ML112790089

Initial Exam 2011-302 Draft SRO Written Exam

ML112790089
Person / Time
Site:	Browns Ferry
Issue date:	10/04/2011
From:	NRC/RGN-II
To:	Tennessee Valley Authority
References
50-259/11-302, 50-260/11-302, 50-296/11-302
Download: ML112790089 (46)

Text

76. 295004AA2.02 OOi/1/1/SROfNW/HJ3fBLCfMAB Unit 1 and 2 are operating at 100% power. Unit 3 is in Mode 5.

The supply breaker to 250V DC RMOV Bd 2A tripped open and cannot be reset. The crew transferred this board to its alternate supply in accordance with 0-Ol-57D, DC Electrical System, Section 8.9, Transfer of Power Supplies to 250V Reactor MOV Bds.

Which ONE of the following completes both statements regarding the 250V DC RMOV Bd 2A being powered from its alternate supply?

The Division I and Division II DC logic for ECCS (ADS, H PCI, Core Spray, and LPCI) being fed from the same battery board.

In accordance with Tech Spec Bases 3.8.7, Distribution Systems

- Operating, the 2A 250V DC RMOV Board operable.

A. are remains B. are NOT remains C are is NOT D. are NOT is NOT

Answer is C.

295004 AA2.02: Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF DC POWER: Extent of partial or complete loss of DC power SRO IR: 3.9 (CFR: 41.10 I 43.5/ 45/13)

Plausibility: 1st part of B and D (DC power supply is different between Div I and II) is plausible if the applicant knows that the 250V RMOV Bd alternate power supply is from Unit 3; however, BB3 is the normal power supply to 250V DC RMOV Bd 2B, which feeds the Div I ECCS logic on Unit 2. 2nd part of A and B is plausible because in Mode 5 the RMOV board is considered operable. (The stem includes one unit in Mode 5.)

SRO only: 10CFR55.43(b)(2): Facility operating limitations in the technical specifications and their bases. 0-Ol-57D, P&L L has the answer to the second fill-in-the-blank item; however, operability determinations are typically SRO level knowledge. Alternatively, can test the SRO applicants knowledge of whether the alternate battery remains operable even though it has been loaded with 250V DC RMOV Bd 2A. (will require applicants to use drawings).

Tech Spec Bases 3.8: The Unit 2 250 V DC RMOV boards 2A, 2B, and 20 have alternate power supplies from another 250 V Unit DC board.

In Modes 1, 2, or 3, the boards are considered inoperable when powered from their alternate feeder breakers because a single failure of the power source could affect both divisions depending on the board alignment. However, if the Unit was in Mode 5, the board is still considered operable.

Lesson Plan OPLI 71.037 Obj 11: Given a hypothetical situation involving loss of a DC power supply, identify the appropriate corrective action in accordance with Ol-57D.

Obj.12: Determine the 250V DC System components that have a Technical Specification Limiting Conditions for Operation (LCO) associated with it.

Obj. 13: Given an interrelated system, determine its interrelationship including how a loss of that system would affect D.C. system

77. 29SQO6G2.441 OOIJ-1/1/SROft4EW/L13!REC/MAW The unit was operating at 100% power when a failure of an automatic and manual scram occurred.

Alternate Rod Insertion (ARI) was unsuccessful and reactor power is currently 8%.

The Shift Manager has declared a Site Area Emergency.

Which ONE of the following identifies an additional threshold criteria listed in the Scram Failure column of EPIP-1, Emergency Classification Procedure Event Classification Matrix, that requires upgrading to a General Emergency declaration?

Ab Heat Capacity Temperature Limit Curve is exceeded.

B. SLC not injected to tank level at or below 43%.

C. Reactor water level intentionally lowered below TAF.

D. SLC not injected to tank level at or below 67%.

No references provided to applicants.

Answer is A.

295006 (SCRAM) G2.4.41 Knowledge of the emergency action level thresholds and classifications. SRO IR: 4.6 (CFR: 41.10/43.5/45.11)

Plausibility: B and D are plausible because these are key levels in the EOl-1 flowchart associated with the SLC cold and hot shutdown boron weight. C is plausible because it is a contingency action in C-5 and closely mirrors the 2nd condition in the Scram Failure column of EPIP-1 that also requires a General Emergency, i.e.,

level cannot be restored and maintained greater than -180.

SRO-only: 10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. The SRO knowledge associated with the E-plan classifications is SRO level knowledge.

EPIP-1, SCRAM FAILURE Column, General Emergency (1.2-G)

Failure of automatic scram, manual scram, and ARI. Reactor power is above 3%

AND Either of the following conditions exists:

Suppression Pool temp exceeds HCTL. Refer to Curve 1 2-G Reactor water level can NOT be restored and maintained at or above -180 inches.

SRO-only:

1 OCFR55.43(b)(5): Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

OPL1 71.075 HLT Objective 2: Given information concerning different types of events, use EPIP-1, Emergency Classification Matrix to determine the correct classification.

78l1aiu3 O1J1ILISRO/NEW/H.12/BLCJMAB The Site Emergency Director (SED) is choosing between Protective Action Recommendation (PAR) #1 and PAR #2, while performing EPIP-5, General Emergency.

Which ONE of the following completes the following statement in accordance with Appendix H, Protective Action Recommendation Flowchart?

In order to determine the PAR, are allowed to be used.

A. only the actual radiation levels at the site boundary (projected doses not allowed)

B. actual radiation levels and projected doses at the site boundary C. only the actual radiation levels at 5 miles or further from the site (projected doses not allowed)

D actual radiation levels and projected doses at 5 miles or further from the site Answer is D.

295038 (High Offsite Release Rate) EA2.03 Ability to determine and/or interpret the following as they apply to HIGH OFFSITE RELEASE RATE: Radiation Levels (CFR41.1O/43.5/45/13) SRO lR: 4.3 Plausibility:

Choice A is plausible because the GE classification criteria utilizes site boundary doses. Not using projected doses during the PAR to the state is plausible because of the artificiality and/or uncertainty associated with computer model predictions.

Choice B is plausible because the GE classification criteria utilizes actual and predicted doses at the site boundary.

Choice C is plausible because of the artificiality and/or uncertainty associated with computer model predictions.

SRO-Only: 10CFR55.43 (b)(1) Conditions and limitations in the facility license and 55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. The SED is an SRO function.

ROs do not make PARs.

References:

EPIP-5, General Emergency, Appendix H, PAR Flowchart Lesson Plan OPLI 71.075 Objective #9: Given that a General Emergency has been declared, determine the appropriate Protective Action Recommendation when given necessary plant data.

79 29502tG243U Q01/1/IJSRO/NEW/H13/BLCJM&R Unit 2 was in Mode 4 with Loop 2 of RHR in Shutdown Cooling with the following parameters:

RPV pressure: 0 psig RPV level: 74 No Recirc Pumps are running RHR B Hx outlet temperature: 135.6 °F RHR B flow rate: 8500 gpm RPV feedwater nozzle temperatures: 150 °F RPV drain to RWCU temperature: 146 °F Subsequently, the 2A RPS MG set tripped. Ten minutes ago, the control room dispatched personnel to transfer RPS to its alternate power supply. The following parameters currently exist:

RPV pressure 0 psig; RPV level 77 RHR SD CLG FLOW LOW (9-3D, Wil) alarming RPV feedwater nozzle temperatures: 158 °F RPV drain to RWCU temperature: 145 °F Which ONE of the following identifies the earliest required NRC notification(s) in accordance with NPG-SPP-03.5, Regulatory Reporting Requirements?

