Text

Initial Exam 2011-301 Draft SRO Written Exam
	ML112650104
Person / Time
Site:	Mcguire, McGuire
Issue date:	09/20/2011
From:	; NRC/RGN-II
To:	; Duke Energy Carolinas
References
	Download: ML112650104 (170)
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2011 MNS SRO NRC Examination QUESTION 76 SYSOO6 2.4.20 Emergency Core Cooling System (ECCS)

S006 GENERIC nowledge of the operational implications of EOP warnings, cautions, and notes. (CFR: 41.10/43.5/ 45.13)

Given the following initial conditions on Unit 1:

A LOCA has occurred inside Containment

Safety Injection and DIG Sequencers have been RESET

ECCS aligned for Recirculation mode in accordance with ES-1.3 (Transfer to Cold Leg Recirc)

lB ND pump is aligned to supply Aux Containment Spray

1A ND pump is aligned for Cold Leg Recirc Subsequently, the following occurs:

A Loss of Off-site Power occurs

The BLACKOUT SEQUENCE is complete on IETA and 1ETB

No subsequent operator actions have been taken Prior to restarting the 1 B ND pump, 1 ND-43A (IA ND Hx Outlet to NS Cont Outside lsol) must be closed to prevent (1) 1A ND pump must be restarted after power is restored to the Emergency Busses because (2)

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

A. 1. pump runout

2. the NV pumps ONLY will be running with NO suction source B. 1. pump runout

2. the NV pumps AND NI pumps will be running with NO suction source C. 1. waterhammer

2. the NV pumps ONLY will be running with NO suction source D. 1. waterhammer

2. the NV pumps AND NI pumps will be running with NO suction source Monday, April 18, 2011 Page 225 of 299

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2011 MNSSRO NRC Examination. QUESTION 7&

eneral Discussion ccordance with a NOTE in ES-1.2:

If an ND pump aligned to aux containment spray ever stops with its spray valve open, the associated spray line will void. If this were to occur, Enclosure 4 (ND Pump Restart Requirement If Aux Spray Is Open) will prevent a water hammer and potential pipe failure in the Annulus.

In accordance with a CAUTION in ES-l.2:

If a B/O occurs, NV pumps will auto sequence on without adequate suction. It is critical to start ND pumps to provide NV pump suction.

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part 1 is plausible because the ND spray header will void if the ND pump trips with the Aux Containment Spray valve open. It is plausible for the applicant to conclude that with the ND spray header voided and the Spray Header Isolation open the ND pump could experience a runout condition due to low flow resistance.

Part 2 is correct. -

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

rt 1 is plausible because the ND spray header will void if the ND pump trips with the Aux Containment Spray valve open. It is plausible for ne applicant to conclude that with the ND spray header voided and the Spray Header Isolation open the ND pump could experience a runout condition due to low flow resistance.

Part 2 is plausible if the applicant does not recall that only the NV pumps will auto sequence on when the DGs load the Emergency Busses.

Answer C Discussion CORRECT: See explanation above.

Answer D Discussion -

INCORRECT: See explanation above.

PLAUSIBLE:

Part 1 is plausible because the ND spray header will void if the ND pump trips with the Aux Containment Spray valve open. It is plausible for the applicant to conclude that with the ND spray header voided and the Spray Header Isolation open the ND pump could experience a runout condition due to low flow resistance.

Part 2 is plausible if the applicant does not recall that only the NV pumps will auto sequence on when the DGs load the Emergency Busses.

Basis for meeting the KA The K/A is matched because the applicant must have knowledge of the operational implications of not performing the actions speciti 1

NOTE in ES-l.3.

Basis for Hi Cog LBasis for SRO only fhis question is SRO-only knowledge linked to 10CFR55.43(b)(5) (Assessment and Selection of Procedures) as described in the Clarification Guidance for SRO-only Question Rev 1 (dated 03/11/2010):

he question can NOT be answered solely by knowing systems knowledge. To answer the question the SRO applicant must have knowledge ot the reasons for performing actions specified in a NOTE in the procedure.

2) The question can NOT be answered solely by knowing immediate operator actions. There are NO immediate actions associated with ES- 1.3

(Transfer to Cold Leg Recirc).

Thursday, April 14, 2011 Page 226 of 299

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2011. MNS SRO NRC Examinatkrn QUESTION 76..

The question can NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs.

a knowledge required is the basis for performing actions in a procedure NOTE.

4) The question can NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigate strategy of the procedure.

The knowledge required is the basis for performing actions in a procedure NOTE and as such constitutes detailed procedure knowledge.

5) The question requires the applicant to have detailed knowledge of the basis for performing actions within the procedure. As such this constitutes SRO-level knowledge.

Job Level Cognitive Level QuestionType Question Source SRO Memory NEW Development References Student References Provided

References:

ES- 1.3 (Transfer to Cold Leg Recirc)

SYSOO6 2.4.20 Emergency Core Cooling System (ECCS)

SYSOO6 GENERIC Knowledge of the operational implications of EOP warnings, cautions, and notes. (CFR: 41.10/43.5 / 45.13) 401-9 Comments: Remarks!Status Thursday, April 14, 2011 Page 227 of 299

Question 76

References:

From ES-I .3 (Transfer to Cold Leg Recirc):

MNS TRANSFER TO flfli fl I Fi RFC1RC PAcF NO EPI1IN5000IES-i 3 14 of 43 Rev. 26 UNIT 1 ACTIOfl/ECTED RZLiPONE P.EONE NC O2TAZND

9. (Continued)

e. EstabJh ND aix containment spray trcm one train that is in Cold Leg Recirc mode as follows:

1) CLOSE Nl-73A(IANDtoA&

B Cold Leqs Cont Outside Isol).

2) OPFN INS-43A (IA Nfl Hx Outlet toNS Cont Outside Isol).

OR

B Train:

1) CLOSE INi 78B (18 NDto C &

D Cold Legs Corn Ou!side Isol).

2) OPEN 1NS38B (18 ND [lx Outlet toNS ont Outside lsol).

AU]1 If an ND pump aligned to aux containment spray ever stops with its spray vahe opens the assocIated spray line will void. if this were to occur. Enclosure 4 (ND Pump Restart Requirement If ux Spray Is Open) will prevent a water hammer and potential pipe failure in the Annulus.

f. WI lEN time allow5, TI l[N place INrO tag onNDpunpccntro switch R hnciosure 4 (ND Pump Restart Requirement If Aux Spray Is Open).

CAUTION Eaiiuie to stop ND iux cuiituiiiuiiit spittywlien iequiied ni.ty iesuit in a neqative containnient pressure.

g. WhEN containrnn: prure ie than 1 PS1G, THEN stop ND aux containment spray ji Enclo3ure 5 (Securing ND Aux Containment Spray.

MNS TRANSFER TO COLD LEG REGRC PAGE NO.

EP/1/N5OOO,ES-I.3 16 of 43 UNIT 1 Rev. 26 ACIION/EXCTED P.E3ONSE EPONsE OT OBTA:NED CUTlQN If a IO occurs, NV pumps will auto sequence on wlthou adequate suction, ft is critical to start ND pumps to provide NV pump suction

10. IF AT ANY TIME a BIO signal occurs, THEN restart S)i equipment previously on.

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2011 MNSSRQ NRC Examination QUESTION 77 SYS1O3 2.4.41 Containment System S103 GENERIC nowledge of the emergency action level thresholds and classifications. (CFR: 41.10 / 43.5 I 45.11)

Given the following conditions:

Unit was operating at 100% RTP

45 minutes ago a large break LOCA occurred on Unit 1

Containment pressure indicates 9 PSIG

Containment hydrogen concentration is 1%

CETs indicate 1100°F

RVLIS lower range level indicates 40%

IEMF- 51A indicates 165 R/hr

Subcooling margin indicates -35°F

All SIG NR levels are off scale low with no CA flow indicated Which ONE (1) of the following is the correct classification and associated EAL Number for this event?

REFERENCE PROVIDED A. Site Area Emergency based on EAL # 4.1 .S.2 B. General Emergency based on EAL # 4.1 .G.2 C. Site Area Emergency based on EAL # 4.1 .S.1 D. General Emergency based on EAL #4.1 .G.1 Thursday, April 21, 2011 Page 228 of 299

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- 21I MNS SRO NRC Examination QUESTION 77 eneraI Discussion The candidate has been given a set of conditions associated with a large break LOCA and given a copy of RP-000, asked to classH, the event.

The conditions given result in the following determinations per RP-000 Enc 4.1 (Fission Product Barrier Matrix):

(Containment Barrier)---0 point Close to Potential loss due to Containment Rad. Monitor EMF-5 lA or 5 lB reading

- time since shutdown (45 mm)> 170 RJhr @ 0.5 2 hr. (Reading 165 R/hr)

(NCS Barrier)--- 5 Points--- Loss due to GREATER THAN available makeup capacity as indicated by a loss of NCS Subcooling (Subcooling margin indicates -35 deg)

(Fuel Clad Barrier)--- 5 Points--- Loss due to Containment radiation monitor EMF 5 1A or 5 lB reading >43 RJhr 45 minutes since shutdown.

The result of the evaluation above is a total of 10 points (4.1 .S. 1) (Loss of Both Nuclear Coolant System and Fuel Clad) and the corresponding classification of Site Area Emergency.

