Text

301 Draft SRO Written Exam
	ML12356A448
Person / Time
Site:	Catawba
Issue date:	12/17/2012
From:	; Division of Reactor Safety II
To:	; Duke Energy Corp
References
	50-413/12-301, 50-414/12-301
	Download: ML12356A448 (133)
	v • d • e

Question 76 01 5AG2.4.46 RCP Malfunctions Ability 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 ah which equipment, if any, must be declared Standby Shutdown Facility (SSF) non-functional non-functional functional C

i.

/

_-7 Page 166 of 235 Catawba 2012 NRC Exam Submittal S.

CNS 2012 NRC Exam 100 Questions Final Submittal

\\

I describes A.

B.

C.

füntional functional non-functional non-functional N

functional

CNS 2012 NRC Exam 100 Questions Final Submittal QUESTION 76 Distractor Analysis A.

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

B.

Incorrect. First part is correct. The Standby Makeup Pump being declared non-functional is plausible:

It is powered by a DIG in the SSF, and is therefore, related to the status of the SSF. If the SSF is non-functional, it is reasonable to believe that the Standby Makeup Pump POWERED FROM the SSF DIG would also be non-functional.

C.

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

D.

Incorrect: Second part is correct.

If the SSF is functional, due to similar reasoning as described in B above, it is plausible that the Standby Makeup Pump (powered from the SSF 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 76 01 5AG2.4.46 RCP Malfunctions Ability to verify that the alarms are consistent with the plant conditions.

Applicant is presented with multiple plant conditions, primarily related to seal leakoff, and then must 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 that function (RCP seals).

Cognitive Level:

High This is a higher cognitive level question because the applicant evaluates multiple plant conditions, including an alarm, and must make a conclusion and a determination regarding functionality of components.

Source of Question:

NEW SRO Only:

Page 167 of 235 Catawba 2012 NRC Exam Submiftal

CNS 2012 NRC Exam 100 Questions Final Submittal 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 Screen Criteria for questions linked to I OCFR55.43(b)(2) (Tech Specs):

1.

It cannot be answered solely by knowing 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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.

Page 168 of 235 Catawba 2012 NRC Exam Submittal

OP/1/B/6100/O1OH PANEL: 1AD-7 Page 19 of 64 NCP #1 SEAL LEAKOFF HI FLOW C/i SETPOINT:

5.0 gpm ORIGIN:

Instrument DCS Description 1NVFT5151 1NVAA5151 NC PUMP A SEAL LEAKOFF FLOW HI 1NVFT5 141 1NVAA5 141 NC PUMP B SEAL LEAKOFF FLOW HI 1NVFT5131 1NVAA5131 NC PUMP C SEAL LEAKOFF FLOW HI 1NVFT5 121 1NVAA5 121 NC PUMP D SEAL LEAKOFF FLOW HI PROBABLE 1.

Damaged #1 Seal CAUSE:

2.

Cocked #1 Seal 3.

Loss of injection water followed by high seal temp 4.

Hi temperature of injection water AUTOMATIC None ACTIONS:

IMMEDIATE 1.

Identify the affected pump from one of the following:

ACTIONS:

DCS graphic 6009, NV

- NC Pump Seal Injection DCS alarm screen 2.

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 System Leakage Calculation).

NOTE:

The SSF is conservatively declared inoperable due to the potential for exceeding the NC Pump 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 16.7-9 2.

CN-1499-NV-3 3.

CNM-1201.Ol-157 4.

CNM 1399.03-0269.001 Drop 6 Sheet 317 5.

CNM 1399.03-0269.001 Drop 8 Sheet 312 6.

CNM 1399.03-0269.001 Drop 12 Sheet 319 7.

CNM 1399.03-0269.001 Drop 7 Sheet 323

CNS 2012 NRC Exam 100 Questions Final Submittal Question 77 038EG2.4.46 Steam Gen. Tube Rupture Ability to verify that the alarms are consistent with the plant conditions.

Given the following Unit I 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 l

23% and increasing I B 15% and decreasing IC 18% and decreasing 7

1D) 0%

Which ONE of the following procedures or enclosures contains the steps for throttling CA flow to the 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)

Page 169 of 235 Catawba 2012 NRC Exam Submittal 7

0

-)

750 psig and increasing 720 psig and decreasing 700 psig and decreasing 200 psig and decreasing

CNS 2012 NRC Exam 100 Questions Final Submittal 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, and EMF 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 77 038EG2.4.46 Steam Gen. Tube Rupture Ability to verify that the alarms are consistent with the plant conditions.

This question matches the KA because the applicant is presented with several radiation monitors 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 the diagnosed plant conditions, reflected by choice of procedure guidance.

Cognitive Level:

High The applicant is presented with several radiation monitors 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 the diagnosed plant conditions, reflected by choice of 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 Screen Criteria 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 system works, 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.

Page 170 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal 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 (involving &O content), and then selecting a section (specific step for a specific purpose) to mitigate a Steam Generator Tube Rupture.

Therefore, this is an SRO only question.

Page 171 of 235 Catawba 2012 NRC Exam Submittal

CNS REACTOR TRIP OR SAFETY INJECTION PAGE NO.

EPIIIAJ5000!E-0 31 of 46

- Page 1 of 2 Revision 40 Foldout 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 between 11%(29% ACC) and 39%.

2.

NC Pump Trip Criteria:

if the following conditions are satisfied, THEN trip all NC pumps while maintaining seal injection 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/G Feedwater).

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 CLOSE I 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 877 Unit 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 Pressure A

17% and stable 750 PSIG and decreasing B

15% and decreasing 720 PSIG and decreasing C

18% and decreasing 700 PSIG and decreasing D

0%

200 PSIG and decreasing Which SIG will have auxiliary feed water throttled to it FIRST and by what procedure and/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 Safety Injection)

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)

CNS 2012 NRC Exam 100 Questions Final Submittal Question 78 054AA2.02 Loss of Main Feedwater Ability to determine and interpret the following as they apply to the Loss of Main Feedwater (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 the loss of ONE CF Pump as compared to the loss of BOTH CF Pumps? (Assume initial conditions at 80% power.)

A.

For ONLY)he loss of BOTH CF Pumps, the notification requirement is within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

B.

For EITHER the loss of ONE or BOTH CF Pumps, the notification requirement is within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

C.

For ONLY the loss ofONE CF Pump, the notification requirement is within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

D.

For ONLY the loss of BOTH CF Pumps, the notification requirement is within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

/

G Page 172 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal 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 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months notification to the NRC Operations Center.

B.

Incorrect. 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is explained in A above. Plausible that the loss of any main feed pump with the plant at 80% power would warrant notification, since it is an operationally significant event and does place the plant in a transient. Notification is not required for either, though it is for both CF pumps ONLY.

C.

Incorrect. The loss of ONE CF Pump is an operationally significant event, and it is plausible to believe that notification is required, and since the time listed is a relatively long time period (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ), an applicant could believe this to be the correct answer.

D.

CORRECT. An RPS actuation is a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months notification; since loss of both CF pumps causes a turbine trip, which results in a reactor trip.

References:

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

Question 78 054AA2.02 Loss of Main Feedwater Ability to determine and interpret the following as they apply to the Loss of Main Feedwater (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 a single main feed pump, as compared to the loss of both main feed pumps, in the context of NRC notifications.

Cognitive Level:

High At first glance, this may appear to be a low cognitive level question, since there is some recall involved. However, there is more than one mental step involved in arriving at the correct answer; applicant must first recall the notification requirements for ONLY the loss of both main feed 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 Screen Criteria for questions linked to IOCFR55.43(b)(1), (Conditions and Limitations in the facility license):

1.

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

Page 173 of 235 Catawba 2012 NRC Exam Submiftal

.3 1/O/B/5ooo/o13 Events 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 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the occurrence becomes known to the licensee:

1 OCFR Section Event Description Reporting Requirement 10CFR5O.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 the ECCS discharge into actuation results from and is part of a pre-planned sequence during testing the Reactor Coolant or reactor operation.

System Valid signal refers to those signals automatically initiated by measurement of an actual physical system parameter that was within the established setpoint band of the sensor that provides the signal to the protection system logic, or manually initiated in response to plant conditions. Valid signals also include passive system actuations that occur as a function of system conditions like differential pressure (i.e.,

cold leg accumulators) whereby no SSPS or other electrical signal is involved. The validity of an ECCS signal may not be determined within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; ECCS signals that result or should have resulted in injections should be considered valid until firm evidence proves otherwise.

Invalid ECCS injections are still considered a System actuation, but are NOT reportable to the NRC per 10 CFR 50.72. It is still reportable under 10 CFR 50.73 as an LER. (Refer to Enclosure 4.8 for guidance as_to_what_constitutes_a_System_actuation.)

10CFR5O.72(b)(2)(iv)(B)

Any event or condition that results in actuation of the reactor protection Notify the NRC Operations Center system (RPS) when the reactor is critical except when the actuation is part RPS Actuation of a pre-planned sequence during testing or reactor operation.

10CFR5O.72(b)(2)(xi)

Any event or situation related to the health and safety of the public or on-Notify the NRC Operations Center 10CFR72.75(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 be Offsite 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 of radioactively contaminated materials.

.4 RP/O/B/5000/o13 Events 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 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after the occurrence becomes known to the licensee:

p

\\ics 1 OCFR Section Event Description Reporting Requirement 10CFR5O.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 or B. The Nuclear Power plant being in an unanalyzed condition that significantly degrades plant safety.

10CFR5O.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 results System Actuation from and is part of a pre-planned sequence during testing or plant

( ISt aoci.

operation.

10CFR5O.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 Function Prevented 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 accident 10CFR72.75(c)(1)

Discovery of a defect in an ISFSI structure, system or component that is Notify the NRC Operations Center important to safety.

ISFSI Defects 10CFR72.75(c)(2)

A significant reduction in the effectiveness of the ISFSI confinement Notify the NRC Operations Center system.

ISFSI Degraded Confinement System

.8 RP/O/B/5000/o13 List of System (ESF) Actuations for Catawba Page 1 of 2 Any 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 40 does 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

.8 RP/O/B/5000/ol3 List of System (ESF) Actuations for Catawba Page 2 of 2 1%

U 6.

Auxiliary feedwater system 1

\\4

+rs s

oiti ii c ijcti A.

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

Example: A feedwater transient is in progress with S/G levels decreasing toward the reactor trip 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 a controlled (documented) test or procedure.

8.

Ice condenser lower inlet door opening as a result of unplanned mass or energy release into containment 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 79 055EA2.02 Station Blackout Ability 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 Unit Offsite Power from Unit 2 D/G IA DIGIB (1)

To establish plant conditions and restore equipment needed for natural circulation the SRO will remain in ECA-0.0 until power is available from (1) of the above 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 natural circulation?

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).

Page 174 of 235 Catawba 2012 NRC Exam Submittal CNS 2012 NRC Exam 100 Questions Final Submittal ccv -

I (2)

CNS 2012 NRC Exam 100 Questions Final Submittal QUESTION 79 Distractor Analysis A.