[Reference provided A NRC Operations Center is NOT required to be notified within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 60 day written LER is required B.

NRC Operations Center is required to be notified within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 60 day written LER is NOT required C. NRC Operations Center is required to be notified within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 60 day written LER is required D. NRC Operations Center is NOT required to be notified within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 60 day written LER is NOT required

Answer is A Provide applicants with 1) NPG-SPP-0.3.5 (entire procedure); 2) NUREG 1022 (entire procedure); 3) EPIP-1 only Section 1.5 (Loss of Decay Heat Removal) page 23 of 206.

295021 (Loss of Shutdown Cooling) G24.30: Knowledge of events related to system operationlstatus that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator. (CFR 41.10/43.5 / 45.11) SRO IR: 4.1 Plausibility: For 1st part, B and C are plausible if the applicant declares an Alert based on the loss of shutdown cooling. (Reactor moderator temp cant be maintained below 212 deg F). For the 2nd part, B and Dare plausible if the applicant determines that the shutdown cooling isolation (as well as all the other isolations) were not a valid actuation signal (due to the MG set tripping vs. actual low water level, etc.)

SRO-only:

1 OCFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NPG-SPP-03.5, Regulatory Reporting Requirements, Appendix A, Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants NUREG 1022, Event Reporting Guidelines 10CFR5O.72 and 50.73, Section 3.2.6, System Actuation Lesson Plan OPLI 71.092 Objective 3: Given a hypothetical event, determine if the event is reportable using SPP-3.5 and NUREG 1022.

Objective 4: Given a hypothetical event, determine if a one, four or eight hour report to the NRC is required using SPP-3.5 and NUREG 1022.

80. 295QaoG2Aa 001!IJ1/SROINEW/HL2JBLC/MAR The Unit Supervisor (US) is implementing 3-EOl-2, Primary Containment Control, due to a lowering torus level following an earthquake. The US has reached the following step:

Which ONE of the following identifies the value of the water level listed in this decision step and the required procedure if the answer to this question is NO?

A. 11.5ft 3-EOI-Appendix 18, Suppression Pool Water Inventory Removal and Makeup B 11.5 ft 3-C-2, Emergency Depressurization C. 12.75 ft 3-EOI-Appendix 18, Suppression Pool Water Inventory Removal and Makeup D. 12.75 ft 3-C-2, Emergency Depressurization Answer is B 295030 (Low Suppression Pool Level) G2.4.6 Knowledge of EOP mitigation strategies (CFR 41.10 /43.5 I 45.13) SRO IR: 4.7 Plausibility: The 1st part of C and D is plausible because 12.75 ft is the torus level value (listed in the SPIL leg as as an override) at which HPCI must be secured. The 2nd part of A and C is plausible because this procedure is required to be implemented in the SP/L leg. The reason that A is incorrect is that the stem asks which procedure is required to be implemented if the answer to the question is NO.

SRO-only: IOCFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NOL SP?L-6 YES L

-Qi-2 Prirnry Coninment Control Lesson Plan OPLI 71.203 Objective 7: Explain the concern relative to the following suppression pooil containment water levels during performance of EOI-2:

a.

11.50feet b.

12.75 feet c.

18 feet d.

105 feet

8t 6flUQOOG2.4A1 JULIS.ROLNEW/LJ2IBLC/MAB Which ONE of the following choices completes both statements, as they relate to an emergency classification due to a fire on site, in accordance with EPIP-1, Emergency Classification Procedure Technical Basis?

If the confirmation of a fire cannot be positively ascertained within

, and symptoms indicative of a fire persist, then confirmation should be assumed.

The escalation from an Unusual Event to an Alert is based on the fire A.

15 minutes involving radiologically contaminated areas B. 20 minutes involving radiologically contaminated areas C. 20 minutes affecting the operability of plant safety systems D 15 minutes affecting the operability of plant safety systems Answer is D No references provided to applicants.

600000 (Plant Fire On Site) G2.4.41 Knowledge of the emergency action level thresholds and classifications. (CFR: 41.10 /43.5 / 45.11) SRO IR 4.6 Plausibility analysis: The 1st part of B and C are plausible because this time value is used in the EPIP-1 classification scheme (Alert declaration following control room evacuation and control not established at the Shutdown Control Panel 25-32). The 2nd part of A and C are plausible because of other emergency classifications involving radiological concerns.

SRO-only: 10CFR55.43(b)(1) Conditions and limitations in the facility license.

Lesson Plan OPL171.075 Objective 2: Given information concerning different types of events, use EPIP-1, Emergency Classification Matrix to determine the correct classification.

EPIP-1, Emergency Classification Procedure Technical Basis, 6.4-UI

81 7QÜ000AO5OOIflJiJSROLNEW/HJ3IBLCiMAB Unit 1, 2, and 3 are operating at 100% power.

4KV Common Board B was transferred to Start Bus I B in accordance with 0-OI-57A, Section 8.23, Control Room Transfer of 4kV COM BD B Power Supplies.

The Trinity 2 (Trico) 500 KV line is out of service for maintenance.

The Transmission Operator subsequently informs the Unit Supervisor that the Athens 161 Ky line is in an unanalyzed condition (Red) due to the total system load exceeding the maximum analyzed system load for the current transmission system conditions.

Which ONE of the following identifies the Unit 3 Tech Spec 3.8.1 actions that are required (if any) for the conditions listed above?

[REFERENCE PROVIDED]

A No required actions V

B. Perform SR-3.8.1.a.1 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and restore the Athens 161KV line to operable status within 7 days.

C. Perform SR-3.8.1.a.1 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and restore EITHER the Athens 161KV line OR the Trinity 2 (Trico) 500 KV line to operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D. Enter Tech Spec 3.0.3 Answer is A.

Provide applicants with Unit 3 Tech Spec 3.8.1 (no bases) 700000 (Generator Voltage and Electric Grid Disturbances) AA2.05: Ability to determine and/or interpret the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: Operational Status of Offsite Circuits.

(CFR: 41.5 and 43.5 / 45.5, 45.7, and 45.8) SRO IR: 3.8 Plausibility: B is plausible if the applicant doesnt understand that the LCO only REQUIRES 2 (of the 3) offsite circuits; therefore, the LCO is still met. C is plausible if the applicant assumes that only one offsite circuit remains operable. D is plausiblie if the applicant assumes that none of the action statements listed in the LCO apply to the condition as a result of degraded grid conditions.

SRO-only: 10CFR55.43(b)(2) Facility operating limitations in the technical specifications and their bases.