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible if the candidate failed to recognize that the conditions given represent a Loss of NCS Barrier. A valid Heat Sink Red Path is represented in the Stem which taken alone would constitute a only 4 Points for the NCS Barrier (Potential Loss) in the Matrix and a resulting in a total of 9 points and a classification of SAE with an EAL of 4.l.S.2.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

is answer is plausible if the applicant incorectly applies the Containment radiation indication and determines that a potential loss of tainment applies. This designation is time dependent and it is possible to miss read this scale when evaluating a given indication.

Mnswer C Discussion CORRECT: See explanation above.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible if the candidate incorrectly determines that a loss of the containment barrier exists but correctly determines the loss of the other two barriers. This would result in a total of 13 points, a General Emergency classification with an EAL# 4.1.0.1.

Basis for meeting the KA KA is matched because most of the evaluations required in c1assif,ing this event require to application of the Containment Radiation levels given in the stem of the question. Two of the three fission product barriers require the applicant to evaluate containment radiation levels and involve addressing concerns with the Containment System).

Basis for Hi Cog question represents a higher cognitive level of application because it involves a multi-part mental process of assembling different combinations of given information to select a correct classification.

Basis for SRO only This question is linked to 10CFR55.43 (b)(7) (Emergency Classification) Per the guidance in 10CFR55.43 and per the MNS objective referenced for this question, assessing plant conditions and determining the proper classification of emergency is considered SRO level.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension MODIFIED Bank 3104 (2009 NRC Exam Q89 MODIFIED) iievelopment References Student References Provided OP-MC-EP-EMP Obj. 10 RP/0/A15700/000 RP/0/A15700/000 (Rev 14)

Enc 4.1 Wednesday, April 20, 2011 Page 229 of 299

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20L11 MNS SRO NRC Exaniination QUESTION 77 EP/l/A15000/F-0 (Rev 4)

YS 103 2.4.4 1 Containment System SYS1O3 GENERIC Knowledge of the emergency action level thresholds and classifications. (CFR: 41.10 I 43.5 I 45.11) 401-9 Comments: Remarks!Status Thursday, April 14, 2011 Page 230 of 299

Question 77

References:

From RPI000 (Fission Product Barrier Matrix) Pg 5 of 5 Evaluation of the Containment Barrier Enclosure 4.1 RP/O/.A15700/GOO Fission Product harrier Matit Page 5 of 5 41.( CONTAUENTBARRR 41.N NCSBkRR1ER 4LF FUELL4DBARIER FOTEtTIAL LOSS - LOSS - ?OTUtTLAL LOSS - LOSS. ?OThNTLi LOSS - LOSS -

(1 Point) (3 Pinrn) (4 Pointc) (5 P.imte) (4 Points) (5 Pothec)

E Sienifirani Radioactive Jnvenrorr In S. Em reemrv Coordinator/EOF Directot Contaunment Judgement Containment Rid Not applicable. Any condition, including inability to niotmitor Monitor EMF51A the battier, that in the opinion of the or 51B Etnerrency CoordinrtorIEOF Director Reading te time indicates LOSS or POTENTLU LOSS of since iimutdowsm: the NCS battier.

S9OPfbr d O-O.Shr l7ORihra O5.2hr END 2-4 hr 9ORibr i 4-8 hr 53 PJhr it 3 hr.

6. Einerencv Coordinator /EOF Dlrecroi Judgeinent Any condition. including inibihty to monitor the battier, that in the opinion of the Emergency CoordinatorfEOP Director indicates LOSS orPOTENTLU, LOSS of the containment bonier.

END

From RP/000 (Fission Product Barrier Matrix) Pg 3 of 5 Evaluation of the NCS Barrier Enclosure 4,1 9 A. 5700 000 Fission Product Barrier Matrix Page 3 of 5 4.1.C CONTAINMENT BARRIER 4.LN NCS BARRIER 4J.F FUEL CLAD BARRIER POTENTIAL LOSS - LOSS - POTENTIAL LOSS - LOSS - POTENTIAL LOSS - LOSS -

(1 Point) (3 Points) (4 Points) (5 Points) (4 Points) (5 Points)

1. Cii&ai Safets- Function Status . Ciitical Safers Function Status 1. Ciitical Safety Function Status Containment-RED.

Not applicable.

NCS Integrity-

Not applicable.

Core Cooling-

Core Cooling-RED.

RED. ORANGE.

Core Cooling -

RED Patio is

Heat Sink-RED.

Heat Sink-RED.

indicated for >15 minutes.

2. Containment Conditions 2. NCS Leak Rate 2. Primary Coolant Activity Level

Containment

Rapid unexplained

Unisolable leak

GREATER THAN

Not applicable.

Coolant Acthaty Pressure 15- decrease in exceeding the available nialceup GREATER THAN PSIG. containment capacity of one capacity as 300 iiCi/cc Dose pressure following charging pump in indicated by a loss Equivalent Iodine

112 concentration initial increase, the normal of NCS subcooling. (DEl) 1-131.

9%. charging mode

Containment with letdown

Containment pressure or sump isolated.

pressure greater than level response not 3 psig with less than consistent with one Bill train ofNS LOCA conditions.

and a VX-CARF operating.

CONTINUED CONTINUED CONTINUED

From RPI000 (Fission Product Barrier Matrix) Pg 4 of 5 Evaluation of the Fuel Clad Barrier Eucloure -1.1 RpIO/vs7oofooo Fission PrGducr iarrier 1atii Page 4 of S 4.i.( CONTAINMENT 11flt. 4.l.Y NC$ARE.IZK 4.1. lUEL CLAD AREIEK

.L LOSS 1

POTENTL - LOSS - POTENTIAL LOSS - LOSS - POTENTIAL LOSS - LOSS -

(1 Pouoo> (3 Pomou) (4 Points) ( Poinu) 4 Poiofu) (5 Poeuta)

3. C t.ini I liie Vat St ,to. .,.ftCr 3.50 Tobo Roploro . Cement R Ihtio efooito.ine Containment Lolatien Actuation Not applicab:e.

Containment Pajama-v-to-

Indication that a Not appLcablo. Containment isolation is Seecudary leak SG i Ruptured and radiation monitur incomplete md a ratO eureeth the han a Non-Lolabe EMF5IAorS1B release path fium capanity of one .eccndaayliue Readingat time contammentexists charring pump in faulL imce chctdown the atunual charpng mode

Indicationthat a 3-0.5 Ian 93EJhr with letdewa SO i: naptared and 0.S-2hra43Rhr auolatod. a paoloaa;od rclcauo -4hra 3IRJSr of c,ntaminated 4-8hz 22ihr

.-.eenncLuay rnnl,fl Rh,- I5thr iz occuni fiom the affeetd SG to the envircument.

4. SG Secendarv Side Re1eao With Primirv-to. I, Containment Radiation Sfonitorin L Emergenc Coordinator EOf Director Secondary Lerkage Judgomeut Not Not j.yli..b1o.

N.L .aplirab1o. Any nodition. doeIaee in.eUliEy to nznt ecomlarv ime to the banjo:, that n the opimot of the the OVU0000Qrd Emareenav CoordinatoriEOF Direcior with primary-to indicates LOSS or POTENTIAL LOSS of econ+/-uy leahage the fuel clad banier.

GREATER THAN Tech Spec allowable. END UtX IL t.fl) (O. ilMhi)

From RPI000 (Fission Product Barrier Matrix) Pg 1 of 5 Evaluation of the EAL #

Use E4.Ls to dctcraiine Fission Product Barrier status (Intact, Potental Loss, or Loss). Add points for all 3 barriers. Classify according to the trble en page 2 of 5 of this enclosure.

Note I: This table is culy app[icable in Modes 1-4.

Sole : Also, an event (ormiiltiple events) cauld occur which results in the coichison that exceeding the Inst cc Poteuia Loss thresholds is imminent (i.e.. within 1 3 hotiri). Inthis imminent loin cituation, sue judgenmit and classify as if the thesholdc are exceeded.

Note 3: When dererninicgFissianProducs Barsier starts, the Fiie: Clad Barrier shotldte corniclerell D be lost orptental:ylos: if the coiditioas for the fuel Clad Barrier loss or poteirial less EALs were met previously (validated and stasrained) thiring the eeir. even if the coadtticns do not currently ealst.

Note : Crirical Safety Function (CSF) indicaticns are not meant to include niisient alarni cocdioas which may appear dvring the start-up of engineered safeguardi equipment. A CSF condition :s satisfied when the alarmed state is valid and sustained. The SEA shouldbe consul:ed to iiuifaiay CS lsrebeeu salidatedpiiui Lu that CS? emg used as Lli basis to ci sfyaineaueagcuuy. (1 xsnipk. If ECA-O.O. Loss of All AC Power, is unpleineaxe5 with an appropriate CS? alarm coedillon valid and sustained, that CS? should be used as the >asis to classtaii emergency raor ils any ttnctaon restoration ricedure bsn nuplemented witlilit the conThies of tUA-O.O.