Incorrect. Second part (procedure transition) is correct. Believing that you need to restore power from at least TWO sources in the list in ECA-0.O Step 17 is plausible by reasoning that it is desirable to have more than a single source of power, particularly during emergency 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 must continue until power is restored from at least ONE of a list of sources, per the step. With the conditions given in the stem meeting all the criteria of Step 42, the crew is then directed 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 since that procedure would be used DURING a natural circ cooldown, but the question is testing procedure transition knowledge by asking where do you go once you have restored power from at least ONE of the listed sources in Step 17 of ECA-0.0. Further, there is NO place in ECA-0.0 that directs a transition directly to ES-0.2 for natural circulation. First, you are directed to transition to ECA-0.1 which contains additional actions for restoring loads important 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 79 055EA2.02 Station Blackout Ability to determine or interpret the following as they apply to a Station Blackout:

RCS core cooling through natural circulation cooling to SIG cooling This KA is matched because the applicant is tested on plant conditions involving a Station Blackout (all SWYD power lost AND no D/Gs available), and then to determine at an SRO level) what is needed, including procedures that are appropriate, for establishing core cooling using natural circulation.

Cognitive Level:

High This is a higher cognitive level question because the applicant must first determine from given conditions that a Station Blackout has occurred, and then apply the effect of that condition to subsequent plant conditions, and then determine which procedures, including needed transitions, will be used to establish natural circulation.

Source of Question:

Bank CNS 503 Page 175 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal 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 Screen Criteria 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 system works, 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, 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 Blackout and to establish natural circulation.

Therefore, this is an SRO only question.

Page 176 of 235 Catawba 2012 NRC Exam Submittal

CNS LOSS OF ALL AC POWER PAGE NO.

EPII/AI5000IECA-O.O 12 of 173 Revision 45 I

ACTION/EXPECTED RESPONSE I

RESPONSE

NOT OBTAINED NOTE

Offsite power may be unavailable for reasons other than switchyard de-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 pump seals will be established by natural circulation cooldown as directed in subsequent procedures.

42.

Select recovery procedure as follows:

a.

Verify NC subcooling based on core exit 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 (Loss Of All AC Power Recovery Without S/I Required).

a.

GOTO EP/1/N5000!ECA-0.2 (Loss Of All AC Power Recovery With S/I Required).

b.

GOTOEP/1/A15000/ECA-0.2 (Loss Of All AC Power Recovery With S/I Required).

c. if any NV pump on, THEN QI EP/1/N5000/ECA-0.2 (Loss Of All AC Power Recovery With S/I Required).

END

rn EXAM 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 to Voids in Reactor Vessel.

Which one of the following is the correct action to control void growth such that natural circulation 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.

CNS 2012 NRC Exam 100 Questions Final Submittal Question 80 058AG2.1.28 Loss of DC Power Knowledge 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 Vjtalbfor when 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 with Tech Spec 3.8.4 (DC Sources Operating), what is the MlNjMUfcIvoltage required for the Vital 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 Page 177 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal QUESTION 80 Distractor Analysis A.

Incorrect. AP/007, (Loss of Normal Power) is plausible since there are numerous instructions 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 battery depletion. There are additional instructions elsewhere in the procedure for Standby Shutdown Facility batteries, and for Switchyard batteries. Enclosure 17, Switchyard Battery 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 bus when battery voltage decays to 105 VDC, by opening the associated battery output breaker.

C.

Incorrect. Plausibility for AP/007 is described in A above. 110 volts could be reasoned a minimum 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 80 058AG2.1.28 Loss of DC Power Knowledge of the purpose and function of major components and controls.

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

Cognitive Level:

Low 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. I dated 03/11/2010) under the Screen Criteria for questions linked to I OCFR55.43(b)(2) (Tech Specs):

1.

It cannot be answered solely by knowing 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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.

Page 178 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal 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.

Therefore, this is an SRO only question.

Page 179 of 235 Catawba 2012 NRC Exam Submittal

CNS LOSS OF NORMAL POWER PAGE NO.

AP111A155001007 26 of 162 Case II Revision 66 Loss of 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/G Level Control).

_2) QIQ.Step14.

b.

IFATANY TIME CF flow control to S/Gs is lost, THEN perform Step 13.

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

IEA[ANY TIME any vital or auxiliary control channel battery charger has been de-energized for greater than I hour, THEN dispatch operators to locally control affected CA flow path.

- A? o7 REFER TO Enclosure 16 (S/G Level Or 1OJi3 0

Control).

CNS LOSS OF VITAL OR AUX CONTROL POWER PAGE NO.

AP111A155001029 31 of 208

- Page 4 of Revision 24 Response To Degraded DC Bus Voltage ACTION/EXPECTED RESPONSE I

RESPONSE

NOT OBTAINED

5. (Continued) c.

Review applicable load list(s) for plant response to a loss of affected control power 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 for recommendations regarding imminent low DC voltage.

NOTE

Indication of Aux Control Power battery voltage is available locally or in Control 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 Power battery voltage decays to 105 VDC, THEN perform the following:

1)

Ensure associated battery output breaker

- OPEN.

2)

IFATANY TIME breaker control power is lost, AND it is desired to operate station breakers, THEN evaluate aligning batteries to breaker control power only.

It (H CNS 2012 NRC Exam 100 Questions Final Submittal Question 81 077AA2.07 Generator Voltage and Electric Grid Disturbances Ability to determine and interpret the following as they apply to Generator Voltage and Electric Grid Disturbances:

Operational status of engineered safety features Given the following Unit I conditions:

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 analysis assumptions 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.

Page 180 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal QUESTION 81 Distractor Analysis A.

CORRECT. Based on the conditions one minute later, both DGs should be running with LOCA loads. The DIG breaker will trip on overcurrent in this situation, and the LOCA will override the blackout and LOCA loads will sequence on. Since B D/G is running already, applicants may think that the B train sequencer wont work correctly or will only load blackout loads.

In that case, the I B ND pump would not be running (or NI) since these are 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 pump would be running. Based on Tech Spec bases for ECCS operation, only one complete train 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 and D/G load sequencer to attempt to start it, and they would also attempt to reset A Train to start 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 the requirement.

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 81 077AA2.07 Generator Voltage and Electric Grid Disturbances Ability to determine and interpret the following as they apply to Generator Voltage and Electric Grid Disturbances:

Operational status of engineered safety features The KA is matched because the question involves a grid disturbance, and testing the ability to evaluate given conditions and determine if LOCA analysis assumptions and criteria are met for ECCS equipment (operational status of engineered safety features)

Cognitive Level:

High This is a higher cognitive level question because the applicant must determine from the given conditions what safeguards equipment should be operating, given the effects of the power disturbance, and then evaluate the status of the equipment in the context of accident analysis assumptions.

Source of Question:

Bank CNS 502 SRO Only:

Page 181 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal 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 Screen Criteria for questions linked to 1 OCFR55.43(b)(2) (Tech Specs):

1.

It cannot be answered solely by knowing 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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 actions and terminology.

Therefore, this is an SRO only question.

Page 182 of 235 Catawba 2012 NRC Exam Submittal

ECCSOperating B 3.5.2 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.2 ECCSOperating BASES BACKGROUND The function of the ECCS is to provide core cooling and negative reactivity to ensure that the reactor core is protected after any of the following accidents:

a.

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

Rod ejection accident; c.

Loss of secondary coolant accident, including uncontrolled steam or feedwater release; and d.

Steam generator tube rupture (SGTR).

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

There are three phases of ECCS operation: injection, cold leg recirculation, and hot leg recirculation.

In the injection phase, water is taken from the refueling water storage tank (RWST) and injected into the Reactor Coolant System (RCS) through the cold legs. When sufficient water is removed from the RWST to ensure that enough boron has been added to maintain the reactor subcritical and the containment sumps have enough water to supply the required net positive suction head to the ECCS pumps, suction is switched to the containment sump for cold leg recirculation. When the core decay heat has decreased to a level low enough to be successfully removed without direct RHR pump injection flow, the RHR cold leg injection path is realigned to discharge to the auxiliary containment spray header. After approximately 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months , part of the ECCS flow is shifted to the hot leg recirculation phase to provide a backflush which, for a cold leg break, would reduce the boiling in the top of 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 heat removal (RHR) (low head). Each subsystem consists of two redundant, 100% capacity trains. The ECCS accumulators and the RWST are also part of the ECCS, but are not considered part of an ECCS flow path as described by this LCO.

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

ECCS Operating B 3.5.2 BASES BACKGROUND (continued)

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

flow to the core.

During the injection phase of LOCA recovery, a suction header supplies water from the RWST to the ECCS pumps. Mostly separate piping supplies each subsystem and each train within the subsystem. The discharge from the centrifugal charging pumps combines, then divides again into four supply lines, each of which feeds the injection line to one RCS cold leg. The discharge from the SI and RHR pumps divides and feeds an injection line to each of the RCS cold legs. Throttle valves in the SI lines are set to balance the flow to the RCS. This balance ensures sufficient flow to the core to meet the analysis assumptions following a LOCA in one of the RCS cold legs. The flow split from the RHR lines cannot be adjusted. Although much of the two ECCS trains are composed of completely separate piping, certain areas are shared between trains. The most important of these are 1) where both trains flow through a single physical pipe, and 2) at the injection connections to the RCS cold legs. Since each train must supply sufficient flow to the RCS to be considered 100% capacity, credit is taken in the safety analyses for flow to three intact cold legs. Any configuration which, when combined with a single active failure, prevents the flow from either ECCS pump in a given train from reaching all four cold legs injection points on that train is unanalyzed and might render both trains of that ECCS subsystem inoperable.

For LOCAs that are too small to depressurize the RCS below the shutoff head of the SI pumps, the centrifugal charging pumps supply water until the RCS pressure decreases below the SI pump shutoff head. During this period, the steam generators are used to provide part of the core cooling function.

During the recirculation phase of LOCA recovery, RHR pump suction is transferred to the containment sump. The RHR pumps then supply the other ECCS pumps.

Initially, recirculation is through the same paths as the injection phase. Subsequently, for large LOCAs, the recirculation phase includes injection into both the hot and cold legs.

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

ECCS Operating B 3.5.2 BASES BACKGROUND (continued)

The high and intermediate head subsystems of the ECCS also functions to supply borated water to the reactor core following increased heat removal events, such as a main steam line break (MSLB). The limiting design conditions occur when the moderator temperature coefficient is highly negative, such as at the end of each cycle.

During low temperature conditions in the RCS, limitations are placed on the maximum number of ECCS pumps that may be OPERABLE. Refer to 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. The actuation of safeguard loads is accomplished in a programmed time sequence.

If offsite power is available, the safeguard loads start immediately in the programmed sequence.

If offsite power is not available, the Engineered Safety Feature (ESF) buses shed normal operating loads and are connected to the emergency diesel generators (EDG5). Safeguard loads are then actuated in the programmed time sequence. The time delay associated with diesel starting, sequenced loading, and pump starting determines the time required before pumped flow is available to the core following a safety injection actuation.

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

APPLICABLE The LCO helps to ensure that the following acceptance criteria for the SAFETY 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 criteria are 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 cladding thickness before oxidation; c.

Maximum hydrogen generation from a zirconium water reaction is 0.01 times the hypothetical amount generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react; Catawba Units 1 and 2 B 3.5.2-3 Revision No. 3

ECCS Operating B 3.5.2 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 following an MSLB event and ensures that containment pressure and temperature limits are met.

Each ECCS subsystem is taken credit for in a large break LOCA event at full power (Refs. 3 and 4). This event has the greatest potential to challenge the limits on runout flow set by the manufacturer of the ECCS pumps. It also sets the maximum response time for their actuation.