TRQ-TQ-SQP-3Q.i28., Browns Feny NucIar Plant (BFN) GrkI QpaingQukie Tech Spec 3.8.1, AC Sources

- Operating PIP-02-03a, Electrical Distribution System Lesson Plan OPL17I.036 Objective 13: Given plant and electrical system status determine the appropriate actions to be performed as stated in Technical Specifications, Cls, ARP, and AOls.

83. 2950119AA2.Q1 QQU1J21SRO!MQDLFIED/H/2JBLC/MAB Unit 3 was operating at 100% power when a loss of all feedwater occurred.

All control rods automatically inserted and reactor pressure is 900 psig and stable.

No source of high pressure feed water is available; however, low pressure pumps are running.

The Unit Supervisor is in EOl-1 assessing the following control room level/temperature INDICATIONS:

Ll-3-53, 3-60, 3:206, 3-253, and 3-208A, B, C, D:

downscale Ll-3-58A and 3-58B

-150 inches Ll-3-52 and Ll-3-62k

-180 inches Ll-3-55 downscale Elevation 621 74-95F 152 °F Elevation 593 74-95 C/D 90 °F Elevation 565 69-835A-D 90 °F Which ONE of the following identifies ACTUAL RPV Water Level and the required procedure?

[REFERENCE PROVIDED]

A. Actual RPV water level is above top of active fuel; Remain in EOl-1 RC/L B Actual RPV water level is above top of active fuel; Exit EOl-1 RC/L and enter C-I C. Actual RPV water level is below top of active fuel; Remain in EOI-I RC/L D. Actual RPV water level is below top of active fuel; Exit EOl-1 RC/L and enter C-I Answer is B Provide applicants with Caution 1 and Level correction PIP at Panel 9-3 295009 (Low Reactor Water Level) AA2.01 Ability to determine and/or interpret the following as they apply to LOW REACTOR WATER LEVEL: Reactor Water Level (CFR: 41.10/43.5/45.13) SRO IR: 4.2 Plausibility analysis: The 1st part of C and D is plausible because the post-accident range instruments are indicating less than -162 (if applicant forgets to compensate the indications). The 2nd part of A is plausible if the applicant incorrectly determines actual RPV water level. The 2nd part of C is plausible if the applicant SRO-only:

I OCFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Significantly modified from Operations Bank item OPL171.205 002 (Old TEGRS 10370) 3-EOI-1, Reactor Vessel Level Control 3-C-I, Alternate Level Control Lesson Plan OPL17I.003 Objective 19: Given a set of plant procedures and unit/plant conditions, correctly apply procedures related to Reactor Vessel Process Instrumentation.

4.oitu.is 00 L/1/2ISROfNEWfH/3JIMA Unit 3 is operating at 98% power.

During a control rod surveillance, the Unit Operator (UO) was required to single notch a control rod from position 14 to 16; however, the UO continuously withdrew the control rod until the following annunciators began alarming:

RBM HIGH/INOP (9-5, W24)

CONTROL ROD WITHDRAWAL BLOCK (9-5, W7)

The control rods final position is 26 and the Unit Operator (UO) observes the following values on Powerplex:

MFLCPR 0.925 MAPRAT 0.754 MFDLRX 1.20 MFLPD 0.00 Which ONE of the following indicates:

1) whether all Tech Spec 3.2, Power Distribution Limits, Limiting Conditions for Operation (LCO) are met and

2) the required classification for this event in accordance with NPG-SPP-1 0.4, Reactivity Management Program?

A. All Tech Spec 3.2 LCOs are met. (no required action statement)

Severe Reactivity Management Event (SL 1)

B. All Tech Spec 3.2 LCOs are NOT met.

Severe Reactivity Management Event (SL 1)

C. All Tech Spec 3.2 LCO5 are met. (no required action statement)

Major Reactivity Management Event (SL 2)

Dv All Tech Spec 3.2 LCOs are NOT met.

Major Reactivity Management Event (SL 2)

Answer: D No reference provided to applicants.

295014 (Inadvertent Reactivity Addition) G2.2.38 Knowledge of conditions and limitations in the facility license. (CFR 41.7/41.10 /45.13) SRO IR: 4.5 Plausibility: The 1st part of A and C are plausible if the applicant does not know that MFDLRX is the same as LHGR. The 2nd part of A and B are plausible because the severe classification examples (BWR) include rods (item 1-1) and Tech Specs (item 1-4)

SRO-only: IOCFR55.43(b)(1) Conditions and limitations in the facility license. Since the operating license requires procedures such as Conduct of Ops and/or Reactivity Management procedures, this meets item (b)(1). This question can also target item 10CFR55.43(b)(2) (Tech Specs) or IOCFR55.43(b)(5) (Procedure selection). The test item deals with the administrative functions associated with the reactivity event, which are SRO responsibilities in the plant.

3AQl-857, Mispositioned Control Rod Tech Specs 3.2.3, Linear Heat Generation Rate NPG-SPP-10.4, Reactivity Management Program Lesson Plan OPLI7I.087 Objective 8: When given situations correctly apply the following definitions:

a.

ACTIONS b.

OPERABLE c.

CORE ALTERATIONS d.

LEAKAGE Less Plan OPLI7I.074 Objective 2: Given specific plant conditions, determine the correct actions to take based on the appropriate AOl.

8& 29Sl32EA2.O2 OOi11/2SRO/NEW/H/3JBLC/MAB.

Unit 2 is operating at 100% power.

A large unisolable steam leak has occurred in the RWCU Heat Exchanger Room and its maximum normal operating temperature has been exceeded.

Which ONE of the following identifies:

1) a Panel 9-5 Normal Range level instrument indication that is affected as the RWCU Heat Exchanger Room temperature approaches the maximum safe operating temperature and

2) the procedure(s) required to be implemented before the room temperature exceeds the maximum safe operating temperature?

[REFERENCE PROVIDED]

A.

Ll-3-208B GOl-1 00-1 2A B. Ll-3-208B EOI-1 C. Ll-3-53 GO 1-100-1 2A D Ll-3-53 EOI-1 Answer is D Provide applicants with Caution I (including Curve 8 and Table 6); do NOT provide any EOI flowchart.

295032 (High Secondary Containment Temperature) EA202 Ability to determine andlor interpret the following as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE: Equipment operability (CFR: 41.10 /43.5/45.13) SRO IR: 3.5 Plausibility: The 1st part of A and B are plausible because this is a Panel 9-5 Normal Range level instrument; however, it is not affected by high RWCU room temperatures because there are no vertical instrument legs in this room. The 2nd part of A and C are plausible because this is the correct procedure if the high temperature condition was not caused by a primary system discharging into secondary containment.

SRO-only: IOCFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

EOl-3, Secondary Containment Control Lesson Plan OPLI71.003 Objective 9: Given a set of plant conditions, describe or identify the errors induced in the level instruments due to density effects, high drywell temperatures, recirculation pump operation, and non condensable gases in the reference legs. (SOER 82-02, Recommendation 7)

Lesson Plan OPL171.201 Objective 11: Given specified plant conditions relative to a Caution in the EOls, determine the basis for the Caution.

86. 211000A2.08 001J2!1!SRO!NEW/H!2/BLC/MAR.

