ML14036A312

2013-11-Final SRO Written Exam

ML14036A312
Person / Time
Site:	Grand Gulf
Issue date:	11/15/2013
From:	Garchow S Operations Branch IV
To:	Entergy Operations
laura hurley
References
Download: ML14036A312 (61)

Text

Revisions to the Grand Gulf Written Exam RO QUESTION # 50 Revision: The answer key was changed to accepting distractors A and B.

Bases for Change: Distractor A for this question states the CRD A breaker will not be capable of any remote or local operation and distractor B states the breaker can be opened, closed, and reopened once manually, at the breaker. The intent of distractor A was that the term local operation would include both the electrical and manual modes of operation at the breaker cubicle. This makes this distractor incorrect since the breaker can be operated manually at the breaker cubicle. However, distractors B and C use the term manually implying this term is distinctly different than the term locally. If distractor A is defined as local electrical operation only, (i.e. does not include manually operating) than distractor A is correct since the breaker cannot be operated electrically from the control room or the breaker.

Because of the way the distractors are written, the distractors are ambiguous and therefore, distractors A and B are both accepted as being correct.

SRO QUESTION # 93 Revision: The answer key was changed to distractor D being correct.

Bases for Change: In this question, the applicant was given a set of plant conditions and was required to select the correct procedure to implement based on those conditions. The licensee recommended accepting distractor D as correct in addition to distractor A. The bases for this recommendation was that the stem states a spurious IP Condenser Hotwell Level Low signaloccurs due to a relay failure. The IP Condenser Hotwell Level Low alarm typically results in a loss of all condensate pumps and the feedwater pumps would then trip on low suction pressure. Consequently, a reactor scram would occur due to a loss of feedwater. The original correct answer was distractor A, implement EOP EP-2, RPV Control.

The licensee argued distractor D, implement the Alarm Response Instruction was also correct since the feedwater pumps would not be lost on the spurious actuation of a single condenser low level relay. This is because there is a two-out-of-three logic required to be met and the actuation of a single relay would not satisfy the logic.

The Chief Examiner referenced electrical drawing E-1148-001, revision 13, Condensate Pump C003A-N, and E-1148, revision 14, Condensate Auxiliary Relays, since these drawings contain the relay contacts in question and are typical of all three condensate pumps.

A review of these drawings indicate the alarm given in the stem, IP Condenser Hotwell Level Low, is actuated by one of three relays; 63X/105, 63X/106, and 63X/107. If two of these relays actuate, then relay 63X-1/105 actuates initiating the sequence of events as described above.

However, the actuation of only one of the referenced relays would result in an IP Condenser Hotwell Level Low alarm but no loss of condensate pumps since the two-out-of-three logic for the condensate pump trip is not satisfied.

The review of the drawings also indicates a failure of relay 63X-1/105 would result in the loss of the condensate pumps, however the IP Condenser Hotwell Level Low alarm would not

annunciate off this relay. Because the stem states the IP Condenser Hotwell Level Low alarm has annunciated and there is only a single relay failure, 63X-1/105 has not actuated, the condensate pumps would not trip, the feedwater pumps would remain in service, and there would be no reactor scram. Because there is no reactor scram, the EOPs would not be entered making distractor A incorrect. Therefore, the Alarm Response Instruction would be the governing procedure for the SRO. Distractor D is the only correct distractor.

GGNS LOT 2013 NRC INITIAL LICENSED OPERATOR WRITTEN EXAMINATION SRO EXAM ANSWER KEY 76 A 77 A 78 C 79 C 80 C 81 D 82 B 83 D 84 C 85 B 86 A 87 C 88 B 89 B 90 C 91 A 92 D 93 A 94 D 95 C 96 B 97 B 98 D 99 A 100 A

Examination Outline Cross-Reference Level SRO 295003 Partial or Complete Loss of A.C. Power AA2. Ability to determine and/or interpret the following as K/A # 295003 they apply to PARTIAL OR COMPLETE LOSS OF A.C. Rating 3.5 POWER : (CFR: 41.10 / 43.5 / 45.13)

Rev / Date 0 AA2.03 Battery status: Plant-Specific Question 76 Use your provided references to answer this question.

Bus 17AC was de-energized for two days due to a bus fault.

The fault has been corrected and bus 17AC is now energized.

Battery Charger 1C4 has NOT been re-energized yet.

Division 3 battery 1C3 pilot cell voltage is now 2.11 VDC.

Which of the following describes the OPERABILITY of Division 3 battery 1C3?

A. Battery 1C3 must be declared INOPERABLE.

B. Battery 1C3 may be considered OPERABLE for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> only if average voltage for each connected cell is a minimum of 2.07 VDC.

C. Battery 1C3 may be considered OPERABLE for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> only if average voltage for each connected cell is a minimum of 2.13 VDC.

D. Battery 1C3 may be considered OPERABLE for 31 days for stated conditions if charger 1C5 is in service.

Answer: A Explanation:

With bus 17B01 de-energized, Div 3 battery charger 1C4 is inoperable, requiring entry into TS 3.8.4 A. Charger 1C4 is the only charger that can be credited for TS 3.8.4 for Div 3. This TS action requires verification of category A battery cell parameters to be within limits or TS 3.8.4 B must be entered, which requires declaring the associated

battery inoperable. Cell voltage is given as 2.11 V, which does not meet the minimum category A limit listed in TS table 3.8.6-1; therefore, TS 3.8.4 is not met, and Div 3 battery is inoperable, as reflected by answer A. All other answers are wrong since they state Div 3 battery may be considered operable given certain other conditions, but with the one .

Answers B and C are plausible if the student only considers the requirements of TS 3.8.6 for battery parameters.

Answer D is plausible if the student assumes charger 1C5 can meet operability requirements.

Technical

References:

04-1-01-L11-1 TS 3.8.4 and Bases TS 3.8.6 References to be provided to applicants during exam: TS 3.8.4, TS 3.8.6 Learning Objective: GLP-OPS-TS001, OBJ. 40 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(10) 55.43(b)(5)

Examination Outline Cross-Reference Level SRO 295006 SCRAM 2.2.39 Knowledge of less than or equal to one hour K/A # 295006 Technical Specification action statements for systems. Rating 4.5 (CFR: 41.7 / 41.10 / 43.2 / 45.13)

Rev / Date 0 Question 77 Tech Spec 3.1.5 Actions require inserting a manual scram in Mode 1 within 20 minutes if charging water header pressure is < 1520 psig and two control rod scram accumulators associated with withdrawn control rods are inoperable.

What is a Tech Spec basis for the allowed completion time of 20 minutes for this action?

A. Allow time to start a CRD pump.

B. Allow time to recharge the inoperable scram accumulators.

C. Allow time to drain water from the inoperable scram accumulator instrument blocks.

D. Allow time to fully insert and disarm the control rods associated with the inoperable scram accumulators.

Answer: A Explanation:

TS 3.1.5 Action B.1 assumes no CRD pumps are running if charging water header pressure is <1520 psig. The completion time for Action B starts when 2 or more scram accumulators for withdrawn control rods are declared inoperable concurrently with charging water header pressure < 1520 psig. If the condition cannot be corrected within 20 minutes, Action B.1 is not met and Action D must be entered, which requires placing the Reactor Mode Switch in SHUTDOWN immediately. The TS Bases states the 20 minute completion time provided for Action B.1 should be adequate for starting a CRD pump, as reflected by Answer A. Answers B and C are plausible because accumulator low pressure and water detection in the accumulator instrument block both cause the CRD accumulator trouble alarm, required operable by TR 3.1.5. But

both answers are wrong. Moisture in the accumulator instrument block does not directly require declaring the accumulator inoperable. Low pressure does require individual accumulators to be declared inoperable, but this is addressed by TS 3.1.5 Actions B.2.1 and B.2.2. Answer D is plausible because inserting and disarming a control rod would be necessary if the rod was declared inoperable, which is an alternative to declaring a control rod slow when its accumulator is inoperable.

However, answer D is wrong because TS 3.1.5 Action B does not address urgency forinserting/disarming a control rod, only for restoring a CRD pump to operation.

Technical

References:

TS 3.1.5 Bases TS 3.1.5 Action B.1 TS 3.1.3 TR 3.1.5 05-1-02-IV-1 References to be provided to applicants during exam: none Learning Objective: GLP-OPS-TS001, OBJ. 39 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7),(10) 55.43(b)(2)

Examination Outline Cross-Reference Level SRO 295018 Partial or Complete Loss of Component Cooling Water K/A # 295018 AA2. Ability to determine and/or interpret the following as Rating 3.4 they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER : Rev / Date 0 (CFR: 41.10 / 43.5 / 45.13)

AA2.01 Component temperatures Question 78 The plant is at rated power.

