodine-131
• Group 1 PCIS Logic A Success light is NOT illuminated Which ONE of the following completes the statement?

These alarms AND indications establish that A. a loss of the Fuel Clad Barrier ONLY exists B. a loss of the Reactor Coolant System Barrier ONLY exists C. a loss of the Containment Barrier AND Fuel Clad Barrier ONLY exists D. a loss of the Reactor Coolant System Barrier AND Fuel Clad Barrier ONLY exists Proposed Answer: D Explanation (Optional): A INCORRECT: Reactor Coolant System fission product barrier is also lost. B INCORRECT: Fuel Clad Barrier is also lost. C INCORRECT: The threshold for Primary Containment fission product barrier loss is considered to be consistent with the following: - Refer to 2.5-U. Unexplained Loss Of Containment Pressure I Exceeding SI-4.7.A.2.a Limits (Excessive N2 Makeup) I Inability To Isolate Any Line Exiting Containment When Isolation Is Required I Venting Irrespective Of Offsite Release Rates Per EOls I SAMGs. D CORRECT: The threshold for fission product barrier loss - a Reactor coolant sample that yields a result of 300 pCi/gm Iodine-131 equivalent is indicative of cladding failure. The threshold for Reactor Coolant System fission product barrier loss is considered to be consistent with Reactor coolant leakage of at least 50 GPM from the primary system.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet PERRY 2007 NRC SRO #10 The plant scrams from 100% power. The following alarms and indications are called to your attention:

• Drywell Pressure 1.7 psig and rising
• Reactor Level at 50 and slowly lowering
• Containment Pressure 2.5 psig and rising
• DW UNIDENT1HED RATE OF CHANGE HIGH, recorder on high peg These alarms and indications establish that A. no loss of a Fission Product Barrier currently exists B. a loss of the Fuel Clad Barrier exists C. a loss of the Reactor Coolant System Barrier exists D. a loss of the Containment Barrier exists

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Level: RO SRO Tir Examination Outline Cross-Reference flrni in K/Alt 2Lfl24 2 4 4 lmnnrfnr I A K&A: Ability to prioritize and interpret the significance of each annunciator or alarm. High Drywell Pressure Explanation (Why the distractors are incorrect): Answer C A Loss of RCS exists B not a loss of fuel barrier D not a loss of containment barrier Technical Reference(s): Reference Attached: EPI-Al Fission Product Barrier Matrix EPI-Al Fission Product Barrier Proposed references to be provided to applicants during examination: None Learning Objective (As available): EPL-0804-O1 4 Question Source: Bank # Modified Bank # New X Question History: Previous NRC Exam Previous Quiz I Test Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Xb.5 Comments: Level of Difficulty = 3

Sample Written Examination Form ES-401-5 ES-401 Question Worksheet Level RO SRQ Examination Outline Cross-reference: 27l0000ffgasSystem Tier# 2 G2.2.40 (IOCFR 55.43.2 SRO Only) 2 Group # Ability to apply Technical Specifications for a system. K!A# 271000G2.2.41 Importance Rating 4.7 L Proposed Question: # 93 out for Unit 3 is operating at 100% Reactor Power. Offgas Hydrogen Analyzer 3A was tagged planned maintenance at 0600 on 1/13/11. er 3B At 0700 on 1/13/11, the Unit Supervisor discovers an error on Offgas Hydrogen Analyz Surveillance completed at 0400 on 1/13/11. Based on the corrected calculation, Offgas 3.7.2 is not Hydrogen Analyzer 3B alarm setpoint is set too high to ensure the limit of TRM LCO exceeded. Which ONE of the following completes the statements below? In accordance with TR 3.7.2, Airborne Effluents, the concentration of hydrogen in Offgas ance with downstream of the recombiners shall be limited to a MAXIMUM of (1). In accord with a start TR 3.3.9, Offgas Hydrogen Analyzer Instrumentation, Condition A must be entered time of (2)on 1/13/11. [REFERENCE PROVIDED] A. (1)1% (2) 0600 B. (1)1% (2) 0700 C. (1)4% (2) 0600 D. (1)4% (2) 0700 Proposed Answer: D Explanation A INCORRECT: Part 1 incorrect. Plausible in that this is the alarm set point (Optional): for the Offgas H2 Analyzers. Part 2 correct Plausibility based on misconception that start time should be when surveillance was complete. B INCORRECT: Part 1 correct See Explanation D. Part 2 incorrect See Explanation A. C INCORRECT: Part 1 incorrect See Explanation A. Part 2 correct See Explanation D.

Sample Written Examination Form ES-401-5 ES-401 Question Worksheet D CORRECT: Part 1 correct In accordance with TR 37.2, Airborne Effluents, the concentration of hydrogen in Offgas downstream of the recombiners shall be limited to 4%. Part 2 correct In accordance with TRM 3.0.2, start time is based on time of discovery KA Justification: The KA is met because the question tests the candidates ability to apply Technical Specifications for the Offgas System SRO Only Justification: This question is SRO Only because it meets the requirements of Clarification Guidance for SRO, Section lI.B Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)J. The question involves application of Required Actions (Section 3) in accordance with rules of application requirements (Section 1). See Attached. Candidate must determine the start time for Offgas Hydrogen Analyzers in accordance with LCO applicability section 3.0.2. Question Cognitive Level: This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. Technical Reference(s): U3 TR 3.3-54 Rev. 16 (Attach if not previously provided) U3 TRM 3.0-1 Rev. 44 U3 TRM 3.7-3 Rev. 0 Proposed references to be provided to applicants during examination: TR 3.3.9 (No SRs and No Bases) Learning Objective: OPL171.087 V.B.10 (As available) Question Source: Bank # Modified Bank # (Note changes or attach parent) New X Question History: Last NRC Exam failure to (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; provide the in formation will necessitate a detailed review of eveiy question.) Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X Comments:

Form ES-401-5 ES-401 Sample Written Examination Question Worksheet tions Clarification Guidance for SROonIy Ques RevI (0311112010) d to 10 CFR 55.43(b)(2) Figure 1: Screening for SRQ-only linke (Tech Specs) ng 1 Yes Can question be answered solely by knowi RD question hour TS/TR M Action ? No ingthe Can question be answeredsoielybvknow LCO/TRM infor rn Yes Can question be answered solely by knowing the R() question TS Safety Limits? ing for TS, Does the question involve one or more of the follow TRM, or ODCM? eillance

• Application of Required Actions (Section 3) and Surv of Requirements (Section 4) in accordance with rules application requirements (Section 1) thru
• Application of generic LCD requirements (LCD 3.0:1 3.0.7 and SR 4.0.1 thru 4.0.4) Yes , SRO-oi ze TS question
• Knowledge of TS bases that is required to analy required actions and terminology Question might not he linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Offgas Hydrogen Analyzer 1 nstrumentation TR 3.3.9 TR :3.3 INSTRUMENTATION TR 3.3.9 Offgas Hydrogen Analyzer instrumentation [CO 3.3.9 There shall be at least one OPERABLE Offgas Hydrogen Analyzer instrument with alarm setpoint set to ensure the limit of TRM LCO 3.7.2 is not exceeded. APPLICABILITY: During main condenser offgas treatment system operation N OTE TRM LCD 3.0.3 is not applicable. ACTI ONS CONDITION REQUIRED ACTION COMPLETION TIME A. No OPERABLE Oftcjas A. 1 Install a temporary 4 hours Hydrogen Analyzer monitor instruments. OR A.2.l Take grab samples 4 hOLIrS from discovery of no OPERABLE AND instrument AND Every 4 hours thereafter A.2.2 Analyze the sample for 4 hours following explosive concentration grab so rnpie of hydrogen. BFN-UNIT 3 3.3-54 TRM Revision 07 16 March 31 2000