An ATWS has occurred on Unit 1 and the Unit Supervisor (US) is implementing EQ I-i, RC/Q, and the crew has initiated SLC in accordance with 1-EOl-Appendix-3A.

RPV level has been lowered in accordance with C-5, Level/Power Control.

The US has subsequently determined that emergency depressurization is required due to high Secondary Containment temperatures.

Which ONE of the following identifies 1) the targeted level in the SLC Tank, which is listed in the RC/Q leg, and 2) describes the procedural implementation of this emergency depressurization?

Note: The title of C-2 is RPV Emergency Depressurization.

A. 43%

C-5 provides all the steps for emergency depressurization; C-2 is not required.

Bb 43%

0-2 is required to be implemented concurrently with the emergency depressurization guidance listed on C-5.

C. 67%

C-5 provides all the steps for emergency depressurization; C-2 is not required.

D. 67%

C-2 is required to be implemented concurrently with the emergency depressurization guidance listed on 0-5.

Answer is B (Ensure no overlap w/ 295006 G2.4.41) 211000 SLC A2.08 Ability to (a) predict the impacts of the following on the STANDBY LIQUID CONTROL SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Failure to SCRAM (CFR: 41.5 / 45.6) SRO IR: 4.2 Plausibility: The 1st part of !ICU and D are plausible because this is the hot shutdown boron weight and occurs before reaching 43%. Also, 67% is listed in the 0-5 level/power control flowchart. The 2nd part of A and C are plausible because 0-5 is normally the procedure used to control depressurization during ATWS conditions, which is different than emergency depressurizing during high Secondary Containment temperatures.

SRO-only: 100FR55.43(b)(5) Assessment of plant conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

1, Reactor Vessel Control Procedure Lesson OPL171.202 Objective 15: Given appropriate plant information, utilize RCIP, RC/L, and RC/Q concurrently to determine appropriate operator actions.

7.

Unit I is operating at 100% power.

A gross failure of the Reactor Feedwater Control system occurs and the following alarm is received:

RFWCS GROSS FAILURE 1-LA-46-5C (9-6C, W7)

The Unit Operator (US) reports that actual reactor water level is rising and that the Narrow Range level instruments are all unreliable.

Which ONE of the following describes:

1) the required actions, in accordance with the alarm response procedure, and

2) the safety analysis for Feedwater and Main Turbine High Water Level Trip Instrumentation, in accordance with Tech Spec Bases B3.3.2.2 and the FSAR?

A. Depress the RFPT Speed Control Raise/Lower Switches to the Manual Governor Position and attempt to control reactor water level.

The indirectly-initiated reactor scram mitigates potential turbine damage.

B. Depress the AUTO/MANUAL pushbutton on RX WATER LVL CONT, 1-LIC-46-5, (Master Controller) and control RFPTs in AUTO with RAISE/LOWER pushbuttons.

The indirectly-initiated reactor scram mitigates potential turbine damage.

C Depress the RFPT Speed Control Raise/Lower Switches to the Manual Governor Position and attempt to control reactor water level.

The indirectly-initiated reactor scram mitigates the reduction in MCPR.

D. Depress the AUTO/MANUAL pushbutton on RX WATER LVL CONT, 1-LIC-46-5, (Master Controller) and control RFPTs in AUTO with RAISE/LOWER pushbuttons.

The indirectly-initiated reactor scram mitigates the reduction in MCPR.

Answer is C 259002 Reactor Water Level, A2.06: Ability to (a) predict the impacts of the following on the REACTOR WATER LEVEL CONTROL SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of controller signal output (CFR: 41.5 I 45.6) SRO IR: 3.4 Plausibility: The 1st part of B and D are plausible because other FW control malfunctions would require placing the Master Controller in AUTO. The 2nd part of A and B is plausible because this is the actual reason for the TURBINE TRIP that is initiated from the Level 8 trip instruments; however, it is NOT how the reactor scram mitigates the transient.

SRO-only: IOCFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. The 1st part of the question is RO knowledge; however, the 2nd part of the question tests the SRO applicants knowledge of the bases for the Tech Spec 3.3.2.2 FW & MT Hi Water Level Trip Instrumentation.

1-ARP-9-6C, Window 7, RFWCS GROSS FAILURE 1-01-3, Reactor Feedwater System Tech Specs and Bases 3.3.2.2 The feedwater and main turbine high water level trip instrumentation is assumed to be capable of providing a turbine trip in the design basis transient analysis for a feedwater controller failure, maximum demand event (Ref. 1). The reactor vessel high water level trip indirectly initiates a reactor scram from the main turbine trip (above 30% RTP) and trips the feedwater pumps, thereby terminating the event. The reactor scram mitigates the reduction in MCPR.

FSAR Section 14.5.8.1 Feedwater Controller Failure Maximum Demand (FWCF)

An event which can cause directly an excess of coolant inventory is one in which makeup water flow is increased without changing other core parameters. The FWCF is the limiting event of the excess coolant inventory type. The FWCF to maximum demand is one of several potentially limiting events normally included in the cycle-specific reload licensing analyses to establish the MCPR operating limits.

8&2u6oouG2L7 Qfl/2J1JSROIMQDIFIEDIW31BLC/MAR The following plant conditions exist on Unit 2:

Unit 2 has suffered a spurious Group 1 Isolation and 20 control rods failed to insert.

HPCI is injecting to maintain RPV level.

Reactor Power is 1% and stable.

Reactor water level is 30 inches and steady.

The following alarm conditions currently exist:

SUPPR CHAMBER WATER LEVEL ABNORMAL (9-3B, W15) alarming SUPPR POOL LEVEL HIGH (9-3F, W12) alarming SUPPR POOL TEMP SINGLE ELEMENT HIGH OR FAILED (9-3B, W29)... alarming SUPPR POOL AVERAGE TEMP HIGH (9-3E, W 12) alarming Which ONE of the following identifies a concern and a required procedure for these plant conditions?

A. Operating HPCI and/or RCIC with suction from the suppression pool may exceed NPSH limits 2-EOI-Appendix-16E, Bypassing HPCI High Suppression Pool Water Level Suction Transfer Interlock B Operating HPCI or RCIC Turbines with suction temperatures above 140 F may result in equipment damage 2-EOI-Appendix-16E, Bypassing HPCI High Suppression Pool Water Level Suction Transfer Interlock C. Operating HPCI and/or RCIC with suction from the suppression pool may exceed NPSH limits 2-EOI-Appendix-4, Prevention of Injection D. Operating HPCI or RCIC Turbines with suction temperatures above 140 F may result in equipment damage 2-EOI-Appendix-4, Prevention of Injection

206000 (H PCI) G2.1.7: Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5 /43.5 /45.12 /45.13) SRO IR: 4.7 Answer is B Plausibility: The 1st part of A and C is plausible because this caution exists for RHR and Core Spray (but not for HPCI/RCIC). The 2nd part of C and D is plausible because this procedure is used to control level during an ATWS; however, in this case, lowering level is not allowed per C-S.

SRO-only: 10CFRS5.43(b)(5): Assessment of plant conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. The 1st part of the question is RO knowledge; however, the 2nd part is SRO knowledge.