Direct flow from the centrifugal charging pumps and SI pumps is credited in a small break LOCA event. The RHR pumps are also credited, for larger small break LOCAs, as the means of supplying suction to these higher head ECCS pumps after the switch to sump recirculation. This event establishes the flow and discharge head at the design point for the centrifugal charging pumps. The MSLB analysis also credits the SI and centrifugal charging pumps. Although some ECCS flow is necessary to mitigate a SGTR event, a single failure disabling one ECCS train is not the limiting single failure for this transient. The SGTR analysis primary to secondary break flow is increased by the availability of both centrifugal charging and SI trains. Therefore, the SGTR analysis is penalized by assuming both ECCS trains are operable as required by the LCO. The OPERABILITY requirements for the ECCS are based on the following LOCA analysis assumptions:

a.

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

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

During the blowdown stage of a LOCA, the RCS depressurizes as primary coolant is ejected through the break into the containment. The nuclear reaction is terminated either by moderator voiding during large breaks or control rod insertion for small breaks. Following depressurization, 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 for in appropriate analyses (Ref. 3). The LCO ensures that an ECCS train will deliver sufficient water to match boiloff rates soon enough to minimize the consequences of the core being uncovered following a large LOCA.

Catawba Units 1 and 2 B 3.5.2-4 Revision No. 3

ECCS Operating B 3.5.2 BASES APPLICABLE SAFETY ANALYSES (continued)

It also ensures that the centrifugal charging and SI pumps will deliver sufficient water and boron during a small LOCA to maintain core subcriticality. For smaller LOCAs, the centrifugal charging pump delivers sufficient fluid to maintain ROS inventory. For a small break LOCA, the steam generators continue to serve as the heat sink, providing part of the required 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 are required to ensure that sufficient ECCS flow is available, assuming a single failure affecting either train. Additionally, individual components within the ECCS trains may be called upon to mitigate the consequences of other transients and accidents.

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

During an event requiring ECCS actuation, a flow path is required to provide an abundant supply of water from the RWST to the RCS via the ECCS pumps and their respective supply headers to each of the four cold leg injection nozzles.

In the long term, this flow path may be switched to take its supply from the containment sump and to supply its flow to the RCS hot and cold legs. The flow path for each train must maintain its designed independence to ensure that no single failure can disable both ECCS trains.

APPLICABILITY In MODES 1, 2, and 3, the ECCS OPERABILITY requirements for the limiting Design Basis Accident, a large break LOCA, are based on full power operation. Although reduced power would not require the same level of performance, the accident analysis does not provide for reduced cooling requirements in the lower MODES. The centrifugal charging pump performance is based on a small break LOCA, which establishes the pump performance curve and has less dependence on power. The SI pump performance requirements are based on a small break LOCA.

For both of these types of pumps, the large break LOCA analysis depends only on the flow value at containment pressure, not on the shape of the flow versus pressure curve at higher pressures. MODE 2 and MODE 3 requirements are bounded by the MODE 1 analysis.

Catawba Units 1 and 2 B 3.5.2-5 Revision No. 3

EXAM BANK - Q 502 Initial 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 indicated o

Both the ON and OFF light on the.E30 pushbutton are DARK IA NI Pump has the following indications:

o No running amps are indicated o

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 per Technical Specification 3.5.2 (ECCS

- Operating)?

2.

When EP/IIAI5000/E-0, Reactor Trip or Safety Injection, is exited, what train(s) of ECCS 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.

CNS 2012 NRC Exam 100 Questions Final Submittal Question 82 OOIAA2.04 Continuous Rod Withdrawal Ability to determine and interpret the following as they apply to the Continuous Rod Withdrawal:.

PReactor power and its trend For a Continuous Rod Withdrawal event initiating from a power level of 15% with IR Nis NOT blocked, AP/15, (Rod Control Malfunction), Case II, Continuous Rod Movement has been implemented.

(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 is the 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)DCSJailure (2)

Site Area Emergency 1

A

)

Page 183 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal QUESTION 82 Distractor Analysis A.

CORRECT. Step 4 of APII 5 contains detailed guidance in attempting to determine if there is a problem with the NC Loop Tavg indications. If a continuous rod withdrawal event initiates from the given power level (15%) and reactor power reaches 31%, and the control room manually trips, this is an indication that an automatic trip did not occur at 25%. This should have occurred by the Intermediate Range NIs since they were not blocked. The Power Range Nis would have already been blocked, and would not be expected to process a 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 trip should have occurred, but did NOT, but the manual trip was successful.

B.

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

C.

Incorrect. Second part is correct. DCS failure is plausible since this system affects numerous 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 82 001 AA2.04 Continuous Rod Withdrawal Ability to determine and interpret the following as they apply to the Continuous Rod Withdrawal:

Reactor power and its trend The KA is matched because given conditions involve a continuous rod withdrawal and then the applicant is tested to determine and interpret a trend given on reactor power, and continue on to provide procedure content, and classification of the event.

Cognitive Level:

High This is a higher cognitive level question since the applicant must diagnose that an automatic reactor trip should have occurred, but did not, and then with that determination, consult a classification matrix to determine the emergency classification.

Source of Question:

NEW Page 184 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal 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 Screen Criteria for questions linked to IOCFR55.43(b)(1), (Conditions and Limitations in the facility license):

1.

It involves classification of an.emergency event.

Page 185 of 235 Catawba 2012 NRC Exam Submittal

jf channel inoperable, THEN perform the following:

a.

WHEN T-Avg is within +/-1°F of T-Ref AND auto rod control is desired, THEN return rod control to AUTO.

b.

Ensure P-12 interlock in required state for existing plant conditions. REFER JQ.Tech Spec 3.3.2.

c.

Have IAE trip bistables associated with the failed loop within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . REFER TO 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.

I AP111A155001015 10 of 11 I

Case II Revision 141 Continuous Rod Movement I

ACTION/EXPECTED RESPONSE

RESPONSE

NOT OBTAINED I

4.

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 of continuous rod movement.

6.

Ensure compliance with appropriate Tech 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 for Criticality).

.4 UNUSUAL EVENT Loss of Shutdown Functions ALERT SITE AREA EMERGENCY 1u/O/A/5000/oo 1 Page 1 of3 GENERAL EMERGENCY 4.4.A.1 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Trip Once a Reactor Protection System Setpoint fins Been Exceeded and Manual Trip Was Successful.

4.4.S.1 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Trip Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Trip Was NOT Successful.

4.4.G.1 Failure of the Reactor Protection System to Complete an Automatic Trip and Manual Trip Was NOT Successful and There is Indication of an Extreme Challenge to the Ability to Cool the Core.

OPERATING MODE:

1 Valid reactor trip signal received or required and automatic reactor trip was not successful.

AND Manual reactor trip from the control room is successful and reactor power is less than 5%

and decreasing.

Valid reactor trip signal received or required and automatic reactor trip was not successful.

AND Manual reactor trip from the control room was not successful in reducing reactor power to less than 5% and decreasing.

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

Valid reactor trip signal received or required and automatic reactor trip was not successful.

AND Manual reactor trip from the control room was not successful in reducing reactor power to less than 5% and decreasing.

(Continued)

EITHER of the following conditions 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

CNS 2012 NRC Exam 100 Questions Final Submittal Question 83 028AG2.1.32 Pressurizer Level Malfunction Ability to explain and apply system limits and precautions.

Given the following Unit I conditions:

With the Unit in Mode 2, a Unit startup is in progress in accordance with OPI1IAI6IOO/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 of OP/I 1A161 00/001.

(1)

What is the basis for the above Limit and Precaution to maintain an outflow on the Pressurizer?

(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, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months the Unit must be placed in L

A (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 Page 186 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal QUESTION 83 Distractor Analysis A.

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

B.

Incorrect. First part plausibility explained in A above. Mode 4 is plausible if the SLC requirement is misapplied.

C.

CORRECT. OPIIIA/61001001, (Controlling Procedure for Unit Startup), Limits and Precaution 2.16 explains that the basis for maintaining an outflow on the PZR is to minimize PZR thermal stratification. The Selected Licensee Commitment (SLC) 16.5-4 requires 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 C I

KA Match:

Question 83 028AG2.1.32 Pressurizer Level Malfunction Ability to explain and apply system limits and precautions.

This KA is matched because the stem conditions involve a PZR level malfunction that requires the operators to take action and restore level. In that process, there is a concern for PZR outflow, as explained in a Limit and Precaution.

Cognitive Level:

High This 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 for oufflow is met, and what action would be required by the SLC.

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. I dated 03/11/2010) under the Screen Criteria for questions linked to IOCFR55.43(b)(2) (Tech Specs):

1.

It cannot be answered solely by knowing < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months TS/ SLC Action.

Page 187 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal 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 actions and terminology.

Therefore, this is an SRO only question.

Page 188 of 235 Catawba 2012 NRC Exam Submittal

OP/1/A16 100/001 Page 5 of 7 2.13 It is recommended that S/G reverse purge flow be maintained at all times when the S/Gs are pressurized and CF flow is NOT aligned to the main feedwater nozzles. This ensures the main feed containment penetration piping is maintained above brittle fracture temperature of 107°F.

If the temperature on both sides of the penetration is greater than 107°F during Mode 1, reverse purge flow can be secured, but reverse purge shall be re-established before temperature reaches 107°F (decreasing) to ensure compliance with the commitment to the NRC on 10CFR5O Appendix A GDC51 (temperature greater than 107°F during power operation-Mode 1). The temperature between the feedwater isolation valves and S/Gs shall be greater than 107°F during Mode 1. (C1AO141, C1A0148, C1A0125, C1A0154, C1A0275, C1AO16O, C1A0815, C1A0166, OAC Group Display GD OPCFTEMP)

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

2.14 When feeding the S/Gs from a source other than main feedwater, Secondary Chemistry needs 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 when the reactor is shutdown or less than or equal to 55°F when the reactor is critical, the RC System 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. PZR outflow 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 level increase, then the PZR level shall be maintained at the elevated level until PZR liquid space temperature recovers. PZR liquid space temperature is directly affected by PZR level during plant conditions requiring a saturated PZR and cooler NC loop temperatures.

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 minutes All Required Actions temperature to within must be completed limits.

whenever this Condition is entered.

AND Pressurizer temperature A.2 Pertorm engineering 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months not within limits, evaluation to determine effects of the out-of-limit condition on the structural integrity of the pressurizer.

AND A.3 Determine that the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months pressurizer remains acceptable for continued operation.

B.

Required Action and B.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

B.2 Reduce pressurizer 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Y.&1 I Tfl pressure to < 500 psig.

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

CNS 2012 NRC Exam 100 Questions Final Submittal

,ji Question 84 068AG2.4.18 Control 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 to Fire or Security Event, is in progress for both Units.

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

A.

All CRD vent fans RUNNING.

4 No 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

}

Page 189 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal QUESTION 84 Distractor Analysis A.

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

B.

CORRECT. Step 6 of AP/1 7 for Control Room evacuation prescribes that all CRD vent fans 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 this is the incorrect basis.

D.

Incorrect. Plausible since this would aid in inventory control, but it is NOT the correct basis for 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 84 068AG2.4.1 8 Control Room Evac.

Knowledge of specific bases for EOPS.

The KA is matched because even though the question does expressly SAY there is an EOP involved, there would be for a loss of control room (E-0). Tripping the reactor is the first action in AP/1 7 for control room evacuation. The applicant is tested on the basis for a specific action associated with these conditions.