BOP Transformer 13 trips.

TBCW heat exchanger outlet temperature is 97°F.

CCW system temperature is 101°F.

One point, RECIRC PUMP A SEAL COOLING WATER DISCHARGE, on Recirc Temperature Recorder 1B33R601 has just reached its alarm setpoint.

No actions have been taken by the crew.

The CRS should NEXT A. Enter 05-1-02-V-1, Loss of CCW, for a complete loss of CCW, and direct placing the Reactor Mode Switch in SHUTDOWN.

B. Enter 05-1-02-V-1, Loss of TBCW, for a partial loss of TBCW, and direct reducing core flow to 70 mlbm/hr.

C. Enter 05-1-02-V-11, Loss of PSW, for a complete loss of PSW, and direct reducing core flow to 70 mlbm/hr.

D. Concurrently enter 05-1-02-V-11, Loss of PSW, for a partial loss of PSW and 05-1-02-V-1, Loss of CCW, for a partial loss of CCW, and direct restarting Radial Well Pumps isolating CCW to FPCCU Heat Exchangers.

Answer: C Explanation:

BOP transformer 13 supplies approximately half of the Radial Well Pumps that are in service at rated power; therefore, trip of BOP Xfmr 13 results in a large reduction in PSW flow to the plant. Conditions stated in the stem reflect a complete loss of PSW, as defined in ONEP 05-1-02-V-11, step 1.1.1, due to CCW temperature >100°F as the result of degraded PSW flow. Both CCW and TBCW cooling water temperatures will rise as a result of loss of PSW flow at rated power. Based on the stem conditions, rising CCW and TBCW temperatures, when considered separately, seem to imply their respective ONEPs may provide appropriate mitigating actions. However, the PSW ONEP is written to address the overall effects of a degraded PSW supply on the plant, including CCW and TBCW effects, and it provides guidance on when concurrent entry into CCW and TBCW ONEPs is required. For this situation, Loss of PSW ONEP is the appropriate ONEP to provide initial response by the crew. PSW ONEP steps 3.2.15 and 3.2.16 provide the criteria at which the Loss of CCW and TBCW ONEPs should be concurrently executed - if mitigating actions being performed per Loss of PSW ONEP are unsuccessful in restoring and maintaining CCW and TBCW temperatures below 100°F. The conditions given in the stem for CCW, TBCW, and Recirc Pump temperatures do not exceed specific limits stated in the loss of CCW and TBCW ONEPs that would require more extensive or severe action than is accommodated by the loss of PSW ONEP.

All answers are plausible because they reflect actions for varying degrees of effects of degraded CCW or TBCW capability.

Answer C is correct since the overall problem is attributable to PSW flow degradation and CCW temperature 101°F requires declaring a complete loss of PSW.

Complete loss of PSW requires power reduction to 60%, beginning with core flow reduction to 70 mlbm/hr. The CRS should select only Loss of PSW, and select the section for a complete loss of PSW, until it is proven that mitigating actions of that ONEP are unsuccessful in restoring TBCW and CCW related temperatures or until the effects of rising CCW or TBCW temperatures reach levels that require declaring total losses of CCW and/or TBCW.

Answer A is wrong because PSW ONEP step 3.2.15 requires entering the CCW ONEP only if PSW ONEP actions are unsuccessful in restoring and maintaining TBCW temperature <100°F and because the alarming Recirc Pump temperature given is not one of the points that require declaring a complete loss of CCW and scramming per the CCW ONEP.

Answer B is wrong because PSW ONEP step 3.2.16 requires entering the TBCW ONEP only if PSW ONEP actions are unsuccessful in restoring and maintaining TBCW temperature <100°F, and TBCW temperature is only 97°F. Loss of TBCW ONEP should not be entered yet.

Answer D is wrong because PSW ONEP step 3.2.15 requires concurrently entering the CCW ONEP only if PSW ONEP actions are unsuccessful in restoring and maintaining TBCW temperature <100°F and because actions for a complete loss of PSW should be taken, not for a partial loss of PSW, which is reflected by this answer.

Technical

References:

05-1-02-V-11, Loss of PSW 05-1-02-V-1, Loss of CCW 05-1-02-V-2, Loss of TBCW 04-1-02-1H13-P680-11A-D6 References to be provided to applicants during exam: none Learning Objective: GLP-OPS-ONEP, OBJ. 3, 34 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(10) 55.43(b)(5)

Examination Outline Cross-Reference Level SRO 295026 Suppression Pool High Water Temperature 2.2.25 K/A # 295026 / 2.2.25 Knowledge of the bases in Technical Specifications for Rating 4.2 limiting conditions for operations and safety limits.

(CFR: 41.5 / 41.7 / 43.2) Rev / Date 0 Question 79 One basis for the Tech Spec ACTION to immediately suspend RCIC flow testing based on high Suppression Pool temperature is.

A. Elevated RCIC turbine lube oil temperature that could damage RCIC turbine.

B. RCIC turbine exhaust check valve chatter that can damage RCIC piping.

C. Ensure the Containment temperature limit is not exceeded.

D. Preclude having to enter EP-3, Containment Control.

Answer: C Explanation: In the stem, RCIC flow testing implies power is above 1%. TS 3.6.2.1 Action C.1 requires immediately suspending testing that adds heat to the Supp Pool when Supp Pool average temperature exceeds 105°F. The implication in the stem is Supp Pool temperature has just exceeded 105°F. TS bases 3.6.2.1 states the Supp Pool temperature limit was developed to address technical concerns that include limiting containment average air temperature to < 185°F during the DBA; therefore, answer C is correct. Answer A is plausible because it describes undesired effects of elevated Supp Pool temperature on RCIC. It is each wrong because 04-1-01-E51-1 step 3.3 states RCIC suction temperature is limited to 140°F for non-emergency operation to prevent bearing/turbine seals overheating, and during emergencies, EP caution 4 states RCIC equipment damage could occur if Supp Pool temperature exceeds 225°F, well above the 105°F condition implied in the stem and because RCIC operability is not a concern in the DBA analysis and is not mentioned in the bases for TS 3.6.2.1. Answer B is plausible because exhaust check valve operation could theoretically be affected mechanistically by Supp Pool parameters, but it is wrong because 04-1-01-E51-1 step 3.2 states this is a concern for operation of RCIC

at speeds below 2000 rpm and because RCIC operability is not a concern in the DBA analysis and is not mentioned in the bases for TS 3.6.2.1. Answer D is plausible because entry into an Emergency Procedure is undesirable and implies elevated risk and challenge to plant safety, but it is wrong because TS 3.6.2.1 Action C that requires suspending testing that is adding heat to the Supp Pool at 105°F, so the Supp Pool Temperature EP-3 entry condition of 95°F would have already been met, so EP-3 entry cannot be avoided.

Technical

References:

TR 3.6.2.1 and bases EP-3 EP Caution 4 References to be provided to applicants during exam: none Learning Objective: GLP-OPS-TS001, OBJ. 39 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(5,7) 55.43(b)(2)

Examination Outline Cross-Reference Level SRO 295028 High Drywell Temperature EA2. Ability to determine and/or interpret the following as K/A # 295028 they apply to HIGH DRYWELL TEMPERATURE : Rating 4.1 (CFR: 41.10 / 43.5 / 45.13)

Rev / Date 0 EA2.03 Reactor water level Question 80 A steam leak in the drywell occurred at rated power.

All but one control rod fully inserted on the resulting scram.

Due to instrument failures, the only Reactor Water Level indicator that is functioning is Upset range, which is indicating +55 inches, slowly trending down.

Drywell temperature is 196°F, slowly trending up.

Containment pressure is 4 psig, slowly trending up.

The CRS should NEXT A. direct overriding Low Pressure ECCS systems.

B. direct terminating injection from all sources outside of primary containment.

C. Exit EP-2 and enter EP-5.

D. Exit EP-2A and enter EP-5A.

Answer: C Explanation: EP Caution 1 states Upset range water level instrumentation may not be used if drywell temperature at elev. 166 is above 195°F and indicated Upset level is <159 inches. The stem states all other water level indication is unavailable, so the stem stating drywell temperature is above 195°F indicates Upset level indication cannot be used, and now all level indication is unavailable. Stating drywell temperature, in general, is above 195°F implies temperature at elev. 166 is above

195°F, since upper elevations will always be the hottest. Conditions in the stem, that a scram has resulted due to a leak in the drywell, imply that either EP-2 or EP-2A would have been entered. Since all but one control rod are fully inserted, EP-2 must have been entered. EP-2 step 3 requires exiting EP-2 and entering EP-5; therefore, only answer C is correct. Answer A is plausible because the indicated level given is high, above the top of the allowed control band in EP-2, 53.5 inches, and 02-S-01-27, Operations Philosophy, requires terminating injection to get back into band if level is actually high, and EP-2 step ED-2 actually requires preventing low pressure ECCS injection if not desired, but here, level has to be assumed to be unknown. Answer B is plausible for the same reasons as answer A, plus EP-2 step L-2 has action that states terminate injection from outside Containment, though it is not applicable here. Answer D is plausible if the student believes an operation is in EP-2A due to one control rod being stuck out, but the reactor is considered shutdown based on shutdown margin.