Question was modified from Brunswick 2010 SRO Q#90 2-C-5, Level/Power Control Flowchart SUPPR CHAMBER WATER LEVEL ABNORMAL (9-3B, W15)

SUPPR POOL LEVEL HIGH (9-3F, W12)

SUPPR POOL TEMP SINGLE ELEMENT HIGH OR FAILED (9-3B, W29)

SUPPR POOL AVERAGE TEMP HIGH (9-3E, W 12) 2-EOI-Appendix-SD, Injection System Lineup HPCI 2-EOI-Appendix-16E, Bypassing HPCI High Suppression Pool Water Level Suction Transfer Interlock OPLI 71.201 Objective 11: Given specified plant conditions relative to a Caution in the EOls, determine the basis for the Caution

1 I2JySRoIN.W1HI3IBLC/MAg All units are operating at 100% power.

The normal supply breaker to 4kV Shutdown Board A tripped open and the A EDG automatically tied to the board.

The alternate supply breaker to the 4kV SD Bd would not close, consequently, the diesel carried the board for a total of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and it remained loaded at 500 kW for five of the six hours.

After temporary repairs to the shutdown board alternate supply breaker, the crew has parallelled the A EDG with the shutdown bus in preparation for shutting down the EDG. The normal supply breaker is still not available.

Which ONE of the following identifies 1) The Tech Spec required actions and

2) an additional action required before shutting down the EDG.

[REFERENCE PROVIDED]

A. Enter the action statement for 4kV SD BD A; however, pursuant to LCO 3.0.6, no action statement entry is required for 480 V SD Bd 1A and 480 V RMOV Bd IA.

Perform 0-GOI-300-4, Switchyard Manual, Section 6.7.2, USST lB Transformer Tap Changer (LTC) Auto Checks B. Separate action statement entries are required for 4kV SD BD A and also for 480v SD Bd 1A and 480 V RMOV Bd IA because they were all de-energized when the supply breaker tripped.

Load the A EDG to greater than 1100 kW for at least 30 minutes prior to engine shutdown C. Separate action statement entries are required for 4kV SD BD A and also for 480v SD Bd IA and 480 V RMOV Bd IA because they were all de-energized when the supply breaker tripped.

Perform 0-GOl-300-4, Switchyard Manual, Section 6.7.2, USST lB Transformer Tap Changer (LTC) Auto Checks.

D Enter the action statement for 4kV SD BD A; however, pursuant to LCO 3.0.6, no action statement entry is required for 480 V SD Bd IA and 480 V RMOV Bd 1A.

Load the A EDG to greater than 1100 kW for at least 30 minutes prior to engine shutdown.

Answer is D (verify this answer with the licensee

- may be 2 correct answers)

Provide applicants with Tech Spec 3.8.7, AC Sources

- Operating, Unit I (no bases) 264000 EDGs, A2.03: Ability to (a) predict the impacts of the following on the EMERGENCY GENERATORS (DIESELIJET); AND (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Operating unloaded, lightly loaded, and highly loaded (CFR: 41.5 I 45.6) SRO IR: 3.4 Plausibility: The 1st part of B and C is plausible because the 480v SD Bd 1A and 480v RMOV Bd 1A were briefly de-energized when the supply breaker to the 4KV SD Bd initially tripped. The NOTE in the required action column for Condition A states to enter the applicable conditions and required actions when the SD Bd results in no power source to a required 480V Board. However, in this question, these board were immediately energized by the EDG; therefore, they do have a power source. The 2nd part of A and C is plausible because the USST I B feeds the 4kV SD BD A and the applicants may reason that verification of the tap changer is warranted because of the tripped supply breaker; however, any tap changer manipulations are NOT to be performed while the diesel is loaded.

SRO-only: 10CFR55.43(b)(2) and 55.43(b)(5): Facility operating limitations in the technical specifications and their bases; Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

0-01-82, Standby Diesel Generator System, P&L B:

Continuous operation of DGs at loads below 550 kW should be avoided to prevent oil and soot accumulation in exhaust system, air box, cylinders, and injection nozzles.

If DG idle time exceeds 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or if diesel operates greater than 4 1/2 hours at full speed (900 rpm) at less than 550 kW, the diesel should be loaded greater than 1100 kW for at least 30 minutes prior to engine shutdown. This will allow the engine to clean out any oil accumulations from the exhaust manifolds.

LCO 3.8.1, AC Sources

- Operating LCO 3.8.7, Distribution Systems

- Operating Lesson Plan OPL17I.038 Objective 14: Given a set of unit conditions, determine the appropriate actions to be performed as stated in BFN Tech Spec, 01, ARP, and E0I.

Lesson Plan OPLI7I.036 Objective 13: Given plant and electrical system status determine the appropriate actions to be performed as stated in Technical Specifications, Cls, ARP, and AOls.

9Q4oGooQG24a QOt/1ISRO/NEW/H/2JBLCI Unit I was operating at 100% when a partial loss of Reactor Building Closed Cooling Water (RBCCW) occurred due to a valve failure. Maintenance personnel are in the process of completing repairs to the valve.

The crew reduced power to 90% to control drywell average temperature, which is currently stable at 133 deg F. All available drywell cooling is in service. Drywell pressure is currently 0.8 psig and stable.

RBCCW PUMP SUCTION HDR TEMP (1-TIS-70-3 at Panell-9-4) is 106 deg F (currently stable)

In accordance with 1-AOI-70-1, Loss of RBCCW, which ONE of the following identifies a required operator action based on the current conditions and the earliest required report to the NRC if the RBCCW suction temperature were to subsequently rise to 110 deg F and the crew manually scrammed the reactor?

[REFERENCE PROVIDED]

A Shutdown the Fuel Pool Cooling System a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> NRC notification would be required B. Shutdown the Fuel Pool Cooling System an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> NRC notification would be required C. Vent the Drywell a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> NRC notification would be required D. Vent the Drywell an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> NRC notification would be required

Answer is A 400000 (CCW), G2.4.1 1: Knowledge of abnormal condition procedures.

(CFR: 41.10/43.5/45.13) SRO IR: 4.2 Provide applicants with NPG-SPP-03.5 (entire copy) and NUREG 1022 (entire copy)

Plausibility: The 1st part of C and D is plausible because venting the drywell could be a reasonable action in response to a loss of RBCCW; however, this action is specified by 1-AOI-64-1 in response to high drywell pressure (which is not high) and the stem specifically asks for the required actions from AOl-70-1. The 2nd part of B and D is plausible if the applicant thinks that the RPS actuation has to be automatic.

However, NUREG 1022 states that the staff considers a manual actuation of a system in response to valid plant conditions as a reportable condition.

SRO-only: 10CFR55.43(b)(1) and 55.43(5) Conditions and limitations in the facility and Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

1 -AC 1-70-1 1-AC1-64-1 NPG-SPP-03.5, Regulatory Reporting Requirements Lesson Plan OPLI7I.074 Objective 2: Given specific plant conditions, determine the correct actions to take based on the appropriate AOl.

Lesson Plan 171.092 Objective 4: Given a hypothetical event, determine if a one, four or eight hour report to the NRC is required using SPP-3.5 and NUREG 1022.

