CNS 2012 NRC Exam 100 Questions Final Submittal Question 76 … · 2014. 10. 14. · CNS 2012 NRC...

Question 7601 5AG2.4.46RCP MalfunctionsAbility to verify that the alarms are consistent with the plant conditions.

Given the following Unit I conditions:

lnitial• The Unit is in Mode 3.• NC System operational leakage is:

• 0.1 gpm unidentified• 1.83 gpm identified

Current:• NC Pump #1 Seal Leakoff flows are:

• IA NCP is 4.0 gpm and slowly increasing.• I B NCP is 2.9 gpm and stable.• IC NCP is 3.0 gpm and stable.• 1D NCP is 3.1 gpm and stable.

• The following alarm then annunciates:

IAD-7, C/i, NCP #1 Seal Leakoff Hi Flow,

If IA NCP #1 Seal Leakoff flow is AT the ahwhich equipment, if any, must be declared’

Standby Shutdown Facility (SSF)

non-functional

non-functional

functional

C —

i. /‘‘

_-7

Page 166 of 235Catawba 2012 NRC Exam Submittal

S.

CNS 2012 NRC Exam 100 Questions Final Submittal

\

I

describes

A.

B.

C.

fünàtional

functional

non-functional

non-functional N

functional


QUESTION 76

Distractor Analysis

A. CORRECT. 1AD-7, C/I, alarm response forjf_#1 seal leakoff high flow containsguidance to conservatively declare the SSR1operle’)as a Supplementary Action of thisalarm response. There is also guidance to declare tfFe Standby Makeup Pump inoperableIF total NC system leakage exceeds 20 gpm. In this case, it has not, and only the SSF isdeclared inoperable.

B. Incorrect. First part is correct. The Standby Makeup Pump being declared non-functionalis plausible: It is powered by a DIG in the SSF, and is therefore, related to the “status” ofthe SSF. If the SSF is non-functional, it is reasonable to believe that the Standby MakeupPump POWERED FROM the SSF DIG would also be non-functional.

C. Incorrect. The applicant could easily confuse and reverse the effect of NCP seal interfaceand select this answer.

D. Incorrect: Second part is correct. If the SSF is functional, due to similar reasoning asdescribed in “B” above, it is plausible that the Standby Makeup Pump (powered from theSSF DIG) would also be not affected (functional).

References:• Tech. Spec. 3.4.13, RCS Operational Leakage• IAD-7, CII, NCP #1 Seal Leakoff Hi Flow• IAD-7, CI4, NCP Seal Water Lo Flow (for plausibility of 7 gpm)• OP-CN-PS-NCP, Lesson Plan for NC Pumps,• SLC 16.7-9, Standby Shutdown System

KA Match:Question 7601 5AG2.4.46RCP MalfunctionsAbility to verify that the alarms are consistent with the plant conditions.Applicant is presented with multiple plant conditions, primarily related to seal leakoff, and thenmust evaluate if an alarm is consistent with the conditions given for the seal leakoffs. Further,must also determine the effect of the alarm condition on additional components related to thatfunction (RCP seals).

Cognitive Level: HighThis is a higher cognitive level question because the applicant evaluates multiple plantconditions, including an alarm, and must make a conclusion and a determination regardingfunctionality of components.

Source of Question: NEW

SRO Only:

Page 167 of 235Catawba 2012 NRC Exam Submiftal


This question meets the following criteria for an SRO only question as described in the“Clarification Guidance for SRO-only Questions (Rev. I dated 03/11/2010) under the ScreenCriteria for questions linked to I OCFR55.43(b)(2) (Tech Specs):

1. It cannot be answered solely by knowing 1 hour TS/ SLC Action.2. It cannot be answered solely by knowing the LCO/SLC information listed above the line.3. It cannot be answered solely by knowing the TS Safety Limits.4. The question involves application of required actions of SLC 16.7-9, Standby Shutdown

System.

Therefore, this is an SRO only question.


OP/1/B/6100/O1OHPANEL: 1AD-7 Page 19 of 64

NCP #1 SEAL LEAKOFF HI FLOW C/iSETPOINT: 5.0 gpm

ORIGIN: Instrument DCS Description1NVFT5151 1NVAA5151 NC PUMP A SEAL LEAKOFF FLOW HI1NVFT5 141 1NVAA5 141 NC PUMP B SEAL LEAKOFF FLOW HI1NVFT5131 1NVAA5131 NC PUMP C SEAL LEAKOFF FLOW HI1NVFT5 121 1NVAA5 121 NC PUMP D SEAL LEAKOFF FLOW HI

PROBABLE 1. Damaged #1 SealCAUSE: 2. Cocked #1 Seal

3. Loss of injection water followed by high seal temp4. Hi temperature of injection water

AUTOMATIC NoneACTIONS:

IMMEDIATE 1. Identify the affected pump from one of the following:ACTIONS: • DCS graphic 6009, NV - NC Pump Seal Injection

• DCS alarm screen2. Refer to AP/1/A/5500/008 (Malfunction of Reactor Coolant Pump).3. Verify total NC leakage is less than 20 gpm to ensure operability of

the standby makeup pump per PT/1/A/4150/OO1D (NC SystemLeakage Calculation).

NOTE: The SSF is conservatively declared inoperable due to the potential for exceeding the NCPump seal cooling capacity of the Standby Makeup Pump. {PIP 96-191 0}

SUPPLEMENTARY 1. Declare the SSF inoperable. {PIP 96-1910}ACTIONS: 2. Notify Engineering to begin Operability determination process per

NSD 203 (Operability/Functionality). {PIP 96-1910}3. Dispatch an operator to 1RFM-12 on 1RFMP1 (AB-574, BB-55, Rm

491) to acknowledge the alarm.

REFERENCES: 1. SLC 16.7-92. CN-1499-NV-33. CNM-1201.Ol-1574. CNM 1399.03-0269.001 Drop 6 Sheet 3175. CNM 1399.03-0269.001 Drop 8 Sheet 3126. CNM 1399.03-0269.001 Drop 12 Sheet 3197. CNM 1399.03-0269.001 Drop 7 Sheet 323


Question 77038EG2.4.46Steam Gen. Tube RuptureAbility to verify that the alarms are consistent with the plant conditions.


The Unit I is initially at 100% power when the I D S/G steam line break)inside the doghouse.

•-

r • Total CA flow = 890 gpm.• I EMF-33 (Condenser Air Ejector Exhaust) Trip 2 actuated.• 1EMF-71 (SIG A Leakage) Trip 2 actuated.• IEMF-74 (S/GD Leakage) Trip 2 actuated c

• S/G indications are:

SIG NIR Level Pressure

lÀ 23% and increasingI B 15% and decreasingIC 18% and decreasing71D) 0%

Which ONE of the following procedures or enclosures contains the steps for throttling CA flow tothe S/G that FIRST requires CA flow throttling?

A. E-0, (Reactor Trip or Safety Injection), Enclosure I (Foldout Page)

B. E-0 (Reactor Trip or Safety Injection), Enclosure 4 (NC Temperature Control)

C. E-3 (Steam Generator Tube Rupture)

D. E-2 (Faulted Steam Generator Isolation)


7

0 -)

750 psig and increasing720 psig and decreasing700 psig and decreasing200 psig and decreasing


QUESTION 77

Distractor Analysis

A. CORRECT. Actions can be taken in E-0 to isolate a suspected ruptured steam generator.There are 2 indications that IA is ruptured. Level is not decreasing like the others, andEMF indications. It requires throttling CA flow first, according to the flow of the procedures.

B. Incorrect. This would be correct if not for the tube rupture.

C. Incorrect. This is how a SGTR is isolated if it is diagnosed later.

D. Incorrect: This is where a faulted SIG would normally be isolated.

References:• E-0, (Reactor Trip or Safety Injection), Enclosure I (Foldout Page), Revision 040• E-0 (Reactor Trip or Safety Injection), Enclosure 4 (NC Temperature Control)• E-3 (Steam Generator Tube Rupture), Revision 040• E-2 (Faulted Steam Generator Isolation), Revision 013

KA Match:Question 77038EG2.4.46Steam Gen. Tube RuptureAbility to verify that the alarms are consistent with the plant conditions.This question matches the KA because the applicant is presented with several radiationmonitors in an alarm condition, and plant parameters involving a steam generator tube rupture.and then must analyze these conditions to determine a course of action that applies to thediagnosed plant conditions, reflected by choice of procedure guidance.

Cognitive Level: HighThe applicant is presented with several radiation monitors in an alarm condition, and plantparameters involving a steam generator tube rupture. and then must analyze these conditions todetermine a course of action that applies to the diagnosed plant conditions, reflected by choiceof procedure guidance

Source of Question: Bank CNS 877

SRO Only:This question meets the following criteria for an SRO only question as described in the“Clarification Guidance for SRO-only Questions (Rev. I dated 03/11/2010) under the ScreenCriteria for questions linked to I OCFR55.43(b)(5) (Assessment and Selection of Procedures):

1. It cannot be answered solely by knowing “systems knowledge”, i.e., how the systemworks, flowpath, logic, component location.

2. It cannot be answered solely by knowing immediate operator actions.3. It cannot be answered solely by knowing entry conditions for AOPs or plant parameters

that require direct entry to major EOPs.



4. It cannot be answered solely by knowing the purpose, overall sequence of events, oroverall mitigative strategy of a procedure.5. The question does involve assessing plant conditions (involving &O content), and thenselecting a section (specific step for a specific purpose) to mitigate a Steam GeneratorTube Rupture.



CNS REACTOR TRIP OR SAFETY INJECTION PAGE NO.EPIIIAJ5000!E-0 31 of 46Enclosure 1 - Page 1 of 2

Revision 40Foldout Page

if any SIG(s) suspected ruptured, THEN perform the following:

• WHEN the following conditions met:

• Total CA flow - GREATER THAN 450 GPM

AND

• All intact SIG(s) N/R level - GREATER THAN 11 %(29% ACC)

THEN THROTTLE feed flow to ruptured SIG(s) to maintain ruptured SIG(s) N/R level between11%(29% ACC) and 39%.

2. NC Pump Trip Criteria:

• if the following conditions are satisfied, THEN trip all NC pumps while maintaining sealinjection flow:

• Any NV or NI pump - ON

• NC subcooling based on core exit T/Cs - LESS THAN OR EQUAL TO 0°F.

3. CA Suction Source Switchover Criterion:

• IF 1AD-8, B/I “UST LO LEVEL” is lit, THEN REFER TOAP/1/A/5500/006 (Loss of S/GFeedwater).

4. Position Criteria for INV-202B and INV-203A (NV Pumps A&B Recirc Isol):

• IF NC pressure is less than 1500 PSIG AND NV S/I flowpath is aligned, THEN CLOSEI NV-202B and I NV-203A.

• if NC pressure is greater than 2000 PSIG, THEN OPEN 1 NV-202B and I NV-203A.

5. Cold Leg Recirc Switchover Criterion:

• IF FWST level decreases to 20% (IAD-9, D/8 “FWST 2/4 LO LEVEL”), AND S/I has occurred,THEN TO EP/1/A15000/ES-1 .3 (Transfer To Cold Leg Recirculation).

-

EXAM BANK - Q 877Unit 1 is at 100% RTP when the ID SIG steam line breaks inside the doghouse.Given the following:

• Required immediate actions have just been completed• TotalCAflow=89Ogpm• I EMF-33 (Condenser Air Ejector Exhaust) Trip 2 actuated• 1 EMF-71 (S/G A Leakage) Trip 2 actuated• 1 EMF-74 (S/G D Leakage) Trip 2 actuated• All equipment responded as expected• S/G indications are as follows:

SIG N/R Level PressureA 17% and stable 750 PSIG and decreasingB 15% and decreasing 720 PSIG and decreasingC 18% and decreasing 700 PSIG and decreasingD 0% 200 PSIG and decreasing

Which SIG will have auxiliary feed water throttled to it FIRST and by what procedureand/or enclosure?A. Throttle to 1A S/G per Enclosure I of EP/1/A!5000/E-0 (Reactor Trip or Safety

Injection)

B. Throttle to ID S/G per Enclosure 4 of EP/1/A/5000!E-0 (Reactor Trip or SafetyInjection)

C. Throttle to IA S!G per EPIIIAI5000IE-3 (Steam Generator Tube Rupture)

D. Throttle to I D S/G per EPII/AI5000IE-2 (Faulted Steam Generator Isolation)


Question 78054AA2.02Loss of Main FeedwaterAbility to determine and interpret the following as they apply to the Loss of MainFeedwater (MFW): -

Differentiation between loss of all MFW and trip of one MFW pump. ‘

Which ONE of te following describes the EARLIEST NRC Notification Requirements for theloss of ONE CF Pump as compared to the loss of BOTH CF Pumps? (Assume initial conditionsat 80% power.)

A. For ONLY)he loss of BOTH CF Pumps, the notification requirement is within 8 hours.

B. For EITHER the loss of ONE or BOTH CF Pumps, the notification requirement is within 8hours. --

C. For ONLY the loss ofONE CF Pump, the notification requirement is within 24 hours.

D. For ONLY the loss of BOTH CF Pumps, the notification requirement is within 4hours.

/ G



QUESTION 78

Distractor Analysis

A. Incorrect. Plausible, since this could be interpreted as the resulting plant conditions(reactor trip) warranting the “ESF Actuation” category, which requires an 8 hour notificationto the NRC Operations Center.

B. Incorrect. 8 hours is explained in “A” above. Plausible that the loss of any main feed pumpwith the plant at 80% power would warrant notification, since it is an operationallysignificant event and does place the plant in a transient. Notification is not required foreither, though it is for both CF pumps ONLY.

C. Incorrect. The loss of ONE CF Pump is an operationally significant event, and it isplausible to believe that notification is required, and since the time listed is a relatively longtime period (24 hours), an applicant could believe this to be the correct answer.

D. CORRECT. An RPS actuation is a 4 hour notification; since loss of both CF pumpscauses a turbine trip, which results in a reactor trip.

References:RP/0/B/5000/1 3, (NRC Notification Requirements), Revision 032

KA Match:Question 78054AA2.02Loss of Main FeedwaterAbility to determine and interpret the following as they apply to the Loss of MainFeedwater (MFW):Differentiation between loss of all MFW and trip of one MFW pump.The KA is matched because the applicant must distinguish the consequences of the loss of asingle main feed pump, as compared to the loss of both main feed pumps, in the context ofNRC notifications.

Cognitive Level: HighAt first glance, this may appear to be a low cognitive level question, since there is some recallinvolved. However, there is more than one mental step involved in arriving at the correctanswer; applicant must first recall the notification requirements for ONLY the loss of both mainfeed pumps, but then apply the knowledge of what it means to the reactor when this happens,and then apply THAT requirements.

Source of Question: NEW -

SRO Only:This question meets the following criteria for an SRO only question as described in the“Clarification Guidance for SRO-only Questions (Rev. 1 dated 03/11/2010) under the ScreenCriteria for questions linked to IOCFR55.43(b)(1), (Conditions and Limitations in the facilitylicense):

1. It involves NRC reporting requirements for a loss of main feedwater.


Enclosure 4.3 1/O/B/5ooo/o13Events Requiring 4-HOUR NRC Notification Page 2 of 3

Complete the reporting requirements for the following events as soon as practical and in all cases within 4 hours after the occurrence becomesknown to the licensee:

1 OCFR Section Event Description Reporting Requirement1OCFR5O.72(b)(2)(iv)(A) Any event that results or should have resulted in ECCS discharge into the Notify the NRC Operations Center

reactor coolant system as a result of a valid signal except when theECCS discharge into actuation results from and is part of a pre-planned sequence during testingthe Reactor Coolant or reactor operation.System

• Valid signal refers to those signals automatically initiated bymeasurement of an actual physical system parameter that was withinthe established setpoint band of the sensor that provides the signal tothe protection system logic, or manually initiated in response to plantconditions. Valid signals also include passive system actuations thatoccur as a function of system conditions like differential pressure (i.e.,cold leg accumulators) whereby no SSPS or other electrical signal isinvolved. The validity of an ECCS signal may not be determinedwithin 1 hour; ECCS signals that result or should have resulted ininjections should be considered valid until firm evidence provesotherwise.

. Invalid ECCS injections are still considered a System actuation, but areNOT reportable to the NRC per 10 CFR 50.72. It is still reportableunder 10 CFR 50.73 as an LER. (Refer to Enclosure 4.8 for guidanceas_to_what_constitutes_a_System_actuation.)

1OCFR5O.72(b)(2)(iv)(B) Any event or condition that results in actuation of the reactor protection Notify the NRC Operations Centersystem (RPS) when the reactor is critical except when the actuation is part

RPS Actuation of a pre-planned sequence during testing or reactor operation.1OCFR5O.72(b)(2)(xi) Any event or situation related to the health and safety of the public or on- Notify the NRC Operations Center1OCFR72.75(b)(2) ISFSI site personnel, or protection of the environment, for which a news release is

planned or notification to other government agencies has been or will beOffsite Notification made. Such an event may include an on-site fatality, transport of an injured

(News Release) or ill employee to a hospital by ambulance, or an inadvertent release ofradioactively contaminated materials.

Enclosure 4.4 RP/O/B/5000/o13Events Requiring 8-HOUR NRC Notification Page 1 of 2

Complete the reporting requirements for the following events as soon as practical and in all cases within 8 hours after the occurrence becomes

known to the licensee: p \ic’s ,

1 OCFR Section Event Description Reporting Requirement

1OCFR5O.72(b)(3)(ii) Any event or condition that results in: Notify the NRC Operations Center

A. the condition of the plant, including its principal safety barriers, being

Degraded Condition seriously degraded orB. The Nuclear Power plant being in an unanalyzed condition that

significantly degrades plant safety.

1OCFR5O.72(b)(3)(iv)(A) Any event of condition that results in valid actuation of any of the systems Notify the NRC Operations Center

listed in Enclosure 4 8 of this procedure, except when the actuation resultsSystem Actuation from and is part of a pre-planned sequence during testing or plant( ISt ao’ci. operation.1OCFR5O.72(b)(3)(v) IOS Any event or condition that at the time of discovery could have prevented Notify the NRC Operations Center

the fulfillment of the safety function of structures or systems needed to:

Safety FunctionPrevented From A. shut down the reactor and maintain it in a safe shutdown condition,Functioning

B. remove residual heat,

C. control the release of radioactive material, or

D. mitigate the consequences of an accident1OCFR72.75(c)(1) Discovery of a defect in an ISFSI structure, system or component that is Notify the NRC Operations Center

important to safety.ISFSI Defects1OCFR72.75(c)(2) A significant reduction in the effectiveness of the ISFSI confinement Notify the NRC Operations Center

system.

ISFSI DegradedConfinement System

Enclosure 4.8RP/O/B/5000/o13

List of System (ESF) Actuations for Catawba Page 1 of 2Any reactor trip (P-4)• Refer to Enclosure 4.3, page 2 of 3, if this trip is an RPS Actuation.

2. Safety injection (UFSAR 6.3.1, 6.3.2)

A. NV charging path

B. NI charging path

C. ND charging path

D. CLA injection

E. D/G sequencer activation

F. Reactor trip signal

0. FWST - containment sump ND suction swap• If a second NV pump is manually started in order to maintain NC inventory, this is also a

system actuation.

3. Containment spray (UFSAR 6.2.2)

A. NS pump start/valve alignment

B. Actual spraydown of containment

4. Containment isolation (UFSAR 6.2.4)

A. Phase A (St)

B. Phase B (Sp)

C. Closure of the VP or VQ valves upon receipt of a high radiation signal from EMF-38, 39, or 40does not constitute a reportable system actuation during any mode.

D. NW system injection

5. Steam line isolation (UFSAR 10.3.2)

A. Individual steam line valve closure*

B. System isolation

C. Actuation of P-12 to close steam dumps is NOT a system actuation

* Individual component activation due to component failure not reportable per this requirement

Enclosure 4.8 RP/O/B/5000/ol3List of System (ESF) Actuations for Catawba Page 2 of 2

1% * U6. Auxiliary feedwater system 1 ‘*\‘4 +rs s oit’i ii c ijcti

A. Auxiliary feedwater pump start, automatic or manual, unless the start was the expected resultof a controlled (documented) test or procedure.

Example: A feedwater transient is in progress with S/G levels decreasing toward the reactortrip setpoint. If the operator starts a CA pump(s) to supplement CF flow and prevent the trip,the start is reportable under the 8-hour NRC notification criterion.

B. Pump suction swap to RN

7. Emergency AC Electrical Power Systems

A. Diesel Generator starts, automatic or manual, unless the start was the expected result of acontrolled (documented) test or procedure.