Sample Written Examination Form ES-401-5 ES-401 Question Worksheet Airborne Effluents TR 3.7.2 TR 3.7 PLANT SYSTEMS YR 3.72 Airborne Effluents LCD :3.7.2 Whenever the SJAE is in service, the concentration of hydrogen in the offaas downstream of the recombiners shall be limited to 4% by volume. APPLICABILITY: During main condenser ofigas treatment system operation NOTE TRM LOG 3.0.3 is not applicable. ACTION S CONDITION REQUIRED ACTION COMPLETION TI ME A. With the concentration A. I Restore the 43 hours of hydrogen >4% by concentration to within volume, the limit. TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY TSR 3.7.2.1 The concentration of hydrogen downstream Continuously by of the recombiners shall be determined to be at least one 4% by volume by monitoring the off-gas OPERABLE whenever the SJAE is in service using Offgas instruments described in Technical Hydrogen Requirement 3.3.9. Analyzer OR As required by TR :3.3.9 when Offgas Hydrogen Analyzer instrumentation is inoperable BEN-UNIT 3 3.7-3 TRM Revision 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet LCD Apphcability

3.0 3.0 LIMITING CONDITION FOR OPERATION (LCD) APPLICA6ILITY LCD 30.1 TRM LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided m TRM LCD 3.0.2.

LCD 3.0.2 Upon discovery of a failure to meet a IRM LCD, the Required Actions of the associated Conditions shall be met. except as provided in TRM LCD 3.0.5 and TRM LCD :3.0.6. If the TRM LCD is met or i.s no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required, unless otherwise stated. LCD 3.0.3 When a TRM LCD is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall I:e placed in a MODE or other specified condition in which the TRM LCD is not applicable. Action shall be initiated within 1 hour to j:lace the unit, as applicable, in:

a. MODE 2 within 10 hours;

b. MODE 3 within 13 hours: and

c. MODE 4 within 37 hours.

Exceptions to this Requirement are stated in the individual Requirements. Where corrective measures are completed that permit operation in accordance with the TRM LCD or ACTIONS. completion of the ACTIONS required by TRM LCD 3.0.3 is not required. TRM LCD 3,0.3 is only applicable in MODES I, 2, and 3. BFN-LJNIT 3 3.0-1 TRM Revision 9-2. 44 March 22. 2004

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet PLAUSIBILITY SUPPORT OPLI 71.030 Revision 18 Page 26 of 74 INSTRUCTOR NOTES (1) Sample flow will be provided by SJAE INLET PIPING routing the discharge of the analyzers to the main condenser and allowing vacuum to produce the motive force. (2) Hydrogen concentration indicates on Flammable range of the local panel and in the control room 2 in air is 4%7$% H in a range of 0 5%. Hydrogen concentrations of 1% or greater will This is an industrial generate local and control room safety concem alarms. (3) Oxygen indicates only locally in a range of 0 50%. c The analyzers and their local controls are Panel 25-588 located on mezzanine level above SJAE Rooms.

d. The HWC local contiols are located near the Panel 25-589 Condensate Booster pumps

e. Ten amber status lights for each analyzer have been added to panel 9-53 in the control roont The analyzer lights al-c:

(1 ) NORMAL OPG MODE (normally lit) This light will illuminate when the local Mode Select Switch (Train A!Normal/Train B) is in the Normal position AND the local Functional Test Switch (Test/Operate) for the respective sample train is in the Operate position.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO G2.1.3 (IOCFR55.43.2SRO Only) Tier# 3 Knowledge of shift or short-term relief turnover practices. Group # KIA# G2.1.3 Importance Rating ---- 3.9 Proposed Question: # 94 Which ONE of the following completes the statements below for Shift Turnover AND Control Board walk down requirements in accordance with OPDP-1 ,Conduct of Operations? During shift turnover, the oncoming Unit Supervisor _(1) required to walk down the Control Boards with an off going RO or SRO. The Unit Supervisor also must walk down Main Control Room panels (2)_. A. (1)is (2) once prior to mid shift brief AND once prior to end of shift turnover B. (1)is NOT (2) once prior to mid shift brief AND once prior to end of shift turnover C. (1) is (2) once every hour during power operations with a 25% grace period D. (1)isNOT (2) once every hour during power operations with a 25% grace period Proposed Answer: A j Explanation A CORRECT: Part I correct In accordance with OPDP-1, the oncoming (Optional): Unit Supervisor will conduct control board walk downs with an off-going Operator. Part 2 correct In accordance with OPDP-1, the Unit Supervisor walks down the main control room panels once each shift prior to the mid-shift brief and once prior to end-of-shift turnover. B INCORRECT: Part 1 incorrect See Explanation D. Part 2 correct See Explanation A. C INCORRECT: Part I correct See Explanation A. Part 2 incorrect See Explanation D. D INCORRECT: Part 1 incorrect Plausible in that this would be the correct answer for the Shift Manager. Part 2 incorrect Plausible in that this would be the correct answer for the Control Room Operators.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification: The KA is met because the question tests the knowledge of shift relief turnover practices for Unit Supervisors. SRO Only Justification: This question is SRO Only because Unit Supervisor turnover and Control Room walk down requirements are knowledge / abilities unique to the SRO position. Question Cognitive Level: Question rated as Fundamental Knowledge. Technical Reference(s): OPDP-1 Rev. 18 (Attach if not previously provided) Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL171.071 V.B.16 (As available) 1 Question Source: Bank # I Modified Bank # (Note changes or attach parent) New X Question History: Last NRC Exam (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.) Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 x Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet NPG Standard Conduct of Operations OPDP-1 Department Rev. 0018 Procedure Page 17of74 4.1 Control Board Monitoring (continued)

3. The restrictions above for the OATC are not applicable for that unit when the reactor vessel is defueled.

C. A walk down of the main control room panels is to be performed a minimum of once every hour duiing power operations (with a 25% grace period) and once per shift or more frequently as determined by the Llnit Supervisor when shutdown to ensure that indications are within established bands.

• The walk down or the panels in the Reactor Controls Area may be conducted by the OATC.
• The walk down of the MCR panels OUtside the Reactor Controls Area will either be conducted by the assigned Control Room Operator OR the OATC wiil he temporarily relieved by another licensed individual prior to leaving the Reactor Controls Area D. The Unit Supervisor walks down the main control room panels once each shift prior to the mid-shift brief and once prior to end-of-shift turnover with a focus on critical parameters with one of those walk downs being paired with a Unit Operator. The Shift Manager should pertonn an end of shift main control room board walk down. The walk down is not a component by component walk down but should concentrate on Safety-Related controls manipulated during the shiW E. When equipment/plant status is chanciing, all applicable indications will be monitored until the equipment/plant stabilizes.

F. During plant operations diverse indications will he used to monitor equiprnentplant performance, determine trends and ensure plant response during evolutions is as expected and correct for conditions. G. During periods such as watchstation turnover, shift turnover or pro-job briefings, the lJnit Supervisor should ensure one Operator maintains the OATC role. 4.2 Equipment Manipulations and Status Control A. All equipment manipulations are performed by qualified personnel in accordance with procedures and/or other documents such as work orders or clearances approved by shift supervision. B. The control of plant equipment status is governed by procedures, work orders. TACFs or tagging. These processes contain specific direction relative to status control. C. In situations where a component is required to be placed in a position differing from its normal alignment, the configuration change must i:e performed in accordance with approved plant specific processes unless the configuration change is immediately necessary to protect personnel, equipment or the public. D. Whenever an activity or evolution is interrupted, ensure affected equipment is placed in a stable condition as soon as prachcable.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet NPG Standard Conduct of Operations OPDP-1 Department Rev. 0018 Procedure Page 48 of 74 7.2 Error Prevention Tools (continued)

c. Control Room Two Minute Drill Prior to performing any Control Room Activity (except for EOI)EOP or AOI/AOP actions) ti-ic Operator performing the activity will pause to ensure the following as a minimum:

• The system line-up supports the evolution.
• Plant announcements made as applicable.
• Operability i risk has been evaluated.
• Radiation Protection contacted for expected dose rate changes.
• Expected response is anticipated.
• A plan exists for getting out of the evolution if needed.
• Other stakeholders notified (i.e. Radiation Protection, Chemistry.