91. 21 2000A2.05 001 /2/2/SRO/NEW/H/2/BLC/MM Unit 2 is at 100% power.

The Instrument Mechanics have been performing 2-SR-3.3.1.1.10, Reactor Protection System (RPS) High Reactor Pressure Instrument Channel Calibration, for each RPS channel.

After reviews of the completed surveillance packages, the Instrument Mechanics notified the Unit Supervisor that PIS-3-22AA and PIS-3-22D are inoperable.

Which ONE of the following identifies the minimum required action in accordance with Tech Spec 3.3.1.1, Reactor Protection System (RPS) Instrumentation?

[REFERENCE PROVIDED]

A. Place PIS-3-22AA and PIS-3-22D in the tripped condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> using 01-99, Illustration 3.

B Place either PIS-3-22AA or PIS-3-22D in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> using 01-99, Illustration 3.

C. Do not place any channel in the tripped condition. Restore either PIS-3-22AA or PIS-3-22D to operable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

D. Do not place any channel in the tripped condition. Be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Answer is B Provide Tech Spec Section 3.3.1.1 (no bases) and 2-01-99, Illustration 3 (only page 5 of 11) 212000 (RPS): A2.05 Ability to predict the impacts of the following on the REACTOR PROTECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Nuclear Boiler Instrument System Failure (CFR: 41.5/45.6) SRO IR: 3.7 Plausibility: A is plausible because TS 3.3.1.1 allows separate condition entry for each channel; therefore an applicant could decided that Condition A is required.

C is plausible because an applicant may incorrectly determine that these instruments are both associated with only the A(B) RPS trip system decide that the trip function has been lost for the A(B) trip system.

D is plausible because this Condition G is listed in Table 3.3.1.1-1 beside the Reactor Vessel Steam Dome Pressure -High function.

SRO-only: IOCFR55.43(b)(2): Facility operating limitations in the technical specifications and their bases.

Tech Spec 3.3.1.1, RPS Instrumentation Lesson Plan OPL1 71.028 Objective 14: Evaluate RPS component status against the requirements in Technical Specifications to determine plant compliance.

2. 2Q2QUJGil44QftI12J2JSROiNEWLH/2JBLC/MAR Unit I is at 8% power with the Mode switch in the Startup/Hot Standby position.

The following alarm and pump seal pressure indications are received:

RECIRC PUMP IA NO I SEAL LEAKAGE ABN (9-4A, W25)

No. I Seal Pressure: 500 psig No. 2 Seal Pressure: 500 psig In accordance with Tech Specs, which ONE of the following identifies:

1) whether the alarm/indications represent RCS pressure boundary leakage and

2) whether Mode 1 can be entered if the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> average TOTAL leakage stabilizes at 31 gpm?

A. The alarm/indications represent pressure boundary leakage Mode 1 cannot be entered The alarm/indications do NOT represent pressure boundary leakage Mode 1 cannot be entered C. The alarm/indications represent pressure boundary leakage Mode 1 can be entered D. The alarm/indications do NOT represent pressure boundary leakage Mode 1 can be entered

Answer is B 202001 Recirculation G2.2.44: Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

(CFR: 41.5/43.5/45.12) SRO IR: 4.4 Plausibility: The first part of A and C are plausible because the Recirc Pump Seals do contain the reactor pressure and because the interpretation of what does/does not constitute pressure boundary leakage is located in the Tech Spec Bases. (However, the 1St part of the question is RO knowledge since the RO knows which systems are routed to the equipment drain sump

- identified leakage). The 2nd part of C and D are plausible if the applicant does not interpret LCO 3.0.4, which is not provided as a reference.

SRO-only: 10CFR55.43(b)(2) Facility operating limitations in the technical specifications and their bases. The 2nd part of the question is SRO knowledge since it requires application of LCO 3.0.4 criteria. The 1st part is RO knowledge because an applicant can determine that the seal leakoff is routed to the equipment drain sump, which is identified leakage and because ROs are required to know the information above-the-line.

1-ARP-9-4A, RECIRC PUMP 1A NO 1 SEAL LEAKAGE ABN LCO 3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made:

a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time; Lesson Plan OPLI 71.007 Objective 8: Using a simplified drawing of the recirc pump seals, describe the various failure modes, including how each would be detected.

Objective 21: Given specified plant conditions and a copy of technical specifications, determine any Limiting Conditions for Operation (LCO5) relative to the recirculation system.

Lesson Plan OPL171.087 Objective 20: dentify the rules for entering the applicability of an LCO (LCO 3.0.4) and, given tech specs, identify examples of otherwise specified.

93. 2 14000A2.0 1 00 1/2/2/SRO/NEW/H/3/BLC/MAB Unit 3 is Ifl Mode 5.

The Unit Operator (UO) is performing 0-Tl-20, CRD System Testing and Troubleshooting, concurrently with 3-SR-3.1.5(B), CRD Coupling Integrity Check After Refueling or Maintenance.

When the UO selected and withdrew control rod 42-55 to position 48, the green full-in light on the full core display remained lit but the 4-rod group display indicators responded normally as the rod traveled from position 00 to 48.

All other control rods are fully inserted.

Which ONE of the following:

1) predicts the RMCS/RPIS logic response if the UO selects a different control rod while control rod 42-55 is at position 48 and

2) identifies the allowable method(s) to disarm the control rod drive to comply with Tech Spec 3.9.4, Control Rod Position Indication, Action Statement A.2.2?

A The Rod Withdrawal Permissive light will illuminate if a different control rod is selected.

The control rod can be disarmed electrically.

B. The Rod Withdrawal Permissive light will NOT illuminate if a different control rod is selected.

The control rod can be disarmed electrically.

C. The Rod Withdrawal Permissive light will NOT illuminate if a different control rod is selected.

Electrically disarming the control rod is NOT allowed.

D. The Rod Withdrawal Permissive light will illuminate if a different control rod is selected.

Electrically disarming the control rod is NOT allowed

Answer is A Since the action statements in Tech Spec 3.9.4, Control Rod Position Indication, are RO knowledge (i.e., immediate action statements), then a reference is not required to be provided.

214000 RPIS A2.01 Ability to (a) predict the impacts of the following on the ROD POSITION INFORMATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Failed reed switches (CFR: 41.5 / 45.6) SRO IR: 3.3 Plausibility: The 1st part of B and C are plausible because there are two reed switches that reflect a fully inserted position (S52 and S00).

If the applicant thinks that S00 provides input to the one-rod-out-interlock logic, then the Rod withdrawal permissive light would be extinguished due to 42-55 being withdrawn. The 2nd part of C and D are plausible since some of the action statements listed in Tech Spec 3.1.3, Control Rod Operability, only allow hydraulically disarming a control rod.

SRO-only: 10CFR55.43(b)(2) Facility operating limitations in the technical specifications and their bases. The 1st part of the question is RO knowledge; however, the 2nd part can only be answered by knowing the Tech Spec 3.9.4 Bases for Action Statement A.2.2.