Cognitive Level:

High At first glance, this may seem like a simple recall question, since it involves recalling what the procedure says, and then what the basis is; but to get the right answer you must apply knowledge of one of the CRD vent fans functions (cooling vessel head area) and apply that to arrive at the correct answer.

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. I dated 03/11/2010) under the Screen Criteria for questions linked to I OCFR55.43(b)(5) (Assessment and Selection of Procedures):

I.

It cannot be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location.

2.

It cannot be answered solely by knowing immediate operator actions.

Page 190 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submiftal 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, or overall mitigative strategy of a procedure.

5.

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

Therefore, this is an SRO only question.

Page 191 of 235 Catawba 2012 NRC Exam Submittal

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 1 ASP to perform Enclosure 1 (ASP Operator Actions).

c.

Dispatch RO!SRO to Unit 1 AFWPTCP to perform Enclosure 2 (AFWPTCP Operator Actions).

2.

Announce the following twice over plant wide communications system: Loss of Unit I Control Room imminent. Assigned operations shift personnel man remote locations.

3.

Trip reactor.

4.

Verify Reactor Trip:

IF reactor will not trip, THEN dispatch operator 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.

CNS 2012 NRC Exam 100 Questions Final Submittal Question 85 076AA2.05 High Reactor Coolant Activity Ability to determineandil terprettbejplLQng as they apply to the High Reactor Coolant Activity:

CVCS letdown flow rate indication N Given the following Unit I conditions:

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 atesaetoo 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) appIicabased on current plant, status?

(N; (2)

Which one.pf the parameters evaluated per Technical Specification 3.4.16 has a lower limit 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

/

A 7

Page 192 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal 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 3 and 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 degrees so it does not apply for the conditions stated. The 2 conditions evaluated are DEl and gross specific activity. When letdown flow is increased above 80 gpm, lower DEl limits are in effect. This is stated in a TS amendment, the NV lesson, and confirmed based on steps in AP/12 and AP/18.

D.

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

References:

Technical Specification 3.4.16 (RCS Specific Activity)

KA Match:

Question 85 076AA2.05 High Reactor Coolant Activity Ability to determine and interpret the following as they apply to the High Reactor Coolant Activity:

CVCS letdown flow rate indication The KA is matched because the question involves a high reactor coolant activity condition, and includes testing on what it would mean to raise the letdown flow rate.

Cognitive Level:

High This is a higher cognitive level question because there is an evaluation of plant conditions and a determination of whether a Tech Spec is applicable. The applicant must also recall and apply knowledge of dose equivalent iodine concerns and the relationship of that with letdown flow rate requirements.

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 Screen Criteria for questions linked to 10CFR55.43(b)(2) (Tech Specs):

1.

It cannot be answered solely by knowing 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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.

Page 193 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal 4.

The question involves application of required actions of Tech Spec 3.4.16, RCS Specific Activity.

Therefore, this is an SRO only question.

Page 194 of 235 Catawba 2012 NRC Exam Submittal

CNS LOSS OF CHARGING OR LETDOWN PAGE NO.

API1IAI5500!012 20 of 32 Case II Revision 31 Loss of Letdown I

ACTION/EXPECTED

RESPONSE

RESPONSE NOT OBTAINED

16. (Continued) k.

WHEN 5 minutes have elapsed, THEN perform the following:

1)

IFATANYTIME letdown flow increased to greater than 80 GPM, THEN perform the following:

a)

Determine current NC Dose Equivalent Iodine concentration (DEl). (OAC Point Cl P0097) b)

Verify DEl specific activity b)

Ensure compliance with Tech LESS THAN 0.18 Ci/GM.

Spec 3.4.16 (RCS Specific Activity).

c)

Notify Primary Chemistry that lower DEl limits are in effect due to NV letdown flows greater than 80 GPM.

2)

Adjust I NV-849 (Letdn Flow Var Orif Ctrl) in 1% increments to desired letdown flow.

3)

WHEN letdown at desired flow, THEN perform the following:

a)

Adjust 1NV-148 (Letdn Press Control) to maintain letdown pressure at 350 PSIG.

b)

Ensure INV-148 (Letdn Press Control)

- IN AUTO.

4)

IFATANY TIME additional letdown flow desired, THEN establish letdown with the 45 or 75 GPM orifice. REFER TO OP/1/A16200/001 (Chemical and Volume Control System).

RCS Specific Activity 3.4.16 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 CONDITION REQUIRED ACTION COMPLETION TIME A.

DOSE EQUIVALENT Note 1-131 > 1.0 aCi/gm.

LCO 3.0.4.c is applicable.

A.1 Verify DOSE EQUIVALENT Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 1-131 within the acceptable region of Figure 3.4.16-1.

AND A.2 Restore DOSE 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months EQUIVALENT 1-131 to within limit.

B.

Gross specific activity of B.1 Be in MODE 3 with 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months the reactor coolant not Tavg < 500°F.

within limit.

(continued)

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

RCS Specific Activity 3.4.16 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

Required Action and C.1 Be in MODE 3 with 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Tavg < 500°F.

Time of Condition A not met.

OR DOSE EQUIVALENT 1-131 in the unacceptable region of Figure 3.4.16-1.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify reactor coolant gross specific activity < bOlE In accordance with iCi/gm.

the Surveillance Frequency Control Program (cbntinued)

Catawba Units 1 and 2 3.4.16-2 Amendment Nos. 263/259

RCS Specific Activity 3.4.16 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.16.2 NOTE Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT 1-131 specific In accordance with activity 1.0 [lCi/gm.

the Surveillance Frequency Control Program AND Between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a THERMAL POWER change of> 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period SR 3.4.16.3 NOTE Not required to be performed until 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for> 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Determine E from a sample taken in MODE 1 after a In accordance with minimum of 2 effective full power days and 20 days of the Surveillance MODE 1 operation have elapsed since the reactor was Frequency Control last subcritical for > 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Program Catawba Units 1 and 2 3.4.16-3 Amendment Nos. 263/259

300 250 I

-J I-0 o

200 U

0 LU U)

O 00 C,)

I.z Ui

-J D0 Ui Ui U)0 O

100 50 0

RCS Specific Activity 3.4.16 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 Catawba Units 1 and 2 3.4.16-4 Amendment Nos. 173/165

Qc EXAM BANK-Q1317 Unit 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 on current plant status?

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

A.

1. Yes

2. Dose Equivalent 1-131 B.

I. Yes

2. Gross specific activity C.

1.No

2. Dose Equivalent 1-131 D.

1.No

2. Gross specific activity

CNS 2012 NRC Exam 100 Questions Final Submittal Question 86 006G2.4.50 Emergency Core Cooling Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

Given the following Unit I conditions:

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 valves inside 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 Page 195 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal QUESTION 86 Distractor Analysis A.

Incorrect. FR-C.2 is plausible sirce given conditions resemble a degradation of core cooling (RVLIS level, PZR level, sump, etc.), but the appropriate procedure entry is for containment flooding. Ensuring power is available is plausible if applicant recognizes that there could be an issue due to flooding, but incorrectly reasons that alternate power sources 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 procedure contains the guidance for using the QAC printout, Valves Subject to Submergence Report for determining last known affected valve positions. This action would be taken due 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 86 006G2.4.50 Emergency Core Cooling Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

This KA is matched because the question is testing the ability to evaluate various parameters which may, or may not have exceeded and setpoint for alarm. The selection of procedures uses the knowledge of the alarm setpoint verification to arrive at the correct answer.

Cognitive Level:

High This is a higher cognitive level question because it involves an array of plant conditions, an evaluation of them, and a determination of which procedure will provide instruction for mitigation.

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 Screen Criteria for questions linked to I OCFR55.43(b)(5) (Assessment and Selection of Procedures):

Page 196 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal 1.

It cannot be answered solely by knowing systems knowledge, i.e., how the system works, 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, or 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 Containment Flooding Event.

Therefore, this is an SRO only question.

Page 197 of 235 Catawba 2012 NRC Exam Submittal

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 as follows:

Spent fuel pool

CA

CF

FWST.

g.

IF leakage into containment is suspected from system(s) listed in previous step, THEN isolate affected system(s) from containment.

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

3.

Use OAC printout, VALVES SUBJECT TO SUBMERGENCE REPORT, to determine last known affected valve positions.

4.

RETURN m procedure and step in effect.

END

Q-EXAM BANK - Q 693 During a large break LOCA, the ECCS suctions have been swapped to the Cold Leg Recirculation alignment. NS suction has been swapped to the containment sump. All equipment 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 above conditions 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 has become uncovered; EPIIIN5000IFR-C.2 (Response to Degraded Core Cooling)

C.

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

D.

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

CNS 2012 NRC Exam 100 Questions Final Submittal Question 87 7-,

013G2.4.2 Engineered Safety Features Actuation Kowledgeofitem 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 Safety Injection or Main Steam Isolation.

(2) are D.

(1)

To permit a normal Unit cooldown and depressurization without actuation of Safety Injection or Main Steam Isolation.

(2) are not

,\\

c_,

Page 198 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal QUESTION 87 Distractor Analysis A.

Incorrect. First part is correct. Plausible to reason that an important permissive interlock such 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 interlock functions to allow reset of ECCS without receiving another SIAS, is also an input to the ability to use steam dumps, and the feed pump runback to prevent overfilling (and overcooling) the S/Gs. Also per the Basis, this function is NOT required in order to meet the 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 permit a normal cooldown and depressurization, because it does allow operations that might otherwise cause an SI, by blocking circuits. However, this is the incorrect function.

References:

TS Bases for TS 3.3.2,.ESFAS, KA Match:

Question 87 01 3G2.4.2 Engineered Safety Features Actuation Knowledge of system set points, interlocks, and automatic actions associated with EOP entry conditions.

The KA is matched because the question tests knowledge of a permissive interlock associated with a reactor trip.

Cognitive Level:

Low 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. I dated 03/11/2010) under the Screen Criteria for questions linked to IOCFR55.43(b)(2) (Tech Specs):

1.

It cannot be answered solely by knowing 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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).

Page 199 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal 5.

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

Therefore, this is an SRO only question.

Page 200 of 235 Catawba 2012 NRC Exam Submittal

ESFAS Instrumentation B 3.3.2 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) a.

Engineered Safety Feature Actuation System InterlocksReactor Trip, P-4 The P-4 interlock is enabled when a reactor trip breaker (RTB) and its associated bypass breaker is open. Operators are able to reset SI 60 seconds after initiation.

If a P-4 is present when SI is reset, subsequent automatic SI initiations will be blocked until the RTBs have been manually closed.

This Function allows operators to take manual control of SI systems after the initial phase of injection is complete while avoiding multiple SI initiations. The functions of the P-4 interlock 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 rejection controller to the unit trip controller; and Prevent opening of the MFW isolation valves if they were closed on SI or SG Water LevelHigh High.

Each of the above Functions is interlocked with P-4 to avert or reduce the continued cooldown of the RCS following a reactor trip. An excessive cooldown of the RCS following a reactor trip could cause an insertion of positive reactivity with a subsequent increase in generated power. To avoid such a situation, the noted Functions have been interlocked with P-4 as part of the design of the unit control and protection system.

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

The RTB position switches that provide input to the P-4 interlock only function to energize or de-energize or open or close contacts. Therefore, this Function has no adjustable Catawba Units 1 and 2 B 3.3.2-26 Revision No. 10

CNS 2012 NRC Exam 100 Questions Final Submittal Question 88 059A2.04 Main Feedwater Ability 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 required 9

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

/

) /

(

Page 201 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal QUESTION 88 Distractor Analysis A.