Maintenance, Area Operator).

• Affects on reactivity understood and discussed.
• Known deficiencies that could effect desired outcome have been evaluated.

7.3 Shift Turnover A. Shift relief and turnover is conducted in a manner such that the oncoming shift has the knowledge to continue safe and efficient operation of the plant Attachment 2 (Form OPDPi-i), Shift Turnover Checklist or similar format is utilized to facilitate turnover. The following watchstations will conduct shift turnover:

• Shift Manager
• Unit Supervisors (MCR)
• Work Control SRO / SRO Designee
• Unit Operators
• Assistant Unit Operators (assigned duty stations)
• Shift Technical Advisor (or position assigned STA function) a Fire OPS Supervisor (FOS)

B. All shift personnel are responsible for reviewing and understanding the shift narrative log for the previous shift and turnover checklist applicable to their shift position before assuming the shift. C. The oncoming Unit Operators and Unit Supervisor will conduct control board walkdowns with an off-going Operator. The Shift Manager will walkdowri control boards as necessary to understand current plant conditions.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT NPG Standard Conduct of Operations OPDP.1 Department Rev. 0018 L Procedure Page l7of 74 4.1 Control Board Monitoring (continued)

3. The restrictions above for the OATC are not applicable for that unit when the reactor vessel is defueled.

C. A walk down of the main control room panels is to be performed a minimum of once every hour durine power operations (with a 25% grace period) and once per shift or more frequently as determined by the Unit Supervisor when shutdown to ensure that indications are within established bands.

• The walk down of the panels in the Reactor Controls Area may be conducted by the OATC.
• The walk down of the MCR panels outside the Reactor Controls Area will either be conducted by the assigned Control Room Operator OR the OATC will be temporarily relieved by another licensed individual prior to leaving the Reactor Controls Area.

D. The lJnit Supervisor walks down the main control room panels once each shift prior to the mid-shift brief and once prior to endof-shift turnover with a focus on critical parameters tith one of those walk downs beinq paired with a Unit Operator. The Shift Manager should perform an end of shift main control room board walk dawn. The walk down is not a component by component walk down but should concentrate on Safety-Related controls manipulated during the shift. E. When equipment/plant status is changing, all applicable indications will be monitored until the equipment/plant stabilizes. F. During plant operations diverse indications will he used to monitor equiprnentplant performance, determine trends and ensure plant response durinq evolutions is as expected and correct for conditions.

0. During periods such as watchstation turnover, shift turnover orpre-job briefings, the Unit Supervisor should ensure one Operator maintains the OATC role.

4.2 Equipment Manipulations and Status Control A. All equipment manipulations are performed by qUalified personnel in accordance with procedures andar other documents such as work orders or clearances approved by shift supervision. B. The control of plant equipment status is governed by procedures, work orders. TACFs or tagging. These processes contain specific direction relative to status control. C. In situations where a component is required to he placed in a position differing from its normal alignment, the configuration change must be performed in accordance with approved plant specific processes unless the configuration change is immediately necessary to protect personnel, equipment or the public. D. Whenever an activity or evolution is interrupted, ensure affected equipment is placed in a stable condition as soon as practicable.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO G2.1.4 (IOCFR55.43.2SRO Only) Tier# --- 3 Knowledge of individual licensed operator responsibilities Grou p # related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, K/A # G2.1 .4 10CFR55, etc. 3.8 Importance Rating ___ Proposed Question: # 95 In accordance with OPDP-10, License Status Maintenance, Reactivation and Proficiency for Non-Licensed Operators, which ONE of the following completes the statements for License Reactivation requirements? Licensee requalification training must be verified current _(1)_ 40 hours of shift functions under instruction. When ALL Reactivation requirements are met, the Licensed individual is authorized to resume licensed activities by the (2). A. (1) prior to standing (2) Plant Manager B. (1) prior to standing (2) Site Licensing Manager C. (1) any time during the (2) Plant Manager D. (1) any time during the (2) Site Licensing Manager Proposed Answer: A Explanation A CORRECT: (1) correct, Licensee requalification training is current, (Optional): including a simulator evaluation within the past 12 months in the position(s) to be assumed and the licensee has had a physical in the last two years. (To be verified prior to standing the 40 hours of shift functions under instruction.) (2) correct, Per OPDP-1 0 Appendix A: The above licensed individual is authorized to resume licensed duties. Date: I / Plant Manager B INCORRECT: (1) correct, (2) incorrect, Plant Manager not Licensing Manager. C INCORRECT: (1) incorrect, must be completed PRIOR to 40 hours. (2) correct, D INCORRECT: Part 1 and 2 incorrect.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification: The KA is met because the question tests the knowledge of individual licensed operator responsibilities associated with maintenance of active license status in accordance with 1 OCRF55.53. SRO Only Justification: This question meets the requirements of Clarification Guidance for SRO-only Questions, Section ll.A- Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)j. The question deals with the requirement of OPDP-10 which is the implementing procedure for license maintenance of license status in accordance with IOCFR55.53 Question Cognitive Level: Question rated as Fundamental Knowledge. Technical Reference(s): OPDP-10 rev 2 (Attach if not previously provided) (Including version / revision number) Proposed references to be provided to applicants during examination: NONE Learning Objective: (As available) Question Source: Bank # BFN 0801 #95 Modified Bank (Note changes or attach parent) Question History: Last NRC Exam Browns Ferry 0801 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of evety question.) Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 X Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Clarification Guidance for SRQ-only Questions Rev 1(03/1112010) H. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.ic]: A. Conditions and limitations in the facility license. [10 CFR 543Wyi)] Some examples of SRO exam items for this topic include:

• Reporting requirements when the maximum licensed thermal power output is exceeded.
• Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems. fire doors. etc.
• The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g.. shift staffing requirements).
• National Pollutant Discharge Elimination System (NPDES) requirements.

if applicable.

• Processes for TS and FSAR changes.

Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(by2) Some examples of SRO exam items for this topic include:

• Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
• Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
• Knowledge of TS bases that are required to analyze TS required actions and terminology.
• Same items listed above for the Technica[ Requirements Manual (TRM) and Offsite Dose Calculation Manual ODCM).