Technical

References:

EP-2 EP-5 References to be provided to applicants during exam: none Learning Objective: GLP-OPS-EP02, OBJ. 3,4,7; GLP-OPS-EP5, OBJ. 5 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(10) 55.43(b)(5)

Examination Outline Cross-Reference Level SRO 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown K/A # 295037 EA2. Rating 4.1 Ability to determine and/or interpret the following as they apply to SCRAM CONDITION PRESENT AND REACTOR Rev / Date 0 POWER ABOVE APRM DOWNSCALE OR UNKNOWN :

(CFR: 41.10 / 43.5 / 45.13)

EA2.04 Suppression pool temperature Question 81 Use the figure on the next page to answer this question.

An ATWS is in progress.

Reactor power is 7%.

Reactor Water Level is -130 inches on wide range.

Reactor Pressure is 910 psig.

Suppression Pool temperature is 160°F.

Suppression Pool level is 18.5 feet..

The CRS should NEXT A. Direct terminating injection to a new RPV level band of -191 inches to -167 inches.

B. Direct lowering reactor pressure to 450 psig to 600 psig.

C. Direct fully opening Main Bypass Valves in anticipation of Emergency Depressurization.

D. Enter the Emergency Depressurization leg.

Answer: D Explanation: With suppression pool temperature 160°F, suppression pool level 18.5 ft, and RPV pressure 910 psig, operation is in the unsafe zone of the HCTL curve. EP-3 step SPT-5 requires emergency depressurization, which is the overriding priority, as stated in EP-2A step P-1. Therefore, answer D is correct, and other answers are wrong since EP-2A step P-1 requires exiting level and pressure legs at that point since they would delay emergency depressurization.

Answer A is plausible because with suppression pool temperature above 110°F

and power above 5%, EP-2A step L-5 requires performance of step L-8 to lower level to reduce power. Answer B is plausible because one strategy to avoid the unsafe zone of HCTL is to reduce reactor pressure, as stated in EP-2A step P-1. But EP-3 step SPT-5 does not accommodate restoring operation to the safe zone of the curve, as does the PSP curve. Once in the unsafe zoneof HCTL, it is too late for that strategy. Answer C is plausible because for non-ATWS situations where emergency depressurization is inevitable, the strategy is to dump the maximum energy possible to the main condenser. But this is only stated in EP-2 step P-1.

Technical

References:

EP-2A EP-3 EP-2 References to be provided to applicants during exam: HCTL curve Learning Objective: GLP-OPS-EP02A, OBJ. 22, GLP-OPS-EP3 Obj. 22 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(10) 55.43(b)(5)

Examination Outline Cross-Reference Level SRO 700000 Generator Voltage and Electric Grid Disturbances 2.1.20 Ability to interpret and execute procedure steps.

K/A # 700000 (CFR: 41.10 / 43.5 / 45.12) Rating 4.6 Rev / Date 0 Question 82 Refer to the surveillance data sheet on the next page.

The plant is at 40% power during start up.

05-1-02-I-4, Loss of AC Power has been entered for grid instability due to grid voltage fluctuations.

06-OP-1R20-W-0001, Plant AC and DC Electrical Power Distribution Weekly Lineup, has been performed in accordance with the ONEP.

Based on this data, the CRS should A. Initiate a Potential LCO, only, for the offsite supply from Baxter Wilson and Franklin lines being inoperable for TS 3.8.1 Condition A.

B. Initiate a Potential LCO, only, for the offsite supply from the 115KV Port Gibson line being inoperable for TS 3.8.1 Condition A.

C. Initiate an actual TS LCO for the offsite supply from Baxter Wilson and Franklin lines being inoperable for TS 3.8.1 Condition A. Perform TS 3.8.1 Action A.1 as specified and restore the sources operable within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

D. Initiate an actual TS LCO for the offsite supply from the 115KV Port Gibson line being inoperable for TS 3.8.1 Condition A. Perform TS 3.8.1 Action A.1 as specified and restore the sources operable within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

GRAND GULF NUCLEAR STATION SURVEILLANCE PROCEDURE 06-OP-1R20-W-0001 Revision 109 Attachment I Page 2 of 12 Page XRef DATA SHEET I PLANT AC AND DC ELECTRICAL POWER DISTRIBUTION WEEKLY LINEUP SAFETY RELATED (Step 5.1.2)

Div 1, 2 & 3 Offsite Feeders OFFSITE ENERGIZED VOLTAGE RECORDED FREQUENCY RECORDED INITIALS FEEDER INDICATOR VOLTAGE INDICATOR FREQUENCY YES/NO (LOCATION) (ACCEPTANCE (LOCATION) (ACCEPTANCE CRITERIA) CRITERIA)

BAXTER 500 kV FREQ

$ WILSON

• YES JACKSON 494 kV SR27-SR-R600 60.5 Hz DISPATCHER (H13-P807)

$ FRANKLIN

• YES (496-525kV)*** or Pine Bluff (58.5-61.8Hz)

Dispatcher

• • 4.05 x 27.64 152-1511 = 111.94 kV 115kV LINE 152-1611 (120.75-

$ PORT

• YES 152-1704 112.13) kV GIBSON

• To determine status of offsite feeders, CONTACT load dispatcher. ENSURE that the feeders are independently energized from the grid, such that the loss of one feeder would NOT result in the loss of another

• • To determine voltage of the Port Gibson 115kV line, RECORD ESF 12 incoming voltage at Bus 15AA, 16AB OR 17AC placing the Sync switch for the designated breaker to ON. MULTIPLY this reading by 27.64 for equivalent feeder voltage. RETURN Sync switch to OFF after taking reading.

• • • Allowable Value of minimum voltage is >491 kV for operability of Offsite Feeders. This value is based on analysis of the Class 1E ESF buses AND includes an allowance for instrument uncertainty associated with the voltage measurement in the switchyard. Extended operation beyond the normal continuous operating limits Should be evaluated AND caution Should be taken when starting large loads under these conditions.

Answer: B Explanation: The value listed for the 115KV Port Gibson line, 111.94KV, is less than the minimum required for operability, 112.13KV; therefore it is inoperable. The value given for the 500KV Baxter and Franklin lines, 494KV is less than the nominal minimum, 496KV, but it is greater than the operability limit, 491KV, listed in note***. Therefore, the 500KV lines meet operability requirements.

This alone makes answer A inicorrect. TS 3.8.1 requires two offsite supplies operable, and since the 500KV feeds are operable, the LCO is met. This makes answers C and D wrong since they imply the LCO is not met. Since the 115KV source is inoperable, only a Potential LCO should be initiated as defined in 02-S-01-17 step 5.1.2b; therefore, answer B is correct. Answer A and C are plausible if the student does not know meeting the allowable value is sufficient for operability. Answer D is plausible if the student thinks inoperability of one offsite source that can in part satisfy LCO 3.8.1 constitutes failure to meet the LCO.

Technical

References:

TS 3.8.1 06-OP-1R20-W-0001 Att I ENS-DC-199 02-S-01-17 References to be provided to applicants during exam: partially completed 06-OP-1R20-W-0001 Att I page 2 (above); TS 3.8.1 first 2 pages Learning Objective: GLP-OPS-TS001, OBJ. 40, GLP-OPS-TSLCO Obj. 3, 7 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(10) 55.43(b)(5)

Examination Outline Cross-Reference Level SRO 295020 Inadvertent Containment Isolation AA2.

K/A # 295020 Ability to determine and/or interpret the following as Rating 3.4 they apply to INADVERTENT CONTAINMENT ISOLATION :

(CFR: 41.10 / 43.5 / 45.13) Rev / Date 0 AA2.02 Drywell/containment temperature.

Question 83 Use your provided references to answer this question.

The coil for relay 1M71-R65 fails, resulting in inadvertent closure of Division 2 Drywell Chilled Water containment isolation valves.

Power has been reduced to 70%.

Average drywell temperature has risen to 145°F.