EAL Unizujal Event EAL Alert EAT s Sit, Area Enieieac EA.L Geaeral Eireny 4.1.13.1 ?otontialLocof .l.l.A.l IorcORlotsntialLos: 4.1.5.1 1osORPotentiallou 4.:.G.l Lou oIA.1 Three Barneic Coatshnent of ofllcth Icuc:en Coohnt Syuer.s lcuc:eas Ceolaut Svitem AND ud Clad 41132 sa of Connusuest 4.l...2 LoOaounm1Loz 4.1.5.2 102; 4..G.2 Loofnyliro3atneic 01 AND r1Cld Pi.,l rf the ThSd cbioo of Both Ceoht &vtcea

.4N11 nod

-L 14 ?tent1T,n 41 55 Clntlismest AND to OR Potential Loz Le Olt Fotnrial Lo,: of Are Cthor 3anier of Any Other 3ainer

Parent Question (2009 NRC Exam Q89):

Examination Outline Cross- Level RO SRO reference: x Tier#

Final Group # 2 KIA# WE16EG2.4.41 Importance Rating 2.9 4.6 High containment Radiation: (Generic) Knowledge of the emergency action level thresholds and classifications Proposed Question: SRO 89 1 Pt Given the following conditions:

45 minutes ago a large break LOCA occurred on Unit I

Containment pressure indicates 9 PSIG

Containment hydrogen concentration is 1%

CETs indicate temperatures of 1100°F

RVLIS lower range level indicates 40%

EMF 51A indicates 185 R/hr

Subcooling margin indicates -35°F

All SIG NR level are off scale low with no CA flow indicated Which ONE (1) of the following is the correct classification and associated EAL Number for this event?

REFERENCES PROVIDED A. Site Area Emergency based on EAL # 4.1.S.2 B. General Emergency based on EAL # 4.1 .G.2 C. Site Area Emergency based on EAL # 4.1 .S.1 D. General Emergency based on EAL # 4.1 .G.1 Proposed Answer: B

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2011 MNS SRO NRC Examination QUESTION 78 SYSO59 A2. 12 Main Feedwater (MFW) System bility to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to

,orrect, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 /45.3 / 45.13)

Failure of feedwater regulating valves Given the following initial conditions on Unit 1:

The unit was operating at 100% RTP

The TD CA pump is out-of-service for maintenance

A DCS failure results in ALL CF Control valves failing closed

As a result, both CF pumps trip on high discharge pressure

Both MD CA pumps fail to automatically start and cannot be started manually

FR-H. 1 (Response to Loss of Secondary Heat Sink) has been implemented In accordance with the FR-H.1 Background document, NC system Feed and Bleed must be initiated within (1) minutes of meeting the criteria in FR-H.1.

The basis for stopping all NC pumps prior to initiating Feed and Bleed is to (2)

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

A. 1.4

2. prevent NC pump impeller damage due to low pressure operation B. 1.8

2. prevent NC pump impeller damage due to low pressure operation C. 1.4

2. conserve secondary inventory by reducing NC system heat input D. 1.8

2. conserve secondary inventory by reducing NC system heat input Thursday, April 14, 2011 Page 231 of 299

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2011 MNS SRO NRC Examination QUESTION 78 7 General Discussion In accordance with FR-H. 1 Background Document:

Feed and bleed should be initiated as quickly as possible after meeting criteria in FR-H. 1, but must be initiated within 4 minutes of meeting criteria. Note that feed and bleed may need to be initiated within 8 minutes after reactor trips on low-low SG level, since the criteria will be quickly met. The NV recirc valve must be closed within 5 minutes of initiating feed and bleed. Although this action is NOT required for a design basis event, this item is included due to its PR.A significance during a loss of secondary heat sink event.

In accordance with FR-H. I Background document:

NC pump operation results in heat addition to the water in the NC system. By tripping the NC pumps, the effectiveness of the remaining water inventory in the SIGs is extended, which extends the time at which the operator action to initiate feed and bleed must occur. This extension of time is additional time for the operator to restore feedwater flow to the SIGs.

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part 1 is correct.

Part 2 is plausible because the applicant may conclude that opening the PORVs during Feed and Bleed would result in a loss of subcooling and potential cavitation of the NC pumps.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

rt 1 is plausible because the basis document states that feed and bleed may need to be initiated within 8 minutes after the reactor trips on low ow SG level.

Part 2 is plausible because the applicant may conclude that opening the PORVs during Feed and Bleed would result in a loss of subcooling and potential cavitation of the NC pumps.

Answer C Discussion CORRECT: See explanation above.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part 1 is plausible because the basis document states that feed and bleed may need to be initiated within 8 minutes after the reactor trips on low-low SG level.

Part 2 is correct.

Basis for meeting the KA The KIA is matched because the SRO applicant must use procedures (FR-H. 1) to mitigate the consequences of a loss of feedwater which is tiated by a failure of the feedwater regulating valves.

Basis for Hi Cog Basis for SRO only This question is SRO-only knowledge linked to 10CFR55.43(b)(5) (Assessment and Selection of Procedures) as described in the Clarification Guidance for SRO-only Question Rev I (dated 03/11/2010):

The question can NOT be answered solely by knowing systems knowledge. To answer the question the SRO applicant must have detailed iwledge of both procedure steps and the procedure background document.

2) The question can NOT be answered solely by knowing immediate operator actions. There are NO immediate actions associated with FR- H.l (Response to Loss of Secondary Heat Sink).

Thursday, April 14, 2011 Page 232 of 299

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2O1IMNS SRO NRC Examination QUESTIQN 78

3) The question can NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs.

-ie knowledge required is detailed procedure step knowledge and background document information.

1

4) The question can NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigate strategy of the procedure.

5) The question requires the applicant to have detailed knowledge of the procedure background document information. As such this constitutes SRO-level knowledge.

Job Level Cognitive Level QuestionType Question Source SRO Memory NEW Development References Student References Provided

References:

FR-H. 1 (Response to Loss of Secondary Heat Sink)

FR-H. 1 Background Document Learning Objectives: OP-MC-EP-FRH Objectives 4 & 7 SYSO59 A2.12 Main Feedwater (MFW) System 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: (CFR: 41.5 / 43.5 / 45.3 / 45.13)

Failure of feedwater regulating valves 401-9 Comments: RemarkslStatus Thursday, April 14, 2011 Page 233 of 299

Question 78

References:

From FR-H.1 Background Document:

8.0 TIME CRITICAL TASKS 8.1 Operator Action to initiate Feed and Bleed once criteria met:

Expectation: Feed and bleed should be initiated as quickly as possible after meeting criteria in FR-H.1, but must be initiated within 4 minutes of meeting criteria. Note that feed and bleed may need to be initiated within 8 minutes after reactor trips on low-low SG level, since the criteria will be quickly met. The NV recirc valve must be closed within 5 minutes of initiating feed and bleed. Although this action is NOT required for a design basis event, this item is included due to its PRA significance during a loss of secondary heat sink event.

From FR-H.1 Background Document:

DUKE ENERGY MCGUIRE OPERATIONS TRAINING FR-H. I Loss of Secondary Heat Sink STEP 8 Check steam dumps:

PURPOSE: Place steam dumps in pressure mode of control before stopping all NC pumps.

BASIS: Provide better control of steam dumps under natural circ conditions. If no NC pump is running, then the NC system average temperature will be higher than the no-load value as natural circulation conditions are established. However, if the steam dump system is working properly, the cold leg temperatures will stabilize at the no-load value.

STEP 9 Stop all NC pumps.

PURPOSE: To stop NC pumps in order to extend the time to restore feed flow to the SIG5.

BASIS: NC pump operation results in heat addition to the water in the NC system.

By tripping the NC pumps, the effectiveness of the remaining water inventory in the S1Gs is extended, which extends the time at which the operator action to initiate feed and bleed must occur. This extension of time is additional time for the operator to restore feedwater flow to the S1Gs.

OP.MC.EP.FRH FOR TRAiNING PURPOSES ONLY REV. 13 Page 31 of 161

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2O1LMNS SRO NRC.Examination QUESTION 79 -

SYSO64 A2.05 Emergency Diesel Generator (ED/G) System

\bility to (a) predict the impacts of the following malfunctions or operations on the ED/G system; and (b) based on those predictions, use iocedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 I 45.3 / 45.13)

Loading the ED/G Given the following conditions:

Both units are at 100% RTP

Testing of Sequencer B on Unit I is in progress

Annunciator lAD-I I I El (SEQ B IN TEST) is illuminated Which ONE (1) of the following describes the current operability status of the lB DG AND SI loading response if an automatic SI actuation occurs before any corrective actions are taken?

A. The lB DG is OPERABLE AND SI loads on both trains will start automatically.

B. The I B DG is OPERABLE AND Train A SI loads will start automatically, but Train B SI loads must be started manually.

C. The 1 B DG is INOPERABLE AND SI loads on both trains must be started manually.

D. The I B DG is INOPERABLE AND Train A SI loads will start automatically, but Train B SI loads must be started manually.

Thursday, April 14, 2011 Page 234 of 299

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2011 MNS SRO NRC Examination QUESTION 79 General Discussion In accordance with TS 3.8.1 Basis:

Each DG must be capable of starting, accelerating to rated speed and voltage, and connecting to its respective ESF bus on detection of bus

undervoltage. This will be accomplished within 11 seconds. Each DO must also be capable of accepting required loads within the assumed loading sequence intervals, and continue to operate until offsite power can be restored to the ESF buses. These capabilities are required to be met from a variety of initial conditions such as DO in standby with the engine hot and DO in standby with the engine at ambient conditions.

Additional DO capabilities must be demonstrated to meet required Surveillance, e.g., capability of the DG to revert to standby status on an ECCS signal while operating in parallel test mode.

Proper sequencing of loads is a function of Sequencer OPERABILITY. Proper load shedding is a function of DO OPERABILITY. Proper tripping of non-essential loads is a function of AC Bus OPERABILITY (Condition A of Technical Specification 3.8.9).