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of 1 hour action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items. SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROTs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet NPG Standard License Status Maintenance. OPDP-10 Department Reactivation and Proficiency for Rev. 0002 Procedure Non-Licensed Positions Page 14 of 21 Appendix A (PageS of 7) Return to Active Status Checklist Dt*ra:cns Tra in 9 Ma.tagn T-r,cm DDea:,ns StpeTeWent Date A. Licensee requalification training is current, including a simulator evaluation within the past 12 months iii the position(s) to be assumed and the licensee has had a physical in the last two years. (To be verified prior to standing the 40 hours of shift functions under instruction.) Deratcai Tm mj Managac B. The qualifications and status of the licensed individual listed above are current and valid, and Standards and Expectations have been discussed, prior to standing the 40 hours of shift functions under instruction. Date: / Dpera:icnai Spenn:enccmt C. If the licensee has a medical restriction requiring corrective lenses, the licensee will verify that he/she has the proper corrective lenses required to Don SCBA available while performing license duties (N/A if corrective lenses are not required). Licensee D. The above licensed individual has completed at least 40 hours of shift functions under the direction of an operator or senior operator, as appropriate, including a complete tour of the plant accompanied by an active licensed RD or SRO, as applicable, and review of all required shift tumover procedures. Date: Licensee Date: Shift Managec Date: Dpezatccic Sc ce,intendac.t Date: / Dparatcc Mananer E. The above licensed individual is authorized to resume licensed activities. nmt Manager F. Complete and Attach Appendix A Page 1, Licensee Documentation Form (SRO & RO) as the cover sheet for this documentation. Licensee cc: Ocea:,cns Marager Trang F/c ED IPJ1 NTh FD° ETU DL e O,,Tl lEP SS TO .7 E ST.nTUS

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO G2.2.23 (IOCFR 5543.2 SRO Only) Tier # 3 Ability to track Technical Specification limiting conditions for operations. Group# K/A # G2.2.23 Importance Rating ----- 4.6 Proposed Question: # 96 Which ONE of the following completes the statement for Completion Time extension for subsequent inoperability of a subsystem in accordance with TS Section 1 .3, Completion Times? If the criteria is met to apply a Completion Time extension, the total Completion Time allowed for completing a Required Action shall be limited to the _(1)_ restrictive of either:

• The stated Completion Time, as measured from the initial entry into the Condition, plus an additional _(2)_ ;OR
• The stated Completion Time as measured from discovery of the subsequent inoperability.

A. (1) more (2) 12 hours B. (1) less (2) 12 hours C. (i)more (2) 24 hours D. (1) less (2) 24 hours Proposed Answer: C Explanation A INCORRECT: Part 1 correct. Part 2 incorrect but plausible in that 12 hours (Optional): is a common Tech Spec criteria / completion time. B INCORRECT: Both are incorrect as explained below C CORRECT: If the subsequent inoperability existed concurrent with the first inoperability and remained inoperable after the first inoperability was resolved, Completion Times may be extended in accordance with TS Section 1 .3, Completion Times. The completion time extension will be the more restrictive of initial entry plus an additional 24 hours or completion time as measured from discovery of the subsequent inoperability. D INCORRECT: Part 1 incorrect but plausible in that when weighing alternative in accordance with Tech Spec use, application and applicability, the less restrictive is sometimes the criteria. Example: SR 3.0.3.

ES-401 Sample Written Examination Form ES-401 -5 Question Worksheet KA Justification: The KA is met because the question tests the candidates ability to track Technical Specification limiting conditions for operations by testing knowledge of Completion Time Extensions. SRO Only Justification: This question is SRO Only because it meets the requirements of Clarification Guidance for SRO, Section ll.B Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]. The question tests knowledge of application of generic Limiting Condition for Operation (LCO) requirements (Section 1 .3, Completion Times) Question Cognitive Level: Question rated as Fundamental Knowledge. Technical Reference(s): Ui TS 1.3-2 Amm 234 (Attach if not previously provided) (Including version I revision number) Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL1 71 .087 V.B.10 (As available) Question Source: Bank # Modified Bank #  : (Note changes or attach parent) New X Question History: Last NRC Exam (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eve,y question.) Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 X Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Clarification Guidance for SRO-only Questions RevI (03111/2010) Figure 1: Screening for SROonly linked to 10 CFR 55.43(b)(2) (Tech Specs) Can question be answered solely by knowing 1 Yes hour TSITRM Action? RO question No Can ctuestion be answered solely by knowing the LCC/TRM information Fisted abovethe-line? q,estior No I, C;an question he answered solely by knowing the Yes TS Safety Limits? RD question v No Does the question involve one or more of the following for TS. TRM, or ODCM? Application of Required Actions (Section 3 and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1) Application of generic LCD requirements (LCD 30 I thru 107 and SR 401 thru 40A) Yes SRO-only Knowledge of TS bases that is required to analyze TS question Lj. required actions and terminology No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Complehon Times [3 1.3 Completion Times DESCRIPTION Once a Condition has been entered, subsequent divisions. (continued) subsystems. components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will result in separate entry into the Condition uniess specifically stated. The Required Actions of the Condition continue to apply to each additional failure, with Completion Times based on initial entry into the Condition. However, when a subsequent division, subsystem, component, or variable expressed in the Condition is discovered to he inoperable or not within limits, the Completion Time(s) may he extended. To apply this Completion Time extension, two criteria must first be met. The subsequent inoperability:

a. Must exist concun-ent with the first inoperahility; and

b. Must remain inoperable or not within limits after the first inoperahility is resolved.

The total Completion Time allowed for completing a Required Action to address the subsequent inoperahility shall be limited to the more restrictive of either:

a. The stated Completion Time, as measured tnm the initial entry into the Condition, plus an additional 24 hours: or

b. The stated Completion Time as measured from discovery of the subsequent inoperahility.

The above Completion Time extensions do not apply to those Specifications that have exceptions that allow completely separate re-entry into the Condition (for each division, subsystem, component or variable expressed in the Condition) and separate tracking of Completion Times based on this re-entrg. These exceptions are stated in individual Specifications. (continued) E{FN-UNIT 1 [3-2 Amendment No. 234

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT SR Applicability 3.0 3.0 SURVEILLANCE REQUIREMENT tSR) APPLICABILITY SR 3.0.1 SR5 shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3. Surveillances do not have to he performed on inoperable equipment or variables outside specified limits. SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as meosured from the time a specified condition of the Frequency is met. For Frequencies specified as once, the above interval extension does not apply. If a Completion Time requires periodic performance on a once per basis, the above Frequency extension applies to each performance after the initial performance. Exceptions to this Specification are stated in the individual Specifications. SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours or up to the limit of the specifled Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall he performed for any Surveillance delayed greater than 24 hours arid the risk impact shall be managed. (continued) BFN-UNIT 1 3.0-4 Amendment No. 23, 243 December 23, 2002

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT 5DM 3.1.1 3.1 REACTIVITY CONTROL SYSTEMS 3:1. 1 SHUTDOWN MARGIN iSDM) LCO 3.1.1 SDM shall be within the limits provided in the COLR. APPLICABILITY: MODES I, 2, 3,4, and 5. ACT IONS CON D moN REQU IRED ACTION COMPLETION TI ME A. SDM not within limits in A. 1 Restore SDM to within 6 hours MODE 1 cr2. limits. B. Required Action and B. 1 Be in MODE 3. 12 hours associated Completion Time of Condition A not met. C. 5DM not within limits in C.l Initiate action to fully Immediately MODE 3. insert all insertable control rods. [). 5DM not within limits in Di Initiate action to fully Immediately MODE 4. insert all insertable control rods. AND (continued) BEN-UNIT 1 3.1-1 Amendment No. 234

ES-401 Written Examination Form ES-401 -5 Question Worksheet Examination Outline Cross-reference: Level RO SRO G2.2.44 (IOCFR 55.43.5 SRO Only) Tier # ----  : 3 Ability to interpret control room indications to verify the status Grou p # and operation of a system, and understand how operator actions and directives affect plant and system conditions. K/A # G2.2.44 Importance Rating 4.4 Proposed Question: # 97 A seismic event has resulted in the following Unit 2 plant conditions:

• ALL control rods are fully inserted
• RPV level is (-) 125 inches and lowering slowly
• RPV pressure is 450 psig with a cooldown in progress at <90 °F/hr
• RHR Loop Ills lined up for Drywell Spray
• ALL other ECCS systems are unavailable
• Drywell pressure is 4.8 psig and lowering
• ADS has been inhibited in accordance with 2-EOl-1, RPV Control step RC/L-7 Which ONE of the following describes the required actions to mitigate this event?