8. Ice condenser lower inlet door opening as a result of unplanned mass or energy release intocontainment

A. Door openings resulting from planned evolutions such as containment ventilation fan starts,personnel entries into containment, etc., do not constitute system actuations.

9. Combustible Gas Control in Containment

A. Containment air return and hydrogen skimmer (VX) operation (UFSAR 6.2.5.2)

1. Any unanticipated system operation

B. Hydrogen Recombiners (UFSAR 6.2.5)

C. Hydrogen Purge (UFSAR 6.2.5)

D. Hydrogen Igniters (UFSAR 6.2.5)

Question 79055EA2.02Station BlackoutAbility to determine or interpret the following as theappIy_toaiQn Blackout:RCS core cooling through natural circulation cooling to S!G cooling

)

Given the following Unit 1 conditions:

Initial:• With the Unit initially at 100% power, a complete loss of switchyard occurred.• Both DIGs FAILED to start, and will not start manually.• ECA-0.0, (Loss of All AC Power) has been implemented.• NO NC Pumps are available.

Current:• NC subcooling is 5°F• PZR level is 19%. /• lNl-9A (NV Pmp CIL lnj lsol) is CLOSED.• lNl-IOB (NV Pmp C/L lnj lsol) is CLOSED.• In accordance with ECA-0.0 the SRO is evaluating the availability of the following power

sources:Offsite Power from UnitOffsite Power from Unit 2D/G IA

• DIGIB

(1) To establish plant conditions and restore equipment needed for natural circulation theSRO will remain in ECA-0.0 until power is available from (1) of theabove powersources.

S

(2) Once the required power sources are available the SRO will then GO TO

implement procedures to establish plant conditions and restore equipment needed for naturalcirculation?

A. (1) AT-LEAST iWO(2) ECA-0.1, (Loss of All AC Power Recovery Without S/I Required).

B. (i)A]LEAS]flrWO ,, ! J.I

(2) ES-O 2, (Naturii Circulation Cooldown). 1/

C. (1) ,ANONE(2) ECA-0.1, (Loss of All AC Power Recovery Without SIl Required).

N

D (1) ANYONE(2) ES-0.2, (Natural Circulation Cooldown).



ccv -

I

(2)


QUESTION 79

Distractor Analysis

A. Incorrect. Second part (procedure transition) is correct. Believing that you need to restorepower from at least TWO sources in the list in ECA-0.O Step 17 is plausible by reasoningthat it is desirable to have more than a single source of power, particularly duringemergency conditions.

B. Incorrect. Plausibility of at least two power sources restored is described in “A” above.Plausibility of ES-O.2 is described in “D” below.

C. CORRECT. ECA-O.O, Step 17, directs that procedure implementation of ECA-0.0 mustcontinue until power is restored from at least ONE of a list of sources, per the step. Withthe conditions given in the stem meeting all the criteria of Step 42, the crew is thendirected to go to ECA-0.1, Loss of All AC Power Recovery Without S/I Required.

D. Incorrect. First part is correct. ES-0.2, Natural Circulation Cooldown, is plausible sincethat procedure would be used DURING a natural circ cooldown, but the question is testingprocedure transition knowledge by asking where do you go once you have restored powerfrom at least ONE of the listed sources in Step 17 of ECA-0.0. Further, there is NO placein ECA-0.0 that directs a transition directly to ES-0.2 for natural circulation. First, you aredirected to transition to ECA-0.1 which contains additional actions for restoring loadsimportant to plant safety, including needed for establishing natural circulation.

References:• ECA-0. 1, (Loss of All AC Power Recovery Without S/I Required), Revision 026• ES-0.2 (Natural Circulation Cooldown), Revision 023• ECA-0.0, (Loss of All AC Power), Step 17, Revision 045

KA Match:Question 79055EA2.02Station BlackoutAbility to determine or interpret the following as they apply to a Station Blackout:RCS core cooling through natural circulation cooling to SIG coolingThis KA is matched because the applicant is tested on plant conditions involving a StationBlackout (all SWYD power lost AND no D/Gs available), and then to determine at an SROlevel) what is needed, including procedures that are appropriate, for establishing core coolingusing natural circulation.

Cognitive Level: HighThis is a higher cognitive level question because the applicant must first determine from givenconditions that a Station Blackout has occurred, and then apply the effect of that condition tosubsequent plant conditions, and then determine which procedures, including neededtransitions, will be used to establish natural circulation.




SRO Only:This question meets the following criteria for an SRO only question as described in the“Clarification Guidance for SRO-only Questions (Rev. 1 dated 03/11/2010) under the ScreenCriteria for questions linked to I OCFR55.43(b)(5) (Assessment and Selection of Procedures):



that require direct entry to major EOPs.4. It cannot be answered solely by knowing the purpose, overall sequence of events, or

overall mitigative strategy of a procedure.5. The question does involve assessing plant conditions for a Station Blackout, and then

selecting a section (specific steps for a specific purpose) to mitigate a Station Blackoutand to establish natural circulation.



CNS LOSS OF ALL AC POWER PAGE NO.EPII/AI5000IECA-O.O 12 of 173Revision 45

I ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED

NOTE • Offsite power may be unavailable for reasons other than switchyardde-energized.

o SATA(B) may be available, even if currently in service on the opposite Unit.

17. Verify at least ne f the following power Perform the following:sources availa

— a. WHEN at least one power source is—• Offsite Power from Unit 1 available, THEN perform Step 18.

OR b. GO]EQStepl9.

—e Offsite Power from Unit 2

OR

• D/G1A

OR

• DIG lB.

CNS LOSS OF ALL AC POWER PAGE NO.EP/1/AI5000IECA-O.0 43 of 173

Revision 45

I ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED

NOTE If NC pump seal cooling was previously isolated, further cooling of the NC pumpseals will be established by natural circulation cooldown as directed insubsequent procedures.

42. Select recovery procedure as follows:

— a. Verify NC subcooling based on coreexit T/Cs - GREATER THAN 0°F.

— b. Verify Pzr level - GREATER THAN 11%(30% ACC).

c. Verify the following valves - CLOSED:

—. 1 Nl-9A (NV Pmp C/L In] Isol)—

. 1NI-IOB (NV Pmp C/L Inj Isol).

— d. GOTOEP/1/N5000/ECA-0.1 (LossOf All AC Power Recovery WithoutS/I Required).

— a. GOTO EP/1/N5000!ECA-0.2 (LossOf All AC Power Recovery With S/IRequired).

— b. GOTOEP/1/A15000/ECA-0.2 (LossOf All AC Power Recovery With S/IRequired).

— c. if any NV pump on, THEN QIEP/1/N5000/ECA-0.2 (Loss Of AllAC Power Recovery With S/IRequired).

END

rnEXAM BANK- Q503

Initial conditions:

• Unit I had a complete loss of switchyard• The crew was performing steps in EPIIIAI5000IES-O.2, NaturalCirculation

Cooldown• Station management recommended a rapid cooldown due to secondary

inventory concerns• The crew transitioned to EPII/N5000IES-O.3, Natural Circulation Cooldown

with Steam Void in the Vessel

Current conditions:

• During the cooldown, a steam bubble formed in the reactor vessel• Reactor vessel Upper Range (UR) level is 92%.• The STA notes a YELLOW path on NC INVENTORY and confers with the

OSM regarding the need to transition to EP/1/A/5000/FR-l.3, Response toVoids in Reactor Vessel.

Which one of the following is the correct action to control void growth such that naturalcirculation is not interrupted, and which procedure will be used to accomplish this?A. Open reactor vessel head vents per EP/IIAI5000IFR-l.3.

B. Open reactor vessel head vents per EP/IIAI5000IES-O.3.

C. Energize pressurizer heaters per EP/I IN5000IFR-l .3.

D. Energize pressurizer heaters per EP/IIA/5000IES-O.3.


Question 80058AG2.1.28Loss of DC PowerKnowledge of the purpose and function of major components and controls.

Given the following conditions on Unit 1:

• The unit was at 100% power when a total loss of onsite and offsite power occurred.)

(1) Which procedure contains the instructions for the voltage value on the DC Vjtalbforwhen the Vital Batteries (EBA, EBB, EBC, EBD) are required to be removd from service?

(2) After power is restored and the battery chargers are placed in service, in accordance withTech Spec 3.8.4 (DC Sources — Operating), what is the MlNjMUfcIvoltage required for theVital Batteries to be OPERABLE-While at-harg3--

—-——-----.

A (1) APIOO7, (Loss of Normal Power) —

(2) 125 volts

B. (1) AP129, (Loss of VitaJr Aux Control Power)(2) 125 volts

C. (1) AP!007, (Loss of Normal Power)(2) 110 volts

D. (1) AP/29, (Loss of Vital orAux Control Power)(2) 110 volts



QUESTION 80

Distractor Analysis

A. Incorrect. AP/007, (Loss of Normal Power) is plausible since there are numerousinstructions for batteries, including in Case II, “Loss of All Power to an Essential Train”,Step I 3.c, the CAUTION just prior to the step which contains instructions regarding batterydepletion. There are additional instructions elsewhere in the procedure for StandbyShutdown Facility batteries, and for Switchyard batteries. Enclosure 17, “SwitchyardBattery Conservation) contains detailed instructions (Step 6 and 7) for batteries.

B. CORRECT. AP129, Enclosure 1, Step 5.e requires separating the battery from the DC buswhen battery voltage decays to 105 VDC, by opening the associated battery outputbreaker.

C. Incorrect. Plausibility for AP/007 is described in “A” above. 110 volts could be reasoned aminimum operability required since it is approx. 10% below the float voltage.

D. Incorrect: Procedure is correct. 110 volts plausibility described in “C” above.

References:• AP/1/A/5500/007, (Loss of Normal Power), Revision 66• AP!1/N5500/029, (Loss of Vital or Aux Control Power), Revision 024

KA Match:Question 80058AG2.1 .28Loss of DC PowerKnowledge of the purpose and function of major components and controls.

The KA is matched because the plant conditions involve a blackout which means the DCsystem is now the source of any power. The applicant is tested on what procedure contains theguidance for removing the batteries from service when they are depleted (below a certainvoltage), which is a form of loss of DC. The system function and purpose aspect is tested at theSRO level by asking what voltage is required to determine battery operability.

Cognitive Level: Low


SRO Only:This question meets the following criteria for an SRO only question as described in the“Clarification Guidance for SRO-only Questions (Rev. I dated 03/11/2010) under the ScreenCriteria for questions linked to I OCFR55.43(b)(2) (Tech Specs):

1. It cannot be answered solely by knowing 1 hour TS/ SLC Action.2. It cannot be answered solely by knowing the LCO/SLC information listed above the line.3. It cannot be answered solely by knowing the TS Safety Limits.



4. The question involves application of required actions of Tech Spec 3.8.4 (DC Sources —

Operating).5. The question involves knowledge of TS bases that is required to analyze TS required

actions and terminology.



CNS LOSS OF NORMAL POWER PAGE NO.AP111A155001007 26 of 162Case IIRevision 66

Lossof All Power to an Essential Train

ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED

13. Control SIG levels as follows:

— a. Verify CF flow - MAINTAINING STABLE a. Perform the following:S/G LEVELS.

— 1) REFER TO Enclosure 16 (S/GLevel Control).

_2) QIQ.Step14.

b. IFATANY TIME CF flow control toS/Gs is lost, THEN perform Step 13.

CAUTION Battery depletion may occur as early as two hours. Battery depletionresults in affected CA control valves failing full open. Failure to takelocal control of S!G level prior to battery depletion may result in SIG

— c. IEA[ANY TIME any vital or auxiliarycontrol channel battery charger hasbeen de-energized for greater than Ihour, THEN dispatch operators tolocally control affected CA flow path.

- A? o7REFER TO Enclosure 16 (S/G Level Or 1OJi3 ‘‘ “ 0Control).

CNS LOSS OF VITAL OR AUX CONTROL POWER PAGE NO.AP111A155001029 31 of 208Enclosure 1 - Page 4 of

Revision 24Response To Degraded DC Bus Voltage

ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED

5. (Continued)

c. Review applicable load list(s) for plantresponse to a loss of affected controlpower buss:

—e Enclosure 6 (1 EPA Load List)

—e Enclosure 7 (1 EPB Load List)

—. Enclosure 8 (1EPC Load List)

—. Enclosure 9 (IEPD Load List)

—. Enclosure 20 (1CDA Load List)

• Enclosure 21 (ICDB Load List).

— d. Contact Station Management forrecommendations regarding imminentlow DC voltage.

NOTE • Indication of Aux Control Power battery voltage is available locally or inControl Room as “ICDA(lCDB) Bus Voltag&’.

• Indication of Vital battery voltage is available locally or in Control Room as“lEDA(lEDB,IEDC,IEDD) Bus Voltage”.

e. WHEN any Vital or Aux Control Powerbattery voltage decays to 105 VDC,THEN perform the following:

— 1) Ensure associated battery outputbreaker - OPEN.

2) IFATANY TIME breaker controlpower is lost, AND it is desired tooperate station breakers, THENevaluate aligning batteries tobreaker control power only.

‘It (H


Question 81077AA2.07Generator Voltage and Electric Grid DisturbancesAbility to determine and interpret the following as they apply to Generator Voltage andElectric Grid Disturbances:Operational status of engineered safety features


Initial:

• The Unit is at 100% power.• 1.B-DieseIerator (DIG) is running at 5750 KW for a periodic test.• IA ND pump has been tagged to repair an emergent oil leak.• A grid disturbance results in a loss of offsite power.• A LOCA initiates concurrent with the loss of offsite power.

One minute later:

• lB NV Pump has the following indications:• NO running amps are indicated.• Both the “ON” and “OFF” light on the E30 pushbutton are DARK.

• IA NI Pump has the following indications:• No running amps are indicated. :• Both the “ON” and “OFF” light on the E30 pushbutton are DARK. ‘

(1) There (1) enough ECCS pumps operating to meet the LOCA analysisassumptions described in Technical Specification 3.5.2 (ECCS - Operating) BASES.

(2) What train(s) of ECCS and DIG load sequencers must be RESET

A. (1) ARE(2) Both “A” and “B” trains must be RESET.

B. (1) ARE(2) Only “B” train must be RESET. \f

C. (1) ARE NOT(2) Both “A” and “B” trains must be RESET.

D. (1) ARE NOT(2) Only “B” train must be RESET.



QUESTION 81

Distractor Analysis

A. CORRECT. Based on the conditions one minute later, both DGs should be running withLOCA loads. The DIG breaker will trip on overcurrent in this situation, and the LOCA willoverride the blackout and LOCA loads will sequence on. Since B D/G is running already,applicants may think that the B train sequencer won’t work correctly or will only loadblackout loads. In that case, the I B ND pump would not be running (or NI) since theseare LOCA-only loads. They should reset to attempt to restart the 1 B NV pump as well.

Ar

Based on the conditions specified, IA NV pump, IA and lB NI pumps and lB ND pumpwould be running. Based on Tech Spec bases for ECCS operation, only one completetrain is required meaning one of each TYPE of pump, not necessarily on the same train.

The 1 B NV pump being off would cause the crew to attempt to reset B Train of ECCS andD/G load sequencer to attempt to start it, and they would also attempt to reset A Train tostart the IA NI pump.

B. Incorrect. First part is correct. Both trains would be reset.

C. Incorrect. Second part is correct. Plausible that criteria not met by poor recall of therequirement.

D. Incorrect. Enough pumps are running, but both trains are reset.

References:• Tech. Spec. B3.5.2, “ECCS - Operating”, Applicable Safety Analysis.

KA Match:Question 81077AA2.07Generator Voltage and Electric Grid DisturbancesAbility to determine and interpret the following as they apply to Generator Voltage andElectric Grid Disturbances:Operational status of engineered safety featuresThe KA is matched because the question involves a grid disturbance, and testing the ability toevaluate given conditions and determine if LOCA analysis assumptions and criteria are met forECCS equipment (operational status of engineered safety features)

Cognitive Level: HighThis is a higher cognitive level question because the applicant must determine from the givenconditions what safeguards equipment should be operating, given the effects of the powerdisturbance, and then evaluate the status of the equipment in the context of accident analysisassumptions.


SRO Only:



This question meets the following criteria for an SRO only question as described in the“Clarification Guidance for SRO-only Questions (Rev. I dated 03/11/2010) under the ScreenCriteria for questions linked to 1 OCFR55.43(b)(2) (Tech Specs):

1. It cannot be answered solely by knowing 1 hour TS/ SLC Action.2. It cannot be answered solely by knowing the LCO/SLC information listed above the line.3. It cannot be answered solely by knowing the TS Safety Limits.4. The question involves application of required actions of Tech Spec 3.5.2 (ECCS —

Operating).5. The question involves knowledge of TS bases that is required to analyze TS required




ECCS—OperatingB 3.5.2

B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.2 ECCS—Operating

BASES

BACKGROUND The function of the ECCS is to provide core cooling and negativereactivity to ensure that the reactor core is protected after any of thefollowing accidents:

a. Loss of coolant accident (LOCA), coolant leakage greater than thecapability of the normal charging system;

b. Rod ejection accident;

c. Loss of secondary coolant accident, including uncontrolled steamor feedwater release; and

d. Steam generator tube rupture (SGTR).

The addition of negative reactivity is designed primarily for the loss ofsecondary coolant accident where primary cooldown could add enoughpositive reactivity to achieve criticality and return to significant power.

There are three phases of ECCS operation: injection, cold legrecirculation, and hot leg recirculation. In the injection phase, water istaken from the refueling water storage tank (RWST) and injected into theReactor Coolant System (RCS) through the cold legs. When sufficientwater is removed from the RWST to ensure that enough boron has beenadded to maintain the reactor subcritical and the containment sumpshave enough water to supply the required net positive suction head to theECCS pumps, suction is switched to the containment sump for cold legrecirculation. When the core decay heat has decreased to a level lowenough to be successfully removed without direct RHR pump injectionflow, the RHR cold leg injection path is realigned to discharge to theauxiliary containment spray header. After approximately 7 hours, part ofthe ECCS flow is shifted to the hot leg recirculation phase to provide abackflush which, for a cold leg break, would reduce the boiling in the topof the core and prevent excessive boron concentration.

The ECCS consists of three separate subsystems: centrifugal charging(high head), safety injection (SI) (intermediate head), and residual heatremoval (RHR) (low head). Each subsystem consists of two redundant,100% capacity trains. The ECCS accumulators and the RWST are alsopart of the ECCS, but are not considered part of an ECCS flow path asdescribed by this LCO.

Catawba Units 1 and 2 B 3.5.2-1 Revision No. 3

ECCS — OperatingB 3.5.2

BASES

BACKGROUND (continued)

The ECCS flow paths consist of piping, valves, heat exchangers, andpumps such that water from the RWST can be injected into the RCSfollowing the accidents described in this LCO. The major components ofeach subsystem are the centrifugal charging pumps, the RHR pumps,heat exchangers, and the SI pumps. Each of the three subsystemsconsists of two 100% capacity trains that are interconnected andredundant such that either train is capable of supplying 100% of the flowrequired to mitigate the accident consequences. This interconnectingand redundant subsystem design provides the operators with the ability toutilize components from opposite trains to achieve the required 100%flow to the core.

During the injection phase of LOCA recovery, a suction header supplieswater from the RWST to the ECCS pumps. Mostly separate pipingsupplies each subsystem and each train within the subsystem. Thedischarge from the centrifugal charging pumps combines, then dividesagain into four supply lines, each of which feeds the injection line to oneRCS cold leg. The discharge from the SI and RHR pumps divides andfeeds an injection line to each of the RCS cold legs. Throttle valves inthe SI lines are set to balance the flow to the RCS. This balance ensuressufficient flow to the core to meet the analysis assumptions following aLOCA in one of the RCS cold legs. The flow split from the RHR linescannot be adjusted. Although much of the two ECCS trains arecomposed of completely separate piping, certain areas are sharedbetween trains. The most important of these are 1) where both trainsflow through a single physical pipe, and 2) at the injection connections tothe RCS cold legs. Since each train must supply sufficient flow to theRCS to be considered 100% capacity, credit is taken in the safetyanalyses for flow to three intact cold legs. Any configuration which, whencombined with a single active failure, prevents the flow from either ECCSpump in a given train from reaching all four cold legs injection points onthat train is unanalyzed and might render both trains of that ECCSsubsystem inoperable.

For LOCAs that are too small to depressurize the RCS below the shutoffhead of the SI pumps, the centrifugal charging pumps supply water untilthe RCS pressure decreases below the SI pump shutoff head. Duringthis period, the steam generators are used to provide part of the corecooling function.