CRD Cavity temperature is 155°F.

All temperatures at other drywell elevations are below 149°F.

What is the limiting action statement at this point?

A. Declare equipment within the drywell inoperable within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

B. Restore drywell temperature to within limit within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or then declare equipment within the drywell inoperable within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

C. Restore drywell temperature to within limit within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or immediately initiate action to provide a record of the cumulative time the limit was exceeded and an analysis to demonstrate continued operability of equipment within the drywell.

D. Restore drywell temperature to within limits within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or be in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Answer: D

Explanation: All answers are plausible, because both TS 3.6.5.5 and TR 6.7.3 impose limits on either average drywell temperature or local drywell temperatures. TS 3.6.5.5 requires drywell average temperature to be maintained 135°F. TR 6.7.3 additionally limits CRD Cavity temperature to 185°F and other local drywell temperatures to 150°F. Only TS 3.6.5.5 limit has been exceeded. Answer D reflects TS 3.6.5.5 Actions; therefore it is correct. Answers A, B, and C reflect TR 6.7.3 actions, and since LCO TR 6.7.3 is still met, no TR 6.7.3 actions are required, these answers are wrong.

Technical

References:

TS 3.6.5.5 TR 6.7.3 06-OP-1000-W-0001 Att. I, data sheet II References to be provided to applicants during exam: TS 3.6.5.5 and TR 6.7.3 Learning Objective: GLP-OPS-TS001, OBJ. 40 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(10) 55.43(b)(5)

Examination Outline Cross-Reference Level SRO 295035 Secondary Containment High Differential Pressure EA2.

K/A # 295035 Ability to determine and/or interpret the following as they Rating 3.9 apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE: Rev / Date 0 (CFR: 41.8 to 41.10)

EA2.01 Secondary containment pressure: Plant-Specific Question 84 Use your provided references to answer this question.

The plant is at rated power.

SGTS A is operating for performance of 06-OP-1T48-M-0001, Standby Gas Treatment System A Operability.

Five hours into the SGTS A surveillance, personnel using door 1A401, elevation 166 ft Turbine Bldg. to Auxiliary Bldg. equipment door, report the door has jammed in the open position.

Auxiliary Bldg D/P rises to -0.05 w.c.

What is the maximum Tech Spec completion time for having to be in Mode 3 if this condition continues?

A. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

B. 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />.

C. 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />.

D. 7 days, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Answer: C Explanation:

Secondary Containment Integrity surveillance requirements SR 3.6.4.1.3 and SR 3.6.4.1.4 require that Secondary Containment be leak tight enough that one SGTS can draw down Secondary Containment pressure to -0.25 w.c. within 180 seconds and maintain -0.266 w.c. for 1 continuous hour. The listed SGTS surveillance tests both proper operation of SGTS and leak tightness of Secondary Containment. For the

stated conditions, SGTS is unable to maintain the required DP, due to excessive Aux.

Bldg. leakage, not due to a failure of SGTS. Therefore, the applicable TS action is TS 3.6.4.1, Action A.1, restore Secondary Containment operable within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or enter TS Condition B and 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 />; so Mode 3 would have to be attained within 16 total hours if the problem is not resolved. This is reflected in correct answer C.

Answer A is plausible if the student incorrectly believes both SGTS are rendered inoperable, incorrectly enters TS 3.6.4.3 Action D, and then misapplies TS 3.0.3, allowing only 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to be in Mode 3 versus the 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> allowed by TS 3.0.3.

Answer B is plausible if the student believes both SGTS are rendered inoperable, incorrectly enters TS 3.6.4.3 Action D, and then applies TS 3.0.3, which allows 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> to be in Mode 3.

Answer D is plausible if the student incorrectly believes only SGTS A is rendered inoperable, since it is the subsystem in operation, and then applies TS 3.6.4.3 Condition A, projecting Condition B.

Technical

References:

TS 3.6.4.1, Secondary Containment TS 3.6.4.3, SGTS References to be provided to applicants during exam: TS 3.6.4.1, TS 3.6.4.3 (including surveillance requirements)

Learning Objective: GLP-OPS-TS001, OBJ. 40 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b) 55.43(b)(5)

Examination Outline Cross-Reference Level SRO 295036 Secondary Containment High Sump/Area Water Level 2.2.42 K/A # 295036 / 2.2.42 Ability to recognize system parameters that are entry-level Rating 4.6 conditions for Technical Specifications.

(CFR: 41.7 / 41.10 / 43.2 / 43.3 / 45.3) Rev / Date 0 Question 85 The plant is in Mode 1.

RHR C room sump pumps are tagged out of service for discharge check valve replacement.

A PSW leak from ADHR heat exchangers results in flooding RHR C pump room with two feet of water.

RHR C Jockey Pump motor is partially submerged but remains operating.

For the initial operability screening of the associated Condition Report, Engineerings judgment based on industry operating experience is that RHR C Jockey Pump motor winding insulation may fail at any time due to submersion.

For the immediate operability determination, the SRO should A. Screen the CR as Non-Functional, and immediately enter TS 3.5.1.

B. Screen the CR as Inoperable, and immediately enter TS 3.5.1.

C. Screen the CR as Functional, and enter TS 3.5.1 after 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> or upon failure of RHR C Jockey Pump, whichever come first.

D. Screen the CR as Operable, and enter TS 3.5.1 after 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> or upon failure of RHR C Jockey Pump, whichever come first.

Answer: B Explanation: There is no direct LCO entry based on Secondary Containment water levels. However, the effect of high water level in a Secondary Containment area can pose operability issues, requiring an immediate operability determination per EN-OP-104, Operability Determination. RHR C Jockey Pump is required for LPCI C OPERABILITY. In this case, the student must recognize Engineering Judgment does not provide a Reasonable Expectation of Operability. EN-OP-104 defines Reasonable Expectation as a high standard and states the supporting basis should provide a high degree of confidence that the SSC remains OPERABLE. Definition 3.0[24] states when system capability is degraded to a point where it cannot perform with Reasonable Expectation or reliability, the system should be judged INOPERABLE. If RHR C may be expected to fail at any time, RHR C may not be considered operable, since ECCS acceptance criterion 5 for long-term cooling cannot be guaranteed. Per EN-OP-104 definitions, the appropriate CR operability coding for equipment listed in Tech Specs is OPERABLE/INOPERABLE versus FUNCTIONAL/NON-FUNCTIONAL, which applies to equipment only required by TRM, E-Plan, Security System, etc. With RHR C inoperable, TS 3.5.1 Condition A must be entered immediately. Therefore, answer B is correct.

Answer A is plausible if the student does not know the required distinction for TS equipment versus other required equipment defined in EN-OP-104, but it is wrong since RHR C Jockey Pump is required by TS.

Answer C is plausible if the student does not consider equipment qualification with respect to operability and only considers that RHR C Jockey Pump continues to run.

But it is wrong because the code FUNCTIONAL cannot be applied to TS equipment and because there is no Reasonable Expectation of Operability as defined in EN-OP-104.

Answer D is plausible if the student does not consider equipment qualification with respect to operability and only considers that RHR C Jockey Pump continues to run.

Technical

References:

EN-OP-104 TS 3.5.1 References to be provided to applicants during exam: none Learning Objective: GLP-OPS-PROC Obj. 42.1, 42.3; GLP-OPS-TS001, OBJ. 40 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7,10) 55.43(b)(2,3)

Examination Outline Cross-Reference Level SRO 203000 RHR/LPCI: Injection Mode (Plant Specific) 2.1.19 K/A # 203000 / 2.1.19 Ability to use plant computers to evaluate system or Rating 3.8 component status.

(CFR: 41.10 / 45.12) Rev / Date 0 Question 86 Refer to the attached figure.

EP-2 and EP-3 are being executed.

RHR A and B are available.

All other injection systems are unavailable.

Operation is in the Unsafe Zone of the PSP curve.

Operation is in the Unsafe Zone of the HCTL curve.

Emergency Depressurization is in progress.

How should RHR A and B be aligned for the plant conditions depicted on the attached SPDS display?

A. Align both RHR A and B for LPCI injection via 1E12F042A and B.

B. Align both RHR A and B for Containment Spray.

C. Align both RHR A and B for Suppression Pool Cooling.

D. Align RHR A for injection via 1E12F053A and align RHR B for Containment Spray.

Answer: A Explanation: Adequate core cooling is the priority objective of the EPs. For the conditions represented on SPDS, adequate core cooling does not exist. Level is below the Minimum Zero Injection RPV Water Level, -204 and Minimum Steam Cooling Water Level, -191. EP-2 step L-14 directs maximizing RPV injection with all available systems due to RPV level <-191, depicted by Fuel Zone (FZ) level on SPDS. Here, only RHR A and B are available. Therefore, answer A is correct.