A. Enter 2-EOl-C1, Alternate Level Control and direct performance of 2-EOI-Appendix 6A, Injection Subsystems Lineup Condensate. B. Enter 2-EOI-C1, Alternate Level Control and direct performance of 2-EOI-Appendix 5A, Injection System Lineup Condensate/Feedwater. C. Enter 2-EOl-C2, Emergency Depressurization and direct performance of 2-EOl-Appendix 6A, Injection Subsystems Lineup Condensate. D. Enter 2-EOl-C2, Emergency Depressurization and direct performance of 2-EOl-Appendix 5A, Injection System Lineup Condensate/Feedwater. Proposed Answer: A Explanation A CORRECT: Part 1 correct With level less than (-) 122 inches and lowering (Optional): with no systems available to turn level for conditions, this is the appropriate leg of the EOls to select. Part 2 correct With the MSIVs closed and conditions not met to re-open MSIVs, this is the appropriate Appendix to select. B INCORRECT: Part 1 correct See Explanation A. Part 2 incorrect See Explanation D. C INCORRECT: Part 1 incorrect See Explanation D. Part 2 correct See Explanation A.

ES-401 Written Examination Form ES-401-5 Question Worksheet D INCORRECT: Part 1 incorrect. Direction to perform Emergency Depressurization based on reactor water level is given from ED I-Cl when RPV level drops below -162 inches. Other conditions given in the stem do not require Emergency Depressurization since Drywell Sprays have been initiated and appear to be effective. Part 2 incorrect. Appendix 5A is a lineup for injection with REPs which require MSIVs open. With RPV level below -122 inches, the MSIVs are closed. In addition, given all rods are in, performance of EDI Appendix 8A to bypass the MSIV low water level isolation is not appropriate KA Justification: The KA is met because the question tests the candidates ability to interpret control room indications to verify the status of injection systems and understand how operator actions and directives affect plant and system conditions. SRO Only Justification: This question is SRO Only because it meets the requirements of Clarification Guidance for SRO, Section ll.E Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)] See Attached. Candidate must assess plant conditions and then select a procedure, 2-EDI-APPENDIX 6A, Injection System Lineup Condensate, due to MSIVs closed and conditions not met to re-open them. Question Cognitive Level: Question rated as C/A because it involves a multi-part mental process of assembling, sorting and integrating the plant conditions given to determine required section of EOls and which Appendix to select. Technical Reference(s): 2-EOl-1 Rev 12 I 2-EOI C-I Rev. 9 (Attach if not previously provided) 2-EQ I Appendix 6A Rev. 4 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL171.205 V.B.l (As available) Question Source: Bank# BEN 07 SRO #18 Modified Bank# (Note changes or attach parent) Nw Question History: Last NRC Exam Browns Ferry 0707 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eveiy question.) Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 x Comments:

ES-401 Written Examination Form ES-401-5 Question Worksheet Clarification Guidance for SRO-only Questions Rev 1(0311112010) Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5) (Assessment and selection of procedures) systems knowledge. Le.. hOW the flowpath. logic, component location? system Can the question be answered solely by knowing works, No 1 immediate operator actions? Can the question be answered solely by knowing j Yes RO question No1 Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOP5? HFJ Can the question be answered solely by knowing the purpose. overall sequence of events, or overall mitigative strategy of a procedure? Ni [)oes the question require one or more of the following?

• Assessing plant conditions (normal, abnormal. or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps SRQ on
• Knowledge of diagnostic steps and decision points in the iiestio EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
• Knowledge of administrative procedures that specify hierarchy, implementation, andior coordination of plant normal, abnormal, and emergency procedures No I Question might not be linked to CFR 55A3(b)(5) for SRQ-only Page 8 of 16

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ES-401 Written Examination Form ES-401-5 Question Worksheet BE FORE VWI.7R LV ThAT C1.N R1OR tAD MICTtY L

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ES-401 Written Examination Form ES-401-5 Question Worksheet 2-101 APPENTIX-SA Rev. 4 Page 2 of 2

5. VERIFY OPEN the fo11ct-.ing heater osoatIorL va1vee

• 2Fov-3.--S6, p m 22 IN :NIET :som VLV
• 2FOV--331, HP HTR 232 1W :NLIT ISOL [LV
• 2Fov324, HP :-ITR 202 1W 2NLII 2301 VLV
• 2101375, HP :-ITR IA! 1W OUTLE: 1501 Vlv
• 2101-3-74, HP HTR 231 1W OUTLET 1501 VI?
• 2101-377, HP HTR 201 1W OUTLET 1501 [IV.

4. VERIFY OPEN the following RIP .SUOtiO:fl va_vest

• 2FOV263, RIP IA SUCTION VALVE
• 2IOV295, RIP 21 SUCTION VALVE
• 2IOV2ICS, RIP 20 SUCTION VALVE.

7. VERIFY at least one condensate oump rannznq.

1. VERIFY at least one condensate booster ojnijm runninq.

      . ADJUST 2L2O3o, RIN START-UP LEVEL OONTPOL,          to control injeotion (Pate! 295:)

IC. VERIFY PIN flo\ to RPV. LAST PAGE

ES-401 Written Examination Form ES-401-5 Question Worksheet 2ED: AIDIX5A. Rev S 3 Df 4

5. RAISE P2 2A{2E) (20: epeed UNTIL RFP iischarge presre i rxinaze1v eue P27 pre5ue ANS fc1Dwin: mezhcds cn Pne1 295

• Ung ivid 2N24E--SA(9A( IDA) , RFPT .(2E (22)

SPEED :o RA:SE /Lo;cER 3It ir iaz COVERNOR, OR

• Usrg 2s:c4ESg o, s: 2A(:3 2C)

SPEED CDNTRCL PDS in MANUAL, OR

• U3ing 2-1.IC-4-5, REACTOR ATER LEVEL CONTROL DS, n MANUAL iai: s::s 1D, RPT 2A(23) (2C SPEED CONTROL PDS in AUTO.

S SLOWLY RAISE soeed of RFPT UNTIL R? flow REV in+/-ctecl sinc ANY of the fci1ownq rrethods Panel 25:

• Using iciivideJ 2Es4eSA(9A) (13A), RFPT 2A(Dp) 2::

sPREt *CDNT RA:SE /LOWER switch in MANUAL GOERNOR, OR

• Using intivica+/- 2SiC46(9) ), REPT 2A(23)(2C)

SPEED CONTROL P05 in MANUAL, OR

• Using ZiIC--65, REACTOR WATER LEVEL CONTROL P05, in MANUAL with inthvithal 25ID4ES() 1o, REST 2A(2P) (2C) SPEED CONTROL 525 in AUTO.

ES-401 Written Examination Form ES-401-5 Question Worksheet 2ED: APPENDIXIA 4 cf 4

17. ADJUST RFP sneed as necessary no control injection usincj ANY of the following methods on Panel 25:

           +  Using individual 2ES46PA(GA) 10A), RFPT 2A(Th} 2C)

SPEED :DONT RA:SE1LONER switch in MANUAL GOVERNOR, OR

• Using individual 2S:C46I9 (1(1, REPT 2A(2Th 2C)

SPEED CONTROL LOS in MANUAL, OR

• Using 2010465, REACTOR PlATER LEVEL ODNTPCL PUS, MANUAL with individual 25:0461:9) fIr), PZPT lA23) (lCf SPEED CONUAOI C5 in AUTO.