OE 11074 (BFPER 00-4248) - Missed Control Rod LCO BFN Unit 3 Tech Spec 3.9.4, Control Rod Position Indication (and Bases)

Tech Spec 3.9.2, Refuel Position One-Rod-Out Interlock (and Bases) 3-01-85, Section 8.28, Bypassing a Failed Full-in Position on RPIS Buffer Card Lesson Plan OPLI7I.029 Objective 5: Describe the conditions that are required to light the Refuel Mode One Rod Permissive light.

Objective 8: Describe the RMCS/RP IS interrelationships with the following: Refueling equipment Lesson PlanOPLl7l.053 ObjectivelO: Given specific refueling conditions determine the actions required by fuel handling personnel or operators in the MCR using appropriate procedures (Gd, dl, ARP, AOl, TS, and TRM).

a42.L500u312..L/sRozNEw/L!2LBLQJMAB Which ONE of the following choices completes both statements in accordance with Unit 1 Tech Spec 5.2.2, Unit Staff?

A non-licensed operator shall be assigned to each and an additional non-licensed operator shall be assigned for each control room from which a reactor is operating in MODES 1, 2, or 3.

Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and Specifications 5.2.2.a and 5.2.2.f for a period of time not to exceed in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

A reactor containing fuel 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> B. reactor containing fuel 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> C. reactor building and turbine building 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> D. reactor building and turbine building 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Answer is A G2.1.5 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

Plausibility: The 1st part of C and D are plausible since this is a total of 9 AUOs, which is close to the OPDP-1 minimum requirement of 8 AUOs. The 2nd part of B and D is plausible/reasonable because 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is one half of a normal eight hour shift.

SRO-only: 10CFR55.43(b)(1) Conditions and limitations in the facility license. The SRO (Unit Supervisor) is responsible for ensuring Shift Staffing functions are met.

Tech Spec Section 5.2.2 Unit Staffing OPDP-1, Conduct of Ops OPL171.087 (Tech specs)

OPL1 71.071 (Conduct of Ops)

95. G2.L4L QOU3I21LSaQ/BANKJLL2LBLCMAB The Fuel Assembly Transfer Form directs placing bundle FBAO4I in U/i SFSP location I 5-C-09, Orientation SW.

B.

C.

D.

Answer is A G2.1.41: Knowledge of the Refueling Process Plausiblity: Choices B, C, and D are all plausible if the applicant does not know which way is north and does not know that the channel fastner is the component which determines the orientation.

Which ONE of the following identifies the required ORIENTATION?

Trolley Refuel Bridge SRO-only: 10CFR55.43(b)(7): Fuel Handling facilities and procedures.

Lesson Plan OPL17I.053 Objective 1: Describe the layout of the following Refuel Floor areas and equipment locations:

a. Reactor Cavities (all units)

b. Spent Fuel pools (all units)

c. Dryer Separator pits (all units)

d. U1/U2 Transfer Canal

e. New Fuel Inspection Stands

f. Fuel prep machines

g. Fuel pool components/racks

h. Rx cavity to SFSP gates OPL1 71.060 Objective 1.c: Identify fuel that is not properly oriented in the core or SFSP.

96. G2.2.17 OO1/3/2.2/SRO/NEW/L/2/BLC/MAB Which ONE of the following completes both statements in accordance with NPG-SPP-7.1, Online Work Management?

The Schedule freeze date is at After this date, and based on the schedule being frozen, new work is treated as and requires special authorization to be added to the schedule.

[REFERENCE PROVIDED]

A. Tminusl6 Weeks Emergent Work B. Tminus6Weeks a High Risk Activity C T minus 6 Weeks Emergent Work D. Tminusl6 Weeks a High Risk Activity Answer is C Provide applicants with NPG-SPP-07.1 Page 34 of 169 (Appendix A Top Level Process Map), with the Schedule Freeze circled and the weeks whited out on all blocks except for T-0.

G2.2.17 Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator (CFR: 41.10/43.5 / 45.13) SRO IR: 3.8 Plausibility: The 1st part of A and D are plausible because at T-16 weeks, this is the scope freeze date. (not the schedule freeze date). The 2nd part of A and C are plausible because this is a defined activity in NPG-SPP-7.3, Work Activity Risk Management Process.

SRO-only:

I OCFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Operations involvement with respect to reviewing preliminary schedules in the Work Control Center is an SRO function.

NPG-SPP-7.1, Online Work Management NPG-SPP-7.3, Work Activity Risk Management Process Lesson Plan OPLI7I.239 Objective 3: Describe the roles and responsibilities of Operations Personnel associated with the Work Activity Risk Management Process.

Objective 5: Describe the basic process for determining work activity risk.

- 97, GZ.Z.25 OO1/3/2.2jSRO/NEW/L/2JC/MA Which ONE of the following identifies the bases for Tech Spec LCO 3.6.3.1, Containment Atmosphere Dilution (CAD) System, Action statement B.1 when two CAD subsystems are inoperable?

B. Two CAD subsystems 6.1 Verify by administrative 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> inoperable means that the hydrogen control function is AND maintained.

Once per 12 AND hours thereafter B.2 Restore one CAD 7 days subsystem to OPERABLE status.

A. The hydrogen control function is the Primary Containment Inerting System.

Reasonable time to allow the operator to replenish nitrogen from outside sources following a LOCA when access to the site may not be immediately available.

B. The hydrogen control function is the Hardened Wet Well Vent System.

Reasonable time to allow continued reactor operation because the hydrogen control function is maintained and because of the low probability of a LOCA generating hydrogen in amounts capable of exceeding the flammability limit.

C The hydrogen control function is the Primary Containment Inerting System.

Reasonable time to allow continued reactor operation because the hydrogen control function is maintained and because of the low probability of a LOCA generating hydrogen in amounts capable of exceeding the flammability limit.

D. The hydrogen control function is the Hardened Wet Well Vent System.

Reasonable time to allow the operator to replenish nitrogen from outside sources following a LOCA when access to the site may not be immediately available.

Answer is C G2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. (CFR: 41.5 / 41/7 / 43.2) SRO IR: 4.2 Plausibility: The 1st part of B and D are plausible because the HWW vent valve can be used to control hydrogen following a LOCA. The 2nd part of A and D is plausible because this is the bases for the DG Fuel Oil Supply.

SRO-only:

1 OCFR55.43(b)(2): Facility operating limitations in the technical specifications and their bases. The hydrogen control function in Action Statement B.1 is not described anywhere other than the Tech Spec Bases.

Lesson Plan OPL1 71.032 Objective 9: Describe the interrelationships between the CAD System and the following:

a. Drywell Control Air System

b. Containment Inerting and Purge Systems Objective 12: Identify the Technical Specifications associated with the Containment Inerting and Purge Systems and CAD.

98. G23.15 OO1/3/2.3/SRO/NEWILI2/BLCIMA Which ONE of the following completes both statements as they pertain to the Wide Range Gaseous Effluent Radiation Monitoring System (WRGERMS)?

Consider each statement separately.