Answer B is plausible because the PSP curve has been exceeded. It is wrong because EP-3 step PCP-5 states only initiate Containment Spray with systems not required for adequate core cooling. Since only RHR A and B are available for injection, adequate core cooling cannot be assured by other means, so RHR A and B should be used for injection, not Containment Spray. Therefore, answer B is wrong.

Answer C is plausible because the HCTL curve has been exceeded. It is wrong because EP-3 step SPT-5 states only initiate Suppression Pool Cooling with systems not required for adequate core cooling. Since only RHR A and B are available for injection, adequate core cooling cannot be assured by other means, so RHR A and B should be used for injection, not Suppression Pool Cooling. Therefore, answer C is wrong.

Answer D is plausible because the PSP curve has been exceeded. It is wrong because EP-3 step PCP-5 states only initiate Containment Spray with systems not required for adequate core cooling. Since only RHR A and B are available for injection, adequate core cooling cannot be assured by other means, so RHR A and B should both be used for injection until adequate core cooling is achieved. Therefore, answer B is wrong Technical

References:

EP-2, EP-3 02-S-01-27, Operations Philosophy References to be provided to applicants during exam: none Learning Objective: GLP-OPS-EP02, OBJ. 8; GLP-OPS-EP3, OBJ. 8, Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X

Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(10)

Examination Outline Cross-Reference Level SRO 211000 Standby Liquid Control System A2.

K/A # 211000 Ability to (a) predict the impacts of the following on the Rating 3.2 STANDBY LIQUID CONTROL SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or Rev / Date 0 mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.07 Valve closures Question 87 The plant is at rated power.

06-OP-1C41-Q-0001, Standby Liquid Control Functional Test, is being performed with SLC A pump running, circulating the SLC Test Tank.

A Loss of Offsite Power occurs.

An ATWS due to multiple stuck control rods exists with power 10%.

Immediate Containment evacuation is directed, and SLC pump A continues to circulate the SLC Test Tank.

(1) What effect will these conditions have on SLC system B injection to the RPV?

(2) What direction should be given by the CRS concerning Power Control for EP-2A?

A. (1) SLC B will inject SLC Boron Tank contents, only.

(2) Trip SLC pump B when SLC tank level drops to 2000 gallons.

B. (1) SLC B will inject SLC Boron Tank contents, only.

(2) Trip SLC pump B when SLC tank level drops to 0 gallons.

C. (1) SLC B will inject SLC Test Tank contents, only.

(2) Direct installation of Attachment 28, Alternate SLC Injection.

D. (1) SLC B will inject SLC Boron Tank contents diluted with SLC Test Tank contents.

(2) Direct installation of Attachment 28, Alternate SLC Injection.

Answer: C

Explanation: During the listed surveillance, SLC pump suction from SLC Test Tank valve 1C41F031 is open. Limit Switches on F031 provide a open permissive to SLC pumps boron tank suction valves 1C41F001A and B and provide a separate start permissive to SLC pumps. With 1C41F031, neither 1C41F001A or B will open when the respective SLC pump is initiated from control room panel 1H13P601, although the respective SLC pump will start since it sees a suction path available. In the situation given, SLC pump B will immediately start with suction from the SLC Test Tank only, and neither suction from the SLC boron tank will open. SLC B injection valve will fire when SLC B is initiated; therefore, SLC B will start and pump SLC Test Tank to the reactor. Since SLC will not inject boron under these conditions, EP-2A step Q-4 requires Att. 28 for alternate boron injection. For these reasons, answer C is correct.

All other answers are plausible since they either pertain to SLC B being functional and include actions that would follow successful SLC injection, as in answers A and B, or they pertain to a disfunctional SLC B and include the associated contingency of Att.

28, as in answer D. Answers A, B and D are fundamentally wrong since they imply SLC boron tank contents would be injected, purely or diluted, as if 1C41F001A/B would open, but F001A/B would not open if F031 is open.

Technical

References:

EP-2A EP Att. 28 E1169-01, 02, 12, 14 References to be provided to applicants during exam: none Learning Objective: GLP-OPS-EP02A, OBJ. 8; GLP-OPS-C4100 OBJ. 10 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(5)

Examination Outline Cross-Reference Level SRO 215005 Average Power Range Monitor/Local Power Range Monitor System K/A # 215005 A2. Rating 3.7 Ability to (a) predict the impacts of the following on the AVERAGE POWER RANGE MONITOR/LOCAL POWER Rev / Date 0 RANGE MONITOR SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.02 Upscale or downscale trips Question 88 The plant is at rated power.

APRM 4 fails to 120%.

The crew places APRM 4 in BYPASS on 1H13P680.

(1) What is the impact of APRM 4 failing to 120%?

(2) In addition to placing APRM 4 in BYPASS on 1H13P680, which of the following are other actions required per plant procedures?

A. (1) Half-trip signal to voters, only. No half-scram.

(2) Initiate a Technical Specification (TS) LCO tracking report in ESOMS for APRM4.

B. (1) Half-trip signal to voters, only. No half-scram.

(2) Initiate a Potential Tech Spec (PTS) LCO tracking report in ESOMS for APRM4.

C. (1) Division 2 half-scram.

(2) Reset the half scram, and initiate a Technical Specification (TS) LCO tracking report in ESOMS for APRM4.

D. (1) Division 2 half-scram.

(2) Reset the half scram, and initiate a Potential Tech Spec (PTS) LCO tracking report in ESOMS for APRM4.

Answer: B

Explanation: This relates to the new PRNM system installed during refueling outage 18 in 2012. Eight APRM channels feeding 2 trip systems were replaced by 4 digital APRM channels feeding a single, VOTER logic trip system. For a single APRM trip, no half scram is produced, but only one vote. Any 2 APRM channels will now produce a full scram, not a half-scram. No single APRM channel will produce a half-scram. This is different than all other RPS logic, which requires a trip in each of 2 trip systems to produce a full scram and where any channel tripping produces a half-scram. TS 3.3.1.1 and TR 3.3.2.1 require only 3 of 4 APRM channels operable in Mode 1. No action statement is entered for only one APRM inoperable. For this situation, 02-S-01-17, Control of Limiting Conditions of Operation requires a Potential LCOTR to be initiated to track the inoperability of equipment potentially required operable so that if another APRM goes inoperable, it would be recognized that 2 APRMS are inoperable and the appropriate LCO conditions entered. Since no half-scram is produced and a PTSLCO is proper, answer B is correct. Other answers are plausible because they each include variations of the concepts discussed above.

Answer A is wrong because it implies actual entry into a TS condition statement.

Answers C and D are wrong because they imply a half-scram was produced by the APRM failure. Additionally, answer C is wrong because it implies actual entry into a TS condition statement.

Technical

References:

TS 3.3.1.1 TR 3.3.2.1 02-S-01-17 References to be provided to applicants during exam: none Learning Objective: GLP-OPS-C5100 Obj. 3.5, 7.2, 14; GLP-OPS-TS001, OBJ. 40; GLP-OPS-TSLCO OBJ. 7 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(5)

Examination Outline Cross-Reference Level SRO 259002 Reactor Water Level Control System 2.4.34 K/A # 259002 / 2.4.34 Knowledge of RO tasks performed outside the main control Rating 4.1 room during an emergency and the resultant operational effects. Rev / Date 0 (CFR: 41.10 / 43.5 / 45.13)

Question 89 The plant is at rated power.

An event occurs that requires Control Room evacuation.

The CRS implementing 05-1-02-II-1, Shutdown from the Remote Shutdown Panel, directs the RO to open Reactor Protection System breakers:

CB2A, CB5A, CB7A, CB8A CB2B, CB5B, CB7B, CB8B (1) For what situation does the CRS direct this action?

(2) What effect will this action have on reactor level control using Feedwater?

A. (1) Control room fire.

(2) Feedwater will NOT be affected by this RO action and will automatically maintain level in the normal band unless disabled by the fire.

B. (1) Control room fire.

(2) Feedwater will become unavailable due to this RO action.

C. (1) Security event.

(2) Feedwater will NOT be affected by this RO action and will automatically maintain level in the normal band unless disabled by the fire.

D. (1) Security event.

(2) Feedwater will become unavailable due to this RO action.