So, Sequencer OPERABILITY and DO OPERABILITY are two separate issues. The Sequencer can be INOPERABLE from the standpoint that it is unable to automatically sequence loads onto the Emergency Bus and the DG can still be OPERABLE provided it can automatically start on a sensed ESF bus undervoltage and is capable of assuming all ESF loads.

With the Sequencer in TEST, the DG is still considered operable. If a valid SI or BO signal is received, the Sequencer will be automatically removed from TEST and the DG will automatically load via the SI or BO sequence.

Answer A Discussion CORRECT: See explanation above.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

st part is correct.

Second part is plausible since the Train B Sequencer is declared INOPERABLE during sequencer testing.

Answer C Discussion iCORRECT: See explanation above.

PLAUSIBLE:

The first part is plausible if the applicant concludes that the DO is INOPERABLE because the sequencer is INOPERABLE. This is plausible because there are times when Sequencer inoperability does affect DO operability. For example, if the Sequencer was de-energized (i.e. SEQ B LOSS OF CONTROL POWER annunciator) not only would the Sequencer be INOPERABLE but its associated DO would be inoperable as well. This is because with the Sequencer de-energized, the DO would not be capable of automatically starting and connecting to its respective Emergency Bus on a sensed undervoltage condition (as the start signal is generated by the Sequencer).

The second part of the answer is plausible if the applicant concludes that both sequencing of both trains is affected by testing of one Sequencei Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

The first part is plausible if the applicant concludes that the DO is INOPERABLE because the sequencer is INOPERABLE. This is plausible because there are times when Sequencer inoperability does affect DO operability. For example, if the Sequencer was de-energized (i.e. SEQ B LOSS OF CONTROL POWER annunciator) not only would the Sequencer be INOPERABLE but its associated DO would be inoperable as well. This is because with the Sequencer de-energized, the DO would not be capable of automatically starting and connecting to its respective Emergency Bus on a sensed undervoltage condition (as the start signal is generated by the Sequencer).

Second part is plausible since the Train B Sequencer is declared INOPERABLE during sequencer testing.

sis for meeting the KA KA is matched because the applicant must understand the difference between Sequencer OPERABILITY (which is related to automatic loading of the DO) and DO OPERABILTY (related to DO start on UV and its ability to assume ESF loads) as defined in TS 3.8.1 basis. By understanding the difference between the two and the exact impact of the Sequencer being in test, the applicant is able to predict the impact and use procedures to control the situation (i.e. use TS basis to determine DO operability).

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2011 MNS SRO NRC Examination QUESTION 79 7 Basis for Hi Cog is is a higher cognitive level question because it requires more than one mental step. First it requires the applicant to analyze plant conditions

.d determine the effect of the Sequencer being in test on both the Sequencer and the DG. Then the applicant must recall from memory the basis for TS 3.8.1 (AC Sources Operating) related to DG operability.

Basis for SRO only This question is SRO-only knowledge linked to 1 OCFR55.43(b)(2) (Tech Specs) as described in the Clarification Guidance for SRO-only Question Rev 1 (dated 03/11/2010):

1) The question can NOT be answered solely by knowing 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months TS/TRM action statements. It requires knowledge of TS basis.

2) The question can NOT be answered solely by knowing the LCO/TRM information listed above-the-line. Again, TS basis knowledge.

3) The question can NOT be answered solely by knowing the TS Safety Limits. TS 3.8.1 (AC Sources Operating)-

4) The question requires the applicant to have detailed knowledge of the TS basis to analyze TS requirements (i.e. DG OPERABILITY). As such this constitutes SRO-level knowledge.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Development References Student References Provided

References:

T.S. 3.8.1 (AC Sources Operating) Basis Learning Objectives: OP-MC-DG-EQB Objectives 5 & 6 S064 A2.05 Emergency Diesel Generator (EDIG) System bility to (a) predict the impacts of the following malfunctions or operations on the ED/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.13)

Loading the ED/G 401-9 Comments: RemarkslStatus Thursday, April 14, 2011 Page 236 of 299

Question 79

References:

From Lesson Plan OP-MC-DG-EQB:

3.0 SYSTEM OPERATION This section will describe Sequencer operation for various plant events I conditions.

This information supports the Operator Fundamentals:

Control: Recognize automatic actions that dont occur and take manual action accordingly.

Knowledge: Integrated plant knowledge; understanding system and component design.

Objective # 10 3.1 Limits and Precautions PT121A14350104A, DIG 2A Load Sequencer Test Manual control of A(B) Train 4160V switchgear is unavailable to Control Room while sequencer is in test mode Basis: This informs the Control Room Operator that while in the test mode he cannot start or stop equipment on the affected bus.

Surveillance testing of DGLSA(B) requires 1A(B) DIG Auto Start.

Basis: The Auto Start Signal will bypass the Manual Mode Circuit thus preventing any Manual Mode Actuation Signal from tripping the Diesel. This also allows those Automatic Mode signals to trip the Diesel if their limits are exceeded.

Circuits inside DGCPI(2)A(B) (Diesel Generator Control Panels 1A, 1B, 2A, 2B) are energized at 120 VAC and 125 VDC.

Basis: This information is used to warn the operator or technician of the dangers inside the control panel and that care must be taken when working near energized circuits.

3.9 Sequencer Testing Initial conditions: Unit on line in a normal electrical line-up.

Test initiation and indication.

Momentarily rotating the key operated test actuate switch to the right, the test position, energizes the 7 test relays TSA1 (TSB1) through TSA7 (TSB7).

Relay TSA5 (TSB5) seals around the test actuate key switch contacts and maintains all the test relays energized. The test actuate light will also be on.

Upon energization, the 7 test relays perform the following functions:

o Manipulates the 4160 volt bus breaker controls which enables the breakers to maintain their present positions while testing the sequencer. Contacts in breaker close and trip circuits disable breaker operation. Indicating lights on the sequencer panel indicate when breaker operation would occur if the sequencer were actually controlling the various loads as the testing progresses.

Relay TSA5 (TSB5) closes a contact in the defeat test relay circuit, DTSA (DTSB), to permit sequencer reset if an actual SI or blackout occurs during testing.

Opens or closes contacts across test pushbuttons (P/B) in the following circuits:

Test P/B 1 Sequence timers ST1A (ST1 B) to ST1 1A (ST1 1 B)

Test P/B 2 ESG aux. relays 1 ESGA)(2 (1 ESGBX2) and 1 ESGAX3 (1 ESGBX3)

Test P/B 3 Accelerated sequence relays AA1 (AB1) and AA2 (AB2)

Test P/B 4 Bypasses the normal feeder breaker permissive contacts in the load shed relay circuit when the P/B is depressed to TEST Test P/B 5 Blackout logic relays LRAI, LRA2, LRA3, and LRA6 (LRB1, LRB2, LRB3, and LRB6)

Test P/B 6 Logic relays LRA4 (LRB4) and LRA5 (LRB5)

Test P/B 7 Bypasses the normal feeder breaker permissive contacts in the reset relay circuit, TRA3 (TRB3), when the P/B is depressed to TEST Blackout signals can now be initiated by depressing test pushbuttons in the UV relay circuits. In this manner, UV logic can be demonstrated to actuate sequencer as in an actual blackout condition.

The safety injection signal is generated by closing input contacts to the sequencer from the SSPS relay K608.

Various sequence features such as accelerated sequence, committed time sequence, etc., are tested by depressing test pushbuttons as necessary during testing.

Following completion of the tests, the sequencer is reset by depressing either reset pushbutton (control room or sequencer panel) and locking the test actuate switch in the left position.

Note: Ensure the test SI signal from the SSPS has cleared before resetting the sequencer. The reset of the sequencer must be done correctly following a test or loads may be inadvertently started. Refer to LER 369/88-09.

Actual accident occurrence while testing.

Initial conditions: Unit on line, test relays energized with blackout or SI loading sequence test in progress.

Either an SI or a loss of voltage condition will energize relay SAA (SAB), which sends a start signal to the diesel generator. The next devices energized will be timer TTA (TTB) (0.5 second) and the latch coil of the defeat test relay, DTSA (DTSB), via a contact from the SSPS (relay K609), or a contact from relay 127AX (127BX). Upon latching, the following events take place:

The 1ESGAX (IESGBX) relays are reset (if energized).

o Relay BOA (BOB) is reset (if energized) through a TRA3 (TRB3) contact.

o The reset relays TRA3 (TRB3) and TRA3X (TRB3X) are energized. This resets the sequencer by de-energizing the ST, ATA (ATB), LTIA (LTIB), and LT2A (LT2B) timers as well as the SAA (SAB), RA (RB), LS, and LR relays.

The DTSA (DTSB) relay is unlatched by a TRA3X (TRB3X) contact. Upon unlatching, the DTSA (DTSB) relay de-energizes the reset relays TRA3 (TRB3) and TRA3X (TRB3X) which maintain a 1 second off-delay and then return to a normal de-energized condition.

o After a 0.5 second time delay, the TTA (TTB) contacts open and de-energize all seven of the test relays. This removes test blocking from the 4160 volt breakers and from the sequencer logic controls. The 0.5 second time delay ensures that the logic controls have reset before test blocking is removed.

Sequencer reset while testing Because the reset relay RRA (RRB) contact in the ESGAX (ESGBX) relays circuit is bypassed by the SI signal from the SSPS (relay K608), the sequencer cannot be manually reset with the SI signal present. This feature accomplishes the following:

Prevents manually resetting the sequencer during an actual safety injection event.