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

B.

Incorrect. Plausibility of first part described in A above. Plausible that a system operating procedure would contain instructions for the desired operation, but FR-H.1 specifically 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 feedwater isolation signal as part of using CF as a feed source for the given conditions. Step 34 requires the SRO to remain in FR-H.1 and continue attempts to establish a secondary heat 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 88 059A2.04 Main Feedwater Ability 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 the consequences of those malfunctions or operations:

Feeding a dry S!G This KA is matched because plant conditions involve S/Gs that are dry (0% level), and the applicant 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 to select which procedure is the correct one to use.

Cognitive Level:

High This 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 of reset vs. bypassing a feedwater isolation signal to determine the correct action and procedure selection.

Source of Question:

NEW SRO Only:

Page 202 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal 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 Screen Criteria 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 system works, 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, or 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 the conditions.

Therefore, this is an SRO only question.

Page 203 of 235 Catawba 2012 NRC Exam Submittal

CNS RESPONSE TO LOSS OF SECONDARY HEAT SINK PAGE NO.

EPI1IAI5000/FR-H.1 40 of 88 Revision 41 I.

ACTION/EXPECTED RESPONSE I

RESPONSE

NOT OBTAINED 32.

Verify KC flow to ND heat exchangers IF AT ANY TIME an ND pump is operating INDICATING FLOW.

with flow less than 1000 GPM to NC loops AND KC to associated ND HX is isolated, THEN stop affected ND pump within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months .

33.

Align CA to establish control of SIG feed as follows:

a.

Ensure CA System valve control RESET.

b.

CLOSE CA flow control valves on S!Gs not presently being fed.

34.

Continue attempts to establish secondary heat sink in at least one SIG as follows:

.. CA. REFER IQ. Steps 6 through 7

CF or CM. REFER]QSteps 10 through 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 RETURNJ 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 ACTION/EXPECTED RESPONSE I

RESPONSE

NOT OBTAINED

11. (Continued) c.

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

d.

Ensure S/I

- RESET:

_i)

ECCS.

_1)

Reset ECCS.REFERTO EP/1/A15000/G-1 (Generic Enclosures), Enclosure 4 (ECCS Master Reset).

2)

D/G load sequencers.

2)

Dispatch operator to open affected sequencer(s) control power breaker:

I EDE-FOl F (Diesel Generator Load 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 equipment previously on.

e. if AT ANY TIME a subsequent Feedwater Isolation occurs, THEN RETURN Step 11.

CNS 2012 NRC Exam 100 Questions Final Submittal

/

Question 89 061A2.06 AuxiliarylEmergency Feedwater Ability 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 the consequences-ofdh6se malfunctions or operations:

Back leakage of MFW Given the following Unit I conditions:

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

-)

Page 204 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal 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 NOTE just 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 operability since 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 is some element of suction head associated with steam binding, but other operations preclude this.

References:

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

Question 89 061A2.06 Auxiliary!Emergency Feedwater Ability 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 the consequences of those malfunctions or operations:

Back leakage of MFW The KA is matched because given conditions involve backleakage of check valves in the feedwater piping and its impact on Aux. Feedwater, and then knowledge of procedure content regarding operability impact on the Aux. Feedwater Pump Cognitive Level:

High This is a higher cognitive level question because the applicant must evaluate plant conditions and 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 Screen Criteria for questions linked to IOCFR55.43(b)(2) (Tech Specs):

1.

It cannot be answered solely by knowing 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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.

Page 205 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal 4.

The question involves application of required actions of Tech Spec 3.7.5 (Auxiliary Feedwater).

5.

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

Therefore, this is an SRO only question.

Page 206 of 235 Catawba 2012 NRC Exam Submittal

.10 OP/11AJ62501002 Cooldown of the Motor Driven CA Pumps Page 3 of 11 Piping NOTE:

OAC point C1P1447 (Primary Thermal Output %) shall be used in determining reactor power 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 maintained 99.5% 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 necessary to cool the piping upstream of the flow control valve(s), perform the following for the applicable 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 for 1CA-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 following for 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 piping at the appropriate pump(s) and record the temperature:

CAPump1A

°F CAPump1B

.10 OP!11AJ6250/002 Cooldown 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 Points S/G 1 A CA Nozzle Feedwater Inlet Temp Rate Cl P1381 S/G lB CA Nozzle Feedwater Inlet Temp Rate C1P1382 S/G 1C CA Nozzle Feedwater Inlet Temp Rate C1P1383 SIG 1 D CA Nozzle Feedwater Inlet Temp Rate Cl P1384 2.

The CA pump discharge valves to the steam generators shall be opened very slowly 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 is initiated, the affected motor driven CA pump shall be shutdown within 10 minutes.

NOTE:

Cooling CA pump downstream piping prevents migration of hot water to CA pump and resultant potential gas formation.

Flow through the check valve is recommended to be increased to approximately 200 gpm. This flowrate is sufficient to remove debris in the check valve 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 to maintain 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 Disch To 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 Disch To 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 Disch To 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 Disch To S/G lD Check), throttle 1CA-40 (CA Pump lB Flow To S/G lD).

Q

- Or--J FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 78 fl578 SYSO61 A2.06 - Auxiliary / Emergency Feedwater (AFW) System Ability to (a) predict the impacts ofthe following malfunctions or operations on the AFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences ofthose malfunctions or operations: (CFR: 41.5 /43.5 /45.3 /45.13)

Back leakage ofMFW 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 method would 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 monitor temperature 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 monitor temperature 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 TD CA 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 TD CA 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

CNS 2012 NRC Exam 100 Questions Final Submittal Question 90 078G2.4.8

\\

I Instrument Air Knowledge 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 0445 VI Pressure (psig) 74 59 54 49 (1)

Which ONE of the following describes the EARLIEST time the SRO must direct the reactor to 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 Page 207 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal QUESTION 90 Distractor Analysis A.

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

B.

Incorrect, Incorrect action (valves are affected but at a lower pressure than the trip) tied with the incorrect time (press). Incorrect time is associated with aligning back up sealing to 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 aligned 60 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 90 078G2.4.8 Instrument Air Knowledge 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 for the topic; however, the abnormal aspect is met by testing the use of the AP and its required actions in conjunction with conditions requiring a manual reactor trip (EOP entry).

Cognitive Level:

High This is a higher cognitive level question because the applicant must analyze and trend of a plant parameter against a time matrix to determine the earliest time that the reactor needs to be tripped.

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 Screen Criteria for questions linked to 1 OCFR55.43(b)(5) (Assessment and Selection of Procedures):

Page 208 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal 1.

It cannot be answered solely by knowing systems knowledge, i.e., how the system works, 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, or 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.

Therefore, this is an SRO only question.

Page 209 of 235 Catawba 2012 NRC Exam Submittal

I PAGE NO.

I CNS LOSSOFINSTRUMENTAIR I

12of87 I

AP101A155001022 I

- Page 1 of 27 I

Revision 331 Unit I Loss Of VI System Actions I

I ACTION/EXPECTED RESPONSE I

RESPONSE

NOT OBTAINED 1.

fJpy TIME VI pressure is less than 55 PSIG AND decreasing, THEN:

a.

Trip reactor.

b.

WHEN reactor power less than 5%,

THEN depress CLOSE pushbutton for all MSIVs.

c.

Continue in this procedure as time permits.

d.

QjQEPI1!AI5OOOIE-O (Reactor Trip Or Safety Injection).

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

Time 0345 0415 0430 0445 VI Pressure 74 59 54 49 Which one of the following describes the time the SRO must direct the reactor to be tripped and what action must be taken following the trip?

A.

1. 0415

2. Close MSIV5 B.

1. 0415

2. Close CF Regs and Bypasses C.

1. 0430

2. Close MSIVs D.

1. 0430

2. Close CF Regs and Bypasses

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

Time 0345 0415 0430 0445 VI Pressure 74 59 54 49 Which one of the following describes the time the SRO must direct the reactor to be tripped and what action must be taken following the trip?

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

CNS 2012 NRC Exam 100 Questions Final Submittal Question 91 01 4A2.06 Rod Position Indication Ability 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 the consequences 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..EDfo 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 Instrumentation Status), which ONE of the following is required?

A.

The shift work manager directs Reactor Engineering to initiate a temporary modification to change the Control Rod Bank Lo Limit (IAD-2, N9) annunciator setpoint to the Control Rod 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 to duplicate the function of the failed annunciator.

D.

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

Page 210 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal QUESTION 91 Distractor Analysis A.

Incorrect. An increased surveillance sheet must be initiated.

If the applicant does not know 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 that alternate 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 91 01 4A2.06 Rod Position Indication Ability 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 the consequences of those malfunctions or operations:

Loss of LVDT Even though this type of KA is sometimes a higher cognitive level KA, this KA has been matched 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 which monitors insertions limits of the control rods (similar function of an LVDT which uses a linear variable differential transformer) to provide an output. The impact is what has to be done as a result of the failure.

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

Cognitive Level:

Low 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 Screen Criteria for questions linked to IOCFR55.43(b)(1), (Conditions and Limitations in the facility license):

I.

It involves administrative requirements for a loss of instrumentation for control rod insertion limits.

Page 211 of 235 Catawba 2012 NRC Exam Submittal

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 the function of the failed instrumentationlannunciator.

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

Technical Specifications, Safety Function, Redundant indications, and alternate monitoring.

6.2.

The Shift Technical Advisor shall be notified of any instrumentation determined to be failed.

7.

Increased Surveillance Sheet Instructions 7.1.

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

7.2.

Detail shall be provided in the Items to Monitor section to give the operator performing 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 Monitored Criteria 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 Supervisors discretion.

This shall be set to ensure any potential problem is discovered in a timely manner.

Engineering guidance may be required to determine proper monitoring frequency.

7.5.

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

7.6.

Active Increased Surveillance Sheets shall be filed in the appropriate section of the Ops Shift Routines Logbook.

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

CNS 2012 NRC Exam 100 Questions Final Submittal Question 92 01 7A2.02 In-core Temperature Monitor Ability to (a) predict the impacts of the following malfunctions or operations on the ITM system; and (b) based on those predictions, use procedures to correct, control, or mitigate the-consequences of those malfunctions or operations:

Core damage Given the following Unit I conditions:

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 will occur is if CET temperatures are NOT LESS than (I)

(2)

If the RNO column of Step 26 for CET temperatures INCREASING applies, what is required?

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 Page 212 of 235 Catawba 2012 NRC Exam Submittal

)(

cc N

jk i

<1

CNS 2012 NRC Exam 100 Questions Final Submittal QUESTION 92 Distractor Analysis A.

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 to restore core cooling.

B.

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

C.

Incorrect. First part is correct. Notifying Engineering to assess core damage is plausible since the listed procedure (RP/15 for Core DamageAssessment) does exist, but would not apply for the listed step.

L D.

CORRECT. Per the quoted reference, 1200°F is the temperature at which core damage begins to occur. The RNO for the listed step requires implementation of the Severe Accident 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 92 01 7A2.02 In-core Temperature Monitor Ability to (a) predict the impacts of the following malfunctions or operations on the ITM system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Core damage The KA is matched because conditions involve core damage conditions, and how that would impact the indications on the CETs. Then the applicant is tested on procedure implementation to mitigate the conditions.