During the recirculation phase of LOCA recovery, RHR pump suction istransferred to the containment sump. The RHR pumps then supply theother ECCS pumps. Initially, recirculation is through the same paths asthe injection phase. Subsequently, for large LOCAs, the recirculationphase includes injection into both the hot and cold legs.



BASES

BACKGROUND (continued)

The high and intermediate head subsystems of the ECCS also functionsto supply borated water to the reactor core following increased heatremoval events, such as a main steam line break (MSLB). The limitingdesign conditions occur when the moderator temperature coefficient ishighly negative, such as at the end of each cycle.

During low temperature conditions in the RCS, limitations are placed onthe maximum number of ECCS pumps that may be OPERABLE. Referto the Bases for LCO 3.4.12, “Low Temperature Overpressure Protection(LTOP) System,” for the basis of these requirements.

The ECCS subsystems are actuated upon receipt of an SI signal. Theactuation of safeguard loads is accomplished in a programmed timesequence. If offsite power is available, the safeguard loads startimmediately in the programmed sequence. If offsite power is notavailable, the Engineered Safety Feature (ESF) buses shed normaloperating loads and are connected to the emergency diesel generators(EDG5). Safeguard loads are then actuated in the programmed timesequence. The time delay associated with diesel starting, sequencedloading, and pump starting determines the time required before pumpedflow is available to the core following a safety injection actuation.

The active ECCS components, along with the passive accumulators andthe RWST covered in LCO 3.5.1, “Accumulators,” and LCO 3.5.4,“Refueling Water Storage Tank (RWST),” provide the cooling waternecessary to meet GDC 35 (Ref. 1).

APPLICABLE The LCO helps to ensure that the following acceptance criteria for theSAFETY ANALYSES ECCS, established by 10 CFR 50.46 (Ref. 2), will be met following a

small break LOCA and there is a high level of probability that the criteriaare met following a large break LOCA:

a. Maximum fuel element cladding temperature is 2200°F;

b. Maximum cladding oxidation is 0.17 times the total claddingthickness before oxidation;

c. Maximum hydrogen generation from a zirconium water reaction is0.01 times the hypothetical amount generated if all of the metal in

the cladding cylinders surrounding the fuel, excluding the claddingsurrounding the plenum volume, were to react;



BASES

APPLICABLE SAFETY ANALYSES (continued)

d. Core is maintained in a coolable geometry; and

e. Adequate long term core cooling capability is maintained.

The LCO also limits the potential for a post trip return to power followingan MSLB event and ensures that containment pressure and temperaturelimits are met.

Each ECCS subsystem is taken credit for in a large break LOCA event atfull power (Refs. 3 and 4). This event has the greatest potential tochallenge the limits on runout flow set by the manufacturer of the ECCSpumps. It also sets the maximum response time for their actuation.Direct flow from the centrifugal charging pumps and SI pumps is creditedin a small break LOCA event. The RHR pumps are also credited, forlarger small break LOCAs, as the means of supplying suction to thesehigher head ECCS pumps after the switch to sump recirculation. Thisevent establishes the flow and discharge head at the design point for thecentrifugal charging pumps. The MSLB analysis also credits the SI andcentrifugal charging pumps. Although some ECCS flow is necessary tomitigate a SGTR event, a single failure disabling one ECCS train is notthe limiting single failure for this transient. The SGTR analysis primary tosecondary break flow is increased by the availability of both centrifugalcharging and SI trains. Therefore, the SGTR analysis is penalized byassuming both ECCS trains are operable as required by the LCO. TheOPERABILITY requirements for the ECCS are based on the followingLOCA analysis assumptions:

a. A large break LOCA event, with loss of offsite power and a singlefailure disabling one ECCS train; and

b. A small break LOCA event, with a loss of offsite power and a singlefailure disabling one ECCS train.

During the blowdown stage of a LOCA, the RCS depressurizes asprimary coolant is ejected through the break into the containment. Thenuclear reaction is terminated either by moderator voiding during largebreaks or control rod insertion for small breaks. Followingdepressurization, emergency cooling water is injected into the cold legs,flows into the downcomer, fills the lower plenum, and refloods the core.

The effects on containment mass and energy releases are accounted forin appropriate analyses (Ref. 3). The LCO ensures that an ECCS trainwill deliver sufficient water to match boiloff rates soon enough to minimizethe consequences of the core being uncovered following a large LOCA.



BASES

APPLICABLE SAFETY ANALYSES (continued)

It also ensures that the centrifugal charging and SI pumps will deliversufficient water and boron during a small LOCA to maintain coresubcriticality. For smaller LOCAs, the centrifugal charging pump deliverssufficient fluid to maintain ROS inventory. For a small break LOCA, thesteam generators continue to serve as the heat sink, providing part of therequired core cooling.

The ECCS trains satisfy Criterion 3 of 10 CFR 50.36 (Ref. 5).

LCO In MODES 1, 2, and 3, two independent (and redundant) ECCS trains arerequired to ensure that sufficient ECCS flow is available, assuming asingle failure affecting either train. Additionally, individual componentswithin the ECCS trains may be called upon to mitigate the consequencesof other transients and accidents.

In MODES 1, 2, and 3, an ECCS train consists of a centrifugal chargingsubsystem, an SI subsystem, and an RHR subsystem. Each trainincludes the piping, instruments, and controls to ensure an OPERABLEflow path capable of taking suction from the RWST upon an SI signal andautomatically transferring suction to the containment sump.

During an event requiring ECCS actuation, a flow path is required toprovide an abundant supply of water from the RWST to the RCS via theECCS pumps and their respective supply headers to each of the four coldleg injection nozzles. In the long term, this flow path may be switched totake its supply from the containment sump and to supply its flow to theRCS hot and cold legs. The flow path for each train must maintain itsdesigned independence to ensure that no single failure can disable bothECCS trains.

APPLICABILITY In MODES 1, 2, and 3, the ECCS OPERABILITY requirements for thelimiting Design Basis Accident, a large break LOCA, are based on fullpower operation. Although reduced power would not require the samelevel of performance, the accident analysis does not provide for reducedcooling requirements in the lower MODES. The centrifugal chargingpump performance is based on a small break LOCA, which establishesthe pump performance curve and has less dependence on power. TheSI pump performance requirements are based on a small break LOCA.For both of these types of pumps, the large break LOCA analysisdepends only on the flow value at containment pressure, not on theshape of the flow versus pressure curve at higher pressures. MODE 2and MODE 3 requirements are bounded by the MODE 1 analysis.


EXAM BANK - Q 502Initial conditions:

• Unit us operating at 100% power.• 1 B diesel generator (DIG) is running at 5750 KW for a periodic test• IA ND pump has been tagged to repair an emergent oil leak• A LOCA and loss of offsite power occurs

One minute later the following conditions are noted:

• I B NV Pump has the following indications:o No running amps are indicatedo Both the “ON” and “OFF” light on the .E30 pushbutton are DARK

• IA NI Pump has the following indications:o No running amps are indicatedo Both the “ON” and “OFF” light on the E30 pushbutton are DARK

Assuming all equipment not specifically addressed operated normally:

1. What is the current status of the ECCS system related to its design basis perTechnical Specification 3.5.2 (ECCS - Operating)?

2. When EP/IIAI5000/E-0, Reactor Trip or Safety Injection, is exited, what train(s) ofECCS and D/G load sequencers must be RESET?

A. 1. There are enough ECCS pumps running to meet :ECCS design criteria.2. Both “A” and “B” trains must be RESET.

B. 1. There are enough ECCS pumps running to meet ECCS design criteria.2. Only “B” train must be RESET.

C. 1. There are not enough ECCS pumps running to meet ECCS design criteria.2. Both “A” and “B” trains must be RESET.

D. 1. There are not enough ECCS pumps running to meet ECCS design criteria.2. Only “B” train must be RESET.


Question 82OOIAA2.04Continuous Rod WithdrawalAbility to determine and interpret the following as they apply to the Continuous RodWithdrawal:.

P—Reactor power and its trend

For a Continuous Rod Withdrawal event initiating from a power level of 15% with IR Nis NOTblocked, AP/15, (Rod Control Malfunction), Case II, Continuous Rod Movement has beenimplemented.

(1) The malfunction that AP/15 attempts to diagnose is (1)

(2) Reactor power reached 31% just prior to a manual trip from the Control Room. What isthe Emergency Classification of this event?

A. (1-)—NC-LoopTavg-failure, /(2) Alert

U— -( -

B. (1)NC-LoopTavgfailure(2) Site Area Emergency ç

C. -(1) DCSIaiIure(2) Alert

D. (1)—DCS Jailure(2) Site Area Emergency

1 A

)



QUESTION 82

Distractor Analysis

A. CORRECT. Step 4 of APII 5 contains detailed guidance in attempting to determine if thereis a problem with the NC Loop Tavg indications. If a continuous rod withdrawal eventinitiates from the given power level (15%) and reactor power reaches 31%, and the controlroom manually trips, this is an indication that an automatic trip did not occur at 25%. Thisshould have occurred by the Intermediate Range NIs since they were not blocked. ThePower Range Nis would have already been blocked, and would not be expected to processa trip for this condition.

Looking at the classification matrix (RP/O1, Enclosure 4.4, “Loss of Shutdown Functions,”the SRO will classify this as an Alert, since the criteria of 4.4.A. I was met: an auto tripshould have occurred, but did NOT, but the manual trip was successful.

B. Incorrect. First part is correct. Site Area Emergency is plausible through misdiagnosis ofthe reactor trip status, and misapplication of the Classification Matrix.

C. Incorrect. Second part is correct. DCS failure is plausible since this system affectsnumerous systems in the plant.

D. Incorrect: Both parts explained in sections above.

References:• AP/1/A15500/15, (Rod Control Malfunction), Case II, “Continuous Rod Movement”,

Revision 014• RP/0/A/5000/001, (Classification of Emergency), Enclosure 4.4, “Loss of Shutdown

Functions,” Revision 027 (Provide to Applicant - 3 pages)

KA Match:Question 82001 AA2.04Continuous Rod WithdrawalAbility to determine and interpret the following as they apply to the Continuous RodWithdrawal:Reactor power and its trendThe KA is matched because given conditions involve a continuous rod withdrawal and then theapplicant is tested to determine and interpret a trend given on reactor power, and continue on toprovide procedure content, and classification of the event.

Cognitive Level: HighThis is a higher cognitive level question since the applicant must diagnose that an automaticreactor trip should have occurred, but did not, and then with that determination, consult aclassification matrix to determine the emergency classification.




SRO Only:This question meets the following criteria for an SRO only question as described in the“Clarification Guidance for SRO-only Questions (Rev. I dated 03/11/2010) under the ScreenCriteria for questions linked to IOCFR55.43(b)(1), (Conditions and Limitations in the facilitylicense):

1. It involves classification of an .emergency event.


jf channel inoperable, THEN perform thefollowing:

a. WHEN T-Avg is within ±1°F of T-RefAND auto rod control is desired, THENreturn rod control to “AUTO”.

— b. Ensure P-12 interlock in required statefor existing plant conditions. REFERJQ.Tech Spec 3.3.2.

c. Have IAE trip bistables associated withthe failed loop within 72 hours. REFERTO Model W!O #00874531:

—. OPDT (Tech Spec 3.3.1)

—. OTDT (Tech Spec 3.3.1)

—. Low T-Avg (Tech Spec 3.3.2).

CNS ROD CONTROL MALFUNCTIONS PAGE NO. IAP111A155001015 10 of 11 I

Case IIRevision 141Continuous Rod Movement I

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED I4. Verify the following channels - NORMAL

FOR EXISTING PLANT CONDITIONS:

—. NC Loop A T-Avg

—. NC Loop B T-Avg

—. NC Loop C T-Avg

—. NC Loop D T-Avg.

— 5. Determine and correct cause ofcontinuous rod movement.

6. Ensure compliance with appropriateTech Specs:

—. 3.1.1 (Shutdown Margin (SDM))

—. 3.1.4 (Rod Group Alignment Limits)

—. 3.1.5 (Shutdown Bank Insertion Limits)

—. 3.1.6 (Control Bank Insertion Limits)

—. 3.3.1 (Reactor Trip Instrumentation)

• 3.3.2 (ESFAS Instrumentation)

—• 3.4.2 (RCS Minimum Temperature forCriticality).

Enclosure 4.4

UNUSUAL EVENT

Loss of Shutdown Functions

ALERT SITE AREA EMERGENCY

1u/O/A/5000/oo 1Page 1 of3

GENERAL EMERGENCY

4.4.A.1 Failure of ReactorProtection SystemInstrumentation to Completeor Initiate an AutomaticReactor Trip Once aReactor Protection SystemSetpoint fins Been Exceededand Manual Trip WasSuccessful.

4.4.S.1 Failure of ReactorProtection SystemInstrumentation to Completeor Initiate an AutomaticReactor Trip Once aReactor Protection SystemSetpoint Has Been Exceededand Manual Trip Was NOTSuccessful.

4.4.G.1 Failure of the ReactorProtection System toComplete an Automatic Tripand Manual Trip Was NOTSuccessful and There isIndication of an ExtremeChallenge to the Ability toCool the Core.

OPERATING MODE: 1

Valid reactor trip signalreceived or required andautomatic reactor tripwas not successful.

AND

Manual reactor trip from thecontrol room is successful andreactor power is less than 5%and decreasing.


AND

Manual reactor trip from thecontrol room was notsuccessful in reducing reactorpower to less than 5% anddecreasing.

4.4.G.1-1 The following conditions exist:


AND

Manual reactor trip from thecontrol room was notsuccessful in reducing reactorpower to less than 5% anddecreasing.

(Continued) EITHER of the followingconditions exist:

• Core Cooling CSF-RED

• Heat Sink CSF-RED.

END

OPERATING MODE: 1,2,3 OPERATING MODE: 1

4.4.A.1-1 The following conditions exist: 4.4.S.1-l The following conditions exist:

(Continued) AND

END


Question 83028AG2.1 .32Pressurizer Level MalfunctionAbility to explain and apply system limits and precautions.


• With the Unit in Mode 2, a Unit startup is in progress in accordance withOPI1IAI6IOO/00l, (Controlling Procedure for Unit Startup).

• A Pressurizer Level malfunction has occurred during the startup.• During the restoration of Pressurizer level, the operators are attempting to maintain an

outflow on the Pressurizer in accordance with the Limits and Precautions ofOP/I 1A161 00/001.

(1) What is the basis for the above Limit and Precaution to maintain an outflow on thePressurizer?

(2) If the operators are unable to maintain an outflow on the Pressurizer, and SLC 16.5-4,(Pressurizer), action and completion time cannot be met for the resulting condition, within6 hours the Unit must be placed in

__________________

LA (1) boron strafication_ - c

(2) MODE 3.

B. (1) boron stratification(2) MODE 4

C. (1) thermal stratification(2) MODE 3.

D. (1) thermal stratification(2) MODE 4.

Ans: C



QUESTION 83

Distractor Analysis

A. Incorrect. Boron stratification is plausible if applicant incorrectly recalls another Limit andPrecaution (Enclosure 4.1, Unit Startup — Limit and Precaution 1.2.3) regarding arequirement for PZR boron concentration compared to NC System boron. But it ismisapplied here, and not the basis for PZR outflow.

B. Incorrect. First part plausibility explained in “A” above. Mode 4 is plausible if the SLCrequirement is misapplied.

C. CORRECT. OPIIIA/61001001, (Controlling Procedure for Unit Startup), Limits andPrecaution 2.16 explains that the basis for maintaining an outflow on the PZR is tominimize PZR thermal stratification. The Selected Licensee Commitment (SLC) 16.5-4requires that the Unit be place in Mode 3 if unable to maintain the condition.

D. Incorrect: First part is correct. Mode 4 plausibility is explained in “B” above.

References:• OP/11A16100/O01, (Controlling Procedure for Unit Startup), Limits and Precaution 2.16,

Revision 225.• SLC 16.5-4, Pressurizer• TS 3.4.5, (RCS Loops - MODE 3), Condition CI

KA Match:Question 83028AG2.1 .32Pressurizer Level MalfunctionAbility to explain and apply system limits and precautions.This KA is matched because the stem conditions involve a PZR level malfunction that requiresthe operators to take action and restore level. In that process, there is a concern for PZRoutflow, as explained in a Limit and Precaution.

Cognitive Level: HighThis is a higher cognitive level question because there is more than one mental step involved.First the applicant must evaluate given conditions and then apply those to if a requirement foroufflow is met, and what action would be required by the SLC.


SRO Only:This question meets the following criteria for an SRO only question as described in the“Clarification Guidance for SRO-only Questions (Rev. I dated 03/11/2010) under the ScreenCriteria for questions linked to IOCFR55.43(b)(2) (Tech Specs):

1. It cannot be answered solely by knowing < 1 hour TS/ SLC Action.



2. It cannot be answered solely by knowing the LCO/SLC information listed above the line.3. It cannot be answered solely by knowing the TS Safety Limits.4. The question involves application of required actions of SLC 16.5-4, Pressurizer.5. The question involves knowledge of SLC bases that is required to analyze SLC required




OP/1/A16 100/001Page 5 of 7

2.13 It is recommended that S/G reverse purge flow be maintained at all times when the S/Gsare pressurized and CF flow is NOT aligned to the main feedwater nozzles. Thisensures the main feed containment penetration piping is maintained above brittle fracturetemperature of 107°F.

• If the temperature on both sides of the penetration is greater than 107°F duringMode 1, reverse purge flow can be secured, but reverse purge shall be re-establishedbefore temperature reaches 107°F (decreasing) to ensure compliance with thecommitment to the NRC on 1OCFR5O Appendix A GDC51 (temperature greaterthan 107°F during power operation-Mode 1). The temperature between the feedwaterisolation valves and S/Gs shall be greater than 107°F during Mode 1. (C1AO141,C1A0148, C1A0125, C1A0154, C1A0275, C1AO16O, C1A0815, C1A0166, OACGroup Display GD OPCFTEMP)

• During Modes 2, 3 and 4 reverse purge can be secured to aid in plant heatup. It isdesirable to have reverse purge at all times during plant heatup to prevent a possibledelay in entering Mode 1.

2.14 When feeding the S/Gs from a source other than main feedwater, Secondary Chemistryneeds to know the source in order to obtain accurate chemistry data.

2.15 If the RC System condenser inlet temperature drops to less than or equal to 60°F whenthe reactor is shutdown or less than or equal to 55°F when the reactor is critical, the RCSystem shall be aligned as follows:

• One RC pump running (throttled).• One tower inlet isolated.• All three riser bypasses open.

2.16 An outflow on the PZR is maintained to minimize PZR thermal stratification. PZRoutflow may be confirmed by the following:

• Extra heater capacity energized.• NC, NV or ND PZR spray indicated by valve positive demand.• PZR surge line temperature and PZR water space temperatures are approximately

equal.• PZR spray valve for idle NC Pumps closed.

2.17 If situations occur causing PZR liquid space temperature to decrease due to PZR levelincrease, then the PZR level shall be maintained at the elevated level until PZR liquidspace temperature recovers. PZR liquid space temperature is directly affected by PZRlevel during plant conditions requiring a saturated PZR and cooler NC looptemperatures.

S LC Pressurizer

16.5 REACTOR COOLANT SYSTEM

16.5-4 Pressurizer

COMMITMENT The pressurizer temperature shall be limited to:

a. A maximum heatup of 100°F in any 1-hour period, and

b. A maximum cooldown of 200°F in any 1-hour period.

APPLICABILITY: At all times.

REMEDIAL ACTIONS

CONDITION REQUIRED ACTION COMPLETION TIME

A. NOTE A.1 Restore pressurizer 30 minutesAll Required Actions temperature to withinmust be completed limits.whenever this Conditionis entered. AND

Pressurizer temperature A.2 Pertorm engineering 72 hoursnot within limits, evaluation to determine

effects of the out-of-limitcondition on the structuralintegrity of the pressurizer.

AND

A.3 Determine that the 72 hourspressurizer remainsacceptable for continuedoperation.

B. Required Action and B.1 Be in MODE 3. 6 hoursassociated CompletionTime not met.

B.2 Reduce pressurizer 36 hoursY.&1 I T’fl pressure to < 500 psig.

s

Catawba Units 1 and 2 1 6.5-4-1 Revision 0


,ji

Question 84068AG2.4.18Control Room Evac /

Knowledge of specific bases for EOPS. I -

Given the following plant conditions:

• Due to a fire event, AP/17, (Loss of Control Room), Case ll,)Loss of Plant Control Due toFire or Security Event, is in progress for both Units.

Which ONE of the following describes the required operation, PRIOR to Control Roomevacuation, of the listed components, AND the BASIS for the requirement?