Answer: B Explanation: For evacuation of the Control Room due to a fire, 05-1-02-II-1 step 3.1.4 specifically requires an RO to perform Att. XXII. The first section of this

procedure requires opening RPS breakers CB-2A,5A,7A,and 8A on panel 1C71P001 in RPS A MG set room located on 189 elevation, Control Bldg and CB-2B,5B,7B,and 8B on panel 1C71P002 in RPS B MG set room located on 148 elevation, Control Bldg. The CB-2A(B) and 8A(B) result in deenergization of RPS A scram solenoids.

The CB-5A(B) and 7(A) deenergize Division 1 and 2 MSIV solenoids. This does not directly affect Feedwater Level Control, but it does result in MSIV closure, which isolates steam to the reactor feed pump turbines, thus disabling Feedwater system.

Since the breakers are in the control building and Feedwater is disabled, answer B is correct and all other answers are wrong. Answer A is plausible because the breakers do not directly affect Feedwater level control. Answers C and D are plausible because there are many actions taken outside of the main control room associated with certain security events, but not these specific actions.

Technical

References:

05-1-02-II-1 step 3.1.4 and Att. XXII References to be provided to applicants during exam: none Learning Objective: GLP-OPS-N2100 Obj. 36; GLP-OPS-C7100 Obj. 4.1, 23; GLP-OPS-C3400, OBJ. 22 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(10) 55.43(b)(5)

Examination Outline Cross-Reference Level SRO 400000 Component Cooling Water System (CCWS) 2.4.2 K/A # 400000 / 2.4.2 Knowledge of system set points, interlocks and automatic Rating 4.6 actions associated with EOP entry conditions.

(CFR: 41.7 / 45.7 / 45.8) Rev / Date 0 Question 90 Use your provided references to answer this question.

(1) Which one of the following signals, including its setpoint, is common to both RPS instrumentation and Containment Isolation instrumentation for Drywell Chilled Water System valves and is an Emergency Procedure entry condition?

(2) What is the completion time for placing the channel in trip if only one channel of this instrumentation is inoperable?.

A. (1) Reactor Water Level Low.

(2) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

B. (1) Reactor Water Level Low.

(2) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

C. (1) Drywell Pressure High.

(2) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

D. (1) Drywell Pressure High.

(2) 24hours.

Answer: C Explanation: Drywell Chilled Water System isolation valves 1P72F121,F122,F123,F124,F125,F126 are part of Group 6. Drywell Pressure High from trip units 1C71N650A,B,C,D is common to both RPS and Group 6 containment isolation instrumentation. With a nominal setpoint of 1.23 psig, it is the basis for EP-2 and EP-3 entry conditions. With one channel of this instrumentation inoperable, TS 3.3.1.1 Action A.1 applies and has a completion time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Also, TS 3.3.6.1 Action A.1 applies, and since this represents TS 3.3.6.1 function 2b, also has a completion time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, as it is for all containment isolation signals that are common to RPS. Therefore, answer C is correct. Answer A and B are plausible because Reactor Level Low (level 3) is a signal common to TS 3.3.1.1, TS 3.3.6.1, and an EP-2 entry condition. But it is wrong in this case because Level 3 is not

specifically common to Group 6, but only Groups 2 and 3. Answer D is plausible because the completion time one channel inoperable of TS 3.3.6.1 instrumentation that is not common to RPS is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per TS 3.3.6.1 Action A.1. Though the associated TS instrumentation tables are not to be provided for this question, the SRO student should recognize the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> completion listed for TS 3.3.6.1 Action A is for instrumentation common to RPS and the TS 3.3.6.1 Action A completion time is and is designed to be consistent with the TS 3.3.1.1 Action A.1 completion time.

Technical

References:

TS 3.3.1.1 TS 3.3.6.1 EP-2 EP-3 17-S-06-5 References to be provided to applicants during exam: TS 3.3.1.1 first page (Condition A), TS 3.3.6.1 first page (Condition A)

Learning Objective: GLP-OPS-TS001 Obj. 40 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level SRO 201003 Control Rod and Drive Mechanism A2.

K/A # 201003 Ability to (a) predict the impacts of the following on the Rating 3.1 CONTROL ROD AND DRIVE MECHANISM ; and (b) based on those predictions, use procedures to correct, control, or Rev / Date 0 mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.06 Loss of CRD cooling water flow Question 91 The plant is at rated power.

Control Rod 32-22 is required to be isolated for maintenance.

04-1-01-C11-1, Control Rod Drive Hydraulic System, section 5.7, Total Isolation of an HCU, is performed for the protective tagging boundary.

(1) Which of the following is a possible consequence of this performing this section of the SOI at rated power?

(2) What is a mitigating action required for this consequence?

A. (1) CRD Mechanism drive seal degradation due to seal temperature above 450°F.

(2) Assign an action to engineering in PCRS to evaluate scram time.

B. (1) CRD Mechanism drive seal degradation due to seal temperature above 450°F.

(2) Assign an action to engineering in PCRS to evaluate HCU Accumulator pre-charge pressure.

C. (1) CRD Mechanism drive seal degradation due to constant application of charging header pressure while the tagout is hanging.

(2) Assign an action to engineering in PCRS to evaluate scram time.

D. (1) CRD Mechanism drive seal degradation due to constant application of charging header pressure while the tagout is hanging.

(2) Assign an action to engineering in PCRS to evaluate HCU Accumulator pre-charge pressure.

Answer: A Explanation: 04-1-01-C11-1, Control Rod Drive Hydraulic System, section 5.7, Total Isolation of an HCU, results in isolation of CRD cooling water to the CRDM.

Precaution 3.9 and 3.11 of 04-1-01-C11-1 warn against this because high CRDM temperature will result if RCS temperature is above 250°F. Alarm 1H13-P680-4A2-A4, CRD HYD TEMP HI would be expected for reduced cooling water flow to a CRDM at rated conditions. ARI 04-1-02-1H13-P680-4A2-A4 step 4.4 states if CRD temperature reaches 400°F, a CR should be initiated for tracking the condition by Engineering. ARI step 3.3 also states Engineering should be contacted to evaluate adding time to the previous scram time if the CRDM temperature reaches 450°F. The administrative mechanism for Engineering evaluations related to operability is the CR process in PCRS. Therefore, answer A is correct. Answer B is plausible because the listed consequence is correct and because HCU Accumulator pressure can mechanistically affect scram time. However, it is wrong because there is no process or requirement to vary HCU Accumulator pressure to quicken or reduce scram time.

Answers C and D are plausible because seal damage is a concern if insert riser valve 101XX is reopened before withdraw riser valve 102XX per 04-1-01-C11-1 step 3.7.

Seal damage is also a concern per 04-S-03-C11-4 if a control rod is scrammed from position 10 or less. Similarly for single rod scram, seal damage might concern the student who does not know the piping configuration for an HCU or that HCU isolation includes isolation of insert riser valve 101XX and withdraw riser valve 102XX, such that the CRDM is isolated from CRD charging water header. The second parts of answers C and D relate to seal health and are, therefore, plausible.

Technical

References:

04-1-01-C11-1 04-1-02-1H13-P680-4A2-A4 04-S-03-C11-4 References to be provided to applicants during exam: none Learning Objective: GLP-OPS-C111B Obj. 4.4,8.1,8.2,9 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X

Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(5)

Examination Outline Cross-Reference Level SRO 219000 RHR/LPCI: Torus/Suppression Pool Cooling Mode A2. Ability to (a) predict the impacts of the following on the K/A # 219000 RHR/LPCI: TORUS/SUPPRESSION POOL COOLING MODE; Rating 3.2 and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those Rev / Date 1 abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.03 Valve closures Question 92 The plant is at rated power.

RHR pump A is tagged out of service for maintenance.

Safety Relief Valve B21-F041F fails open.

RHR B is placed into Suppression Pool Cooling mode.

A spurious Division 2 ECCS initiation signal is received.

The RHR B/C Initiation signal can NOT be reset.

Suppression Pool temperature is 94°F.

(1) What is the status of RHR B for Suppression Pool Cooling mode?

(2) What action should be taken NEXT by the CRS for current conditions?

A. (1) Interlocks prevent opening 1E12F024B, RHR B TEST RTN TO SUPP POOL.

(2) Direct placing SRV hand switches to OFF for B21-F041F.

B. (1) Interlocks prevent opening 1E12F024B, RHR B TEST RTN TO SUPP POOL.

(2) Direct placing the Reactor Mode Switch to SHUTDOWN.

C. (1) 1E12F024B, RHR B TEST RTN TO SUPP POOL can be opened from 1H13P601.

(2) Direct placing the Reactor Mode Switch to SHUTDOWN.

D. (1) 1E12F024B, RHR B TEST RTN TO SUPP POOL can be opened from 1H13P601.

(2) Direct placing SRV hand switches to OFF for B21-F041F.