Cognitive Level:

High This is a higher cognitive level KA because the applicant must analyze plant conditions and determine that these represent core damage conditions, and then make a determination on which procedure flowpath to use.

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. I dated 03/11/2010) under the Screen Criteria for questions linked to I OCFR55.43(b)(5) (Assessment and Selection of Procedures):

Page 213 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal 1.

It cannot be answered solely by knowing systems knowledge, i.e., how the system works, 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, or 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.

Therefore, this is an SRO only question.

Page 214 of 235 Catawba 2012 NRC Exam Submittal

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 associated loops available, THEN RETURN m Step 23.

c.

IF core exit temperatures increasing, THEN EG/1/AICSAM/SACRG1 (Severe Accident Control Room Guideline 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 all CLA discharge isolation valves.

REFER TO EPI1IA/50001G-1 (Generic Enclosures), Enclosure 9 (Power Alignment for CLA Valves).

c.

Ensure S/I

- RESET:

_1)

ECCS.

1)

Perform thefollowing:

a) if either reactor trip breaker is closed, THEN dispatch operator to open Unit 1 reactor trip breakers.

b)

WHEN reactor trip breakers open, THEN reset ECCS.

CNS 2012 NRC Exam 100 Questions Final Submittal Question 93 034G2.1.27 Fuel Handling Equipment Knowledge of system purpose andlor function.

In accordance with the BASIS for Selected Licensee Commitment (SLC) 16.9-20 (Refueling Operations - 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 Page 215 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal 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 the load restriction. The basis is correct.

B.

Incorrect. Plausibility of first part is described in A above. Second part is plausible since the 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 control rod assembly and associated handling tool over other fuel assemblies 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 fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the safety analyses.

Loads in excess of 3000 pounds shall be prohibited from travel over fuel assemblies in the storage 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 93 034G2.1.27 Fuel Handling Equipment Knowledge of system purpose andlor function.

The KA is matched because the question tests knowledge of the weight of fuel handling equipment in the context of load restrictions over the Spent Fuel Pool.

Cognitive Level:

Low Source of Question:

NEW SRO Only:

Page 216 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal 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 Screen Criteria for questions linked to IOCFR55.43(b)(7):

1.

It involves assessment of fuel handling equipment surveillance requirement acceptance criteria (load limitations).

Page 217 of 235 Catawba 2012 NRC Exam Submittal

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

- Crane Travel

- Spent Fuel Storage Pool Building COMMITMENT NOTE Spent fuel pool weir gates may be moved by crane over the stored fuel provided the spent fuel has decayed for> 19.5 days since last being part of a core at power.

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

APPLICABILITY:

With fuel assemblies in the storage pool.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

COMMITMENT not met.

A.1 Place the crane load in a Immediately safe condition.

TESTING_REQUIREMENTS TEST FREQUENCY TR 16.9-20-1 NOTE Spent fuel pool weir gates may be moved by crane over the stored fuel provided the spent fuel has decayed for 19.5 days since last being part of a core at power.

Verify that the weight of each load, other than a fuel Prior to moving assembly 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 Building 16.9-20 BASES The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies 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 fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the safety analyses.

REFERENCES 1.

Letter from NRC to Gary R. Peterson, Duke, Issuance of Improved Technical Specifications Amendments for Catawba, September 30, 1998.

2.

Letter from NRC to Gary R. Peterson, Duke, Issuance of Amendments 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 94 G2.1.23 Conduct of Operations Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Given the following Unit I conditions:

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 NON functional.

Which ONE of the following completes the statements above?

A.

(1) 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months (2) will

/

B.

(1) 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months I

(2) will not i)

C.

(1) 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (2) will D.

(1) 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (2) will not 7

A/f I

Page 218 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal 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 and determine leak size.

Leak rate is 78-45-12= 21 gpm, making the Standby Makeup Pump INOPERABLE in accordance with SLC 16.7-9. Step 8 of AP-1 0 Case 1, directs an SRO to evaluate if leakage 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 would be exceeded.

If this leakage is exceeded, Condition B requires the unit be in Mode 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . 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 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months 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 seal return flow (12 GPM) from Charging flow along with subtracting Letdown flow (45 GPM). If that 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 be less than 20 GPM, the Standby Makeup Pump would NOT have to be declared INOPERABLE.

C.

Incorrect. Part (1) Plausible because this is correct per the requirement of Condition B of TS 3.4.13 (RCC Operational Leakage) which requires the unit to be in Mode 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

It would be reasonable for the applicant to believe this would also be the required time specified in AP-1 0.

Part (2) is correct.

D.

Incorrect.

. Part (1) Plausible because this is correct per the requirement of Condition B of TS 3.4.13 (NC Operational Leakage) which requires the unit to be in Mode 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . It would be reasonable for the applicant to believe this would also be the required time specified in AP-lO.

Part (2) is plausible if the applicant subtracts actual seal injection (20 GPM) instead of seal return flow (12 GPM) from Charging flow along with subtracting Letdown flow (45 GPM). If that 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 be less than 20 GPM, the Standby Makeup Pump would NOT have to be declared INOPERABLE.

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 Page 219 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal KA Match:

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

Cognitive Level:

High This is a higher cognitive level question because the applicant must perform calculation (solve a problem) and then perform a level of analysis concerning the given indications and predict the impact and determine the correct procedural course of action.

Source of Question:

Bank CNS 4440 SRO Only:

This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev Idated 03/11/2010 for screening questions linked 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 the actions described are immediate actions.

3) The question can NOT be answered solely by knowing entry conditions for AOP or direct entry 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 of events, or overall mitigative strategy of the procedure. This is detailed knowledge of procedure content related to knowing the plant shutdown requirements.

5) The question also requires the applicant to recall a below the bar TS (SLC) limit associated with the S/G tube leakage. Therefore, it is SRO knowledge.

Page 220 of 235 Catawba 2012 NRC Exam Submittal

CNS I

REACTOR COOLANT LEAK I

PAGE NO. I AP/1/N5500/O1O I

I 11 of 158 I

Case I

Revision 561 Steam Generator Tube Leak I

I ACTION/EXPECTED RESPONSE I

I

RESPONSE

NOT OBTAINED I

7.

Minimize Secondary contamination as follows:

a.

Remove CM polishing demineralizers from service as follows:

1)

Ensure POLSH DEMIN BYP CTRL

- IN MANUAL.

2)

Ensure POLSH DEMIN BYP CTRL

- OPEN.

3)

Notify Secondary Chemistry CM polishing demineralizers have been bypassed.

b.

Align auxiliary systems to minimize secondary side contamination. REFER TO EPI1IAI5000IG-1 (Generic Enclosures), Enclosure 2 (Minimizing Secondary Side Contamination).

c.

Stop any transfer of water between both Units CSTs.

8.

Ensure compliance with appropriate Tech Specs and Selected Licensee Commitments 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!11N55001010 Case I

l9of 158 Steam Generator Tube Leak Revision 561 ACTION/EXPECTED RESPONSE I

RESPONSE

NOT OBTAINED 14.

Determine unit shutdown requirements as follows:

a. !EAINY TIME leak rate is greater than or equal to 100 gpd, THEN perform the following:

1)

Ensu eactor power less than 50%

with I h hrs 2)

Ensure unit Mode 3 within the foIlowir

3) Observe Note prior to Step 15 and QIQ Step 15.

b. if leak rate is greater than or equal to 75 gpd and less than 100 gpd, THEN perform the following:

1) IEAIANY TIME the following conditions are met:

. Any main steam line N-16 radiation monitor

- INOPERABLE AND Cl P0187 (Estimated Total Pri To Sec Leakrate)

- INVALID.

THEN perform the following:

a)

Ensure reactor power less than 50% within 1 hr.

b)

Ensure unit in Mode 3 within the following 2 hrs.

c)

Observe Note prior to Step 15 and QIQ Step 15.

c.

if leak rate is greater than or equal to 75 gpd and less than 100 gpd sustained for one hour, THEN ensure unit in Mode 3 within 24 hrs.

SSS 16.7-9 16.7 INSTRUMENTATION 16.7-9 Standby Shutdown System (SSS)

COMMITMENT The SSS shall be FUNCTIONAL.

APPLICABILITY:

MODES 1, 2, and 3.

REMEDIAL ACTIONS NOTE SLC 16.2.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A.

SSS non-functional.

A.1 Restore SSS to 7 days FUNCTIONAL status.

B.

Total accumulative B.1 Declare the standby Immediately LEAKAGE from makeup pump non-unidentified LEAKAGE, functional and enter identified LEAKAGE, Condition A.

and reactor coolant pump seal LEAKAGE>

20 gpm.

C.

More than one cell in a C.1 Enter Condition A.

Immediately 24-Volt battery bank is <

1.36 volts on float charge with no other cells jumpered.

D.

Required Action and D.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not AND met.

D.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Catawba Units 1 and 2 16.7-9-1

-Revis1on--7

9-- 0 J

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 conditions specified is (1)

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

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

A.

1. 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months

2. will B.

1. 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months

2. will not C.

1. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

2. will D.

1. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

2. will not

l(2 CNS 2012 NRC Exam 100 Questions Final Submittal Question 95 G2.1.35 Conduct of Operations Knowledge 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 rods under 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 the Refueling SRO in the Control Room ONLY..

C.

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

D.

The Fuel Handling Maintenance Supervisor must establish communications witifhe Refueling SRO in the Control Room AND containment integrity must be restored.

/)

(

Page 221 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal 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 integrity must be established when core alterations are in progress. Plausible:

If the candidate does 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 be maintained.

D.

Incorrect. The SRO must be physically present in containnient when core alterations are in progress. Plausible:

If the candidate does not realize that latching control rods is a core alteration 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 95 G2.1.35 Conduct of Operations Knowledge of the fuel-handling responsibilities of SROs.

The KA is matched because it tests knowledge of fuel handling responsibilities of the Refueling SRO.

Cognitive Level:

Low Source of Question:

Bank CNS 1299 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 Screen Criteria for questions linked to IOCFR55.43(b)(7):

1.

It involves assessment of fuel handling equipment surveillance requirement acceptance criteria (load limitations).

Page 222 of 235 Catawba 2012 NRC Exam Submittal

O S C)

EXAM BANK - Q 1299 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 requirements (if any) must be met to proceed with latching -control rods 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 -communications with the Refueling SRO in the Control Room.

C.

The Refueling SRO must be in containment and containment integrity must be restored.

D.

The Fuel-Handling Maintenance Supervisor must establish communications with the Refueling SRO in the Control Room and containment integrity must be restored.

fj CNS 2012 NRC Exam 100 Questions Final Submittal Question 96 G2.2.1 Equipment Control Ability to perform pre-startup procedures for the facility, including operating those controls 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 to Criticality), 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 be operated?

(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

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Page 223 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal 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 would be used for a dropped rod situation. In this case, the Limits and Precautions of the approach to criticality takes precedence over other procedures, and this is explicitly stated in the PT procedure.

It is plausible to for an applicant to confuse this guidance, since the procedure does contain guidance for1nsertingafl Control and Shutdown Banks, but this is for a malfunction on rods in the Shutdown Banks C.

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

IF a control rod fails to withdraw or a single rod is dropped during approach to criticality perform 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 Rod Misalignment), and therefore shall take precedence. {PIP C-06-4287}

D.