A. All CRD vent fans RUNNING. —‘ , 4No control of these fans is available from the ASP

B. All CRD vent fans RUNNING.Cooling of the reactor vessel head during natural circulation.

C. NV-IOA (Letdn Orif B OtIt Cont Isol) CLOSED. ‘

To ensure NC inventory is conserved in case the event also involves a LOCA.

D. NV-bA (Letdn Orif B OtIt Cont Isol) OPEN.To ensure inventory control is available when transferring controls to theS.

(- 1

}



QUESTION 84

Distractor Analysis

A. Incorrect. First part is correct. Second part is plausible since it is true, but it is NOT thebasis for the requirement.

B. CORRECT. Step 6 of AP/1 7 for Control Room evacuation prescribes that all CRD ventfans should be ON. The basis for this is explained in the Lesson Plan for AP/17, Case I,Step 6, on page 11 of 34 of OP-CN-AP-1 7, Lesson Plan for APII 7, Loss of Control Room.

C. Incorrect. Plausible, since closing this valve would isolate and conserve inventory, but thisis the incorrect basis.

D. Incorrect. Plausible since this would aid in inventory control, but it is NOT the correct basisfor the requirement.

References:• AP/IIA/55001017, Loss of Control Room, Rev. 055• page 11 of 34 of OP-CN-AP-1 7, Lesson Plan for AP/1 7, Loss of Control Room.

KA Match:Question 84068AG2.4.1 8Control Room Evac.Knowledge of specific bases for EOPS.The KA is matched because even though the question does expressly SAY there is an EOPinvolved, there would be for a loss of control room (E-0). Tripping the reactor is the first actionin AP/1 7 for control room evacuation. The applicant is tested on the basis for a specific actionassociated with these conditions.

Cognitive Level: HighAt first glance, this may seem like a simple recall question, since it involves recalling what theprocedure says, and then what the basis is; but to get the right answer you must applyknowledge of one of the ‘CRD vent fans functions (cooling vessel head area) and apply that toarrive at the correct answer.



I. It cannot be answered solely by knowing “systems knowledge”, i.e., how the systemworks, flowpath, logic, component location.

2. It cannot be answered solely by knowing immediate operator actions.


CNS 2012 NRC Exam 100 Questions Final Submiftal

3. It cannot be answered solely by knowing entry conditions for AOPs or plant parametersthat require direct entry to major EOPs.

4. It cannot be answered solely by knowing the purpose, overall sequence of events, oroverall mitigative strategy of a procedure.

5. The question does involve assessing plant conditions (involving AP/17 content), andthen selecting a section (specific step for a specific purpose) to mitigate a Loss ofControl Room event.



CNS LOSS OF CONTROL ROOM PAGE NO.AP111A155001017 2 of 78

Control Room Revision 55

I ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

C. Operator Actions

1. Available SRO performs the following:

a. Take over as OATC.

— b. Dispatch RO with key boxes to Unit 1ASP to perform Enclosure 1 (ASPOperator Actions).

— c. Dispatch RO!SRO to Unit 1 AFWPTCPto perform Enclosure 2 (AFWPTCPOperator Actions).

— 2. Announce the following twice over plantwide communications system: “Loss ofUnit I Control Room imminent. Assignedoperations shift personnel man remotelocations”.

— 3. Trip reactor.

4. Verify Reactor Trip: IF reactor will not trip, THEN dispatchoperator to open the following:

—. All rod bottom lights - LIT

• Reactor trip breakers—

. All reactor trip and bypass breakers -

OPEN • Reactor trip bypass breakers.

• hR power - DECREASING.

5. Verify all turbine stop valves - CLOSED. — Trip turbine.

6. Ensure all CR0 vent fans - ON.

7. Trip CF pumps.


Question 85076AA2.05High Reactor Coolant ActivityAbility to determineandil terprettbejplLQng as they apply to the High Reactor CoolantActivity: CVCS letdown flow rate indication N


• The Unit is heating up following a short forced outage following a reactor trip 3 days ago.• NC pressure is 1500 psig. 7

• NC temperature is 385°F. /‘

• Due to High NC Activity, Radiation Protection discovered dose ates’aetoo high toallow work in an area near the letdown line.

• To reduce dose rates, smaller micron NC filters were installed and letdown flow wasincreased from 75 gpm to 95 gpm to aid in cleanup.

(1) Is Technical Specification 3.4.16 (RCS Specific Activity) appIicabased on current plant,status? (N;

(2) Which one.pf the parameters evaluated per Technical Specification 3.4.16 has a lowerlimit instatedbased on increases in letdown flow to greater than 80 gpm?

A. (1) Yes(2) Dose Equivalent 1-131

B. (1) Yes(2) Gross specific activity

C. (1) No(2) DoseEquivalentl-131

D. (1) No(2) Gross specific activity

7 ,/

A7



QUESTION 85

Distractor Analysis

A. Incorrect. The student may believe that TS 3.4.16 is applicable due to being in Mode 3.DEl is the correct parameter with the lower limit

B. Incorrect. The student may believe that TS 3.4.16 is applicable due to being in Mode 3and that gross specific activity is the parameter with the lower limit.

C. CORRECT. Tech spec 3.4.16 applies in Modes 1,2, and 3 *when Tavg is >500 degreesso it does not apply for the conditions stated. The 2 conditions evaluated are DEl andgross specific activity. When letdown flow is increased above 80 gpm, lower DEl limits arein effect. This is stated in a TS amendment, the NV lesson, and confirmed based on stepsin AP/12 and AP/18.

D. Incorrect. TS applicability is correct. The student may believe that gross specific activity isthe parameter with the lower limit.

References:Technical Specification 3.4.16 (RCS Specific Activity)

KA Match:Question 85076AA2.05High Reactor Coolant ActivityAbility to determine and interpret the following as they apply to the High Reactor CoolantActivity: CVCS letdown flow rate indicationThe KA is matched because the question involves a high reactor coolant activity condition, andincludes testing on what it would mean to raise the letdown flow rate.

Cognitive Level: HighThis is a higher cognitive level question because there is an evaluation of plant conditions and adetermination of whether a Tech Spec is applicable. The applicant must also recall and applyknowledge of dose equivalent iodine concerns and the relationship of that with letdown flow raterequirements.

Source of Question: Bank

SRO Only:

This question meets the following criteria for an SRO only question as described in the“Clarification Guidance for SRO-only Questions (Rev. I dated 03/11/2010) under the ScreenCriteria for questions linked to 1OCFR55.43(b)(2) (Tech Specs):




4. The question involves application of required actions of Tech Spec 3.4.16, RCS SpecificActivity.



CNS LOSS OF CHARGING OR LETDOWN PAGE NO.API1IAI5500!012 20 of 32Case II

Revision 31Loss of Letdown

I ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

16. (Continued)

k. WHEN 5 minutes have elapsed, THENperform the following:

1) IFATANYTIME letdown flowincreased to greater than 80 GPM,THEN perform the following:

— a) Determine current NC DoseEquivalent Iodine concentration(DEl). (OAC Point Cl P0097)

— b) Verify DEl specific activity - — b) Ensure compliance with TechLESS THAN 0.18 Ci/GM. Spec 3.4.16 (RCS Specific

Activity).

c) Notify Primary Chemistry thatlower DEl limits are in effect dueto NV letdown flows greater than80 GPM.

— 2) Adjust I NV-849 (Letdn Flow VarOrif Ctrl) in 1% increments todesired letdown flow.

3) WHEN letdown at desired flow,THEN perform the following:

— a) Adjust 1NV-148 (Letdn PressControl) to maintain letdownpressure at 350 PSIG.

— b) Ensure INV-148 (Letdn PressControl) - IN AUTO.

— 4) IFATANY TIME additional letdownflow desired, THEN establishletdown with the 45 or 75 GPMorifice. REFER TOOP/1/A16200/001 (Chemical andVolume Control System).

RCS Specific Activity3.4.16

3.4 REACTOR COOLANT SYSTEM (RCS)

3.4.16 RCS Specific Activity

LCO 3.4.16 The specific activity of the reactor coolant shall be within limits.

APPLICABILITY: MODES 1 and 2,MODE 3 with RCS average temperature (Tavg) 500°F.

ACTIONS


A. DOSE EQUIVALENT Note1-131 > 1 .0 aCi/gm. LCO 3.0.4.c is applicable.

A.1 Verify DOSE EQUIVALENT Once per 4 hours1-131 within the acceptableregion of Figure 3.4.16-1.

AND

A.2 Restore DOSE 48 hoursEQUIVALENT 1-131 towithin limit.

B. Gross specific activity of B.1 Be in MODE 3 with 6 hoursthe reactor coolant not Tavg < 500°F.within limit.

(continued)

Catawba Units 1 and 2 3.4.16-1 Amendment Nos. 213/207


ACTIONS (continued)


C. Required Action and C.1 Be in MODE 3 with 6 hoursassociated Completion Tavg < 500°F.Time of Condition A notmet.

OR

DOSE EQUIVALENT1-131 in theunacceptable region ofFigure 3.4.16-1.

SURVEILLANCE REQUIREMENTS

SURVEILLANCE FREQUENCY

SR 3.4.16.1 Verify reactor coolant gross specific activity < bOlE In accordance withiCi/gm. the Surveillance

Frequency ControlProgram

(cbntinued)



SURVEILLANCE REQUIREMENTS (continued)

______________

SURVEILLANCE FREQUENCY

SR 3.4.16.2 NOTEOnly required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT 1-131 specific In accordance withactivity 1 .0 [lCi/gm. the Surveillance

Frequency ControlProgram

AND

Between 2 and6 hours after aTHERMALPOWER changeof> 15% RTPwithin a 1 hourperiod

SR 3.4.16.3 NOTENot required to be performed until 31 days after aminimum of 2 effective full power days and 20 days ofMODE 1 operation have elapsed since the reactor waslast subcritical for> 48 hours.

Determine E from a sample taken in MODE 1 after a In accordance withminimum of 2 effective full power days and 20 days of the SurveillanceMODE 1 operation have elapsed since the reactor was Frequency Controllast subcritical for > 48 hours. Program


300

250

I

-J

>I-0

o 200U0LU

U)

_

O

00

C,)

I.zUi-J

>D0Ui

UiU)0O

100

50

0


20 30 40 50 60 70 80 90 100

PERCENT OF RATED THERMAL POWER

Figure 3.4.16-1 (page 1 of 1)Reactor Coolant DOSE EQUIVALENT 1-131 Specific Activity

Limit Versus Percent of RATED THERMAL POWER


Qc

EXAM BANK- Q1317Unit I is heating up following a short forced outage following a reactor trip 3 days ago.Given the following:

• Current NC pressure is 1500 pisg• Current NC temperature is 385°F• Two days ago, Radiation Protection discovered dose rates were too high to

allow work in an area near the letdown line• To reduce dose rates, smaller micron NC filters were installed and letdown flow

was increased from 75 gpm to 95 gpm to aid in cleanup

1. Is Technical Specification 3.4.16 (RCS Specific Activity) applicable based oncurrent plant status?

2. Which one of the parameters evaluated per Technical Specification 3.4.16 has alower limit instated based on increases in letdown flow to greater than 80 gpm?

A. 1. Yes2. Dose Equivalent 1-131

B. I. Yes2. Gross specific activity

C. 1.No2. Dose Equivalent 1-131

D. 1.No2. Gross specific activity


Question 86006G2.4.50Emergency Core CoolingAbility to verify system alarm setpoints and operate controls identified in the alarmresponse manual.


• A large break LOCA occurred.• The ECCS suctions have been swapped to the Cold Leg Recirculation alignment.• NS suction has been swapped to the containment sump.

Current Conditions:

• FWST level is at 4%.• Containment Sump level is off scale high.• Containment pressure is 7 psig and slowly decreasing.• CETs are 560°F.• RVLIS level is 57% and slowly decreasing.• PZR Level indicates 0%.

Which ONE of the following describes:

(1) The procedure for implementation which contains the required actions?

(2) What is contained in this procedure that will aid the crew in the operation of certain valvesinside containment?

A. (I) FR-C.2 (Response to Degraded Core Cooling).(2) Instructions on ensuring power is available to valves needed for mitigation.

B. (1) FR-C.2 (Response to Degraded Core Cooling)(2) Use the OAC ‘Valves Subject to Submergence Report” to determine last known

valve positions.

C. (1) FR-Z.2 (Response to Containment Flooding)(2) Instructions on ensuring power is available to valves needed for mitigation.

D. (1) FR-Z.2 (Response to Containment Flooding)(2) Use the OAC “Valves Subject to Submergence Report” to determine last

known valve positions.‘N



QUESTION 86

Distractor Analysis

A. Incorrect. FR-C.2 is plausible sirce given conditions resemble a degradation of corecooling (RVLIS level, PZR level, sump, etc.), but the appropriate procedure entry is forcontainment flooding. Ensuring power is available is plausible if applicant recognizes thatthere could be an issue due to flooding, but incorrectly reasons that alternate powersources may be available.

B. Incorrect. Second part is correct. FR-C.2 is described in “A” above.

C. Incorrect. First part is correct. Second part plausibility described in “A” above.

D. CORRECT. FR-Z.2 is the Response to Containment Flooding. Step 3 of this procedurecontains the guidance for using the QAC printout, “Valves Subject to SubmergenceReport” for determining last known affected valve positions. This action would be takendue to an excessively high level in the containment sump for the conditions.

References:• FR-Z.2 (Response to Containment Flooding), Revision 005• FR-C.2 (Response to Degraded Core Cooling), Revision 022

KA Match:Question 86006G2.4.50Emergency Core CoolingAbility to verify system alarm setpoints and operate controls identified in the alarmresponse manual.This KA is matched because the question is testing the ability to evaluate various parameterswhich may, or may not have exceeded and setpoint for alarm. The selection of proceduresuses the knowledge of the alarm setpoint verification to arrive at the correct answer.

Cognitive Level: HighThis is a higher cognitive level question because it involves an array of plant conditions, anevaluation of them, and a determination of which procedure will provide instruction formitigation.

Source of Question: Bank CNS 693 - Sig Mod

SRO Only:SRO Only:This question meets the following criteria for an SRO only question as described in the“Clarification Guidance for SRO-only Questions (Rev. I dated 03/11/2010) under the ScreenCriteria for questions linked to I OCFR55.43(b)(5) (Assessment and Selection of Procedures):






overall mitigative strategy of a procedure.5. The question does involve assessing plant conditions (involving FR-Z.2 content), and

then selecting a section (specific step for a specific purpose) to mitigate a ContainmentFlooding Event.



I

CNS RESPONSE TO CONTAINMENT FLOODING PAGE NO.EPI1IAI5000IFR-Z.2 4 of 5

Revision 5

ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED

2. (Continued)

f. Evaluate other possible sources asfollows:

• Spent fuel pool•CA• CF• FWST.

— g. IF leakage into containment issuspected from system(s) listed inprevious step, THEN isolate affectedsystem(s) from containment.

NOTE Water penetrating limit switch assembly on valves without leakproof rotor switchhousings may cause an erroneous valve indication on OAC and control boardinstrumentation.

— 3. Use OAC printout, “VALVES SUBJECTTO SUBMERGENCE REPORT”, todetermine last known affected valvepositions.

— 4. RETURN m procedure and step ineffect.

END

Q-EXAM BANK - Q 693During a large break LOCA, the ECCS suctions have been swapped to the Cold LegRecirculation alignment. NS suction has been swapped to the containment sump. Allequipment is running as expected. The following conditions are present:

• FWSTlevel=4%• Containment Sump level is off scale high• Containment pressure = 7 PSIG and slowly decreasing• CET=56OdegF• RVLIS Level is 57% and slowly decreasing• PZR Level = 0%

Which one of the following statements correctly states the concern for the aboveconditions and the procedure the CRS must enter?A. The level of water in the core region has been reduced such that core cooling

has been lost; EPII/AI5000IFR-C.2 (Response to Degraded Core Cooling)

B. The level of water in the core region has been reduced such that the core hasbecome uncovered; EPIIIN5000IFR-C.2 (Response to Degraded CoreCooling)

C. Containment sump level is higher than would be expected due to a damagedRN or KC pipe; EPIIIAI5000/FR-Z.2 (Response to Containment Flooding)

D. Containment sump level is high due to the input from the reactor coolantsystem and Refueling Water Storage Tank; EP/1/N5000!FR-Z.2 (Response toContainment Flooding)


Question 87 7-,013G2.4.2Engineered Safety Features ActuationKowledgeofitem set points, interlocks, and automatic actions associated with EOP )

--éntry conditions

In accordance with Tech. Spec. 3.3.2 BASESJESFAS Instrumentation):

_\(1) What is the BASIS for the f&19wiT interlock: Reactor Trip P-4?

(2) In accordance with Tech Spec 3.3.2 BASES, the related functions provided by P-4(2) required in order to meet the unit licensing basis safety analysis

acceptance criteria.

A. (1) To avert a continued cooldown upon a reactor trip.(2) are

B. (1) To avert a continued cooldown upon a reactor trip.(2) are not

C. (1) To permit a normal Unit cooldown and depressurization without actuation of SafetyInjection or Main Steam Isolation.

(2) are

D. (1) To permit a normal Unit cooldown and depressurization without actuation of SafetyInjection or Main Steam Isolation.

(2) are not

,\ ‘c_, -



QUESTION 87

Distractor Analysis

A. Incorrect. First part is correct. Plausible to reason that an important permissive interlocksuch as P.4 would be required in order to meet safety analysis acceptance criteria,however the basis document does not support this.

B. CORRECT. As described in the TS Bases for ESFAS, the P-4 permissive interlockfunctions to allow reset of ECCS without receiving another SIAS, is also an input to theability to use steam dumps, and the feed pump runback to prevent overfilling (andovercooling) the S/Gs. Also per the Basis, this function is NOT required in order to meetthe unit licensing basis safety analysis acceptance criteria.

C. Incorrect. Plausibility of first part is described in “D” below. Second part described in “A”above.

D. Incorrect. Second part is correct. Plausible that the P-4 permissive interlock would permita normal cooldown and depressurization, because it does allow operations that mightotherwise cause an SI, by blocking circuits. However, this is the incorrect function.

References:• TS Bases for TS 3.3.2, .ESFAS,

KA Match:Question 8701 3G2.4.2Engineered Safety Features ActuationKnowledge of system set points, interlocks, and automatic actions associated with EOPentry conditions.The KA is matched because the question tests knowledge of a permissive interlock associatedwith a reactor trip.



SRO Only:This question meets the following criteria for an SRO only question as described in the“Clarification Guidance for SRO-only Questions (Rev. I dated 03/11/2010) under the ScreenCriteria for questions linked to IOCFR55.43(b)(2) (Tech Specs):

1. It cannot be answered solely by knowing 1 hour TS/ SIC Action.2. It cannot be answered solely by knowing the LCO/SLC information listed above the line.3. It cannot be answered solely by knowing the TS Safety Limits.4. The question involves application of required actions of Tech Spec 3.3.2 (ESFAS

Instrumentation).



5. The question involves knowledge of TS bases that is required to analyze TS requiredactions and terminology.



ESFAS InstrumentationB 3.3.2

BASES

APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

a. Engineered Safety Feature Actuation SystemInterlocks—Reactor Trip, P-4

The P-4 interlock is enabled when a reactor trip breaker(RTB) and its associated bypass breaker is open. Operatorsare able to reset SI 60 seconds after initiation. If a P-4 ispresent when SI is reset, subsequent automatic SI initiationswill be blocked until the RTBs have been manually closed.This Function allows operators to take manual control of SIsystems after the initial phase of injection is complete whileavoiding multiple SI initiations. The functions of the P-4interlock are:

• Trip the main turbine;

• Isolate MFW with coincident low Tavg;

• Prevent reactuation of SI after a manual reset of SI;

• Transfer the steam dump from the load rejectioncontroller to the unit trip controller; and

• Prevent opening of the MFW isolation valves if theywere closed on SI or SG Water Level—High High.

Each of the above Functions is interlocked with P-4 to avertor reduce the continued cooldown of the RCS following areactor trip. An excessive cooldown of the RCS following areactor trip could cause an insertion of positive reactivity witha subsequent increase in generated power. To avoid such asituation, the noted Functions have been interlocked with P-4as part of the design of the unit control and protectionsystem.

None of the noted Functions serves a mitigation function inthe unit licensing basis safety analyses. Only the turbine tripFunction is explicitly assumed since it is an immediateconsequence of the reactor trip Function. Neither turbinetrip, nor any of the other four Functions associated with thereactor trip signal, is required to show that the unit licensingbasis safety analysis acceptance criteria are not exceeded.