Answer: D Explanation: A division 2 ECCS signal will cause RHR B Return to Supp Pool valve E12-F024B to close, interrupting RHR B Supp Pool Cooling flow. The automatic valve closure can be manually overridden by placing the 1H13P601 hand switch for E12-F024B to Open. Therefore, RHR B Supp Pool Cooling can be aligned.

Ops Philosophy allows ROs to take actions without direction for a limited number of conditions. In the case above ROs and SROs should have the system knowledge required to know placing the SRV hand switches in off will (or should) cause the SRV to de-energize and shut. However, the SRO must make an operational decision in this case because this action will require entry into TS 3.5.1 Condition H which is a LCO 3.0.3 shutdown statement. Although the plant will begin preps for shutdown, this will allow the plant to perform a controlled shutdown vice scram the plant.

EP-3 step SPT 3 and 4 requires entry into EP-2 to effect a manual scram BEFORE, i.e.no later than when, Supp Pool temperature reaches 110°F. At the current suppression pool temperature, the SRO is expected to take the actions required to prevent an unnecessary plant transient (scram); however, EP-3 will be entered at 95F and the applicant may mistake that Entry condition for a need to scram the plant.

Technical

References:

EP-3 EP-2 E1181-68, 69 References to be provided to applicants during exam: none Learning Objective: GLP-OPS-E1200 Obj. 8.10, 20; GLP-OPS-EP3 Obj. 3,7,8 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(5)

Examination Outline Cross-Reference Level SRO 256000 Reactor Condensate System 2.4.1 Knowledge of EOP entry conditions and immediate K/A # 256000 / 2.4.1 action steps. Rating 4.8 (CFR: 41.10 / 43.5 / 45.13)

Rev / Date 0 Question 93 The plant is at rated power.

A spurious IP Condenser Hotwell Level Low signal, <1-11 wc, occurs due to a relay failure, indicated by amber alarm 1H13-P680-2A-E9, CONDSR HTWL LVL LO.

As a result, the highest priority for the CRS is to implement A. EP-2, RPV Control.

B. 05-1-02-III-3, Reduction in Recirculation System Flow Rate.

C. 05-1-02-V-7, Feedwater Systems Malfunctions.

D. ARI 04-1-02-1H13-P680-2A-E9, CONDSR HTWL LVL LO.

Answer: A Explanation: Low IP hotwell level is sensed by N19 level switches N105, N106, and N107, which actuate auxiliary relay N19 63X1/N105. This relay is a single point failure component that provides a common trip signal to all three Condensate Pumps.

If the relay contacts fail to the energized state, i.e. closed, all running Condensate Pumps trip, resulting in trip of all Condensate Booster Pumps and Reactor Feed Pumps. Plant data shows for this transient, reactor water level rapidly falls to around -

80 inches wide range, due to loss of all feed flow and due to shrink following the scram at reactor Level 3. EP-2 entry condition <11.4 reactor water level will be met within seconds of the initiating event. There would be no time for the RO or SRO to refer to the ARI before EP-2 execution is required. Additionally, no actions of the ARI would provide immediate, rapid recovery of Condensate in order to avert an automatic scram or HPCS/RCIC initiations. Therefore, any time before an automatic scram on low water level should be spent immediately effecting a manual scram per EN-OP-115 to protect the reactor when it is in jeopardy. EP-2 entry condition on low water level will be met for either an automatic or a manual scram, due to shrink. Emergency Procedure execution is higher priority than lower tier procedures. The lower tier procedures should only delay EP execution when they are needed to accomplish specific EP steps. Answer A is correct. All other answers are plausible because entry requirements for the listed procedures are met. However, for answers B and C, the reactor operators are responsible for performing immediate ONEP actions without requiring direction from the CRS, and though those actions may seem important to the initial license candidate, EP-2 is more important in the onset of an event to ensure the overall plant is stabilized. Answers B and C are wrong because ONEP subsequent action execution in this case is of lower priority than restoring and stabilizing reactor water level using the more comprehensive strategies given in EP-2.

Answer D is plausible because the alarm is the initial manifestation of the problem and because the ARI provides instructions for defeating the failed relay. However, EP-2 begins with particularly arranged steps necessary to ensure reactor and personnel safety under all conditions that must be addressed before focus is dedicated to individual systems or components. No actions of the ARI would provide immediate, rapid recovery of Condensate in order to avert an automatic scram or HPCS/RCIC initiations. Recovery of Condensate will be prioritized according EP-2 itself, and then the ARI would be executed to compliment/facilitate EP-2 steps. Therefore, answer D is wrong.

Technical

References:

04-1-02-1H13-P680-2A-E9 EP-2 05-1-02-III-3 05-1-02-V-7 E1148-15 References to be provided to applicants during exam: none Learning Objective: GLP-OPS-N1900 Obj. 13; GLP-OPS-EP2 Obj. 1,2,3,5,8 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(10) 55.43(b)(5)

Examination Outline Cross-Reference Level SRO 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

K/A # 2.1.23 (CFR: 41.10 / 43.5 / 45.2 / 45.6) Rating 4.4 Rev / Date 0 Question 94 Following a LOCA with limited fuel damage, the Shift Technical Advisor reports containment hydrogen concentration has reached 3%.

Based on this; A. EP-2, RPV Control, is exited and EP-3, Containment Control is continued.

B. EP-2, RPV Control, is exited, EP-5, RPV Flooding is entered, and EP-3, Containment Control, is continued.

C. EP-2, RPV Control, is continued, EP-3, Containment Control is exited, and the SAPs are entered.

D. All EPs are exited and all SAPs are entered.

Answer: D Explanation: This is incorrect because the EPs state if containment or drywell hydrogen concentration exceeds 2.9%, all the EPs are exited and all the SAGs are entered.

B. This is incorrect because the EPs state if containment or drywell hydrogen concentration exceeds 2.9%, all the EPs are exited and all the SAGs are entered C. This is incorrect because the EPs state if containment or drywell hydrogen concentration exceeds 2.9%, all the EPs are exited and all the SAGs are entered D. This is correct based on the discussion above..

Technical

References:

EP-3, Containment Control, step H-1.

References to be provided to applicants during exam: none Learning Objective: GLP-OPS-EP3 Obj. 8 Question Source: Bank # 2009 NRC Q#95 (note changes; attach parent) Modified Bank #

New Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(10) 55.43(b)(5)

Examination Outline Cross-Reference Level SRO 2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management.

K/A # 2.1.37 (CFR: 41.1 / 43.6 / 45.6) Rating 4.6 Rev / Date 0 Question 95 Power is to be reduced from 100% to 65% for a control rod sequence exchange.

Who is responsible for reviewing the Reactivity Maneuver Plan and ensuring that the control rod pull sheets are highlighted to emphasize areas of concern?

A. Reactor Engineering B. Field Support Supervisor C. Reactivity Management SRO D. Operations Manager Answer: C Explanation: 02-S-01-27, Operations Philosophy, classifies the described power change as a Type 3 power maneuver per step 6.8.1a(3), which requires staffing an additional SRO, the Reactivity Management SRO (RMSRO). Step 6.8.1b(2) states the RMSRO is responsible for reviewing the Reactivity Maneuver Plan and ensuring that the control rod pull sheets are highlighted to emphasize areas of concern. Step 6.8.2d states the RMSRO is responsible for instructing ROs on the execution of the specific pull sheets. Therefore, answer C is correct. Answer A is plausible since Reactor Engineers develop the control rod pull sheets and are involved in during the sequence exchange. Answer B is plausible since this is an on shift SRO who does not normally have the control room command function and might be considered capable of assuming dedicated reactivity management duties. Answer D is plausible because Ops Management is present in the control room for management oversight during power reductions for sequence exchanges.

Technical

References:

02-S-01-27, Operations Philosophy References to be provided to applicants during exam: none Learning Objective: GLP-OPS-PROC Obj. 4.10 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(1) 55.43(b)(6)

Examination Outline Cross-Reference Level SRO 2.2.35 Ability to determine Technical Specification Mode of Operation.

K/A # 2.2.35 (CFR: 41.7 / 41.10 / 43.2 / 45.13) Rating 4.5 Rev / Date 0 Question 96 Use your provided references to answer this question.

A refueling outage is in progress.

The following conditions exist:

RPV head bolt de-tensioning is in progress.

Primary and Secondary containment will be relaxed on the next shift.

A complete loss of Shutdown Cooling occurred.

The refuel floor supervisor has directed that all head bolts be re-tensioned Recirc loop temperature reached 201°F at 0630.