Incorrect: The procedure is correct. It is plausible to for an applicant to confuse this guidance, since the procedure does contain guidance for inserting all Control Shutdown 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 96 G2.2.1 Equipment Control Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity.

The KA is matched because the question conditions involve a startup, but it is testing knowledge of PRE-startup since it is an approach to criticality condition. The operating controls aspect is met because the question involves knowledge of how the rods will be operated for the given conditions.

Cognitive Level:

High Page 224 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal This is a higher cognitive level question because conditions are given for an approach to criticality, and a subsequent dropped rod. Analyzing the conditions is required in order to make the correct selection of procedure.

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. I dated 03/11/2010) under the Screen Criteria 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 system works, 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, or 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) to mitigate the condition.

Therefore, this is an SRQ only question.

Page 225 of 235 Catawba 2012 NRC Exam Submittal

PT/O/A/4 150/019 Page 4 of 13 6.4 IF a control rod fails to withdraw or a single rod is dropped during approach to criticality 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, Rod Control Malfunction, and therefore shall take precedence. {PIP C-06-4287}

6.5 IF more than one control rod drops during the approach to criticality, MANUALLY TRIP 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 rod withdrawal to be halted AND IAE cannot determine cause for alarm and repair problem or determine if further rod withdrawal is permissible, reinsert all Control Banks.

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

jstartup is xenon-free (xenon worth < 100 pcm), take action within 60 minutes of 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 between 555 and 559 °F during approach to criticality.

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

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 be suspended. Rod withdrawal may be resumed only after the source of the abnormality has been identified and it has been determined that it will not jeopardize plant 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)/Rod Insertion Limits, insert all control banks and contact Reactor Systems Engineering Supervisor or designee.

6.11 Ensure that the NC boron sample used for reactivity balance calculations is representative of current NC system boron (i.e. taken with all four NC pumps operating, 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 97 G2.2.5 Equipment Control Knowledge 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 for security guards.

B.

Revision to the Emergency Plan changes the designated assembly areas for accountability.

C.

System modification that adds a backup Nitrogen accumulator to an air operated containment isolation valve.

D.

Change to the Nuclear Quality Assurance Plan Page 226 of 235 Catawba 2012 NRC Exam Submiftal

CNS 2012 NRC Exam 100 Questions Final Submittal QUESTION 97 Distractor Analysis A.

Incorrect. Security Guard staffing is not covered under Tech Specs but under I 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 is similar in concept to I OCFR5O.59 but not the same. Plausible:

If the applicant is not familiar with the requirements for USQs.

C.

CORRECT. Containment isolation valves are Tech Spec SSCs D.

Incorrect. Does not require a 10CFR5O.59 evaluation. The QA Plan is not a Tech Spec SSC. Plausible: If the applicant is not familiar with the requirements for USQs. NQA is covered under I OCFR5O Appendix B.

References:

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

Question 97 G2.2.5 Equipment Control Knowledge 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 situations this type of review would be required.

Cognitive Level:

Low Source of Question:

Bank CNS 1292 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 Screen Criteria for questions linked to I OCFR55.43(b)(3) (Facility licensee procedures required for operating changes in the facility):

I.

It involves processes for changing the plant or plant procedures.

Therefore, this is an SRO only question.

Page 227 of 235 Catawba 2012 NRC Exam Submittal

VERIFY HARD COPY AGAINST WEB SITE IMMEDIATELY PRIOR TO EACH USE NSD 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 performed commensurate 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 for technical accuracy by Checkers. Furthermore, they are not required to be approved by Approvers nor are they required to be reviewed by Peer Reviewers. Such activities are optional unless deemed appropriate by responsible supervisors or managers (or designees). Notwithstanding, some operability determinations and functionality assessments may rely upon calculations or other engineering documents that require independent verification and approval. In such cases, the requirements for independent verification and approval are provided in the applicable governing procedure(s). For example, if an operability determination relies upon a calculation, then that calculation must be checked and approved in accordance with EDM-101.

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

4 REVISION 24 VERIFY HARD COPY AGAINST WEB SITE IMMEDIATELY PRIOR TO EACH USE

VERIFY HARD COPY AGAINST WEB SITE IMMEDIATELY PRIOR TO EACH USE Nuclear Policy Manual Volume 2 NSD 203 T0CFR5O.59 Regulation that establishes the conditions under which Part 50 licensees may make changes to the facility or procedures and conduct tests or experiments without prior NRC approval. Refer to NSD 209, 10CFR5O.59 Process, for guidance regarding application of 10CFR5O.59.

2.

1 OCFR72.48 Regulation that establishes the conditions under which Part 72 licensees may make changes to the facility or procedures and conduct tests or experiments without prior NRC approval. Refer to NSD 211, IOCFR72.48 Process, for guidance regarding application of 10CFR72.48.

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; or 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 otherwise compensate 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 10CFR54.3, the CLB is the set ofNRC requirements applicable to a specific plant and a licensees written commitments for ensuring compliance with and operation within applicable NRC requirements and the plant-specific design basis (including all modifications and additions to such 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, Exemptions c.

Technical Specifications (TSs) and TS Bases d.

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 relied on to grant, amend, or modify the operating license and technical specifications and to ensure continued compliance 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) in which one or more of the following conditions exist:

There has been any loss of required quality or functional capability. Examples of degraded conditions are failures, malfunctions, deficiencies, deviations, and defective material and equipment. Examples of conditions that can reduce functional capability of a system are aging, erosion, corrosion, improper operation, 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 functional capability 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 the specific values or ranges of values chosen for controlling parameters as reference bounds for design. These values may be (1) restraints derived from generally accepted state of the art practices for achieving functional goals, (2) requirements derived from analysis (based on calculation or experiment) of the effects of a postulated accident for which a SSC must meet its functional goals. Design bases information, defined by 10CFR 50.2, is documented 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 10CFR5O.2 Design Bases, endorses Appendix B to Nuclear Energy Institute (NET) document NET 97-04, Guidance and Examples for Identifying 10CFR5O.2 Design Bases.

REVISION 24 5

VERIFY HARD COPY AGAINST WEB SITE IMMEDIATELY PRIOR TO EACH USE

VERIFY 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 provided prior 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 SSCs described in TSs can perform their specified safety functions but may not be required in the winter. If an operability determination concludes that the ventilation system does not currently perform a necessary 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 the ventilation system is not required remains valid. In addition, corrective actions should be established to ensure timely restoration of the ventilation system.

g.

Facility operation should be consistent with the CLB for the facility. Thus, degradedlnonconforming conditions should be resolved in a time frame commensurate with their safety significance (refer to Appendix D.3)

Operability and functionality are separate from corrective action to restore full compliance to the CLB, including all applicable codes and standards, design criteria, safety analyses assumptions and specifications, and licensing commitments. Corrective actions to restore full compliance to the CLB should be addressed through the corrective action process and are not completely addressed by this directive. The treatment of operability and functionality as a separate issue from the restoration of full compliance emphasizes that the operability determination and functionality processes are focused on safe plant operation and should not be impacted by decisions or actions necessary 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 those documents 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 preliminary conservative 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 to otherwise 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 otherwise compensate for degraded/nonconforming conditions that call into question functionality.

Compensatory measures should have minimal impact on plant operations and should be relatively simple to implement. Implementing compensatory measures for SSCs that have been determined to be degraded or nonconforming may restore plant operating margins.

In general, there are two types of compensatory measures: (1) those that do g constitute changes and (2) those that constitute changes. According to NEI 96-07, Revision 1, change means a modification or addition to, or removal from, the facility or procedures that affects: (1) a design function, (2) method of performing or controlling the function, 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 operating procedure.

Using an alternate means to monitor Reactor Coolant Pump lower reservoir oil levels as outlined in an approved alarm manual.

Taking actions specified by Test Acceptance Criteria (TAC) sheets or similar type documents that have been evaluated under 10CFR 50.59.

Increasing the frequency of operator rounds or monitoring activities.

32 REVISION 24 VERIFY HARD COPY AGAINST WEB SITE IMMEDIATELY PRIOR TO EACH USE

- a c EXAM BANK - Q 1292 Which one of the following changes will require a I OCFR5O.59 review?

A.

Change to the Physical Security Plan that reduces the shift staffing requirements for security guards.

B.

Revision to the Emergency Plan changes the designated assembly areas for accountability.

C.

System modification that adds a backup Nitrogen accumulator to an air operated containment isolation valve.

D.

Changes to the Nuclear Quality Assurance Plan

CNS 2012 NRC Exam 100 Questions Final Submittal Question 98 G2.3.1 5 Radiation Control Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, 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 the required 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.

Page 228 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal QUESTION 98 Distractor Analysis A.

CORRECT. Note prior to Step 1.1 of RP/07 The four Reactor Coolant Leakage Detection Systems are not seismically qualified and must be assumed to be inoperabJe-followirg any sejsmicevenLEME3S(L) can be verified to be operable based orçpoweravailabllityandiample pump operatiorL )

Technical Specification 3.0.3 is entered until one of the Reactor Coolant leakage detection systems 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 Leakage Detection Systems (listed below) as inoperable:

1.1.1 Containment Floor and Equipment Sump Level Monitors and the Incore Instrument Sump 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 is required FIRST, per RP/07.

C.

Incorrect. Applicant confuses guidance for an alternate alignment, which actually renders the EMF inoperable.

D.

Incorrect: Plausible, since EMF38(L) is assumed to be INOPERABLE, and other EMFs in the plant when inoperable, require manual sampling.

References:

RP/0/A15000/007, Natural Disaster and Earthquake, Enclosure 4.4 (Earthquake), Rev.

033 Question 98 G2.3.1 5 Radiation Control Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

The KA is matched because it tests knowledge of the effect of a seismic event on a particular radiation monitor (operability, etc.)

Coinitive Level:

Low Source of Question:

NEW Page 229 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal 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 Screen Criteria for questions linked to I OCFR55.43(b)(2) (Tech Specs):

1.

It cannot be answered solely by knowing < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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.

Therefore, this is an SRO only question.

Page 230 of 235 Catawba 2012 NRC Exam Submittal

.4 RPJO/AI5000/oo7 Earthquake 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 qualified and must be assumed to be inoperable following any seismic event. EMF38(L) can be verified to be operable based on power availability and sample pump operation.

3.

Reactor Coolant Leakage Detection Systems are not required to be operable during Cold Shutdown.

4.

Technical Specification 3.0.3 is entered until one of the Reactor Coolant leakage detection 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 a seismic system actuation. This alarm is in addition to event indications present on 1IEECS1000NCC on 1MC8.

1.1 Following any earthquake that is felt in the protected area or is recorded on instrumentation, including earthquakes smaller than OBE, declare all four Reactor Coolant Leakage Detection Systems (listed below) as inoperable:

{23}

1.1.1 Containment Floor and Equipment Sump Level Monitors and the Incore Instrument 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 following methods:

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

CNS 2012 NRC Exam 100 Questions Final Submittal Question 99 G2.3.6 Radiation Control Ability to approve release permits.

Given the following conditions:

An Auxiliary Buildingwaste monitor tank !iquid waste release,(LWR) package has been delivered to-thControl 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 Dishirge 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 Standby Nuclear Service Water Pond B.

The CRS should allow this release since RL flow is greater than required for the WL release ONLY.

C.

The CRS should allow this release since EMF 49 is operable AND RL flow is greater than required for the WL release.

D.

The CRS should NOT allow this release since EMF 57 is NOT operable.