The RTB position switches that provide input to the P-4interlock only function to energize or de-energize or open orclose contacts. Therefore, this Function has no adjustable



Question 88059A2.04Main FeedwaterAbility to (a) predict the impacts of the following malfunctions or operations on the MFW;and (b) based on those predictions, use procedures to correct, control, ormjttethe_consequences of thosenaIfunctions or operations:Feeding a dry

Gwen the following Unit I initial conditions:

• The Unit was at 100% power.

Subsequent:• A Loss of Secondary Heat Sink event has occurred.• All Tcolds are approximately 365°F.• SI has initiated.• FR-H.1, (Response to Loss of Secondary Heat Sink) is being implemented.• NC temps are slowly rising.• All SG levels are indicating 0% WR. “

• The SRO has decided to use CF as a feed source.

In accordance with FR-H.1:

(1) In order to use CF as a feed source the CF isolation signal must be:

(2) If that action is NOT successful, what procedure implementation is required9

A. (1) Reset AND then Bypassed.(2) Remain in FR-H.land continue attempts to restore secondary heat sink.

B. (1) Reset AND then Bypassed.(2) Refer to OP/I /A/6250/001, (Condensate and Feedwater System) and attempt to place

CM System in service.

C. (1) Bypassed ONLY(2) Remain in FR-H.1 and continue attempts to restore secondary heat sink.

D. (1) ypssedON1(2) Rto 0P111A162501001, (Condensate and Feedwater System) and attempt to place

,M Syteminservice

/ ) / (



QUESTION 88

Distractor Analysis

A. Incorrect. Reset and bypass is plausible, since reset OR bypassed is in a subsequentstep (Step 12) and does not apply for Step 11 which is specific to resetting feedwaterisolation. Second part is correct.

B. Incorrect. Plausibility of first part described in “A” above. Plausible that a systemoperating procedure would contain instructions for the desired operation, but FR-H.1specifically requires continuing attempts to restore a heat sink while remaining in FR-H.1.

C. CORRECT. Step I 1.c of FR-H.1 requires that IAE be called to BYPASS the feedwaterisolation signal as part of using CF as a feed source for the given conditions. Step 34requires the SRO to remain in FR-H.1 and continue attempts to establish a secondaryheat sink in at least one SIG.

D. Incorrect. First part is correct. Second part plausibility is described in “B” above.

References:• EM/I /A152001009, Bypassing Feedwater Isolation• FR-H.1, (Response to Loss of Secondary Heat Sink)• OPII/A162501001, (Condensate and Feedwater System)

KA Match:Question 88059A2.04Main FeedwaterAbility to (a) predict the impacts of the following malfunctions or operations on the MFW;and (b) based on those predictions, use procedures to correct, control, or mitigate theconsequences of those malfunctions or operations:Feeding a dry S!GThis KA is matched because plant conditions involve S/Gs that are dry (0% level), and theapplicant must predict how the operation to feed these SIGs affects the feedwater system (i.e.,reset vs. bypass an isolation signal), and then use detailed knowledge of procedure content toselect which procedure is the correct one to use.

Cognitive Level: HighThis is a higher cognitive level question because the applicant must analyze a set of conditions,determine that they involve feeding a dry SIG, and then apply detailed system knowledge ofreset vs. bypassing a feedwater isolation signal to determine the correct action and procedureselection.


SRO Only:



This question meets the following criteria for an SRO only question as described in the“Clarification Guidance for SRO-only Questions (Rev. I dated 03/11/2010) under the ScreenCriteria for questions linked to I OCFR55.43(b)(5) (Assessment and Selection of Procedures):




overall mitigative strategy of a procedure.5. The question does involve assessing plant conditions for a Loss of Secondary Heat

Sink, and then selecting a section (specific steps for a specific purpose) to mitigate theconditions.



CNS RESPONSE TO LOSS OF SECONDARY HEAT SINK PAGE NO.EPI1IAI5000/FR-H.1 40 of 88

Revision 41I.


— 32. Verify KC flow to ND heat exchangers - — IF AT ANY TIME an ND pump is operatingINDICATING FLOW. with flow less than 1000 GPM to NC

loops AND KC to associated ND HX isisolated, THEN stop affected ND pumpwithin 3 hours.

33. Align CA to establish control of SIG feedas follows:

— a. Ensure CA System valve control -

RESET.

— b. CLOSE CA flow control valves on S!Gsnot presently being fed.

34. Continue attempts to establishsecondary heat sink in at least one SIGas follows:

.. CA. REFER IQ. Steps 6 through 7

• CF or CM. REFER]QSteps 10through 17.

35. Verify N/R level in at least one SIG - RETURN IQ. Step 34.GREATER THAN 11% (29% ACC).

36. Verify NC System temperatures as — RETURN J Step 34.follows:

—• Core exit T/Cs - DECREASING—• All NC T-Hots - DECREASING.

CNS RESPONSE TO LOSS OF SECONDARY HEAT SINK PAGE NO.EP/i/A15000/FR-H.1 11 of 88

Revision 41


11. (Continued)

— c. Notify IAE to bypass FeedwaterIsolation. REFER TO EM/i /A/5200/009(Bypassing Feedwater Isolation).

d. Ensure S/I - RESET:

_i) ECCS. _1) Reset ECCS.REFERTOEP/1/A15000/G-1 (GenericEnclosures), Enclosure 4 (ECCSMaster Reset).

— 2) D/G load sequencers. 2) Dispatch operator to open affectedsequencer(s) control power breaker:

—. I EDE-FOl F (Diesel GeneratorLoad Sequencer Panel I DGLSA)(AB-577, BB-46, Rm 496)

—e 1EDE-FO1F (Diesel Generator

Load Sequencer Panel 1 DGLSB)(AB-560, BB-46, Rm 372).

3) IEAT ANY TIME a B/O occurs,THEN restart S/I equipmentpreviously on.

— e. if AT ANY TIME a subsequentFeedwater Isolation occurs, THENRETURN Step 11.


/Question 89061A2.06AuxiliarylEmergency FeedwaterAbility to (a) predict the impacts of the following malfunctions or operations on the AEW;and (b) based on those predictions, use procedures to correct, control, or mitigate theconsequences-ofdh6se malfunctions or operations:Back leakage of MFW -


• The Unit is at 12% power and preparing to roll the main turbine.• CIAI4I I (UI CA Temp at Chk Vlv ICA-37) alarms on the OAC.• I CA-37 is the #1 CA to S/G I D Check.• In accordance with 0P1I1A162501002 (Auxiliary Feedwater System), one of the required

actions the crew takes is to CLOSE I CA-36 (UI CA Pump Disch to ID S/G Control).

(I) What concern is addressed by the required action for the QAC alarm?

(2) Does this actionffect the operability of the CA Pump?

— (I) I

A (1) Loss of NPSH-(2) ‘-NQ———--

B. (1) Steam binding.(2) YES

j) — —

C. (1) Steam binding.(2) NO

D. (1) Loss of NPSH.(2) YES -)



QUESTION 89

Distractor Analysis

A. Incorrect. Plausibility of first part described in “D” below. Second part described in “C”below.

B. CORRECT. Closing the discharge valve makes the CA pump inoperable per the NOTEjust prior to Step 3.3 of Enclosure 4.10, “Cooldown of the Motor Driven CA Pumps Piping”in 0P111A16250/002, (Auxiliary Feedwater System). The concern is for steam binding.

C. Incorrect. First part is correct. Plausible that valve operation would not affect operabilitysince the action is being taken to restore a problem with the piping and the pump.

D. Incorrect: Plausible that loss of NPSH would be a concern for this condition, since there issome element of suction head associated with steam binding, but other operationspreclude this.

References:OP/1/N6250/002 (Auxiliary Feedwater System), Revision 145

KA Match:Question 89061A2.06Auxiliary!Emergency FeedwaterAbility to (a) predict the impacts of the following malfunctions or operations on the AFW;and (b) based on those predictions, use procedures to correct, control, or mitigate theconsequences of those malfunctions or operations:Back leakage of MFWThe KA is matched because given conditions involve backleakage of check valves in thefeedwater piping and its impact on Aux. Feedwater, and then knowledge of procedure contentregarding operability impact on the Aux. Feedwater Pump

Cognitive Level: HighThis is a higher cognitive level question because the applicant must evaluate plant conditionsand determine the effect they have on the operability of a safety related pump.

Source of Question: Bank MNS 2010 #78 Sig Mod

SRO Only:This question meets the following criteria for an SRO only question as described in the“Clarification Guidance for SRO-only Questions (Rev. I dated 03/1 1/201 0) under the ScreenCriteria for questions linked to IOCFR55.43(b)(2) (Tech Specs):




4. The question involves application of required actions of Tech Spec 3.7.5 (AuxiliaryFeedwater).

5. The question involves knowledge of TS bases that is required to analyze TS requiredactions and terminology.



Enclosure 4.10 OP/11AJ62501002Cooldown of the Motor Driven CA Pumps Page 3 of 11

Piping

NOTE: OAC point C1P1447 (Primary Thermal Output %) shall be used in determining reactorpower while this enclosure is in effect.

3.2 Control reactor power as follows:

3.2.1 IF an extended run is anticipated, ensure reactor power is maintained 98%and stable while this procedure is in effect. (R.M.)

3.2.2 IF an extended run is NOT anticipated, ensure reactor power is maintained99.5% and stable while this procedure is in effect. (R.M.)

NOTE: Closing the pump discharge valve will make the pump inoperable.

3.3 IF Enclosure 4.13 (Checking Pipe Surface Temperatures) has determined it is necessaryto cool the piping upstream of the flow control valve(s), perform the following for theapplicable pump:

_____

3.3.1 Log the applicable CA pump(s) in TSAIL.SRO

3.3.2 IF cooling the piping associated with CA Pump lA, perform the following for1CA-87 (CA Pump 1A Disch To S/G Isol) (AB-533, BB-49, Rm 256):

El Unlock the valve.El Close the valve.

3.3.3 IF cooling the piping associated with CA Pump 1B, perform the followingfor 1CA-88 (CA Pump lB Disch To S/G Isol) (AB-533, BB-50, Rm 255):

El Unlock the valve.El Close the valve.

3.3.4 Take a manual temperature reading with a pyrometer of the discharge pipingat the appropriate pump(s) and record the temperature:

• CAPump1A °F

• CAPump1B

Enclosure 4.10 OP!11AJ6250/002Cooldown of the Motor Driven CA Pumps Page 9 of 11

Piping

CAUTION: 1. Do NOT exceed a 150°F/mm rate of change at the CA nozzle feedwater inlet:

OAC PointsS/G 1 A CA Nozzle Feedwater Inlet Temp Rate Cl P1381S/G lB CA Nozzle Feedwater Inlet Temp Rate C1P1382S/G 1C CA Nozzle Feedwater Inlet Temp Rate C1P1383SIG 1 D CA Nozzle Feedwater Inlet Temp Rate Cl P1384

2. The CA pump discharge valves to the steam generators shall be opened veryslowly to prevent a water hammer in the piping.

3. No more than two check valves shall be cooled at a time.

4. If reactor power is being maintained greater than 98%, then once CA flow isinitiated, the affected motor driven CA pump shall be shutdown within 10minutes.

NOTE: • Cooling CA pump downstream piping prevents migration of hot water to CA pumpand resultant potential gas formation.

• Flow through the check valve is recommended to be increased toapproximately 200 gpm. This flowrate is sufficient to remove debris in the checkvalve seat which may have prevented the valve from seating.

• A maximum of two substeps of Step 3.11 may be performed simultaneously.

3.11 Open appropriate valve(s) very slowly to cool the CA System piping and position tomaintain the following: (R.M.)

• Proper S/G level

• CA nozzle temperature

• Approximately 200 gprn

3.11.1 IF cooling piping associated with check valve 1CA-61 (CA Pump #1A DischTo S/G 1A Check), throttle 1CA-60 (CA Pump 1A Flow To S/G 1A).

3.11.2 IF cooling piping associated with check valve 1CA-57 (CA Pump #1A DischTo S/G lB Check), throttle 1CA-56 (CA Pump lA Flow To S/G 1B).

3.11.3 IF cooling piping associated with check valve 1 CA-45 (CA Pump #1 B DischTo S/G lC Check), throttle 1CA-44 (CA Pump lB Flow To S/G 1C).

3.11.4 j cooling piping associated with check valve 1 CA-4 1 (CA Pump #1 B DischTo S/G lD Check), throttle 1CA-40 (CA Pump lB Flow To S/G lD).

Q - Or--JFOR REVIEW ONLY - DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 78 fl578

SYSO61 A2.06 - Auxiliary / Emergency Feedwater (AFW) SystemAbility to (a) predict the impacts of the following malfunctions or operations on the AFW; and (b) based on those predictions, use procedures tocorrect, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 /43.5 /45.3 /45.13)Back leakage of MFW

Given the following conditions on Unit I:

• The unit is at 12% RTP preparing to roll the main turbine• Ml Al 276 (UI CA Temp at Chk Vlv I CA-37) alarms on the QAC• ICA-37(#ITDCAtoS/GD)

Based on the above conditions:

I. In accordance with OP/11A16250/002 (Auxiliary Feedwater System), what methodwould FIRST be used to reduce the temperature at the check valve?

2. How would this action affect the operability of the TD CA Pump?

A. 1. Close ICA-36 AB (UI TD CA Pump Disch to ID SIG Control) and monitortemperature for 15 mm.

2. The U-I TD CA Pump remains OPERABLE.

B. 1. Close ICA-36 AB (UI TD CA Pump Disch to ID S/G Control) and monitortemperature for 15 mm.

2. The U-I TD CA Pump shall be declared INOPERABLE.

C. 1. Close ICA-38B (UI TD CA Pump Disch to ID SIG Isol) and start the TDCA pump aligned for recirculation to the UST.

2. The U-I TD CA Pump remains OPERABLE.

D. 1. Close ICA-38B (UI TD CA Pump Disch to ID S/G Isol) and start the TDCA pump aligned for recirculation to the UST.

2. The U-I TD CA Pump shall be declared INOPEPABI...E.

Tuesday, August 24,2010 Page 227 of 295


Question 90078G2.4.8 \ IInstrument AirKnowledge of how abnormal operating procedures are use in conjunctionwith EOP5.)

Given the following conditions:

• AP/0/A/5500/022 (Loss of Instrument Air) is in progress.• VI pressure response is as shown below:

Time 0345 0415 0430 0445VI Pressure (psig) 74 59 54 49

(1) Which ONE of the following describes the EARLIEST time the SRO must direct the reactorto be tripped, in accordance with AP/022?

(2) What MANUAL action is required following the trip?(7

A. (1) 0415(2) Close MSIVs

B. (1) 0415(2) Close CF Regs and Bypasses

C. (1) 0430(2) Close MSIVs

-

D. (1) 0430(2) Close CF Regs and Bypasses



QUESTION 90

Distractor Analysis

A. Incorrect. Correct action but time (press) corresponds to when backup sealing air isaligned to VI.

B. Incorrect, Incorrect action (valves are affected but at a lower pressure than the trip) tiedwith the incorrect time (press). Incorrect time is associated with aligning back up sealingto the compressors.

C. CORRECT. At 80 PSIG N2 becomes the primary motive force for the SIG PORVs.At 80 PSIG the CA valve accumulators will begin to discharge.At 80 PSIG the PZR PORVs may be affected.75 is when VS aligned60 is when backup sealing must be aligned to the VI Compressor.55 is when reactor trip is required and closure of the MSIV.50 CF reg erratic.S/G Level decreasing in an uncontrolled manner.MSIV goes closed due to loss of VI.

D. Incorrect. This time is correct but the incorrect action given.

References:AP/0/A/5500/022 (Loss of Instrument Air), Revision 033

KA Match:Question 90078G2.4.8Instrument AirKnowledge of how abnormal operating procedures are used in conjunction with EOPs.The KA is challenging to precisely match, since there is little direct involvement in the EOPs forthe topic; however, the abnormal aspect is met by testing the use of the AP and its requiredactions in conjunction with conditions requiring a manual reactor trip (EOP entry).

Cognitive Level: HighThis is a higher cognitive level question because the applicant must analyze and trend of a plantparameter against a time matrix to determine the earliest time that the reactor needs to betripped.

Source of Question: Bank - CNS 880

SRO Only:This question meets the following criteria for an SRO only question as described in the“Clarification Guidance for SRO-only Questions (Rev. I dated 03/11/2010) under the ScreenCriteria for questions linked to 1 OCFR55.43(b)(5) (Assessment and Selection of Procedures):






overall mitigative strategy of a procedure.5. The question does involve assessing plant conditions for a loss of instrument air, and

then selecting a section (specific steps for a specific purpose) to mitigate the conditions.



I PAGE NO. ICNS LOSSOFINSTRUMENTAIRI 12of87 I

AP101A155001022 I Enclosure 3 - Page 1 of 27 I Revision 331Unit I Loss Of VI System Actions I I


1. fJpy TIME VI pressure is less than55 PSIG AND decreasing, THEN:

— a. Trip reactor.

— b. WHEN reactor power less than 5%,THEN depress “CLOSE” pushbutton forall MSIVs.

— c. Continue in this procedure as timepermits.

— d. QjQEPI1!AI5OOOIE-O (Reactor TripOr Safety Injection).

EXAM BANK - Q 880AP101A155001022 (Loss of Instrument Air) is in progress. VI pressure is as shown

below:

Time 0345 0415 0430 0445VI Pressure 74 59 54 49

Which one of the following describes the time the SRO must direct the reactor to betripped and what action must be taken following the trip?A. 1. 0415

2. Close MSIV5

B. 1. 04152. Close CF Regs and Bypasses

C. 1. 04302. Close MSIVs

D. 1. 04302. Close CF Regs and Bypasses

- o_v_J

EXAM BANK - Q 880AP/0/A/5500/022 (Loss of Instrument Air) is in progress. VI pressure is as shown

below:

Time 0345 0415 0430 0445VI Pressure 74 59 54 49

Which one of the following describes the time the SRO must direct the reactor to betripped and what action must be taken following the trip?A. 1. 0415

2. Close MSIVs

B. 1.04152. Close CF Regs and Bypasses

C. 1. 04302. Close MSIVs

D. 1. 04302. Close CF Regs and Bypasses


Question 9101 4A2.06Rod Position IndicationAbility to (a) predict the impacts of the following malfunctions or operations on the RPIS;and (b) based on those predictions, use procedures to correct-control, ‘or mitigate theconsequences of those malfunctions or operations: Loss of LVDT )

Given the following Unit I conditions ,i

• The Unit is at 65% power.• During annunciator testing, Annunciator IAD-2, B/9 (Control Rod Bank Lo-Lo Limit)

FM..ED’fo illuminate.• (lAhas reported that a failed annunciator card must be replaced.• the part will not be available until next week.

In accordance with Operations Management Procedure 2-31 (Control Room InstrumentationStatus), which ONE of the following is required?

A. The shift work manager directs Reactor Engineering to initiate a temporary modificationto change the Control Rod Bank Lo Limit (IAD-2, N9) annunciator setpoint to the ControlRod Bank Lo-Lo rod insertion limit.

B. The unit supervisor initiates a control panel information tag for 1AD-2, B/9.

C. The operations shift manager ensures that alternate indications are monitored toduplicate the function of the failed annunciator.

D. The reactor operator enters the requirement to verify Rod Insertion Limits manuallyduring transients in the shift turnover log.



QUESTION 91

Distractor Analysis

A. Incorrect. An increased surveillance sheet must be initiated. If the applicant does notknow the requirement, this is a logical alternative.

B. Incorrect. If the applicant does not know the requirement, this is a logical alternative.

C. CORRECT. In accordance with the OMP 2-31, Section 6, the OSM will ensure thatalternate indications are monitored to duplicate the function of the failed annunciator.

D. Incorrect. If the applicant does not know the requirement, this is a logical alternative.

References:OMP 2-31, Section 6 .(Action on lailed lnstrumentationlAnnunciators, Revision 029

KA Match:Question 9101 4A2.06Rod Position IndicationAbility to (a) predict the impacts of the following malfunctions or operations on the RPIS;and (b) based on those predictions, use procedures to correct, control, or mitigate theconsequences of those malfunctions or operations: Loss of LVDTEven though this type of KA is sometimes a higher cognitive level KA, this KA has beenmatched with a question that is primarily at the lower cognitive level in the following manner:

Predicting the impact aspect is met by testing the effect of a loss of failed instrument whichmonitors insertions limits of the control rods (similar function of an LVDT which uses a linearvariable differential transformer) to provide an output. The impact is what has to be done as aresult of the failure.

Using procedures is met by recall of what the procedure requires as a result of the failure.