Prevents resetting the sequencer completely while in test and initiating safety injection loading with an SI test signal.

Allows logic reset in test without necessitating an SI test signal removal prior to each sequencer segment test.

From TS 3.8.1 (AC Sources Operating) Basis:

AC Sources-Operating B 3.8:

BASES CC) (ccntimid) switchgear from eitner urits 6.9 kV slstem. A key inteiock scherre is provided to precluc the aossibilitv of connectng the two units together at efher the 6.9or4.16 kV level.

Each tr2in of thc 4.16 kV Excntk Auxiliary Powr Syctorn i aba provid1 with srnrlfc nd irdpendcnt rrrgcnry di.l ge ratcx to supply :he Chss 1 E loads required to safely shut dcrnn the unit :ollowhg a design basis acident Each DG must be capable of starting, acceleratinq to rated speed and vultaqe, and o riediriq to its lespeclive ESF bus on detetiun ul bus undcrioltago. Thic wi I b accomplithod wthin 11 cocondc. Each DG must also be capable of accoptinq required loads within the assumed loadng sequence intevs. and continue to operate urtil offte power can be restored to the ESF buses. These capabilities are required to be met from a vanety of initial conditions sLch as DG in stanoby wilh he engine hot and L)C3in standby witli the enqneatantient conditons.

Aldioiial DG capbil.ies rriusi be deinorislmiLed to rriee[ equired Surveillance, e.g., capability of the DC to revert to slandby status on an ECCS signal wfile operatin9 in parallel test mode.

Proper sequencinq of loads is a funcor of Sequencer OFERABILITY.

Proper load sheddiKj is a ft.nction of DO OPERABILITY. Proper trippinç of non-essential loads is s function of AC Dos OP[RADILITV (Condition A of Tochnical Spccification 3.8.9).

The AC sources in one train must be seDarate and independent (tc the extent possible) of the AC sources in the other train. For the DOs, separation and independence are compete.

Both nomial and emergency power must be OPERABLE for shared component to be OPERABLE. If normal or emergency pcrner supplying a shared component becomes inoperable then the Required Actions of the affected shared component LCD must be entered irependently for each unit that is in the MUUh ot apphcability ot the shared component LCU.

The sharel conponent LCO5 are:

3.7.7 Juclear Service Water System (l.ISWS),

37 9 Control Poorn Ar Vntilntinn S/strn (CRAVS),

3.7. 0- Control Room Area Chilled V/ater System (CRtCWS), and 3.7:1 Auxiliary Bilcing Fitered Ventilation Exhaust System (ABFVES.

McGuire Urits 1 and 2 B 3.8.1-5 Revision ND. lii

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2011 MNS SRO NRC Exminafion QUESTION 80 SYSO73 A2.02 Process Radiation Monitoring (PRM) System

\bility to (a) predict the impacts of the following malfunctions or operations on the PRM system; and (b) based on those predictions, use ocedures to cor- rect, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 I 43.5 / 45.3 / 45.13)

Detector failure Given the following conditions:

Unit 1 is releasing the Ventilation Unit Condensate Drain Tank (VUCDT) to the RC Discharge using the Continuous Release Method

Shortly after the release was initiated, 1 EMF-44 (Ventilation Unit Condensate Drain Tank) count rate indication fails to a reading of less than background Which ONE (1) of the following statements correctly describes the effect on this LWR (Liquid Waste Release) in accordance with 0P111A165001001 A (Ventilation Unit Condensate Drain Tank Operation)?

A. Continue the release. Document that OEMF-49 (Waste Liquid Disc) is both OPERABLE and monitoring release flow path.

B. Continue the release. Request an updated LWR release and update the OP paperwork.

C. Stop the release. Initiate a Batch Release with updated LWR and OP paperwork.

D. Stop the release. Notify IAE to restore 1EMF-44 to OPERABLE prior to any VUCDT release.

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2O11 MNS SRQ NRC Examination QUESTION 80 General Discussion In accordance with OP/l/A16500/00l A (Ventilation Unit Condensate Drain Tank Operation):

IF any component required by SLC (1EMF44, VUCDT flow totalizer or composite sampler) is inoperable, release shall be made using batch release method.

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible since EMF-49 is a liquid waste release monitor which will close several of the same valves as EMF-44 to isolate a release.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible if the applicant confuses a Continuous Release with a Batch Release and does not recall that EMF-44 is required to be operable for a Continuous Release.

Answer C Discussion CORRECT: See explanation above.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

us answer is plausible if the applicant does not recall that a Batch Release can still be performed with EMF-44 inoperable.

basis for meeting the KA The K/A is matched because an EMF process radiation monitor detector has failed and the applicant must know the procedural requirements for performing both Continuous and Batch Releases to determine the correct answer.

Basis for Hi Cog This is a higher cognitive level question because it requires more than one mental step. First the SRO applicant must analyze the given conditions to determine what type of release is in progress and the status of the system. The applicant must then recall from memory the procedural requirements for performing Continuous and Batch releases and apply those requirements to the given conditions.

Basis for SRO only This question is SRO-only knowledge linked to 10CFR55.43(b)(5) (Assessment and Selection of Procedures) as described in the Clarification Guidance for SRO-only Question Rev 1 (dated 03/11/2010):

1) The question can NOT be answered solely by knowing systems knowledge. To answer the question the SRO applicant must have knowledge of the different procedural requirements contained the OP relative to Continuous and Batch releases.

2) The question can NOT be answered solely by knowing immediate operator actions. There are NO immediate actions associated OP/1/A16500/00I A (Ventilation Unit Condensate Drain Tank Operation).

3) The question can NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs.

The knowledge required is the specific procedural requirements contained the OP relative to Continuous and Batch releases.

4) The question can NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigate strategy of the procedure.

It requires detailed procedure knowledge.

5) The question requires the applicant to analyze plant conditions and determine a correct course of action based on that analysis. As such the constitutes SRO-level knowledge.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension BANK Bank 3704 (2005 NRC Exam Q80)

Thursday, April 14, 2011 Page 238 of 299

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2O11 MNS SRO NRC Examination QUESLTION 81)

Development References Student References Provided rferences:

OP/1/A/6500/OO1 A (Ventilation Unit Condensate Drain Tank Operation)

Lesson Plan OP-MC-WE-WL Section 2.14 SYSO73 A2.02 Process Radiation Monitoring (PRM) System Ability to (a) predict the impacts of the following malfunctions or operations on the PRM system; and (b) based on those predictions, use procedures to cor- rect, control, or mitigate the consequences of those malfi.mctions or operations: (CFR: 41.5 /43.5/45.3 / 45.13)

Detector failure 401-9 Comments: RemarkslStatus

-J Thursday, April 14, 2011 Page 239 of 299

Question 80

References:

From Lesson Plan OP-MC-WE-RLR:

From 0P111A165001001 A (Ventilation Unit Condensate Drain Tank Operation):

EiicLosure op!1fA/65 00,001 A.

Puinpingvt.TCDT tu RC D&lmige tiu Pago 4 of 1 I Contlirnotis Release Method 3.4 WBZEN VTJCDT level reaches greater than S0 OR it is desired to pump V(JCDT.

perform the foliowiag and record oa Attachient 1:

34.1 NoifyRadwa%tc C iiistry of pending rclenc so appropriate icnistry personnel can pefform valve a1inmen: anc assist in release.

?tsunNu(ifled Dii(e Tiui NOTE: if any conipontnt required by SIC (1EMF44. VUCOT flow totalizer or cmpo&tc sampler) is nrperable, re1ese shall be made using batch release methzcl.

i4. ii Al ANY ILdL any ofthe lbllowingbecome mopersble. go to i1ncir 4 1 (Pumping VUCT)T to Rf 1rgi Using Thwl Refrce Method):

thMM4 (L (Cent \ent Urn ank Our)

VUCDT Minimum Flow Device (I WLFS7O 10)

VtXDT L5olock Conipos.te Snipkr Q 34.3 ccrd ThitiIs and Prinie Naine onAttacLment 2 (:nitials c1Jr.e1Ltthon.

NOTE: VIJCDT level monitoring is inoperable turing puinpiiig ofVLtCDT.

3.4.4 va1uateTS .4.1S.

3.4. II all Unit I R.C Pumps are off check lltC-21 RC Crossover Disch to Jnit 1)closed.

3.4.6 all Unit 2 RC Pumps re o check 1R.C-22 RC Crossover Dich to Juit 2) closed.

3,41 NotiirKadwaste Chenustryto ensure RC ps1<ecuired tbrL)ihtion set greater than or equal to tb immber of m.ims required by discharge document.

/________

?efsonNotiflecl Date Time 3A. Notify Radwiste Cheniistryto unlock and open 1WM 222 (VUCD to RC Disch ilcir).

/_________

crson Notified Date Time Unit 1

From Lesson Plan OP-MC-WE-WL Section 2.14:

2.14 Ventilation Unit Condensate Drain Tank Subsystem Objective #2, 3 Refer to Drawing 7.12, VUCDT Subsystem. The VUCDT collects non-recyclable condensation from various ventilation units. The sources of input to the VUCDT include:

Upper CONT. Ventilation Units

Lower CONT. Ventilation Units

Incore Instrument Air Handling Units

Auxiliary Building Air Handling Units **

- Condensation from the VA AHUs is normally routed to WZ Groundwater Sump A. It can be routed to the Unit I VUCDT if needed.