It is now 0743 with:

- Recirc loop temperature reached 202°F

- Two RPV head bolts remain de-tensioned (1) What EAL entry is required?

(2) What is the current Tech Spec MODE of Operation?

A. (1) CA3 - Alert (2) Hot Shutdown B. (1) CA3 - Alert (2) Refueling C. (1) CU3 - Unusual Event (2) Hot Shutdown D. (1) CU3 - Unusual Event (2) Refueling

Answer: B Explanation:

The given stem conditions places the plant in EAL CA3 because even though it suggests maintaining Cold Shutdown (Mode 4), it is applicable in Modes 4 & 5 (10-S-01-1 Att. 2 page 30 of 113, Rev 122).

The EAL is met because the RCS temperature has been above 200°F for 60 minutes.

CU3 is plausible based on misinterpreting the EAL conditions for CA3.

Hot Shutdown is plausible based on temperature being above 200°F; however, Mode 5 has no temperature requirements.

Technical

References:

10-S-01-1 flowchart TS table 1.1-1 References to be provided to applicants during exam: EAL flowchart page 2 of 2 Learning Objective: GLP-OPS-TS001 Obj. 5; GLP-EP-EPTS6, Obj. 1 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7,10) 55.43(b)(2)

Examination Outline Cross-Reference Level SRO 2.2.36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting K/A # 2.2.36 conditions for operations. Rating 4.2 (CFR: 41.10 / 43.2 / 45.13)

Rev / Date 0 Question 97 The plant is at rated power.

Corrective Maintenance is scheduled to replace an ECCS jockey pump.

Which of the following is required to be performed before the tagout lineup to drain the ECCS jockey pump piping necessary for this work?

A. Quarterly functional surveillance for the associated ECCS system valves B. Closed loop leakage test for jockey pump containment boundary valves C. Monthly lineup and fill and vent surveillance for the associated ECCS D. Quarterly functional surveillance for the associated jockey pump Answer: B Explanation: per 02-S-01-17 section 6.12, a closed loop leakage test is required before breaching the closed system outside containment to ensure the overall containment leak rate limit, La, is not exceeded. Otherwise, an LCO for Containment Integrity would have to be entered. Therefore, answer B is correct. Other answers are plausible because they are all necessary retests possibly required for this type of maintenance, but they are wrong since they are required post-maintenance, not pre-maintenance.

Technical

References:

TS 3.8.7 and bases TS 3.8.1 References to be provided to applicants during exam: none Learning Objective: GLP-OPS-TSLCO Obj. 14

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(10) 55.43(b)(2)

Examination Outline Cross-Reference Level SRO 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or K/A # 2.3.14 activities. Rating 3.8 (CFR: 41.12 / 43.4 / 45.10)

Rev / Date 0 Question 98 Operators are in EP-3 (CTMT Control) attempting to control a rising CTMT pressure.

The CRS has determined it is now necessary to vent the CTMT in order to keep CTMT pressure below 22.4 psig.

The EOF is fully operational.

Just before opening the 20 vents, the CRS/Shift Manager is required to directly notify the ________(1)________ that ________(2)_________ release of radioactivity will occur.

A. (1) Radiological Assessment Coordinator (2) a Filtered B. (1) Radiological Assessment Coordinator (2) an Un-filtered C. (1) Emergency Director (2) a Filtered D. (1) Emergency Director (2) an Un-filtered Answer: D Explanation: See EP-1, Attachment 13 (page 3 of 10), step 2.7, where direction is given to notify the Emergency Director that an Un-filtered release will begin, not the Radiological Assessment Coordinator and not filtered. Therefore, answer D is correct Radiological Assessment Coordinator is a plausible distracter because that position is

responsible for overall dose assessment.

Filtered release is a plausible distracter because the smaller 6 containment vent path (Low Volume Purge) for EP Att. 14 vents through the Ctmt exhaust charcoal filter train.

Technical

References:

EP-1, Attachment 13 - Containment Venting/Defeating Containment Vent Path Isolation Interlocks References to be provided to applicants during exam: none Learning Objective: GLP-EP-EPTS26 Obj. 40 Question Source: Bank #

(note changes; attach parent) Modified Bank # 408 X New Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(12) 55.43(b)(4)

Examination Outline Cross-Reference Level SRO 2.4.11 Knowledge of abnormal condition procedures.

(CFR: 41.10 / 43.5 / 45.13)

K/A # 2.4.11 Rating 4.2 Rev / Date 0 Question 99 Use your provided references to answer this question.

The plant is operating at rated power.

Feedwater and Condensate conductivity is trending up but do not exceed limits.

Reactor Water Conductivity is 1.5 umho/cm.

The CRS will A. (1) Enter Condensate/Reactor Water High Conductivity ONEP and TRM 6.4.1 (2) Restore reactor chemistry to within limits in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B. (1) Enter Condensate/Reactor Water High Conductivity ONEP and TRM 6.4.1 (2) Manually scram the reactor C. (1) Enter TRM 6.4.1, Only (2) Restore reactor chemistry to within limits in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> D. (1) Enter Condensate/Reactor Water High Conductivity ONEP, Only (2) Manually scram the reactor Answer: A Explanation:

TRM 6.4.1 is entered in Mode 1 with Conductivity 1.0 umho/cm. The Condensate/Reactor Water High Conductivity ONEP directs actions with as little as 0.3 umho/cm so it also requires entry.

The TRM action in this case is to restore within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The ONEP requires a scram when conductivity exceeds 5 umho/cm.

Technical

References:

TRM 6.4.1, Chemistry 05-1-02-V-12, Condensate/Reactor Water High Conductivity ONEP References to be provided to applicants during exam: TRM 6.4.1 without TR Table 6.4.1-1 Learning Objective: GLP-OPS-TS001 obj 4 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b) 55.43(b)(5)

Examination Outline Cross-Reference Level SRO 2.4.25 Knowledge of fire protection procedures.

(CFR: 41.10 / 43.5 / 45.13)

K/A # 2.4.25 Rating 3.7 Rev / Date 0 Question 100 Use your provided references to answer this question.

Painters are covering one smoke detector in the Control Building per an approved work order.

All fire rated assemblies in the area are OPERABLE.

All other smoke detectors and all heat detectors in the Control Building are OPERABLE.

What is the minimum action (i.e. least manpower and material required) required by the TRM for this activity?

A. Establish an hourly fire watch patrol, only, for the affected area within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

B. Establish an hourly fire watch patrol with backup fire suppression for the affected area within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

C. Establish a continuous fire watch, only, for the affected area within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

D. Establish a continuous fire watch with backup fire suppression for the affected area within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Answer: A Explanation: Per 10-S-03-1, Fire Protection System Impairment, step 2.4, the Shift Manager is responsible for ensuring LCO requirements are met. A smoke detector is a Function A device per TR 6.2.1, and covering one causes that detector to be inoperable. TR 6.2.8 Action A.1 requires a hourly fire watch patrol; therefore, answer

A is correct. Answer C is plausible because a continuous firewatch may be required per TR 6.2.8 based on fire rated assembly operability in association with fire detection instrumentation operability. But here, fire rated assemblies are operable, so only an hourly FW is required. And although a continuous FW would satisfy the purpose of an hourly FW patrol, the stem asks for the minimum action, which would be an hourly patrol rather than an individual dedicated to one specific area. Answers B and D are plausible because inoperability of some fire detection instrumentation also inops fire suppression systems, such as CO2, sprinklers, and halon, which would require backup fire suppression, if affected, per TR 6.2.3, TR 6.2.4, TR 6.2.5. However, only Function B detectors initiate these systems and smokle detectors are Function A; therefore, answers B and D are wrong.

Technical

References:

10-S-03-1, Fire Protection System Impairment TR 6.2.1 TR 6.2.8, TR 6.2.3, TR 6.2.4, TR 6.2.5 References to be provided to applicants during exam: TR 6.2.1 Learning Objective: GLP-OPS-FIRELCO Obj. 1,19,34 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(10) 55.43(b)(5)

Filename: Final SRO NRC exam Rev 1.docx Directory: C:\NRC EXAMS\2013\Initial\NRC\Written Template: C:\Documents and Settings\Administrator\Application Data\Microsoft\Templates\Normal.dotm

Subject:

Author: NRC EXAM Keywords:

Comments:

Creation Date: 9/17/2013 1:47:00 PM Change Number: 28 Last Saved On: 10/8/2013 11:28:00 AM Last Saved By: NRC EXAM Total Editing Time: 1,295 Minutes Last Printed On: 10/8/2013 2:17:00 PM As of Last Complete Printing Number of Pages: 58 Number of Words: 10,909 (approx.)

Number of Characters: 62,183 (approx.)