Source of Question: Bank - CNS 289

SRO Only:This question meets the following criteria for an SRO only question as described in the“Clarification Guidance for SRO-only Questions (Rev. I dated 03/11/2010) under the ScreenCriteria for questions linked to IOCFR55.43(b)(1), (Conditions and Limitations in the facilitylicense):

I. It involves administrative requirements for a loss of instrumentation for control rodinsertion limits.


Operations Management Procedure 2-31 Page 3 of 4

6. Action on Failed Instrumentation/Annunciators

6.1. The OSM shall ensure that alternate indications are monitored to duplicate thefunction of the failed instrumentationlannunciator.

The OSM should evaluate if an increased surveillance is necessary byconsidering, but not limited to, the following elements:

• Technical Specifications, Safety Function, Redundant indications, andalternate monitoring.

6.2. The Shift Technical Advisor shall be notified of any instrumentationdetermined to be failed.

7. Increased Surveillance Sheet Instructions

7.1. An Increased Surveillance Sheet shall be initiated when plant parameters orconditions are deemed necessary to monitor on an increased frequency.

7.2. Detail shall be provided in the Items to Monitor section to give the operatorperforming the increased surveillance specific direction.

• Specific setpoints should be included in this section if applicable.

• Engineering guidance may be required for setpoint determination.

7.3. Specific instructions shall be listed in the Actions To Perform If MonitoredCriteria Exceeded section to provide direction if setpoints are exceeded.The Actions to Perform section shall NOT be used to operate the plant.{C-1 1-6679}

7.4. Monitoring Frequency shall be set at the Unit Supervisor’s discretion.

• This shall be set to ensure any potential problem is discovered in atimely manner.

• Engineering guidance may be required to determine proper monitoringfrequency.

7.5. Increased surveillance monitoring should be terminated when theindication/annunciator/parameter/condition is restored to normal.

7.6. Active Increased Surveillance Sheets shall be filed in the appropriate sectionof the Ops Shift Routines Logbook.

• A copy of the Active Increased Surveillance Sheet shall be supplied tothe individual responsible for performing the surveillance.


Question 9201 7A2.02In-core Temperature MonitorAbility to (a) predict the impacts of the following malfunctions or operations on the ITMsystem; and (b) based on those predictions, use procedures to correct, control, ormitigate the-consequences of those malfunctions or operations:Core damage


• A small break LOCA is in progress.• FR-C.1, (Response to Inadequate Core Cooling) is in progress.• BOTH NC Pumpsinihe iWO available loops are running..jhe TSC is NOT yet staffe• ThS is at Step 26 in FR-C.I for assessing GET indications.

(1) In accordance with the Basis for Step 26 of FR-C.l, an indication that core damage willoccur is if CET temperatures are NOT LESS than (I)

(2) If the RNO column of Step 26 for CET temperatures INCREASING applies, what isrequired?

A. (1) 700°F(2) Continue attempts in FR-C.I to restore core cooling.

B. (1) 700°F(2) GO TO SACRGI, (Severe Accident Control Room Guideline Initial Response)

C. (I) i2O0°F.(2) Notify Engineering to assess core damage in accordance with RP/015, (Core Damage

- - - -

D (1) 1200°F(2) GO TO SACRGI, (Severe Accident Control Room Guideline Initial Response)

d2


)(

cc

N

jk i

<1


QUESTION 92

Distractor AnalysisA. Incorrect. First part plausibility is described in “B” below. Plausible, that with the given

conditions, and the incorrect temperature, to remain in FR-C. 1 and continue attempts torestore core cooling.

B. Incorrect. Second part is correct. 700°F is plausible since it is listed in Step 7, but it ismisapplied here, since it is related to verifying CETs indication as adequate, and not ascriteria for SACRG implementation.

C. Incorrect. First part is correct. Notifying Engineering to assess core damage is plausiblesince the listed procedure (RP/15 for Core DamageAssessment) does exist, but would notapply for the listed step. L’

D. CORRECT. Per the quoted reference, 1200°F is the temperature at which core damagebegins to occur. The RNO for the listed step requires implementation of the SevereAccident Control Room Guideline.

References:• E-1, (Loss of Reactor or Secondary Coolant), Revision 027• ES-I .2, (Post LOCA Cooldown and Depressurization), Revision 031• FR-Cl, (Response to Inadequate Core Cooling), Revision 022• FR-C.I Basis Document

KA Match:Question 9201 7A2.02In-core Temperature MonitorAbility to (a) predict the impacts of the following malfunctions or operations on the ITMsystem; and (b) based on those predictions, use procedures to correct, control, ormitigate the consequences of those malfunctions or operations:Core damageThe KA is matched because conditions involve core damage conditions, and how that wouldimpact the indications on the CETs. Then the applicant is tested on procedure implementationto mitigate the conditions.

Cognitive Level: HighThis is a higher cognitive level KA because the applicant must analyze plant conditions anddetermine that these represent core damage conditions, and then make a determination onwhich procedure flowpath to use.








overall mitigative strategy of a procedure.5. The question does involve assessing plant conditions for core damage, and then

selecting a severe accident procedure for mitigating the conditions.



CNS RESPONSE TO INADEQUATE CORE COOLING PAGE NO.EPIIIAI5000IFR-C.1 26 of 40

Revision 22

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

— 26. Verify Core Exit TCs - LESS THAN Perform the following:1200°F.

a. IF core exit temperatures decreasing,THEN RETURN] Step 23.

— b. j additional NC pumps and associatedloops available, THEN RETURN mStep 23.

— c. IF core exit temperatures increasing,THEN EG/1/AICSAM/SACRG1(Severe Accident Control RoomGuideline Initial Response).

27. Isolate CLAs as follows:

— a. Verify any ND Pump - INDICATING AT — a. Step 29.LEAST INTERMITTENT FLOW.

— b. Dispatch operator to restore power to allCLA discharge isolation valves.REFER TO EPI1IA/50001G-1 (GenericEnclosures), Enclosure 9 (PowerAlignment for CLA Valves).

c. Ensure S/I - RESET:

_1) ECCS. 1) Perform thefollowing:

a) if either reactor trip breaker isclosed, THEN dispatch operatorto open Unit 1 reactor tripbreakers.

— b) WHEN reactor trip breakersopen, THEN reset ECCS.


Question 93034G2.1 .27Fuel Handling EquipmentKnowledge of system purpose andlor function.

In accordance with the BASIS for Selected Licensee Commitment (SLC) 16.9-20 (RefuelingOperations - Crane Travel - Spent Fuel Storage Pool Building:

(I) The MAXIMUM load limit for loads over the Spent Fuel Pool is based on the weight of(1)

(2) ONE of the BASES for this limitation is to ensure that

______________________________

A. (1) One fuel and control rod assembly ONLY.(2) any distortion of fuel in the storage racks will not result in a critical array.

B. (1) One fuel and control rod assembly ONLY. PT ‘ C

(2) 10 CFR 20, (Standards for Protection Against Radiation) requirements are mel.

C. (1) One fuel and control rod assembly AND associated fuel handling tool(2) any distortion of fuel in the storage racks will not result in a critical array.

D. (1) One fuel and control rod assembly AND associated fuel handling tool(2) 10 CFR 20, (Standards for Protection Against Radiation) requirements are met.

‘S

____

I /

/ 2



QUESTION 93

Distractor Analysis

A. Incorrect. Fuel and control rod assembly is plausible as the two factors in the load limit,since they are the actual LOAD. Applicant fails to recognize that the tool is also part of theload restriction. The basis is correct.

B. Incorrect. Plausibility of first part is described in “A” above. Second part is plausible sincethe CFR document relates to release concerns and protection of the public.

C. CORRECT. 16.9 AUXILIARY SYSTEMS

16.9-20 Refueling Operations - Crane Travel - Spent Fuel Storage Pool Building

The restriction on movement of loads in excess of the nominal weight of a fuel and controlrod assembly and associated handling tool over other fuel assemblies in the storage poolensures that in the event this load is dropped: (1) the activity release will be limited to thatcontained in a single fuel assembly, and (2) any possible distortion of fuel in the storageracks will not result in a critical array. This assumption is consistent with the activityrelease assumed in the safety analyses.

Loads in excess of 3000 pounds shall be prohibited from travel over fuel assemblies in thestorage pool.

D. Incorrect. First part is correct. Plausibility of second part described in “B” above.

References:• BASIS for Selected Licensee Commitment (SLC) 16.9-20 (Refueling Operations - Crane

Travel - Spent Fuel Storage Pool Building:• Catawba UFSAR Section 15.7.4, “Fuel Handling Accidents in the Containment and

Spent Fuel Storage Buildings”

KA Match:Question 93034G2.1 .27Fuel Handling EquipmentKnowledge of system purpose andlor function.The KA is matched because the question tests knowledge of the weight of fuel handlingequipment in the context of load restrictions over the Spent Fuel Pool.



SRO Only:



This question meets the following criteria for an SRO only question as described in the“Clarification Guidance for SRO-only Questions (Rev. I dated 03/11/2010) under the ScreenCriteria for questions linked to IOCFR55.43(b)(7):

1. It involves assessment of fuel handling equipment surveillance requirement acceptancecriteria (load limitations).


Refueling Operations — Crane Travel — Spent Fuel Storage Pool Building16.9-20

16.9 AUXILIARY SYSTEMS

16.9-20 Refueling Operations - Crane Travel - Spent Fuel Storage Pool Building

COMMITMENT NOTES pent fuel pool weir gates may be moved by crane over the storedfuel provided the spent fuel has decayed for> 19.5 days since lastbeing part of a core at power.

Loads in excess of 3000 pounds shall be prohibited from travel overfuel assemblies in the storage pool.

AND

The requirements of Technical Specification 3.8.2, AC Sources —

Shutdown, shall be met whenever loads are moved over the storagepool.

APPLICABILITY: With fuel assemblies in the storage pool.

REMEDIAL ACTIONS


A. COMMITMENT not met. A.1 Place the crane load in a Immediatelysafe condition.

TESTING_REQUIREMENTS

TEST FREQUENCY

TR 16.9-20-1 NOTESpent fuel pool weir gates may be moved by crane overthe stored fuel provided the spent fuel has decayed for19.5 days since last being part of a core at power.

Verify that the weight of each load, other than a fuel Prior to movingassembly and control rod, is 3000 pounds. the load over fuel

assemblies

Catawba Units 1 and 2 16.9-20-1 Revision 0

Refueling Operations — Crane Travel — Spent Fuel Storage Pool Building16.9-20

BASES The restriction on movement of loads in excess of the nominal weight of afuel and control rod assembly and associated handling tool over other fuelassemblies in the storage pool ensures that in the event this load is dropped:(1) the activity release will be limited to that contained in a single fuelassembly, and (2) any possible distortion of fuel in the storage racks will notresult in a critical array. This assumption is consistent with the activityrelease assumed in the safety analyses.

REFERENCES 1. Letter from NRC to Gary R. Peterson, Duke, Issuance ofImproved Technical Specifications Amendments for Catawba,September 30, 1998.

2. Letter from NRC to Gary R. Peterson, Duke, Issuance ofAmendments 198 and 191 to Facility Operating Licenses,April, 2002.

Catawba Units 1 and 2 16.9-20-2 Revision 0

CNS 2012 NRC Exam 100 Questions Final Submittal -\ J

Question 94G2.1 .23Conduct of OperationsAbility to perform specific system and integrated plant procedures during all modes ofplant operation.


• The Unit is at 100% power.• 1 EMF-33 (Condenser Air Ejector Exhaust) is in Trip 2 alarm.• IEMF-71 (N-16 Leakage) is in Trip 2 alarm.• Pressurizer level has been stabilized in accordances with APIl/AI5500I0i (Reactor

Coolant Leak), Case I (Steam Generator Tube Leak).• Letdown flow is 45 gpm.• Charging flow is 78 gpm.

j

(1) The MAXIMUM time that AP/lO allows for the unit to reach MODE 3 for these conditions is(1)

(2) In accordance with SLC 16.7-9 [Standby Shutdown System (SSS)], Condition B(Leakage), the Standby Makeup Pump (2) have to be declared NONfunctional.

Which ONE of the following completes the statements above?

A. (1) 3 hours(2) will /

B. (1) 3 hours I(2) will not i)

C. (1) 6 hours(2) will

D. (1) 6 hours(2) will not 7

A/f’I



QUESTION 94

Distractor Analysis

A. CORRECT. With the indications given, the crew will enter AP-1 0 (Reactor Coolant Leak),Case I, SG Tube Leak. This procedure directs the crew to stabilize PZR level anddetermine leak size.

Leak rate is 78-45-12= 21 gpm, making the Standby Makeup Pump INOPERABLE inaccordance with SLC 16.7-9. Step 8 of AP-1 0 Case 1, directs an SRO to evaluate ifleakage exceeds SLC 16.7-9 limits. The limit is defined as >20 GPM. PerTS 3.4.13(RCS Operational Leakage), the limit for an individual SIG tube leakage of 150 GPD wouldbe exceeded. If this leakage is exceeded, Condition B requires the unit be in Mode 3 in 6hours. Per Step 14 of AP-1 0, Case 1, if the leakage in one SIG is greater than 100 GPD,the unit is required to be in Mode 3 within 3 hours of exceeding 100 GPD.

B. Incorrect. Part (1) is correct and therefore plausible.

Part (2) is plausible if the applicant subtracts actual seal injection (20 GPM) instead of sealreturn flow (12 GPM) from Charging flow along with subtracting Letdown flow (45 GPM). Ifthat were the case the applicant would determine that total leakage would be 13 GPM (78-45-20) instead of 21 GPM (78-45-12). Since the applicant would determine leakage to beless than 20 GPM, the Standby Makeup Pump would NOT have to be declaredINOPERABLE.

C. Incorrect. Part (1) Plausible because this is correct per the requirement of Condition B ofTS 3.4.13 (RCC Operational Leakage) which requires the unit to be in Mode 3 in 6 hours.It would be reasonable for the applicant to believe this would also be the required timespecified in AP-1 0.

Part (2) is correct.

D. Incorrect. . Part (1) Plausible because this is correct per the requirement of Condition B ofTS 3.4.13 (NC Operational Leakage) which requires the unit to be in Mode 3 in 6 hours. Itwould be reasonable for the applicant to believe this would also be the required timespecified in AP-lO.

Part (2) is plausible if the applicant subtracts actual seal injection (20 GPM) instead of sealreturn flow (12 GPM) from Charging flow along with subtracting Letdown flow (45 GPM). Ifthat were the case the applicant would determine that total leakage would be 13 GPM (78-45-20) instead of 21 GPM (78-45-12). Since the applicant would determine leakage to beless than 20 GPM, the Standby Makeup Pump would NOT have to be declaredINOPERABLE.

References:• AP/1/A/5500/010, (Reactor Coolant Leak), Case I (Steam Generator Tube Leak),

Revision 056• SLC 16.7-9 [Standby Shutdown System (SSS)], Condition B (Leakage), Revision 7



KA Match:This K/A is matched because the applicant is first required to demonstrate the ability todetermine the actual SIG tube leakage, and then required to interpret this information as itapplies to procedural direction from AP-1 0 for leakage being greater than tech specs and theapplication of SLC 16.7-9 limit on leakage.

Cognitive Level: HighThis is a higher cognitive level question because the applicant must perform calculation (solve aproblem) and then perform a level of analysis concerning the given indications and predict theimpact and determine the correct procedural course of action.


SRO Only:This question meets the following criteria for an SRO only question as described in theClarification Guidance for SRO-only Questions Rev Idated 03/11/2010 for screening questionslinked to I OCFR55.43(b)(5) (Assessment and selection of procedures):1) The question can NOT be answered solely by knowing systems knowledge.2) The question can NOT be answered by knowing immediate operator actions. Neither of theactions described are immediate actions.3) The question can NOT be answered solely by knowing entry conditions for AOP or directentry conditions for EOPs. These are detailed procedure steps from AP-lO.4) The question can NOT be answered solely by knowing the purpose, overall sequence ofevents, or overall mitigative strategy of the procedure. This is detailed knowledge of procedurecontent related to knowing the plant shutdown requirements.5) The question also requires the applicant to recall a “below the bar” TS (SLC) limit associatedwith the S/G tube leakage. Therefore, it is SRO knowledge.


CNS I REACTOR COOLANT LEAK I PAGE NO. IAP/1/N5500/O1O I I 11 of 158 ICase IRevision 561Steam Generator Tube Leak I

I ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I7. Minimize Secondary contamination as

follows:

a. Remove CM polishing demineralizersfrom service as follows:

— 1) Ensure “POLSH DEMIN BYP CTRL”- IN MANUAL.

— 2) Ensure “POLSH DEMIN BYP CTRL”- OPEN.

— 3) Notify Secondary Chemistry CMpolishing demineralizers have beenbypassed.

— b. Align auxiliary systems to minimizesecondary side contamination. REFERTO EPI1IAI5000IG-1 (GenericEnclosures), Enclosure 2 (MinimizingSecondary Side Contamination).

— c. Stop any transfer of water betweenboth Unit’s CSTs.

8. Ensure compliance with appropriateTech Specs and Selected LicenseeCommitments Manual:

—. 3.4.13 (RCS Operational Leakage)

—. 3.4.14 (RCS Pressure Isolation Valve(PIV) Leakage)

—. 3.5.5 (Seal Injection Flow)

• 3.7.17 (Secondary Specific Activity)

• SLC 16.7-9 (Standby Shutdown System).

CNS REACTOR COOLANT LEAK I PAGE NO.AP!11N55001010Case I l9of 158

Steam Generator Tube Leak Revision 561


14. Determine unit shutdown requirementsas follows:

a. !EAINY TIME leak rate is greaterthan or equal to 100 gpd, THENperform the following:

— 1) Ensu eactor power less than 50%with I h hrs

— 2) Ensure unit Mode 3 within thefoIlowir

— 3) Observe Note prior to Step 15 andQIQ Step 15.

b. if leak rate is greater than or equal to75 gpd and less than 100 gpd, THENperform the following:

1) IEAIANY TIME the followingconditions are met:

—. Any main steam line N-16

radiation monitor - INOPERABLE

AND

Cl P0187 (Estimated Total Pri ToSec Leakrate) - INVALID.

THEN perform the following:

— a) Ensure reactor power less than50% within 1 hr.

— b) Ensure unit in Mode 3 within thefollowing 2 hrs.

— c) Observe Note prior to Step 15and QIQ Step 15.

— c. if leak rate is greater than or equal to75 gpd and less than 100 gpdsustained for one hour, THEN ensureunit in Mode 3 within 24 hrs.

SSS16.7-9

16.7 INSTRUMENTATION

16.7-9 Standby Shutdown System (SSS)

COMMITMENT The SSS shall be FUNCTIONAL.

APPLICABILITY: MODES 1, 2, and 3.

REMEDIAL ACTIONS

NOTESLC 16.2.3 is not applicable.


A. SSS non-functional. A.1 Restore SSS to 7 daysFUNCTIONAL status.

B. Total accumulative B.1 Declare the standby ImmediatelyLEAKAGE from makeup pump non-unidentified LEAKAGE, functional and enteridentified LEAKAGE, Condition A.and reactor coolantpump seal LEAKAGE>20 gpm.

C. More than one cell in a C.1 Enter Condition A. Immediately24-Volt battery bank is <

1 .36 volts on floatcharge with no othercells jumpered.

D. Required Action and D.1 Be in MODE 3. 6 hoursassociated CompletionTime of Condition A not ANDmet.

D.2 Be in MODE 4. 12 hours

Catawba Units 1 and 2 16.7-9-1 -Revis1on--7—

9-- 0J

EXAM BANK - Q 4440

Unit I is operating at 100% RTP. Given the following:

• I EMF-33 (Condenser Air Ejector Exhaust) is in Trip 2 alarm• I EMF-71 (SIG A Leakage) is in Trip 2 alarm• Pressurizer level has been stabilized using AP-1 0 (NC Leakage Within the

Capacity of Both NV Pumps)• Letdown flow is 45 GPM• Charging flow is 78 GPM

The MAXIMUM time that AP-lO allows for the unit to reach MODE 3 for the conditionsspecified is (1)

In accordance with SLC 16.9.7 (Stby S/D System) Condition C (Leakage), theStandby Makeup Pump (2) have to be declared INOPERABLE.

• Which ONE (1) of the following completes the statements above?

A. 1. 3 hours2. will

B. 1. 3 hours2. will not

C. 1. 6 hours2. will

D. 1. 6 hours2. will not

l(2CNS 2012 NRC Exam 100 Questions Final Submittal

Question 95G2.1 .35Conduct of OperationsKnowledge of the fuel-handling responsibilities of SROs.