Contents of VUCDT are released to RC Discharge periodically utilizing the following process:

VUCDT is placed in Recirc by starting pump and throttling WL-359 (VUCDT Pump Recirc) to obtain 25-30 psig on Discharge Pressure Gage.

Radwaste samples contents of VUCDT.

1WL-1312 (VUCDT Pump Disch Isol.) is opened.

(Unit 1 only valve does not exist on Unit 2)

NOTE: For Unit 2 the air supply is opened to 2WL-320 (VUCDT EMF Isol.) to align the discharge flowpath. In both cases (UI and U2), EMF-44 Trip 2 will cause WL-320 to isolate, terminating the release.

To begin release, Operators open:

- 1WP-35 (WMTNUCDT to RC)

- IWP-37 (Liquid Waste to RC)

- WL-324 (VUCDT EMF Bypass) to obtain 3-5 GPM through EMF-44 while throttling WL-359 to obtain 29-31 psig on pump discharge.

EMF-44 Trip 2 will cause WL-320, WP-35, and WM-46 to isolate, terminating the release. (Refer to Drawing # 7.7)

7.12 VUCDT Subsystem (09110197 FLCOR&

.SUMPIB WL324 ftC

Parent Question (2005 NRC Exam Q80):

Unit I is releasing the Ventilation Unit Condensate Drain Tank (VUCDT) using continuous release per approved station procedures. Just after the release was initiated, 1 EMF-44 (Ventilation Unit Condensate Drain Tank) count rate indication fails to a reading of less than background.

Which one of the following statements correctly describes the effect on this LWR (Liquid Waste Release) In Accordance With 0P111A165001001 A, Ventilation Unit Condensate Drain Tank Operation?

A. Continue the release. Document OEMF-49 (Waste Liquid Disc) both OPERABLE and monitoring release flow path B. Continue the release. Request updated LWR release and OP paperwork.

C. Stop release. Initiate batch release with updated LWR and OP paperwork.

D. Stop Release. Notify IAE to restore 1 EMF-44 to OPERABLE prior to any VUCDT release.

ANSWER: C

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2011 MNS SRO NRC Examination QUESTION 81 SYSO34 Al .02 Fuel Handling Equipment System (FHES) ility to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the Fuel Handling System ntrols including: (CFR: 41.5 / 45.5)

Water level in the refueling canal Given the following conditions on Unit 1:

The unit is in MODE 6 with fuel movement in progress

A leak has developed which is causing Spent Fuel Pool level to decrease

The Spent Fuel Pool Level Low computer (OAC) alarm is activated Which ONE (1) of the following will be the FIRST action(s) directed by the CRS to mitigate the loss of level in the Spent Fuel Pool in accordance with (AP-40 Loss of Refueling Cavity Level)?

A. Makeup to the Spent Fuel Pool from the FWST.

B. Makeup to the Spent Fuel Pool from the RF Header.

C. Install the Spent Fuel Pool Weir Gate AND inflate the seals.

D. Move the fuel transfer cart to the spent fuel (pit) side AND close 1 KF-122 (Fuel Transfer Tube Block valve).

Wednesday, April 20, 2011 Page 240 of 299

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2011 MNS SRONRC Examination QUESTION 81.

General Discussion In accordance with AP-40 the first action which will be directed by the CRS is to move the fuel transfer cart to the spent fuel pit side and close 1KF-122.

Answer A DIscussion INCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible because the Operators are directed to makeup to the spent fuel pooi in AP-40. Makeup to the SFP via the FWST is the method directed by AP-40.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible because the Operators are directed to makeup to the spent fuel pool in AP-40. Makeup to the SFP from the RF header is one of the possible flow paths.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible because this would be the correct answer if 1KF-122 could not be closed. However, the first attempt is to close lKFl22.

Answer D Discussion CORRECT: See explanation above.

sis for meeting the KA KA is matched because the SRO applicant must have knowledge of the actions to be taken based on monitoring of changing parameters (i.e.

pent Fuel Pool Level). The to prevent exceeding design limits is met by the fact that taking the appropriate actions reduces the potential to damage fuel due to the lowering level in the SFP/refueling canal. For this particular set of circumstances, the refueling cavity and SPF would be tied together since the transfer tube block valve would be open. Therefore, refueling canal level would be lowering at the same time as the SFP level would be lowering.

Basis for Hi Cog

This is a higher cognitive level question because it requires more than one mental step. First the applicant must recall the difference in the actions between AP-40 and AP-41. Initial entry in this particular case would be into AP-4 1. Since the fuel transfer tube blind flange is NOT installed, KF-l22 is open, and Weir Gate is NOT installed, transition would be made to AP-40. The applicant must then recall that the first action directed by AP-40 is to move the transfer cart and close KF-l22.

Basis for 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 for screening questions linked to 10CFR55.43(b)(5) (Assessment and selection of procedures):

1) The question can NOT be answered by knowing systems knowledge alone. This is detailed procedure content knowledge from AP-40 and AP 41.

2) The question can NOT be answered by knowing immediate Operator actions. There are no immediate operator actions associated with AP-40 or AP-41.

3) The question can NOT be answered by knowing AOP or EOP entry conditions. Knowledge of AP-40 entry conditions will not enable the applicant to correctly answer this question.

4) The question can NOT be answered by knowing the purpose, overall sequence of events, or overall mitigative strategy of AP-40 or AP-4 1.

5) The question requires the applicant to assess plant conditions and then prescribing a procedure or section of a procedure to mitigate the consequences of the event. Specific to this event, initial entry would be into AP-41 (Loss of SFP Cooling or Level). However, since 1KF-122 is open the operator is directed out ofAP-41 and into AP-40 (Loss of Refueling Canal Level) where they arc directed to perform the appropriate actions.

ob Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Wednesday, April 20,2011 Page 241 of 299

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2011 MNS SRONRC Examination QUESTIQN 81 eveIopment References Student References Provided ferences:

Lesson Plan OP-MC-FH-FC Section 3.2.2 AP-40 (Loss of Refueling Canal Level)

SYSO34 Al .02 Fuel Handling Equipment System (FHES)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the Fuel Handling System controls including: (CFR: 41.5/45.5)

Water level in the refueling canal 401-9 Comments: RemarkslStatus Thursday, April 14, 2011 Page 242 of 299

Question 81

References:

From Lesson Plan OP-MC-FH-FC Section 3.2.2:

The Symptoms include:

EMF36 UNIT VENT GAS HI RAD alarm

EMF38 CONTAINMENT PART HI RAD alarm

EMF39 CONTAINMENT GAS HI RAD alarm

EMF4O CONTAINMENT IODINE HI RAD alarm

EMF42 FUEL BLDG VENT HI RAD alarm

EMFI6 CONTAINMENT REFUELING BRIDGE alarm (2 EMF3 on Unit 2) -

EMFI7 SPENT FUEL BLDG REFUEL BRDG alarm (2 EMF4 on Unit 2)-

Gas bubbles originating from the damaged assemblies

Visible evidence of damage with the potential of radioactive releases Operator Actions CAUTION Damage to the rubber Reactor Vessel Cavity Seal may occur if an assembly is dropped on or near it.

Announce on page. If in containment, evacuate containment, assemble in contaminated change room and refer to RPIO/A1570011 1, Conducting a Site Assembly, Site Evacuation, or Containment Evacuation. Isolate containment: stop VP fans, ensure VP valves close, stop any VQ release, ensure equipment hatch closed, ensure one airlock door closed, dispatch Operator to move conveyor to Spent Fuel Pool Building, dispatch Operator to close KF-122. If high containment radiation exists, place Aux Carbon Filters in service per OP. Place Refueling Cavity in purification per OP.

If in Spent Fuel Building, evacuate Spent Fuel Pool area, assemble in contaminated change room. Isolate Spent Fuel Pool area: Check if VF EXH BYP DAMPER closed lite lit, and if not, place its control switch to CLOSE, and close the doors to the Spent Fuel Pool area. Ensure KF purification loop in service per OP.

Refer to RPIOIAI5700IOO, Classification of Emergency.

3.2.2 AP111A15500140, LOSS OF REFUELING CANAL LEVEL The purpose is to provide actions in the event of loss of water in the refueling canal.

The Symptoms include:

Spent Fuel Pool Level Low computer alarm

Decreasing level in refueling canal

Incore Inst Room Sump Hi Level alarm

EMF16 CONTAINMENT REFUELING BRIDGE alarm (2 EMF3 on Unit 2) -

EMF17 SPENT FUEL BLDG REFUEL BRDG alarm (2 EMF4 on Unit 2)-

Operator Actions NOTE Any available core location may be used when lowering a fuel assembly during emergency conditions.

If fuel movement is in progress: lower any assembly in the reactor building crane to fully down in the core, any assembly in the spent fuel crane to fully down, and any assembly in the upender to fully down. If they wont lower otherwise, manually release the brake and hand crank the hoist down. NOTE: The sequence for lowering the hoist manually should be to put the emergency handwheel on the end of the hoist motor, hold it steady, while another person screws in the brake release (star shaped knob on a threaded stud) which when threaded in forces the brake disengaged. Care should be taken to remove the handwheel before electric operation of the hoist motor. The upender is similar. The bridge and trolly brake release is a lever, otherwise similar. Dispatch Operator to locally move fuel transfer cart to the spent fuel (pit) side. Stop FWST Pump and close FW-13, and dispatch Operator to locally close KF-122. If KF-122 cannot be closed, then notify RP to begin surveys, consider installing the weir gate, and isolate the Spent Fuel Building (VF in filter mode and doors closed). Evacuate nonessential personnel from containment and Spent Fuel Building.