1.1. TEEN RPV level is approxonately equal to des iced level AND automatic level control s desired, TEEN PLACE 2-LI:-46---5, PEACTCR WANE?. LrIEL ODUCROL LOS, in AUO with individual ZSI:468 9) (10), ?RNT 2A(23) (20) SPEED CONTROL LOS in AuTO. LAST PAGE

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO G2.3.12 (IOCFR 55A3.4 SRO On!y) Tier# Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel Group # handling responsibilities, access to locked high-radiation areas, aligning filters, etc. KIA# G2.3.12 Importance Rating 3.7 I Proposed Question: # 98 Which ONE of the following completes the statements below in accordance with 1-GOI-200-2, Primary Containment Initial Entry and Closeout? Initial Drywell Entry with the Reactor at Power must be approved by the _(1)_. A member of _(2)_ will remain at the Personnel Airlock in continuous communication with the Control Room AND with the persons in the Drywell. A. (1) Shift Manager ONLY (2) Rad Protection B. (1) Shift Manager AND Plant Manager (2) Rad Protection C. (1) Shift Manager ONLY (2) Operations D. (1) Shift Manager AND Plant Manager (2) Operations Proposed Answer: D Explanation A INCORRECT: Part 1 incorrect This is plausible in that if the entry is made (Optional): with the Reactor Mode switch is in SHUTDOWN or REFUEL position, Plant Manager authorization is not required and this would be the correct answer. Part 2 incorrect Plausible in that Rad Protection has several responsibilities and communications requirements associated with Drywell Entry in accordance with 1-GOl-200-2. B INCORRECT: Part 1 correct See Explanation D. Part 2 incorrect See Explanation A. C INCORRECT: Part I incorrect See Explanation A. Part 2 correct See Explanation D. D CORRECT: Part 1 correct Initial entries are permitted only when the Reactor Mode switch is in SHUTDOWN, REFUEL, or STARTUP/HOT STANDBY position, unless drywell entry at power has been authorized by the Plant Manager. Shift Manager approval is required for all initial entries. Part 2 correct In accordance with 1-GO 1-200-2, if Primary Containment is required, a member of Operations will remain at the Personnel Airlock during drywell entry. This person will be in continuous communication with the Control Room and with the persons in the Drywell.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification: The KA is met because the question tests knowledge of radiological safety principles pertaining to licensed operator duties associated with containment entry requirements SRO Only Justification: This question is SRO Only because it meets the requirements of Clarification Guidance for SRO, Section ll.D Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. [10 CFR 55.43(b)(4)] The question tests knowledge of radiological safety requirements associated with Drywell Entry with the reactor at power. Question Cognitive Level: Question rated as Fundamental Knowledge. Technical Reference(s): 1 -GOl-200-2 Rev. 11 (Attach if not previously provided) Proposed references to be provided to applicants during examination: NONE Learning Objective: (As available) Question Source: Bank # Mod ified Bank # BFN 1006 #98 (Note changes or attach parent) New:  : Question History: Last NRC Exam Browns Ferry 1006 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of evety question.) Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 x Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Clarification Guidance for SRO-only Questions Rev 1(03111/2010> C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. 110 CFR 55.43(by3)] Some examples of SRO exam items for this topic include:

• 10 CFR 50,59 screening and evaluation processes.
• Administrative processes for temporary modifications.
• Administrative processes for disabling annunciators.
• Administrative processes for the installation of temporary instrumentation.
• Processes for changing the plant or plant procedures.

Section IV provides an example of a satisfactory SRC-only question related to this topic.

      ) D. Radiation hazards that may arise durino normal and abnormal situations.

including maintenance activities and various contamination conditions. [10 CFR 5543(b)(4)J Some examples of SRO exam items for this topic include:

• Process for gaseous/liquid release approvals, i.e., release permits.
• Analysis and interpretation of radiation and activity readings as they pertain to selection of administrattve. normal, abnormal, and emergency procedures.
• Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.

SRO-on[y knowledge should not be claimed for questions that can he answered solely based on RO knowledge of radiological safety principles: e.g., RWP requirements, stay-time, DAC-hours. etc. E. Assessment of facility conditions and selection of appropriate procedures during normal. abnormal. and emergency situations. [10 CER 5543(b)(5)} This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal. abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose. The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item. for example: Page 6 of 16

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Primary Containment Initial Entry and 1-GOI-200-2 Unit I Closeout Rev. 0011 Page 8 of 94 3.0 PRECAUTIONS AND LIMITATIONS 3.1 General A. This procedure contains numerous conditional action steps. If the condition part of the step does not exist, is not or can not be met, then the operator should mark the step N/A and continue to the next step in the procedure. B Drywell access will be in accordance with this procedure and MCI-O-064-HLTOO1. C. All personnel are required to follow all RWP instructions and ALARA guidelines during Drywell entry and during the closeout walkdowns. D Nuclear Power Safety anti Health Manual should be followed to control heat stress if diywell temperature is elevated. E. All M&TE used in the drywell, with the exception of vendor supplied M&TE, should receive a post-use check when possible. F. No Containment entry is permitted \\qthout special breathing equipment unless a natural air atmosphere has been established (oxygen greater than or equal to 19.5%), as verified by Chemistry obtaining a grab sample lAW Cl-403. G When personnel are present in the Drywell, Radiation Protection should be notified of any change in Drjwell configuration :NER; and prior to any power or control rod pattern changes mmc IE 03-039. F-h Permitting access to the Drywell for leak inspections during a startup is judged prudent in teniis of the added plant safety offered without significantly reducing the margin of safety Thus, to preclude the possibility of starting the Reactor and operating for extended periods with significant leaks in the Primary System. leak inspections are scheduled during startup periods, when the Primary System is at or near rated operating temperature and pressure. These entries require Plant Manager permission. Initial Primary Containment Entry A. Shift Manager approval is required for all initial entries. r B. At least one of the airlock doors is required to be closed whenever Primary Containment is required C. If Primary Containment is required, a member of Operations will remain at the Personnel Aiiiock during drywell entry. This person will he in continuous communication with the Control Room and with the persons in the Drywell.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Primary Containment Initial Entry and 1-GOl-200-2 Unit I Closeout Rev. 0011 Page 9 of 94 3.2 Initial Primary Containment Entry (continued)

0. Initial entries will be preplanned with the best available data, including dose rates external to the inner airlock door. remote radiation monitor inside of containment and primary containment CAM readings to ensure the safety of the entry team. Any indication of unusual conditions will be further investigated prior to initial entry.

E. Initial entries are permitted only when the Reactor Mode switch is in SHUTDOWN, REFU EL, or STARTUP/HOT STANDBY position, unless drywell entry at power has been authorized by the Plant Manager. F. Entry team personnel are required to meet the requirements set forth in RCI-i 1.1 for SCBA use, when required. Detennination for usage of SCBA will be made by the Shift Manager, Radiation Protection Manager and Safety Manager.

0. Both the Suppression Chamber AND the Drywell are required to be purged UNTIL the 02 concentration is greater than or equal to 19.5% as verified by Chemistry cibtaining a grab sample lAW Cl-403. prior to entry into the Drqwell OR Suppression Chamber, UNLESS otherwise appnved by the Plant Manager or authorized representative.