Unit I is shutdown in Mode 6. Given the following events and conditions:

• Containment airlock doors are all open.• A full shift of qualified maintenance personnel are inside containment.• The Refueling SRO is in the Control Room.• The Fuel Handling Maintenance Supervisor is inside containment.• Refueling has been completed and the Maintenance Supervisor requests permission to

begin control rod latching.

What additional FQNlUrequirement(s) must be met to proceed with latching control rodsunder the direction of the Fuel Handling Maintenance Supervisor? 1

&

A. The Refueling SRO must be present in the Reactor Building ONLYV

B. The Fuel Handling Maintenance Supervisor must establish communications with theRefueling SRO in the Control Room ONLY..

C. The Refueling SRO must be present in containment AND the containment closuremust be maintained.

D. The Fuel Handling Maintenance Supervisor must establish communications witifheRefueling SRO in the Control Room AND containment integrity must be restored. /)

(



QUESTION 95

Distractor Analysis

A. Incorrect. If the candidate does not realize that latching control rods is a core alteration.Partially correct — the SRO must be present in containment.

B. Incorrect. The SRO must be physically present in containment and containment integritymust be established when core alterations are in progress. Plausible: If the candidatedoes not realize that latching control rods is a core alteration.

C. CORRECT. Latching rods is considered to be a “core alteration” under Tech Specs.Therefore, the Refueling SRO must be present AND containment closure must bemaintained.

D. Incorrect. The SRO must be physically present in containnient when core alterations arein progress. Plausible: If the candidate does not realize that latching control rods is a corealteration — or does not know the requirements.

References:• Lesson Plan Objective: FH-FHS-6• SLC 16.9-18• PT/0/A/41 50/022 page 8• PT/l/A/4550/OOIC page 1• MP/0/A17150/067 page 23

KA Match:Question 95G2.1 .35Conduct of OperationsKnowledge of the fuel-handling responsibilities of SROs.The KA is matched because it tests knowledge of fuel handling responsibilities of the RefuelingSRO.



SRO Only:This question meets the following criteria for an SRO only question as described in the“Clarification Guidance for SRO-only Questions (Rev. I dated 03/11/2010) under the ScreenCriteria for questions linked to IOCFR55.43(b)(7):

1. It involves assessment of fuel handling equipment surveillance requirement acceptancecriteria (load limitations).


O S - C)

EXAM BANK - Q 1299Unit I is shutdown in Mode 6. Given the following events and conditions:

• Containment airlock doors are all open.• A full shift of qualified maintenance personnel are inside containment.• The Refueling SRO is in the Control Room.• The Fuel Handling Maintenance Supervisor is inside containment. -

• Refueling has been completed and the Maintenance Supervisor requestspermission to begin control rod latching.

What additional requirements (if any) must be met to proceed with latching -controlrods under the direction of the Fuel Handling Maintenance Supervisor?A. The Refueling SRO must be in containment.

B. The Fuel Handling Maintenance Supervisor must establish -communicationswith the Refueling SRO in the Control Room.

C. The Refueling SRO must be in containment and containment integrity must berestored.

D. The Fuel-Handling Maintenance Supervisor must establish communicationswith the Refueling SRO in the Control Room and containment integrity mustbe restored.

fj


Question 96G2.2.1Equipment ControlAbility to perform pre-startup procedures for the facility, including operating thosecontrols associated with plant equipment that could affect reactivity.

During the approach to criticality in accordance with PT/0/A/41 50/019, (1/M Approach toCriticality), rod status is as follows:

• All shutdown banks are fully withdrawn.• Control Bank A are fully withdrawn.• Control Bank B is approaching fully withdrawn.• SR count rate has doubled once.

Subsequent:• During a subsequent rod withdrawal a single rod in Control Bank B drops fully into the

core.

(1) Which procedure will be used that contains the specific guidance for how the rods are to beoperated?

(2) The required action is to insert (2)

A. (1) AP/1/N5500/014, (Control Rod Misalignment), Case II, (Dropped Control Rod)(2) all control banks ONLY

B. (1) AP111A155001014, (Control Rod Misalignment), Case II, (Dropped Contro’ Rod)(2) all control banks AND all shutdown banks

C. (1) PT101A141501019, (IIM Approach to Criticality)(2) all control banks ONLY

D. (1) PT/0/A14150/019, (1/M Approach to Criticality)(2) all control banks AND all shutdown banks -)

r — —

.ju / (.



QUESTION 96

Distractor Analysis

A. Incorrect. Second part is correct. Plausibility of first part is described in “B” below.

B. Incorrect. It is plausible that a procedure with a section titled, “Dropped Control Rod” wouldbe used for a dropped rod situation. In this case, the Limits and Precautions of theapproach to criticality takes precedence over other procedures, and this is explicitly statedin the PT procedure. It is plausible to for an applicant to confuse this guidance, since theprocedure does contain guidance for1nsertingafl Control and Shutdown Banks, but this isfor a malfunction on rods in the Shutdown Banks

C. CORRECT. The procedure for approach to criticality (PT/01A14150/O1 9) contains thefollowing guidance:

IF a control rod fails to withdraw or a single rod is dropped during approach to criticalityperform one of the following:

IF malfunction is in Qpntrpi Bank,jflsert all Control Banks.IF malfunction is in Shutdown Bank, insert all Control and Shutdown Bank

\

This guidance is more conservative than that given in APII(2)15500/014, Control RodMisalignment), and therefore shall take precedence. {PIP C-06-4287}

D. Incorrect: The procedure is correct. It is plausible to for an applicant to confuse thisguidance, since the procedure does contain guidance for inserting all ControlShutdown Banks, but this is for a malfunction on rods in the Shutdown Banks.

References:• PTIO/A141501019, (IIM Approach to Criticality), Limit and Precaution 6.4, Revision 038• AP/1155001015, (Rod Control Malfunction), Revision 014• AP/1!A/5500/014, (Control Rod Misalignment), Case II, (Dropped Control Rod), Revision

016

KA Match:Question 96G2.2.1Equipment ControlAbility to perform pre-startup procedures for the facility, including operating thosecontrols associated with plant equipment that could affect reactivity.The KA is matched because the question conditions involve a startup, but it is testingknowledge of PRE-startup since it is an approach to criticality condition. The operating controlsaspect is met because the question involves knowledge of how the rods will be operated for thegiven conditions.

Cognitive Level: High



This is a higher cognitive level question because conditions are given for an approach tocriticality, and a subsequent dropped rod. Analyzing the conditions is required in order to makethe correct selection of procedure.






overall mitigative strategy of a procedure.5. The question does involve assessing plant conditions for a dropped rod during an

approach to critical, and then selecting a section (specific steps for a specific purpose) tomitigate the condition.

Therefore, this is an SRQ only question.


PT/O/A/4 150/019Page 4 of 13

6.4 IF a control rod fails to withdraw or a single rod is dropped during approach tocriticality perform one of the following:

• IF malfunction is in Control Bank, insert all Control Banks.• IF malfunction is in Shutdown Bank, insert all Control and Shutdown Banks.

This guidance is more conservative than that given in AP/1(2)/5500/015, RodControl Malfunction, and therefore shall take precedence. {PIP C-06-4287}

6.5 IF more than one control rod drops during the approach to criticality, MANUALLYTRIP the reactor per AP/1(2)/A/5500/14, Control Rod Misalignment.

6.6 IF an alarm is received on the Rod Control System or DRPI requiring rodwithdrawal to be halted AND IAE cannot determine cause for alarm and repairproblem or determine if further rod withdrawal is permissible, reinsert all ControlBanks.

• IF Startup is NOT xenon-free (xenon worth 100 pcm), take action within 30minutes of malfunction.

• jstartup is xenon-free (xenon worth < 100 pcm), take action within 60 minutesof malfunction.

• Obtain ICRR data at 10-minute intervals for the duration of the delay.

6.7 To reduce uncertainties in achieving criticality, T-AVG shall be maintained between555 and 559 °F during approach to criticality.

6.8 IF diluting with BDMS enabled, periodically monitor and reset the BDMS actuationsetpoint.

6.9 IF abnormal changes in count rate (i.e., irregular count rates, instrument drift, etc.)are observed on either Source Range or BDMS detector, rod withdrawal shall besuspended. Rod withdrawal may be resumed only after the source of theabnormality has been identified and it has been determined that it will not jeopardizeplant safety.

6.10 IF it is expected that criticality will be achieved above the Upper Allowable Limit(UAL)/Rod Withdrawal Limits OR below the Lower Allowable Limit (LAL)/RodInsertion Limits, insert all control banks and contact Reactor Systems EngineeringSupervisor or designee.

6.11 Ensure that the NC boron sample used for reactivity balance calculations isrepresentative of current NC system boron (i.e. taken with all four NC pumpsoperating, sufficient time allowed for mixing after last boron change, etc.).

6.12 Prerequisite steps in Sections 4, 7, and 8 may be signed off in any order.

\CNS 2012 NRC Exam 100 Questions Final Submittal

Question 97G2.2.5Equipment ControlKnowledge of the process for making design or operating changes to the facility.

Which one of the following changes requires a I OCFR5O.59 review?

A. Change to the Physical Security Plan that reduces the shift staffing requirements forsecurity guards.

B. Revision to the Emergency Plan changes the designated assembly areas foraccountability.

C. System modification that adds a backup Nitrogen accumulator to an air operatedcontainment isolation valve.

D. Change to the Nuclear Quality Assurance Plan



QUESTION 97

Distractor Analysis

A. Incorrect. Security Guard staffing is not covered under Tech Specs but underI OCFR5O.72. Plausible: If the applicant is not familiar with the requirements for USQs.Some station staffing requirements are covered under Tech Specs.

B. Incorrect. The emergency plan is changed under the IOCFR5O.54q process which issimilar in concept to I OCFR5O.59 but not the same. Plausible: If the applicant is notfamiliar with the requirements for USQs.

C. CORRECT. Containment isolation valves are Tech Spec SSCs

D. Incorrect. Does not require a 1OCFR5O.59 evaluation. The QA Plan is not a Tech SpecSSC. Plausible: If the applicant is not familiar with the requirements for USQs. NQA iscovered under I OCFR5O Appendix B.

References:NSD 203, (Operability/Functionality), Page 9, Revision 024

KA Match:Question 97G2.2.5Equipment ControlKnowledge of the process for making design or operating changes to the facility.The KA is matched because it tests knowledge of I OCFR5O.59 reviews and for which situationsthis type of review would be required.



SRO Only:This question meets the following criteria for an SRO only question as described in the“Clarification Guidance for SRO-only Questions (Rev. 1 dated 03/11/2010) under the ScreenCriteria for questions linked to I OCFR55.43(b)(3) (Facility licensee procedures required foroperating changes in the facility):

I. It involves processes for changing the plant or plant procedures.



VERIFY HARD COPY AGAINST WEB SITE IMMEDIATELY PRIOR TO EACH USENSD 203 Nuclear Policy Manual — Volume 2• Ensuring that any proposed compensatory actions are evaluated per the appropriate process(es), such as

10CFR5O.59, and reviewed by the Plant Operations Review Committee when required; and• Requesting a Challenge Board in accordance with 203.6.7 when desired.

203.4.4 REGULATORY COMPLIANCE

Regulatory Compliance is responsible for:

• Assessing the timeliness of operability determinations to provide additional assurance that they are performedcommensurate with safety significance;

• Supporting others, as needed, in establishing the current licensing basis for affected SSCs;• Reviewing and providing assistance with operability determinations when requested;• Interfacing with the NRR Project Manager, Region II personnel and the NRC resident inspector, as needed;• Providing the overall lead for NSD 203, monitoring performance, and implementing actions to improve

performance; and

• Assisting others with the management of OBDN items per Appendix D, “Corrective Action Considerations.”

203.4.5 PREPARERS

Preparers are responsible for:

• Preparing operability determinations and Formal Functionality Assessments per this directive;• Initiating action, as appropriate, to address extent of condition following preparation of an immediate

determination of operability (refer to Appendix F. 10)

• Obtaining Manager/Supervisor (or designee) concurrence when appropriate; and• Obtaining OSM concurrence on operability determinations and Formal Functionality Assessments.

203.4.6 MANAGERS AND SUPERVISORS

Managers and Supervisors (or their designees) are responsible for:

• Concurring with operability determinations and Formal Functionality Assessments when directed.

203.4.7 CHECKERS, APPROVERS, AND PEER REVIEWERS

Operability determinations and Formal Functionality Assessments are not required to be independently verified fortechnical accuracy by Checkers. Furthermore, they are not required to be approved by Approvers nor are theyrequired to be reviewed by Peer Reviewers. Such activities are optional unless deemed appropriate by responsiblesupervisors or managers (or designees). Notwithstanding, some operability determinations and functionalityassessments may rely upon calculations or other engineering documents that require independent verification andapproval. In such cases, the requirements for independent verification and approval are provided in the applicablegoverning procedure(s). For example, if an operability determination relies upon a calculation, then that calculationmust be checked and approved in accordance with EDM- 101.

203.5 DEFINITIONS

To facilitate identification, definitions are listed in alphabetical order below:

4REVISION 24

VERIFY HARD COPY AGAINST WEB SITE IMMEDIATELY PRIOR TO EACH USE

VERIFY HARD COPY AGAINST WEB SITE IMMEDIATELY PRIOR TO EACH USENuclear Policy Manual — Volume 2 NSD 203

T0CFR5O.59 — Regulation that establishes the conditions under which Part 50 licensees may make changes tothe facility or procedures and conduct tests or experiments without prior NRC approval. Refer to NSD 209,“1OCFR5O.59 Process,” for guidance regarding application of 1OCFR5O.59.

2. 1 OCFR72.48 — Regulation that establishes the conditions under which Part 72 licensees may make changes tothe facility or procedures and conduct tests or experiments without prior NRC approval. Refer to NSD 211,“IOCFR72.48 Process,” for guidance regarding application of 1OCFR72.48.

3. Compensatory Measure — An interim action that may be used during the Period of Interim Operation to:a. Enhance, maintain or restore the capability of TS SSCs to perform their specified safety function(s) or to

otherwise compensate for degradedJnonconforming conditions that call into question operability; orb. Enhance, maintain, or restore the capability of non-TS SSCs to perform their specified function(s)

including those functions required to meet associated SLC Manual COMMITMENTs or to otherwisecompensate for degraded or nonconforming conditions that call into question functionality;

Refer to Appendix D,2, “Compensatory Measures,” for additional information.

4. Current Licensing Basis (CLB) — According to 1OCFR54.3, the CLB is the set of NRC requirements applicableto a specific plant and a licensee’s written commitments for ensuring compliance with and operation withinapplicable NRC requirements and the plant-specific design basis (including all modifications and additions tosuch commitments over the life of the facility operating license. The CLB includes, but is not limited to:a. NRC regulations contained in 10 CFR parts 2, 19, 20, 21,26,30,40,50,51,54,55,70,72,73, 100 and

appendices thereto

b. Commission orders, License conditions, Exemptionsc. Technical Specifications (TSs) and TS Basesd. Plant-specific design-basis information defined in 10 CFR 50.2 as documented in the most recent updated

final safety analysis report (UFSAR) as required by 10 CFR 50.71.e. Commitments remaining in effect that were made in docketed licensing correspondence (such as licensee

responses to NRC bulletins, generic letters, enforcement actions, licensee event reports) and those reliedon to grant, amend, or modify the operating license and technical specifications and to ensure continuedcompliance and operation within applicable NRC requirements.

f. Commitments documented in NRC safety evaluations.

5. Degraded/Nonconforming Condition (DNC) — A condition of an SSC (within the scope of Section 203.2) inwhich one or more of the following conditions exist:

• There has been any loss of required quality or functional capability. Examples of degraded conditions arefailures, malfunctions, deficiencies, deviations, and defective material and equipment. Examples ofconditions that can reduce functional capability of a system are aging, erosion, corrosion, improperoperation, and maintenance [Degraded Condition]

• There is failure to confonu to all aspects of the CLB [Nonconforming Condition]

The term DNC only has meaning relative to the CLB. It does not apply to losses of quality or functionalcapability that are not credited in the CLB nor does it apply to nonconformances outside the CLB.

6. Design Bases - That information which identifies the specific functions to be perfonned by a SSC and thespecific values or ranges of values chosen for controlling parameters as reference bounds for design. Thesevalues may be (1) restraints derived from generally accepted “state of the art” practices for achieving functionalgoals, (2) requirements derived from analysis (based on calculation or experiment) of the effects of a postulatedaccident for which a SSC must meet its functional goals. Design bases information, defined by 1OCFR 50.2, isdocumented in the UFSAR as required by 1 OCFR5O.7 1. The design basis of safety-related SSCs is established

‘ NRC Regulatory Guide 1.186, “Guidance and Examples for Identifying 1OCFR5O.2 Design Bases,” endorsesAppendix B to Nuclear Energy Institute (NET) document NET 97-04, “Guidance and Examples for Identifying1OCFR5O.2 Design Bases.”

REVISION 24 5VERIFY HARD COPY AGAINST WEB SITE IMMEDIATELY PRIOR TO EACH USE


NSD 203 Nuclear Policy Manual — VoLume 2

corrective actions cannot be completed as planned, then appropriate justification shall be providedprior to entry into a mode or other specified condition in the Applicability.

• A ventilation system not described in TSs maybe required in the summer to ensure that SSCsdescribed in TSs can perform their specified safety functions but may not be required in the winter. Ifan operability determination concludes that the ventilation system does not currently perform anecessary and required support function, then the supported systems in TSs are currently operable.However, adequate controls should be established to ensure that the basis for determining that theventilation system is not required remains valid. In addition, corrective actions should be establishedto ensure timely restoration of the ventilation system.

g. Facility operation should be consistent with the CLB for the facility. Thus, degradedlnonconformingconditions should be resolved in a time frame commensurate with their safety significance (refer toAppendix D.3)

Operability and functionality are separate from corrective action to restore full compliance to the CLB, including allapplicable codes and standards, design criteria, safety analyses assumptions and specifications, and licensingcommitments. Corrective actions to restore full compliance to the CLB should be addressed through the correctiveaction process and are not completely addressed by this directive. The treatment of operability and functionality as aseparate issue from the restoration of full compliance emphasizes that the operability determination andfunctionality processes are focused on safe plant operation and should not be impacted by decisions or actionsnecessary to plan and implement corrective actions (i.e., restore full compliance).

D.2 COMPENSATORY MEASURES

The guidance in this section should be used in concert with Appendix AS, “Compensatory Measure Flowchart,”NSD 209 ‘10 CFR 50.59 Process,” and NSD 228, “Applicability Determination.” Summary information from thosedocuments is included here for convenience only per PIP M-08-3735 and other operating experience documents.Always refer to the controlling procedures for current guidance.

Compensatory measures (also called compensatory actions, interim actions, prudent measures, or preliminaryconservative measures) are interim actions/measures that may be used during the Period of Interim Operation to:

a. Enhance, maintain or restore the capability of TS SSCs to perform their specified safety function(s) or tootherwise compensate for degraded/nonconforming conditions that call into question operability.

b. Enhance, maintain, or restore the capability of non-TS SSCs to perform their specified function(s)including those functions required to meet associated SLC Manual COMMITMENTs or to otherwisecompensate for degraded/nonconforming conditions that call into question functionality.

Compensatory measures should have minimal impact on plant operations and should be relatively simple toimplement. Implementing compensatory measures for SSCs that have been determined to be degraded ornonconforming may restore plant operating margins.

In general, there are two types of compensatory measures: (1) those that do g constitute changes and (2) those thatconstitute changes. According to NEI 96-07, Revision 1, “change” means a modification or addition to, or removalfrom, the facility or procedures that affects: (1) a design function, (2) method of performing or controlling thefunction, or (3) an evaluation that demonstrates that intended functions will be accomplished.

Compensatory measures that do not constitute changes include, but are not limited to, the following:

• Adding additional inventory to the Refueling Water Storage Tank in accordance with an approved operatingprocedure.

• Using an alternate means to monitor Reactor Coolant Pump lower reservoir oil levels as outlined in an approvedalarm manual.

• Taking actions specified by Test Acceptance Criteria (TAC) sheets or similar type documents that have beenevaluated under 1OCFR 50.59.

• Increasing the frequency of operator rounds or monitoring activities.

32 REVISION 24


- a c

EXAM BANK - Q 1292Which one of the following changes will require a I OCFR5O.59 review?

A. Change to the Physical Security Plan that reduces the shift staffingrequirements for security guards.

B. Revision to the Emergency Plan changes the designated assembly areas foraccountability.

C. System modification that adds a backup Nitrogen accumulator to an airoperated containment isolation valve.

D. Changes to the Nuclear Quality Assurance Plan


Question 98G2.3.1 5Radiation ControlKnowledge of radiation monitoring systems, such as fixed radiation monitors andalarms, portable survey instruments, personnel monitoring equipment, etc.