The WRGERMS (RM-90-306) is required to be operable in accordance with IF the WRGERMS gaseous release rate indication is the reason for a Notice of Unusual Event (NOUE) classification, THEN in accordance with EPIP-1, Emergency Classification Procedure, A TRM 3.3.5, Surveillance Instrumentation the gaseous release rate is required to be assessed by another method, provided the assessment can be accomplished within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, BEFORE the NOUE declaration.

B. ODCM Section 1/2.1.2 Radioactive Gaseous Effluent Monitoring Instrumentation the NOUE is required to be declared immediately. No other assessment method is required to be performed before the declaration is made, even if completed within I hour.

C. ODCM Section 1/2.1.2 Radioactive Gaseous Effluent Monitoring Instrumentation the gaseous release rate is required to be assessed by another method, provided the assessment can be accomplished within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, BEFORE making the NOUE classification.

D. TRM 3.3.5, Surveillance Instrumentation the NOUE is required to be declared immediately. No other assessment method is required to be performed before the declaration is made, even if completed within I hour.

Answer is A G2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instrments, personnel monitoring equipment, etc. (CFR: 41.12 I 43.4 / 45.9) SRO IR: 3.1 Plausibility: The 1st part of C and D are plausible because the Stack Rad Monitor (RM-90-147B & -148B) is listed in the ODCM ODCM Section 1/2.1.2. The WRGERMS is only required in TR 3.3.5, Surveillance Instrumentation. The 2nd part of B and D is plausible because the WRGERMS indication is valid.

SRO-only: 10CFR55.43(b)(4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. Also 55.43(b)(1) Conditions and limitations in the facility license. The knowledge of EPIP requirements is an SRO responsibility.

Lesson Plan OPL1 71.224 Objective 3: Describe the use and application of BEN ODCM Section 1 I 2, Controls and Surveillance Requirements.

Objective 4: Given the BEN ODCM and plant conditions, determine if there is non-compliance with the BEN ODCM Controls and Surveillance requirements.

Lesson Plan OPL1 71.075 Objective 2: Given information concerning different types of events, use EPIP-1, Emergency Classification Matrix to determine the correct classification.

Lesson Plan OPL1 71.033 Objective 5: Given plant and Process Radiation Monitoring System status, determine the appropriate actions to be performed as stated in Technical Specifications, Ols, ARPs, and AOIs.

ODCM Section 1/2.1.2, Radioactive Gaseous Effluent Monitoring Instrumentation TR 3.3.5, Surveillance Instrumentation EPIP-1, Emergency Classification Procedure, Section 11-4, Radioactivity Release Prior to making this emergency classification based upon the gaseous release rate indication, assess the release by either of the following methods:

1. Actual field measurements exceed the limits

2. 0-SI-4.8.B.1.a.1 release fraction (OR Projected or actual dose assessments)

99. G2.4.5 00 1/3/2.4/SROINEW/L/2/BLCiNM A unisolable main steam line break has occurred in the Unit 3 Turbine Building and entry to 0-EOl-4, Radioactivity Release Control, is required.

TURBINE BLDG AREA RADIATION HIGH (9-3A, W29) is alarming Which ONE of the following identifies:

1) the required 0-EOl-4 flowchart implementation and

2) the procedure required to be implemented before the offsite radioactivity release reaches the General Emergency classification value?

A. ONLY Unit 3 is required to enter and execute 0-EOI-4.

C-2, Emergency Depressurization B Units 1, 2, and 3 are all required to independently enter and execute 0-EOI-4.

C-2, Emergency Depressurization C. ONLY Unit 3 is required to enter and execute 0-EOl-4.

EDMG-06, Rapid Evacuation of the Protected Area (REPA)

D. Units 1, 2, and 3 are all required to independently enter and execute O-EOI-4.

EDMG-06, Rapid Evacuation of the Protected Area (REPA)

Answer is B G2.4.5 Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions.

(CFR: 41.10/43.5/45.13) SRO IR: 4.3 Plausibility: The 1st part of A and C is plausible because the stem states that the steam line break is on Unit 3. The 2nd part of C and D is plausible because the stem of the question deals with a General Emergency radioactive release, which an applicant may think requires an evacuation of the protected area. However, the EDMGs are only implemented due to a large loss of ares in the station due to a fire or explosion.

SRO-only: 10CFR55.43(b)(5): Assessment of plant conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. The SRO is responsible for entry and exit to this flowchart. The entry conditions are all SRO knowledge.

EOIPM O-VIll-A, Users Guide for Emergency Operating Instructions Section D. Entry and Use of Radioactivity Release Control O-EOl-4 Unlike all other EOls, EOl-4 is a unit 0 (common) EOl whose entry condition is the declaration of any Emergency Action Level (EAL) for Radiological Release (EPIP-1, Section 11-4). Both unit 2 and unit 3 are required to independently enter and execute the applicable steps of EOl-4, even if the radiological event is believed to be associated with only one of the units. When a step contains a reference to systems discharging OUTSIDE PRIMARY AND SECONDARY CONTAINMENT, this applies only to systems on the unit on which EOI-4 is being executed. For example, should unit 3 have a main steam line leak outside containment, the operators executing EOl-4 on unit 3 would evaluate that a primary system IS discharging. At the same time, the operators executing EOl-4 on unit 2 would evaluate that a primary system IS NOT discharging.

Lesson Plan OPL17I.204 Objective 7: Given appropriate plant status, determine whether a reactor shutdown, reactor scram, or emergency depressurization is required.

Objective 9: Given a list of plant parameters, identify which would require entry into E01-4.

Lesson Plan OPLI7I.201 Objective 9: Recognize the conditions required for exiting an EOI.

100. G2.4.38 OOi/3I2.4ISiO/NWV2_/LC1MA Which ONE of the following choices completes both statements in accordance with the EPIPs?

If an emergency action level for a higher classification was exceeded, but the present situation indicates a lower classification, then the higher classification declared.

If a person volunteers to receive a dose in excess of 25 rem in order to carry out lifesaving operations or to avoid extensive exposures to large populations, then the signature is required on EPIP-15, Emergency Exposures, Appendix B, Acknowledgement and Authorization to Exeed Occupational Dose Limits.

A. should NOT be; Technical Support Center (TSC) Radiation Protection Managers B

should NOT be; Site Emergency Directors (SED)

C. should still be; TSC Radiation Protection Managers D. should still be; SEDs Answer is B G2.4.38: Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required.

(CFR: 41.10/43.5/45.11) SRO IR: 4.4 Plausibility: The 1st part of C and D is plausible because the higher classification should still be REPORTED to the NRC, but not declared. The 2nd part of A and C is plausible because the Radiation Protection section is reponsible for briefing the volunteer on EPIP-15, Appendix A, EPA Emergency Exposure Risk Information.

SRO-only: IOCFR55.43 (B)(5): Assessment of plant conditions and selection of appropriate procedures during normal, abnormal, and emergency conditions. This question tests the applicants knowledge of only emergency plan information that the SROs are required to perform.

EPIP-i, Emergency Classification EPIP-15, Emergency Exposures Lesson Plan OPLI 71.075 Objective 6: Given that an emergency action level for a higher classification was exceeded, but the present conditions indicate a lower classification, state what should be reported and/or declared.

Objective 15: Determine the circumstances under which excessive exposure limits may be approved.