Page 231 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal QUESTION 99 Distractor Analysis A.

CORRECT. With RN aligned to the Standby Nuclear Service Water Pond, release cannot be approved.

B.

Incorrect. Cannot approve due to RN being aligned to the SNSWP. Plausible because the distracter makes a true statement. RN is aligned to the SNSWP.

C.

Incorrect. Correct approval but the reason is flawed. Discharge to RN is allowed with EMF-49 mop provided Radiation Protection takes action per their procedure.

D.

Incorrect. Plausible, since EMF-57 is a radiation monitor associated with releases, but its operability does not affect this release.

References:

SLCI6.11-2 KA Match:

Question 99 G2.3.6 Radiation Control Ability to approve release permits.

The KA is matched because the SRO applicant is presented with conditions involving a proposed liquid radwaste release, and then tested on whether the release should be allowed, and why or why not.

Cognitive Level:

High This is a higher cognitive level question because it involves analysis of plant conditions, and a conclusion on whether radwaste release should be authorized.

Source of Question:

Bank CNS 898 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 Screen Criteria for questions linked to I OCFR55.43(b)(4) (Radiation Hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions):

1.

It involves the process for a liquid release approval.

Page 232 of 235 Catawba 2012 NRC Exam Submittal

OP/O/B/6500/1 13 Page 2 of 2 Operations Liquid Waste Release 1.

Purpose To aid the operator in the correct methods of performing steps in Radwaste procedure OP/0/B/6500/0 15 (Discharging a Monitor Tank to the Environment) and Radiation Protection procedure HP/0/B/1004/004 (Radioactive Liquid Waste Release). Also to aid the operator as to limits 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 drops below the setpoint for ORLP5O8O (RL Discharge Total Flow), 1WL-l24 (Waste Monit Tnk Pmps Disch) will trip closed.

2.4 Releases that are interrupted by EMF-49 HI-RAD trips maybe initiated up to a maximum of three times, including original initiation, without re-sampling per HP/O/B/1 004/004 (Radioactive Liquid Waste Release).

2.5 Turbine Building Sump releases are secured if the pre-set levels are exceeded on l/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 898 An auxiliary building waste monitor tank liquid waste release (LWR) package has been 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 Room Supervisor 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 Nuclear Service Water Pond.

C.

The CRS should NOT approve this release; a release is NOT allowed when MF 49 is inoperable.

D.

The CRS should approve this release; a release is allowed when EMF 57 is operable.

CNS 2012 NRC Exam 100 Questions Final Submittal Question 100 G2.4.23 Emergency Procedures!Plans Knowledge of the bases for prioritizing emergency procedure implementation during emergency 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 CURRENT conditions?

(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 Owners Group Background Documents C.

(1)

Discontinue implementation of the originally selected CSF procedure and GO TO FR H. 1.

(2)

RP/O1, Classification of Emergency, Fission Product Barrier Matrix D.

(1)

Discontinue implementation of the originally selected CSF procedure and GO TO FR-H.1.

(2) Westinghouse Owners Group Background Documents Page 233 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal QUESTION 100 Distractor Analysis A.

Incorrect. FR-C.2, step 13 for plausibility of staying in C.2:. IF AT ANY TIME a red path on 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 procedure is completed. RP/O1, Classification of Emergency, Fission Product Barrier Matrix is plausible since that document contains prioritization related items, but it is for the purpose of emergency classification.

B.

Incorrect. Plausibility of staying in FR-C.2 per description in A above. Second part is correct.

C.

Incorrect. First part is correct. Plausibility of second part is described in A above.

D.

CORRECT. From EAL Basis Document: 4.1.N.1 EOPs 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 of multiple events is inherently addressed. The EOPs contain detailed instructions regarding the monitoring of these functions and provides a scheme for classifying the significance of the challenge to the functions. In providing EALs based on these schemes, the emergency classification can flow from the EOP assessment rather than being based on a separate EAL assessment. This is desirable as it reduces ambiguity and reduces the time necessary 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 100 G2.4.23 Emergency ProcedureslPlans Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations.

The KA is matched because the applicant must analyze conditions involving critical safety function status changes, and the how the emergency procedures are implemented based on that prioritization.

Cognitive Level:

High This is a higher cognitive level question because the applicant analyzes conditions to determine the priority of emergency procedure implementation.

Source of Question:

NEW SRO Only:

Page 234 of 235 Catawba 2012 NRC Exam Submittal

CNS 2012 NRC Exam 100 Questions Final Submittal 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 Screen Criteria 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 system works, 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, or 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: to mitigate the Critical Safety Function that has the highest priority per the rules of usage.

Therefore, this is an SRQ only question.

Page 235 of 235 Catawba 2012 NRC Exam Submittal

Operations Management Procedure 1-7 Page 12 of 31 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 for challenges 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 be monitored and implemented as directed by E-O or ECA-O.O.

CSF Status Trees may also be used when EPs are the controlling procedure to determine or identify abnormal conditions.

C.

CSF procedures shall not be implemented until the entry conditions are met (PPRB OPS-12571).

D.

The responsibility for monitoring the CSF Status Trees is pre-assigned to the Shift Technical Advisor (STA). However, the Operations Shift Manager may reassign this responsibility based on available resources.

Operations Management Procedure 1-7 Page 13 of3l E.

After the CSF Status Trees have been implemented, the following rules of usage apply:

Use of EPs to restore a critical safety function is based on a two factor priority system. The first factor considers the relative importance of the safety function in an accident scenario. On the OAC alarm video, the order of priority is established by position along the bottom of the screen. SUBCRITICALITY at the left side of the screen has the highest priority. The others in decreasing priority, are:

CORE COOLING, HEAT SINK, NC INTEGRITY, CONTAINMENT (Integrity) and NC INVENTORY on the right side of the screen.

The second factor considers the degree of severity to which the critical safety function is being challenged.

The order of priority is designated by color in decreasing 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 this validation process.

it If a valid red path is encountered, the operator shall immediately implement the corresponding EP. If during the performance of any red path procedure, a red condition of higher priority arises, then the higher priority condition shall be addressed first, and the lower priority red path procedure suspended.

NOTE:

An instrument or computer related failure that causes an erroneous SPDS indication is the only example of an invalid CSF Status Tree condition.

Operations Management Procedure 1-7 Page 14 of 31 If a valid orange path is encountered, the operator is expected to scan all of the remaining trees, and then, if no red path is encountered, to promptly implement the corresponding EP. If during the performance of an orange path procedure, any red condition or higher priority orange condition arises, then the red or higher priority orange condition shall be addressed first, and the original orange path procedure suspended.

Once a procedure is entered due to a valid red or orange condition, that procedure shall be performed to completion unless preempted by some higher priority condition. It is expected that the actions in the procedure will clear the red or orange condition before all the operator actions are complete. However, these procedures shall be performed to the point of the defined transition to a specific procedure. At this point, any lower priority red or orange paths currently indicating or previously started but completed shall be addressed.

If a CSF procedure directs the operator to return to the procedure and step in effect and the corresponding status tree continues to display the off normal condition, then the corresponding CSF procedure does not have to be implemented again since all recovery actions have already been completed. However, if the same status tree subsequently changes to a valid higher priority condition, then the corresponding CSF procedure shall be implemented as required by its priority.

Certain CSF procedures are used to address both orange and red path conditions for the same parameters. If the procedure is already in progress due to the orange path condition, it is not required to return to the first step if the condition becomes red. Also, at the completion of the procedure, the procedure does have to be implemented again, since all recovery actions have already been implemented.

Operations Management Procedure 1-7 Page 15 of 31 If a CSF procedure is implemented and it is subsequently determined that the indication is not valid, the crew should:

1.

Consult with the TSC and EOF if they are manned.

2.

Evaluate actions and system realignments performed in the invalid procedure.

3.

Realign systems as needed, address any valid red or orange paths, and return to the procedure and step in effect.

If Critical Safety Function (CSF) Status Trees are implemented during a reactor trip event and a subsequent safety injection occurs, the CSF Trees are still in effect. If either a RED or ORANGE priority on a status tree is evident, the operator must transition to the appropriate FRG as soon as it is validated to address the challenge to the critical safety function before completing any steps in E-O (DW-92-065).

If a valid yellow path is encountered, the operator is expected to scan all of the remaining trees, then if no higher red or orange path is encountered, consider implementation of the corresponding EP. Implementation is by the Operations Shift Managers judgement, based on time, resources available and whether the yellow condition will be cleared by actions in progress or is a warning of a more serious event to occur.

If during the performance of a yellow path procedure, any red or orange condition arises, then the red or orange condition shall be addressed first, and the yellow path procedure suspended.

Operations Management Procedure 1-7 Page 16 of 31 Yellow path procedures are to be performed concurrent with the non-critical safety function EP in effect when the yellow path is implemented. While performing the actions of the yellow path, continuous actions or foldout page items of the non-critical safety function EP in effect are still applicable and shall be monitored by the operator. (DW-95-043)

If a red or orange condition indicates and then clears prior to implementation of the corresponding procedure, the procedure shall j be performed. The CSF procedure is considered to be implemented when the CRS reads the first step to the crew.

The STA shall keep the Operations Shift Manager informed of all off normal CSFs. The Operations Shift Manager shall ensure the crew is updated as appropriate, typically by allocating time during updates for the STA. (SOER 94-1)

F.

Normally, the condition of the CSF Status Trees is continuously displayed by SPDS on the OAC. Control room indications shall be used to validate any off normal alarm and to determine which procedure to implement. Once status tree monitoring is initiated, the STA should periodically monitor the status trees and compare against control board indications to ensure SPDS is functioning properly. Status tree monitoring shall be continuous if an orange or red condition exists.

Otherwise, monitoring frequency shall be every 10 to 15 minutes. (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 the pressurizer safety and reliefvalves, and other connections up to and including the primary isolation valves.

4.1.N.1 Critical Safety Function Status NCS Integrity - RED indicates NCS pressure and temperature conditions which may challenge the Reactor Vessel integrity. Heat Sink - RED indicates the ultimate heat sink function is under extreme challenge. Either of these conditions indicate a potential loss ofthe NCS Barrier.

There is no Loss EAL associated with this item.

If a steamline break occurs that challenges reactor coolant system integrity (PTS),

implementation ofthe functional restoration procedure directs the throttling of CA flow to the S/Gs (creating the heat sink CSF alarm) to mitigate the NCS cooldown.

It would be inappropriate to upgrade the severity ofthe emergency classification based on the directed action to minimize NCS cooling.

The heat sink functional restoration guidance specifically provides a relief from implementing the loss ofheat sink actions, when the operator is responding to the NCS overcooling situation. Thus the loss ofheat sink is considered a controlled operator response, not a loss ofthe heat sink.

If, however, the heat sink is made physically unavailable for some other reason, then the upgrade in emergency classification severity is appropriate.

EOPs are designed to maintain and/or restore a set of CSFs during accident conditions. By monitoring the CSFs instead ofthe individual system component status, the impact of multiple events is inherently addressed. The EOPs contain detailed instructions regarding the monitoring ofthese functions and provides a scheme for classifying the significance of the challenge to the functions. In providing EALs based on these schemes, the emergency classification can flow from the EOP assessment rather than being based on a separate EAL assessment. This is desirable as it reduces ambiguity and reduces the time necessary to classify the event.

D-7 Rev. 09-1 April, 2009

ML12356A448

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