H. SCBAs will be required until O concentrations are known to be greater than or equal to 19.5 percent as verified by Chemistry obtaining a grab sample lAW C 1-403. Emergency life support apparatus will be worn by all personnel for initial entry when SCBA is NOT used. I. SCBA bottles to he taken into the airlock are required to he secured in a safe manner. 3.3 Torus Entry A. Life preservers and pikes should be located at four locations in the Torus. B. Signs designating the locations of the preservers and pikes should be properly located in the torus. C. Life preservers are required for work or inspections off the catwalk. D. Employees working over water in the torus are required to wear life vest or a safety harness with the lifeline secured properly. This means work outside of the catwalk guardrails. E. A sign requiring use of safety harnesses or life vests when performing work off of the catwalk should be posted at the torus access on Elevation 565.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BROWNS FERRY 1006 #98 Which ONE of the following completes the statements in accordance with 1 -GOl-2OO2, Primary Containment Initial Entry and Closeout, AND RCI-17, Control of High Radiation Areas and Very High Radiation Areas? Initial Drywell Entry with the Reactor at Power must be approved by the _(1)_. The (2)_ that ALL keys are accounted for. A. (1) Shift Manager ONLY (2) Shift Manager AND Rad Protection Shift Supervisor shall verify DAILY B. (1) Shift Manager AND Plant Manager (2) Shift Manager OR designee shall verify SHIFTLY C. (1) Shift Manager ONLY (2) Shift Manager OR designee shall verify SHIFTLY D. (1) Shift Manager AND Plant Manager (2) Shift Manager AND Rad Protection Shift Supervisor shall verify DAILY ANSWER: B

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level G2.3.7 (IOCFR 55.43.415 SRO Only) Tier # Ability to comply with radiation work permit requirements Grou p # during normal or abnormal conditions. KIA# G2.3.7 Importance Rating 36 Proposed Question: # 99 In accordance with RCDP-3, Administration of Radiation Work Permits, for normal and emergency situations, which ONE of the following completes the statements? During NORMAL situations, RADPRO Supervision _(1)_ authorize short term deviation from RWP requirements (for example, verbally requiring additional protective clothing), without revising the RWP. If the Shift Manager authorizes IMMEDIATE entry into a High Radiation Area during emergency situations, then RADPRO escort _(2). A. (1)may (2) is required B. (1) may NOT (2) is required C. (1) may NOT (2) is NOT required D. (1)may (2) is NOT required Proposed Answer: A Explanation A CORRECT: Part 1 = correct Per RCDP-3, Administration of Radiation (Optional): Work Permits, RADCON Supervision may authorize short term deviations (excluding regulatory and procedural deviations) from RWP requirements without revising the RWP. Part 2 = correct Per RCDP-3, Administration of Radiation Work Permits, in emergency situations where the Shift Manager authorizes immediate entry to an area, RADPRO is required to escort. B INCORRECT: Part 1 = incorrect but plausible in that the candidate may assume that ALL RWP requirements need to be written within the RWP. Part 2 = correct for reasons detailed in A. C INCORRECT: Part I = incorrect, as detailed in A. Part 2 = incorrect for reasons detailed in A and plausible in that the candidate may assume that since approval has been granted, only normal dosimetry is required w/o the need of an escort.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet D INCORRECT: Part 1 = correct Per RCDP-3, Administration of Radiation Work Permits, RADCON Supervision may authorize short term deviations (excluding regulatory and procedural deviations) from RWP requirements without revising the RWP. Part 2 = incorrect for reasons detailed in A and plausible in that the candidate may assume that since approval has been granted, only normal dosimetry is required wlo the need of an escort. KA Justification: The KA is met because the question tests the ability to comply with radiation work permit requirements during normal or abnormal conditions. SRO Only Justification: This question is SRO Only because it meets the requirements of Clarification Guidance for SRO, Section ll.D Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. [10 CFR 55.43(b)(4)] The question involves RWP requirements associated with radiation hazards. Question Cognitive Level: Question rated as Fundamental Knowledge. Technical Reference(s): RCDP-3 Rev 2 (Attach if not previously provided) (Including version I revision number) Proposed references to be provided to applicants during examination: NONE Learning Objective: (As available) Question Source: Bank # BFN 0801 #99 Modified Bank # (Note changes or attach parent)

                                               /   New Question History:                   Last NRC Exam               Browns Ferry 09 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 X Comments:

ES-401 Sample Written Examination Form ES-401 -5 Question Worksheet Clarification Guidance for SRO.only Questions Rev 1(0311112010) C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)J Some examples of SRO exam items for this topic include:

• 10 CFR 50.59 screening and evaluation processes.
• Administrative processes for temporary modifications.
• Administrative processes for disabling annunciators.
• Administrative processes for the installation of temporary instrumentation.
• Processes for changing the plant or plant procedures.

Section IV provides an example of a satisfactory SRO-only question related to this topic. D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. [10 CFR 5543(b)(4)} Some examples of SRO exam items for this topic include:

• Process for gaseous/liquid release approvals, i.e., release permits.
• Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.
• Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.

SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g.. RWP requirements, stay-time. DAC-hours, etc. E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal. and emergency situations. [10 CFR 5543(by5)J This 10 C.FR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose. The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item. for example: Page 6 of 16

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet WAN STANDARD RCDP-3 DEPARTMENT ADMINISTRATION OP RADIATION WORK PERMiTS Rev. 2 PROCEDURE Page 6 of 11 3.6.5 RWPs describe the minimum requirements for Per orniing radiological work. RADCON job covera-ge personnel or supervision may verbally require additionai protective requirements for certain aspects of a work activity without revising the RWP. RADCON supervision may also authorize short-term deviations (exciudincj regulatory and procedural deviahons) from RWP requirements without revising the RWP. Any deviations shall be documented in the RADCON Corn puter System R.WP logbook. TVAN STANDARD RCDP-3 DEPARTMENT ADMINISTRATION OF RADIATION WORK PERMITS Rev. 2 PROCEDURE Page 7 of 11 3.6.7 The use of the RADCON Computer System to log RWP entries and exits may be suspended during emergency conditions. In emergency situations where the b Shift Manager authorizes immediate entry to an area, the prior approval requirements of a RWP will be waived. If the RWP approval requirement is waived. RADCON and the personnel escorted by RADCON must comply with radiation protection procedures for entry into high radiation areas (La., RA.DCON individual is equipped with radiation dose rate monitoring dev;ce and provides posihve control over activities within the area to include protective recommendations for the personnel being escorted for the duration of the emergency). Radiation surveillance by virtue of RADCON escort s considered to be continuous coverage. The RWP must be completed when the emergency entry is completed or the emergency is over. At the completion of the exempt work, actions will be taken to document (in the RADCON Computer System) the work, entries: exits: dose accrued, etc. Per WBN Tech specs and FSAR, this step does not apply to WBN.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO G2.4.22 (IOCFR 55.43.5) SRO ONLY Tier # ------ 3 Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations. Group # KIA# G2.4.22 Importance Rating 4.4 Proposed Question: # 100 With an ATWS, Emergency Operating Instructions (EOls) require operators to reduce Recirc Pump speeds to minimum prior to tripping them if Reactor Power is above 5%. Which ONE of the following identifies the (1) bases for this action AND (2) the EQI leg which requires it? A. (1) To allow time for ARI to actuate thus allowing the Recirc Pumps to stay in operation for coolant circulation. (2) C-5, Level I Power Control B. (1) To allow time for ARI to actuate thus allowing the Recirc Pumps to stay in operation for coolant circulation. (2) EOI-l, RPV Control, RC/Q leg C. (1) To prevent tripping the turbine on high water level AND exceeding the capacity of the bypass valves. (2) C-5, Level I Power Control D. (1) To prevent tripping the turbine on high water level AND exceeding the capacity of the bypass valves. (2) EQ I-I, RPV Control, RCIQ leg Proposed Answer: D Explanation A INCORRECT: Part 1 incorrect Plausible in that ARI is designed to dump (Optional): air to HCU banks and SDV to atmosphere, ensuring rod insertion begins within 15 seconds and completes within 25 seconds. Therefore, the delay would provide time for ARI to complete the Scram, lowering power to less than 5% would possibly prevent need to trip Recirc Pumps. However, this is not the EOI Bases for this action. Part 2 incorrect Plausible in that EOl 1 RC/L leg is exited and C-5 is entered with an ATWS and Reactor Power> 5%. However, the requirement to reduce Recirc to minimum prior to tripping is addressed in EOl-1 RC/Q leg. B INCORRECT: Part 1 incorrect See Explanation A. Part 2 correct See Explanation D. C INCORRECT: Part 1 correct See Explanation D. Part 2 incorrect See Explanation A.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet D CORRECT: Part 1 correct a recirculation flow runback is performed prior to tripping recirculation pumps in order to effect a more controlled reduction in reactor power. Even though the quickest reactor power reduction is achieved by tripping recirculation pumps, if a recirculation pump trip is initiated from a high reactor power level, the resulting plant transient may cause a main turbine trip due to rapid changes in steam flow, RPV pressure, and RPV water level. If reactor power is above turbine bypass valve capacity and the main turbine trips, RPV pressure will increase until one or more MSRVs open. Heatup of the suppression pool then begins. KA Justification: The KA is met because the question test knowledge of bases for prioritizing safety functions, i.e. Reactivity Control / CTMT Control with an ATWS condition present. The K/A requests knowledge of the bases for prioritizing safety functions during EOP operations and the question asks for the bases and emergency procedure. SRO Only Justification: This question is SRO Only because it meets the requirements of Clarification Guidance for SRO, Section ll.E Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)] See Attached. Question involves knowledge of decision points in the EOls that involve transitions to event specific contingency procedures. Question Cognitive Level: Question rated as Fundamental Knowledge. Technical Reference(s): 1 -EOl-1, Rev 0 I EOIPM 0-V-C Rev 1 (Attach if not previously provided) OPL171.204 Rev7 Proposed references to be provided to applicants during examination: NONE Learning Objective: (As available) Question Source: Bank# Modified Bank # BEN 04 #98 (Note changes or attach parent) New Question History: Last NRC Exam Browns Eerry 2004 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eveiy question.) Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 X Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Clarification Guidance for SRO-only Questions RevI (03/1112010) Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5) (Assessment and selection of procedures) Can the question be answered solely by knowing systerns knowledge. Le., how the system works, flovath, logic, component location? Oqeslion No Can the question be answered solely by knowing 1 immediate operator actions? j Yes L,. RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? No Can the question be answered solely by knong the purpose, overall sequence of events, or overall mitigative strategy of a procedure? No Does the question require one or more of the following?

• Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps SRO onl Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
• Knowledge of administrative procedures that specify hierarchy, implementation. and/or coordination of plant normal, abnormal, and emergency procedures No j

I Question might not be linked to 10 CFR 55,43(b)(5) for SRO-onlv PageS of 16

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet L L L EXZCUTE  :-i AN) RCQ CON RNCNLY L 1EOl-1 PAGE 1 OF I RPV CONTROL UNIT 1 BROWNS FERRY NUCLEAR PLANT REV: 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet EO1-l, RPV CONTROL BASES EOI PROGRAM MANUAL SECTION 0-V-C DISCUSSION: STEP RC/Q-7 This action step directs the operator to reduce reactor power by manually running back recirculation pump flow to minimum, if an automatic runback has occurred, the operator need only confirm the action. This step is reached only if the main turbine-generator is still synchronized. Therefore, to avoid a main turbine trip and its associated complications, a recirculation flow runback is performed prior to tripping recirculation pumps in order to effect a more controlled reduction in reactor power. Even though the quickest reactor power reduction is achieved by tripping recirculation pumps, if a recirculation pump trip is initiated from a high reactor power level, the resulting plant transient may cause a main turbine trip due to rapid changes in steam flow, RPV pressure, and RPV water level. If reactor power is above turbine bypass valve capacity and the main turbine trips, RPV pressure will increase until one or more MSRVs open. Heatup of the suppression pool then begins and, if not adequately controlled, boron injection may ultimately be required. REVISION 1 PAGE 109 OF 127 SECTION 0-V-C

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT OPL171 .0 05 Revision 17 Page 33 of 79 INSTRUCTOR NOTES (2) This assumes an ATWS occurs with or without RPS de-energization. ATWS!AR I (3) The associated valves are controlled implementation was through ATU (ECCS) powered logic, required by 10 CFR independent of the RPS System. 50.62 and DCR 3125, Rev. I (a) AR! Valve FSV-85-730 - 25OVDC, Pnl A 25/4 18 Obj. \f.B.30.e (b) AR! Valve FSV-85-731 - Obj. V.E.29.e 25OVDC. Pnl A 25/419 (4) Two 3-way AR! scram valves were added to the HCU air supply These are actually header. to both iso[ate the header one pair of the old and also vent the scram air header. back-up scram (5) Six 2-way air vent valves were added valves that were re to the HCU banks and SDV air supply powered from the for fast dumping of air to atmosphere, ECCS ATU cabinet ensuring rod insertion begins within 15 seconds and completes within 25 seconds. (6) A coincident trip of either 2 lo.*v levels or 2 high pressures in the same trip channel causes ATWSIARI/RPT; actual trip values are 1148 psig or level 2. <45 RWL. Both channels must trip to trip both recirc pumps and realign all the AR! valves. (7) Manual ARI pushbuttons A/B on pnl. 9-5 are supplied to actuate either trip channel, These PB will only initiate the ARI system. The RPT wont manually initiate. ATWS/ARI Logic Manual Initiation

                                               -                               Obj. V.B.31 (1)   Turn collar on HS-68-1 1 9A to arm           Ob). V.0.16 ARI.                                         TP-5 (2)   RLY-68-1 1 SAl energizes and the             ARI cannot be white ARMED light on pnl. 9-5              manually initiated energizes and RPT is blocked.                unless armed (3)   Depress HS-68-1 iSA to initiate OH. A ARI.

(4) RLY 1 19A2 energizes. (a) Contact T2-M2 closes. TP-6

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT RC)L-2 WHILE EXECUTING THE FOLLOWING STEPS: IF THEN 1 HAS NOT BEEN DLrERMINEDTHAT THE REOR WILL REMAIN SUBCRITICAL EXIT RC/L AND WL1HOu BORON UNDER ALL CONDITIONS ENTER CS. LEVEL/POWER CONTROL (SEE NOl B) EXIT RC/L AND RPV WATER LVL CANNOT BE DETERMINED ENTER C4. RPV FLOODING PC WAIER LVL CANNOT BE MAIN tANBD DELOW 105 FT STOr INJ INTO THE RPV FROM SOURCES OR EXTERNAL TO THE PC NOT REQUIRED FOR ADEQUATE CORE COOLING. SUPPR CHMBR PRESS CANNO1 BE MAIN LdNED BELOW 55 PSIG L RC/L-3 fr 7 1EOl1 PAGE 1 OF I R.PV COTROL UNIT 1 BROWNS FERRY CLEAR PLANT REV: 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTIONS REPORT for BF Initial Exam 98, 024.22 001 EOI1, RPV Control, RC!Q leg dfrects the operators to reduce Recc Pump speeds to minimum prior to tripping them if Rx Power is above 5%. Which ONE of the following is the bases for this action? A. To minimize power oscillations that may result from tripping Recirc Pumps at higher speeds. B To prevent tripping the turbine on high water level and exceeding the capacity of the bypass valves. C. To allow time for ARI to actuate thus allowing the Reciro Pumps to stay in for coolant circulation. operation I

0. To prevent RPV level from reaching + 2 inches as a result of tripping Recirc Pumps at higher speeds and initiating PCIS.

K/A G2.4.22 Knowledge of the hoses for prioritizing safety functions during abnormal/emergency operations. (3.0/4.0)

References:

OPLI71.102, Rev.6, Pg 61 and $2 of 67 A. Incorrect since a rapid power reduction is required at this time. B. Correct answer. C. Incorrect since every means possible is used to reduce reactor power regardless of how long it takes ARI to actuate. D. incorrect since tripping the recirc pumps causes swell, not shrink, Thursday, Apili 08, 2004 2:08:54 PM 102}}