Try to identify and correct the cause of decreasing level. Verify seal integrity and air pressure to the Rx Vessel cavity seal and the Rx Vessel nozzle inspection port seals, and if not, reestablish VI to seals. Dispatch an Operator to locally ensure the Refueling Cavity Drains are closed. Check the SIG Nozzle Dams. Refer to API1 9, Loss of ND or ND System Leakage, while continuing with this procedure.

Makeup to the canal per OP111A16200113. CAUTION: Makeup to the SFP could dilute NC system boron concentration.

Monitor the Spent Fuel Pool level. If it gets to minus two feet, stop the KF Pump and turn off the lights. Initiated makeup per OP. If pool level low enough for radiation hazard, makeup from RN.

Ensure Containment Integrity with equipment hatch and airlock doors closed. If time permits, turn off canal underwater lights before they become uncovered. If necessary due to increasing radiation levels, consider using ND or NS to transfer water from the containment sump to the FWST for additional makeup capability.

Refer to RPIOIAI5700IOO, Classification of Emergency.

From AP-40:

MNS LOSS OF REFUELING CAVITY LEVEL PAGE NO.

AP/i/A/5500/40 2 of 16 Rev. 7 UNIT 1 ACTION/EKPECTED RESPONSE RESPONSE NOT OBORINED B. npjms

SPENT FUEL POOL LEVEL LOW computer alarm

Level in refueling cavity going down

INCORE INST ROOM SUMP HI LEVEL alarm

IEMF-16 CONTAINMENT REFUELING BRDG alarm

1 EMF-1T SPENT FUEL BUILDING BROG alarm.

C. Operator Actions

1. Announceoccurrenceonpage.

2. Check FUEL MOVEMENT IN

- Perfarm the following:

PROGRESS.

a. IF any radioactive component is being handled in the spent fuel pool or refueling covity, TI-lEN have fuel handling crew lower component to fully down.

b. IF cevity level is dropping more than one inch per minute, A 1 FW-27A (Unit 1 FWST to ND Pumps 301) is open, THEN initiate makeup .PER Enclosure 3 (Refueling Cavity Makeup Using ND Pump) while continuing in this AR

_c. OIOStep4.

MNS LOSS OF REFUELING CAVITY LEVEL PAGE NO.

APJ1/A15500140 3 of 18 Rev. 7 UNIT 1 I ACTIONJEXPECcED P.ESPONSE REPPONSE NOT O3TINED MOTE Any availab[e core location may be used when lowering a fuel assembly during emergency conditions.

3. Contact fuel handling SRO to have fuel handling crew perform the following:

a. Lower any fuel assembly in the reactor a. Release brake and hand crank hoist building manipulator crane to fully down down.

in the core.

b. Lower any fuel assembly in the spent b. Release brake and hend crank hoist fuel manipulator crane to fully down, down.

c. Lower any fuel assembly in either c. Reiease brake and hand crank upender upender to fully down. down.

-L

d. Move fuel transfer cart to the spent fuel d. Release brake and hand crank transfer (Pit) side. cart to spent fuel (Pit) side.

e. Lower any radioactive component in the e. Perform the following:

spent fuel pool or refueling cavity to fully down. . Reinstall component.

OR

. Place component as fr below the water as safely possible.

4. WHEN fuel transfer cart is in the spent fuel bldg, THEN dispatch 2 operators to CLOSE IKF-122 (Unit I Fuel Transfer Tube Isol) (spent fuel bldg, 780, PP-51, top of fuel pool at south east corner).

5. Notify Containment Closure Coordinator to initiate containment closure PER PTII)N42001002 C (Containment Closure).

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2011 MNS SRO NRC Examimitiow. QUESTION 82 -

SYSO72 2.4.31 Area Radiation Monitoring (ARM) System YS072 GENERIC now1edge of annunciator alarms, indications, or response procedures. (CFR: 41.10 I 45.3)

Regarding the use of FR-Z.3 (Response To High Containment Radiation):

1) At what MINIMUM reading on IEMF 51A (Containment High Range) is the YELLOW path for Containment High Radiation valid?

2) What does this procedure direct to reduce activity in the Containment atmosphere?

A. 1. 35RIhr

2. Start the Containment Auxiliary Charcoal Filter Unit.

B. 1. 15R/hr

2. Start the Containment Auxiliary Charcoal Filter Unit.

C. 1. 35RIhr

2. Ensure the VE system is in service and purge containment to the annulus D. 1. l5RIhr

2. Ensure the VE system is in service and purge containment to the annulus Thursday, April 14, 2011 Page 243 of 299

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2011 MNS SRO NRCExamination QUESTION 82

.neraI Discussion In this question, the applicant is asked what reading from the containment Area radiation monitor (1 EMF-5 IA) would give a valid yellow path on SPDS for High Containment Radiation. The yellow path for containment radiation is 35 R/hr per FRi FO. The procedural direction for mitigation is to place the containment Charcoal Filter unit in-service. While purging the containment atmosphere to the annulus would effectively lower Rad levels in containment, it is not directed for this situation.

Answer A Discussion CORRECT: See explanation above.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part (1) is plausible because the value of lSmr/hour is the limit for an area monitor outside containment (Control room) which requires the declaration of an Alert because it is considered a value which would impede the operation of systems required to maintain safe operations. The SRO applicant may confuse this value (15) with that required to implement corrective actions for high containment rad levels.

Part (2) is correct and therefore plausible.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part (I) is correct and therefore plausible.

Part (2) is plausible because this strategy would be effective in lowering containment radiation levels by performing a controlled release of the containment atmosphere to the Annulus Ventilation (VE) filter system.

Thswer D Discussion CORRECT: See explanation above.

PLAUSIBLE:

Part (1) is plausible because the value of 1 Smr/hour is the limit for an area monitor outside containment (Control room) which requires the declaration of an Alert because it is considered a value which would impede the operation of systems required to maintain safe operations. The SRO applicant may confuse this value (15) with that required to implement corrective actions for high containment rad levels.

Part (2) is plausible because this strategy would be effective in lowering containment radiation levels by performing a controlled release of the containment atmosphere to the Annulus Ventilation (VE) filter system.

Basis for meeting the KA The K/A is matched because the implementation of FRP Z.3 (Response to High Containment Rad) is a response procedure for a abnormal indication associated with the containment ARIVI (1EMF-51A). This is also a direct alarm response for the Yellow path for SPDS Containment which would come in alarm at 35 Rihr. The Alarm response on the OAC directs implementation of the FR Z.3. The question requires the applicant to demonstrate both what reading would be required to implement and detailed knowledge of what is directed by this procedure.

Basis for Hi Cog Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev ldated 03/11/2010 for screening questions linked to 10CFR55.43(b)(5) (Assessment and selection of procedures):

1) This question can NOT be answered by knowing systems knowledge alone. This is strict procedure knowledge. This is not covered during systems training or discussed in a systems lesson plan.

2) This question can NOT be answered by knowing immediate operator actions.

3) This question can NOT be answered by knowing the ntry conditions for AOPs. The steps to be taken by the crew are not based on the entry ditions provided.

fhis question can NOT be answered by knowing the purpose, overall sequence of events, or overall mitigative strategy of the AOPs. The question is based on knowledge of specific procedure content.

5) The question requires the applicant to have in-depth knowledge of specific steps within FRP Z.3. Specifically, it requires the applicant to recall that FR Z.3 requires the crew to place the containment aux filter in service vs. another set of actions which would provide mitigation but directed for the indications provided.

____j Thursday, April 14, 2011 Page 244 of 299

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2011 MNS SRO.NRC Examination. QUESTION S2 Job Level Cognitive Level QuestionType Question Source SRO Memory BANK Bank 591 Development References Student References Provided Lesson Plan OP-MC-EP-FRZ Objectives 2 & 3 (Rev 20)

Lesson Plan OP-MC-EP-FRZ Pg 53 of 89 EP/FR-Z.3 (Response to High Containment Rad Level)

SYSO72 2.4.31 Area Radiation Monitoring (ARM) System SYSO72 GENERIC Knowledge of annunciator alarms, indications, or response procedures. (CFR: 41.10 I 45.3) 401-9 Comments: Remarks/Status Thursday, April 14, 2011 Page 245 of 299

Question 82

References:

From Lesson Plan OP-MC-EP-FRZ Objectives 2 & 3 (Rev 20)

From Lesson Plan OP-MC-EP-FRZ Pg 53 of 89 (Rev 20)

DUKE ENERGY MCGUIRE OPERATIONS TRAIrllUc3 FR-2. Repne tQ Higt Ccntainrnent P.diricn Level 5.0 FR-Z.3, RESPONSE TO I4IGH CONTAINMENT RADIATION LEVEL 5.1. Purpose This procedure provides actions to respond to high containment radiation level.

5.2. SymptomslEntry Co ndit ions This procedure is entered, based on op&atorjudgenient, from EPI1IAJ5OaO/F-O (Critical Safety Function Status Trees) (Contain,ent). on a Yellow condition This condition is:

I. Containment radiation greater than 35 RJhr(1EMF 51A orB)

From EPIFR-Z.3 (Response to High Containment Rad Level) Pg 2 of 3 (Rev 3)

MNS RESPONSE TO 1-HG H CONTAINMENT RADIATION LEVEL PAGE NO.