A seismic event has occurred that was felt in the plant.

In accordance with RPIO/A150001007, (Natural Disaster and Earthquake), the SRO makes therequired initial assumption regarding operability of EMF38(L) (Containment Particulate).

What course of action is specifically required by the procedure (RP/07)?

A. Enter Tech Spec 3.0.3

B. Enter Tech Spec 3.4.15, RCS Leakage Detection Instrumentation

C. Align EMF38(L) to Upper Containment ONLY.

b. Implement manual sampling of containment particulate activity.



QUESTION 98

Distractor Analysis

A. CORRECT. Note prior to Step 1.1 of RP/07The four Reactor Coolant Leakage Detection Systems are not seismically qualified andmust be assumed to be inoperabJe-followirg any sejsmicevenLEME3S(L) can be verifiedto be operable based orçpoweravailabllityandiample pump operatiorL )

Technical Specification 3.0.3 is entered until one of the Reactor Coolant leakage detectionsystems identified in step 1.1 is declared operable.

1.1 Following any earthquake that is felt in the plant or is recorded on instrumentation,including earthquakes smaller than OBE, declare all four Reactor Coolant LeakageDetection Systems (listed below) as inoperable:

1.1.1 Containment Floor and Equipment Sump Level Monitors and the Incore InstrumentSump Level Alarm

1.1.2 VUCDT Level Monitoring System)

1.1.3 EMF3S(L)

B. Incorrect. Plausible, since EMF38(L) is assumed to be INOPERABLE, but TS 3.0.3 isrequired FIRST, per RP/07.

C. Incorrect. Applicant confuses guidance for an alternate alignment, which actually rendersthe EMF inoperable.

D. Incorrect: Plausible, since EMF38(L) is assumed to be INOPERABLE, and other EMFs inthe plant when inoperable, require manual sampling.

References:• RP/0/A15000/007, Natural Disaster and Earthquake, Enclosure 4.4 (Earthquake), Rev.

033

Question 98G2.3.1 5Radiation ControlKnowledge of radiation monitoring systems, such as fixed radiation monitors andalarms, portable survey instruments, personnel monitoring equipment, etc.The KA is matched because it tests knowledge of the effect of a seismic event on a particularradiation monitor (operability, etc.)

Coinitive Level: Low




SRO Only:This question meets the following criteria for an SRO only question as described in the“Clarification Guidance for SRO-only Questions (Rev. I dated 03/11/2010) under the ScreenCriteria for questions linked to I OCFR55.43(b)(2) (Tech Specs):

1. It cannot be answered solely by knowing < 1 hour TS/ SLC Action.2. It cannot be answered solely by knowing the LCO/SLC information listed above the line.3. It cannot be answered solely by knowing the TS Safety Limits.4. The question involves application of required actions of Tech Spec 3.0.3 for a seismic

event and the requirements for declaring a particular radiation monitor inoperable.



Enclosure 4.4 RPJO/AI5000/oo7Earthquake Page 1 of 6

1. Immediate Actions

NOTE: 1. Immediate Actions may be performed simultaneously.

2. The four Reactor Coolant Leakage Detection Systems are not seismically qualifiedand must be assumed to be inoperable following any seismic event. EMF38(L) canbe verified to be operable based on power availability and sample pump operation.

3. Reactor Coolant Leakage Detection Systems are not required to be operable duringCold Shutdown.

4. Technical Specification 3.0.3 is entered until one of the Reactor Coolant leakagedetection systems identified in step 1.1 is declared operable. {6}

5. An OAC Alarm at point C1D 2252 and C2D2252 indicates that there has been aseismic system actuation. This alarm is in addition to event indications present on1IEECS1000NCC on 1MC8.

1.1 Following any earthquake that is felt in the protected area or is recorded oninstrumentation, including earthquakes smaller than OBE, declare all four ReactorCoolant Leakage Detection Systems (listed below) as inoperable: {23}

1.1.1 Containment Floor and Equipment Sump Level Monitors and the IncoreInstrument Sump Level Alarm

1.1.2 VUCDT Level Monitoring System)

1.1.3 EMF38(L)

1.2 Determine the operable status of l(2)EMF38(L) from the Control Room by the followingmethods:

1.2.1 Perform a source check to verify that power is available:

• 1EMF38(L)• 2EMF38(L)

1.2.2 Visually verify that the sample pumps are operational:

• “ON” indicating light for “1EMF38 CONTAINMENT PAR” - LIT

• One indicating light on “SAMPLE FLOW SELECT PANEL” (Unit 1) - LIT

• “ON” indicating light for 2EMF 38 CONTAINMENT PAR” - LIT

• One indicating light on “SAMPLE FLOW SELECT PANEL” (Unit 2) - LIT


Question 99G2.3.6Radiation ControlAbility to approve release permits.

Given the following conditions:

• An Auxiliary Buildingwaste monitor tank !iquid waste release,(LWR) package has beendelivered to-thêControl Room.

• All RN pumps are on.• IAD-12 B/I RN Pump Intake Pit A Lo Level is LIT.• 1AD-12 E/2 RN Pit-A Swap to SNSWP is LIT.• A and B RL pumps are on and RL flow is greater than the flow required for the WL

release. —-------

• EMF-57 (MnitcrJank Buii Liquid Discharge Monitor) is NOT operable.• EMF-49 (UqWaé Disähirge Lo Range) is operable.

Should the CRS allow the release and why or why not? .

A. The CRS should NOT allovthis release because RN is aligned to the StandbyNuclear Service Water Pond

B. The CRS should allow this release since RL flow is greater than required for the WLrelease ONLY.

C. The CRS should allow this release since EMF 49 is operable AND RL flow is greater thanrequired for the WL release.

D. The CRS should NOT allow this release since EMF 57 is NOT operable.



QUESTION 99

Distractor Analysis

A. CORRECT. With RN aligned to the Standby Nuclear Service Water Pond, release cannotbe approved.

B. Incorrect. Cannot approve due to RN being aligned to the SNSWP. Plausible because thedistracter makes a true statement. RN is aligned to the SNSWP.

C. Incorrect. Correct approval but the reason is flawed. Discharge to RN is allowed withEMF-49 mop provided Radiation Protection takes action per their procedure.

D. Incorrect. Plausible, since EMF-57 is a radiation monitor associated with releases, but itsoperability does not affect this release.

References:• SLCI6.11-2

KA Match:Question 99G2.3.6Radiation ControlAbility to approve release permits.The KA is matched because the SRO applicant is presented with conditions involving aproposed liquid radwaste release, and then tested on whether the release should be allowed,and why or why not.

Cognitive Level: HighThis is a higher cognitive level question because it involves analysis of plant conditions, and aconclusion on whether radwaste release should be authorized.


SRO Only:This question meets the following criteria for an SRO only question as described in the“Clarification Guidance for SRO-only Questions (Rev. 1 dated 03/11/2010) under the ScreenCriteria for questions linked to I OCFR55.43(b)(4) (Radiation Hazards that may arise duringnormal and abnormal situations, including maintenance activities and various contaminationconditions):

1. It involves the process for a liquid release approval.


OP/O/B/6500/1 13Page 2 of 2

Operations Liquid Waste Release

1. Purpose

To aid the operator in the correct methods of performing steps in Radwaste procedureOP/0/B/6500/0 15 (Discharging a Monitor Tank to the Environment) and Radiation Protectionprocedure HP/0/B/1004/004 (Radioactive Liquid Waste Release). Also to aid the operator as tolimits and results expected while these procedures are being performed.

2. Limits and Precautions

2.1 Ensure that RN is discharging through at least one RL header.

2.2 Ensure that RN is NOT discharging to SNSWP.

2.3 If the pre-set radiation levels are exceeded on EMF-49 or the dilution flow rate dropsbelow the setpoint for ORLP5O8O (RL Discharge Total Flow), 1WL-l24 (Waste MonitTnk Pmps Disch) will trip closed.

2.4 Releases that are interrupted by EMF-49 “HI-RAD trips maybe initiated up to amaximum of three times, including original initiation, without re-sampling perHP/O/B/1 004/004 (Radioactive Liquid Waste Release).

2.5 Turbine Building Sump releases are secured if the pre-set levels are exceeded onl/2EMF-3 1.

3. Procedure

Refer to Section 4 (Enclosures)

4. Enclosures

4.1 Liquid Waste Release from a Monitor Tank

4.2 Discharging a Contaminated Turbine Building Sump to Holdup Pond

o J JC

EXAM BANK - Q 898An auxiliary building waste monitor tank liquid waste release (LWR) package hasbeen delivered to the Control Room.

The following conditions exist:

• All RN pumps are on• 1AD-12 B/I RN Pump Intake PitA Lo Level is LIT• lAD-I 2 E12 RN Pit-A Swap to SNSWP is LIT• A and B RL pumps are on and RL flow is greater than the flow required for the

WL release• EMF-49 (Liquid Waste Discharge Lo Range) is NOT operable• EMF-57 (Monitor Tank Building Liquid Discharge Monitor) is operable

Based upon the conditions given above describe the actions the Control RoomSupervisor should take regarding the release and the reason for that action?A. The CRS should NOT approve this release; RN is aligned to the Standby

Nuclear Service Water Pond.

B. The CRS should approve this release; RN is aligned to the Standby NuclearService Water Pond.

C. The CRS should NOT approve this release; a release is NOT allowed whenMF 49 is inoperable.

D. The CRS should approve this release; a release is allowed when EMF 57 isoperable.


Question 100G2.4.23Emergency Procedures!PlansKnowledge of the bases for prioritizing emergency procedure implementation duringemergency operations.

During a Unit I emergency event, the Primary SPDS Display indicates:

Initial:• CORE COOL ORANGE• RADIATION RED• The remaining Critical Safety Functions indicate GREEN.• The SRO selects the appropriate CSF procedure and begins implementation.

Current:

• The Primary SPDS Display NOW indicates:• CORE COOL ORANGE• RADIATION RED• HEAT SINK RED• The remaining CSFs indicate GREEN.

(1) How does the SRO prioritize implementation of the CSF procedures for the CURRENTconditions?

(2) What is the basis for that prioritization?

A. (1) Continue implementation of the originally selected CSF procedure until it is completed.(2) RPIOI, Classification of Emergency, Fission Product Barrier Matrix

B. (1) Continue implementation of the originally selected CSF procedure until it is completed.(2) Westinghouse Owner’s Group Background Documents

C. (1) Discontinue implementation of the originally selected CSF procedure and GO TO FRH. 1.

(2) RP/O1, Classification of Emergency, Fission Product Barrier Matrix

D. (1) Discontinue implementation of the originally selected CSF procedure and GO TOFR-H.1.

(2) Westinghouse Owner’s Group Background Documents



QUESTION 100

Distractor Analysis

A. Incorrect. FR-C.2, step 13 for plausibility of staying in C.2:. IF AT ANY TIME a red pathon NC Integrity occurs while in this procedure, THEN do not implement EP/1/N5000/FR-P.1 (Response To Imminent Pressurized Thermal Shock Condition) until this procedureis completed. RP/O1, Classification of Emergency, Fission Product Barrier Matrix isplausible since that document contains “prioritization related” items, but it is for thepurpose of emergency classification.

B. Incorrect. Plausibility of staying in FR-C.2 per description in “A” above. Second part iscorrect.

C. Incorrect. First part is correct. Plausibility of second part is described in “A” above.

D. CORRECT. From EAL Basis Document: 4.1.N.1EOPs are designed to maintain and/or restore a set of CSFs during accident conditions.By monitoring the CSFs instead of the individual system component status, the impact ofmultiple events is inherently addressed. The EOPs contain detailed instructions regardingthe monitoring of these functions and provides a scheme for classifying the significance ofthe challenge to the functions. In providing EALs based on these schemes, the emergencyclassification can flow from the EOP assessment rather than being based on a separateEAL assessment. This is desirable as it reduces ambiguity and reduces the timenecessary to classify the event.

References:• Lesson Plan for CSF, Section 1.2, Rev. 100• RP/01, Classification of Emergency, Fission Product Barrier Matrix• OMPI-7• From EAL Basis Document: 4.1.N.1

KA Match:Question 100G2.4.23Emergency ProcedureslPlansKnowledge of the bases for prioritizing emergency procedure implementation duringemergency operations.The KA is matched because the applicant must analyze conditions involving critical safetyfunction status changes, and the how the emergency procedures are implemented based onthat prioritization.

Cognitive Level: HighThis is a higher cognitive level question because the applicant analyzes conditions to determinethe priority of emergency procedure implementation.

Source of Question: NEWSRO Only:



This question meets the following criteria for an SRO only question as described in the“Clarification Guidance for SRO-only Questions (Rev. I dated 03/11/2010) under the ScreenCriteria for questions linked to I OCFR55.43(b)(5) (Assessment and Selection of Procedures):




overall mitigative strategy of a procedure.5. The question does involve assessing plant conditions (involving F-0 content and rules of

usage), and then selecting a procedure with specific content for a specific purpose: tomitigate the Critical Safety Function that has the highest priority per the rules of usage.

Therefore, this is an SRQ only question.



7.3. Functional Recovery or Critical Safety Function Procedures

A. This group of function related emergency procedures (EPs)covers the diagnostic, mitigating and recovery actions forchallenges to the following critical safety functions:

• Subcriticality

• Core cooling

• Heat sink

• NCS integrity

• Containment integrity

• NCS inventory

B. The Critical Safety Function (CSF) Status Trees shall bemonitored and implemented as directed by E-O or ECA-O.O.CSF Status Trees may also be used when EPs are thecontrolling procedure to determine or identify abnormalconditions.

C. CSF procedures shall not be implemented until the entryconditions are met (PPRB OPS-12571).

D. The responsibility for monitoring the CSF Status Trees ispre-assigned to the Shift Technical Advisor (STA). However,the Operations Shift Manager may reassign this responsibilitybased on available resources.

Operations Management Procedure 1-7 Page 13 of3l

E. After the CSF Status Trees have been implemented, thefollowing “rules of usage” apply:

• Use of EPs to restore a critical safety function is basedon a two factor priority system. The first factorconsiders the relative importance of the safety functionin an accident scenario. On the OAC alarm video, theorder of priority is established by position along thebottom of the screen. SUBCRITICALITY at the leftside of the screen has the highest priority. The others indecreasing priority, are:

CORE COOLING, HEAT SINK, NC INTEGRITY,CONTAINMENT (Integrity) and NC INVENTORY onthe right side of the screen.

• The second factor considers the degree of severity towhich the critical safety function is being challenged.The order of priority is designated by color indecreasing order, as follows:

1. Red - Extreme Challenge

2. Orange - Severe Challenge

3. Yellow - Off-Normal

4. Green - Satisfied.

• Prior to transitioning to any CSF procedure,shall be validated (SOER 94-1).

• The OSM should provide concurrence of thisvalidation process.

it

• If a valid red path is encountered, the operatorshall immediately implement the correspondingEP. If during the performance of any red pathprocedure, a red condition of higher priorityarises, then the higher priority condition shall beaddressed first, and the lower priority red pathprocedure suspended.

NOTE: An instrument or computer related failure that causes an erroneousSPDS indication is the only example of an invalid CSF Status Treecondition.


• If a valid orange path is encountered, theoperator is expected to scan all of the remainingtrees, and then, if no red path is encountered, topromptly implement the corresponding EP. Ifduring the performance of an orange pathprocedure, any red condition or higher priorityorange condition arises, then the red or higherpriority orange condition shall be addressedfirst, and the original orange path proceduresuspended.

• Once a procedure is entered due to a valid red ororange condition, that procedure shall beperformed to completion unless preempted bysome higher priority condition. It is expectedthat the actions in the procedure will clear thered or orange condition before all the operatoractions are complete. However, theseprocedures shall be performed to the point of thedefined transition to a specific procedure. Atthis point, any lower priority red or orange pathscurrently indicating or previously started butcompleted shall be addressed.

• If a CSF procedure directs the operator to returnto the procedure and step in effect and thecorresponding status tree continues to displaythe off normal condition, then the correspondingCSF procedure does not have to be implementedagain since all recovery actions have alreadybeen completed. However, if the same statustree subsequently changes to a valid higherpriority condition, then the corresponding CSFprocedure shall be implemented as required byits priority.

• Certain CSF procedures are used to address bothorange and red path conditions for the sameparameters. If the procedure is already inprogress due to the orange path condition, it isnot required to return to the first step if thecondition becomes red. Also, at the completionof the procedure, the procedure does have tobe implemented again, since all recovery actionshave already been implemented.


If a CSF procedure is implemented and it issubsequently determined that the indication isnot valid, the crew should:

1. Consult with the TSC and EOF if theyare manned.

2. Evaluate actions and systemrealignments performed in the invalidprocedure.

3. Realign systems as needed, address anyvalid red or orange paths, and return tothe procedure and step in effect.

• If Critical Safety Function (CSF) Status Treesare implemented during a reactor trip event anda subsequent safety injection occurs, the CSFTrees are still in effect. If either a RED orORANGE priority on a status tree is evident, theoperator must transition to the appropriate FRGas soon as it is validated to address the challengeto the critical safety function before completingany steps in E-O (DW-92-065).

• If a valid yellow path is encountered, theoperator is expected to scan all of the remainingtrees, then if no higher red or orange path isencountered, consider implementation of thecorresponding EP. Implementation is by theOperations Shift Manager’s judgement, based ontime, resources available and whether theyellow condition will be cleared by actions inprogress or is a warning of a more serious eventto occur.

• If during the performance of a yellow pathprocedure, any red or orange condition arises,then the red or orange condition shall beaddressed first, and the yellow path proceduresuspended.


• Yellow path procedures are to be performedconcurrent with the non-critical safety functionEP in effect when the yellow path isimplemented. While performing the actions ofthe yellow path, continuous actions or foldoutpage items of the non-critical safety function EPin effect are still applicable and shall bemonitored by the operator. (DW-95-043)

• If a red or orange condition indicates and thenclears prior to implementation of thecorresponding procedure, the procedure shallj be performed. The CSF procedure isconsidered to be “implemented” when the CRSreads the first step to the crew.

• The STA shall keep the Operations ShiftManager informed of all off normal CSFs. TheOperations Shift Manager shall ensure the crewis updated as appropriate, typically by allocatingtime during updates for the STA. (SOER 94-1)

F. Normally, the condition of the CSF Status Trees iscontinuously displayed by SPDS on the OAC. Control roomindications shall be used to validate any off normal alarm andto determine which procedure to implement. Once status treemonitoring is initiated, the STA should periodically monitorthe status trees and compare against control board indicationsto ensure SPDS is functioning properly. Status tree monitoringshall be continuous if an orange or red condition exists.Otherwise, monitoring frequency shall be every 10 to 15minutes. (SOER 94-1)

Catawba Nuclear Site

REACTOR COOLANT SYSTEM (NCS) BARRIER EALs:

The NCS Barrier includes the NCS primary side and its connections up to and including thepressurizer safety and reliefvalves, and other connections up to and including the primaryisolation valves.

4.1.N.1 Critical Safety Function Status

NCS Integrity - RED indicates NCS pressure and temperature conditions which maychallenge the Reactor Vessel integrity. Heat Sink - RED indicates the ultimate heatsink function is under extreme challenge. Either of these conditions indicate apotential loss of the NCS Barrier.

There is no “Loss” EAL associated with this item.

If a steamline break occurs that challenges reactor coolant system integrity (PTS),implementation of the functional restoration procedure directs the throttling of CAflow to the S/Gs (creating the heat sink CSF alarm) to mitigate the NCS cooldown.It would be inappropriate to upgrade the severity of the emergency classificationbased on the directed action to minimize NCS cooling.

The heat sink functional restoration guidance specifically provides a relief fromimplementing the “loss of heat sink actions,” when the operator is responding to theNCS overcooling situation. Thus the loss of heat sink is considered a controlledoperator response, not a loss of the heat sink.

If, however, the heat sink is made physically unavailable for some other reason, thenthe upgrade in emergency classification severity is appropriate.

EOPs are designed to maintain and/or restore a set of CSFs during accidentconditions. By monitoring the CSFs instead of the individual system componentstatus, the impact of multiple events is inherently addressed. The EOPs containdetailed instructions regarding the monitoring of these functions and provides ascheme for classifying the significance of the challenge to the functions. Inproviding EALs based on these schemes, the emergency classification can flow fromthe EOP assessment rather than being based on a separate EAL assessment. This isdesirable as it reduces ambiguity and reduces the time necessary to classify theevent.

D-7 Rev. 09-1April, 2009

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