EP/1/A/5000/FR-Z.3 2 of 3 Rev. 3 UNIT 1 CT c::i-:

C. LActois I. Check Containment ventilaUon isolation as follows:

a. Check the following isolation valves - a. CLOSE valve(s).

CLOSED:

IVQ-1A (UI Cont Air Release Inside isol)

IVQ-6A (Ui Cont Air Addition lnide Isol)

IVQ-2B (Ui Cont Air Release Outside Isol)

. iVQ-5(Ui Cont Air Addition Outside lsol)

a. a. Start fans as follows:

C2 and D2 1) Select ON.

2) Retim switch to AUTO.

select switches in AUTO.

. 1AVS-D-T Mode Select

. 1AVS-D-8 Mode Select

. 1AVS-D-2 Mode Select IAVS-D-3 Mode Select.

From EPIFR-Z.3 (Response to High Containment Rad Level) Pg 3 of 3 (Rev 3)

MNS RESPONSE TO HIGH CONTAINMENT RADIATION LEVEL PAGE NO.

EPII/A/5000!FR-Z.3 3 of 3 Rev. 3 urr i c:oN/Ex?Ec:ED pONE tESOE I;oT cm:1ED

3. Check if Containment Aux Carbon Filter Fan can be placed in service as follows:

a. Check containment sunip level LESS

- a. GOTOStep4.

THAN OR EQUAL TO 0.5 FT.

b. Start IA Containment Aux Carbon Filter Fan.

4. Notify station management of containment radiation level to obtain recommended actionS

5. RETURN TO procedure and step in effect, END

From RPI000 (Classification of Emergency) Enc 4.3 Pg 5 of 5 (Rev 17)

Enclosure 4.3 Abnormal Rad. Leve1s/RathologicaI Effiue]it tNUSUAL EVENT ALERT SITE AREA EMERGENCY 4.3.A.3 Release of Radioactive Material or Iiicreases in Radiation LeTels Within the Iadility That Impeles Operation of Systems Reqnired to Maintain Safe Operations oi to Establish or Maintain Cold Slim tdowm OPERATING MODE ALL 4.3.A.3-1 Valid reading on EMF-12 greaierthan1mRfbrinthe Control Room.

4.3.A..3-2 Valid indication of radiation levels greater than 15 mRfbr inthe Central Alann Station (CA.S) or Secondary Mann Station (SAS).

4.3.A..3-3 Valid area EMP reading exceeds the levels sho in Enclosure 4.10.

END

Parent Question (Bank 591):

Regarding the use of EPIIIAI5000IFR-Z.3 (Response To High Containment Radiation):

1. At what minimum reading on 1 EMF 53A (Containment High Range) is the YELLOW path for Containment High Radiation valid?

2. What mitigative strategy does this procedure direct to reduce activity in the containment atmosphere?

A. 1. 35R/hr

2. Start Containment Auxiliary Charcoal Filter Units.

B. 1. l5RIhr

2. Start Containment Auxiliary Charcoal Filter Units.

C. 1. 35R/hr

2. Ensure the VE system is in service and vent containment to the annulus using the VY system.

D. 1. 15R/hr

2. Ensure the VE system is in service and vent containment to the annulus using the VY system.

ANSWER: A

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2011 MNS SRO NRC Examination. QUESTION 83 SYSO86 A2.Ol Fire Protection System (FPS) bility to (a) predict the impacts of the following malfunctions or operations on the Fire Protection System; and (b) based on those predictions, se procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 /43.5/45.3 / 45.13)

Manual shutdown of the FPS Given the following:

Unit 1 was operating at 100% RTP

While performing maintenance activities in the 1A DIG Room, one of the heat detectors associated with the Halon system was inadvertently actuated

The Halon actuation was successfully aborted by the fire watch depressing the ABORT/OFF pushbutton prior to actual Halon discharge Which ONE (1) of the following describes how the 1A D/G Halon system is affected AND what MINIMUM actions are required per SLC 16.9.3 (Halon Systems)?

A. Auto actuation ONLY is blocked Establish an Hourly fire watch within one hour B. BOTH Manual and Auto actuation of the system is blocked Establish a Hourly fire watch within one hour C. Auto actuation ONLY is blocked Establish a Continuous Fire Watch within one hour D. BOTH Manual and Auto actuation of the system is blocked Establish a Continuous Fire Watch within one hour Thursday, April 14, 2011 Page 246 of 299

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2011 MNS. SRO NRC Examination QUESTION -83 General Discussion The DIG rooms are protected by an installed Halon system. These systems are actuated by any one of 4 installed heat detectors in each DIG room. The system is equipped with a time delay device which will delay the discharge of Halon for 20 seconds after an automatic actuation.

During this time an operator can abort the discharge utilizing a local Abort switch. This button locks into place and will block auto actuation only. A manual actuation cannot be blocked.

Per SLC 16.9.3 basis, if the Halon system associated with the 1A DIG room becomes inoperable, Condition A is applied because there is impact on redundant equipment. (1CA-42B power cable). Condition A requires the establishment of a continuous fire watch within one hour.

Answer A Discussion LI INCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible because all of the other DIG room Halon inoperability would only require an Hourly fire watch.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible because the candidate may not remember Manual cannot blocked but it would be reasonable to assume that it could.

Also, If Halon is determined to be inoperable in any of the 3 remaining DIG rooms; entry is required into Condition B of SLC 16.9.3 which only requires the establishment of an hourly fire watch Answer C Discussion CORRECT: See explanation above.

Answer D Discussion TCORRECT: See explanation above.

rLAUSIBLE:

This answer is plausible because the candidate may not remember Manual cannot blocked but it would be reasonable to assume that it could.

Also, If Halon is determined to be inoperable in any of the 3 remaining DIG rooms; entry is required into Condition B of SLC 16.9.3 which only requires the establishment of an hourly fire watch.

Basis for meeting the KA KA is matched because DIG Halon is inadvertently actuated due to welding near a detector requiring an manual shutdown. The candidate must evaluate (Predict) the resulting impact on the Halon system and mitigate the consequences of the resulting inoperability by applying the appropriate actions in SLC 16.9.3. (Use procedures to mitigate the consequence)

Basis for Hi Cog Basis for SRO only This Question is linked to 10CFR55.43 (b)(2) (Facility Operating limitations in the tech specs and their basis). Justification for SRO level lies in the fact that the candidate cannot successfully answer this question without the knowledge contained in the Basis of SLC 16.9.3. In that basis, the lA DIG room is identified as the only DIG room where Condition A is applied due to presence of redundant train equipment.

Job Level Cognitive Level QuestionType Question Source SRO Memory BANK Bank 3107 (2009 NRC Exam Q92)

Development References Student References Provided OP-MC-SS-RFY Obj: 22

,son Plan OP-MC-SS-RFY (Rev 26) Pg 51 0

L C 16.9.3, Halon Systems and Basis I SYSO86 A2.0l Fire Protection System (FPS)

Ability to (a) predict the impacts of the following malfunctions or operations on the Fire Protection System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 I 43.5 I 45.3 I 45.13)

Thursday, April 14, 2011 Page 247 of 299

FOR REVIEW ONLY DO NOT DISTRIBUTE 2011 MNS SRO NRC ExaminatiQu QUESTIQ S3 83 Manual shutdown of the FPS 1-9 Comments: RemarkslStatus Thursday, April 14, 2011 Page 248 of 299

Question 83

References:

From Lesson Plan OP-MC-SS-RFY Pg 51 of 101 (Rev 28)

The Transfer Switch (main/reserve positions) located at the halon cylinders is used to select the bank of halon cylinders that will automatically dump into a selected diesel room during a fire. Two solenoids automatically open to allow actuation to occur. One solenoid, mounted on a nitrogen bottle pressurized to 1000 psig, opens to admit nitrogen up to the first main or reserve cylinder ( refer to Drawing 7.10). This nitrogen then forces open a valve that permits halon to exit the bottle onto the header. Then the room select pilot solenoid located under one of the Fire Protection Control Panel permits a pilot halon pressure signal to be sent back to the room select valve for the room in question. This pilot pressure signal opens the appropriate room select valve permitting halon to enter the room. Both solenoids will close when the fire detection signal has cleared (temperature below thermostat setting).

Objective #23 The manual actuator lever and pull key are used in conjunction with the halon supply control valve (pilot valve) and pull key to manually actuate halon. First the pull key is removed and the pilot valve is opened. This allows a pilot pressure to be sent to the room select valve when halon is released to the header. Then the pull key for the manual lever is removed and the lever pulled downward. This releases halon to the header. Once the pilot pressure signal has returned to the room select valve, the valve opens. Halon is then dumped to suppress the fire.

Once all the halon has dumped the pilot valve should be closed. This will allow the remaining bank of halon bottles to be used to suppress fires in the other diesel room or same room if another fire occurs.

Objective #22 The MAIN and RESERVE actuator pushbuttons, located on each of the actuation panels, in each diesel room are used to electrically actuate either bank of halon cylinders regardless of MAINIRESERVE switch positions (refer to Drawing 7.11 ).

By depressing the MAIN or RESERVE pushbuttons (and holding for five seconds ) the appropriate solenoid valve is opened at the nitrogen bottle. The solenoid valve, located on the actuator panel in the diesel room, is also opened. This allows halon to be sent to whichever room the MAIN or RESERVE actuator pushbutton is located in. For proper discharge, the button must be held for no less than five seconds. The ABORTIOFF switch, located on each of the actuation panels, in each diesel room is used to stop any automatic halon actuation. This button locks into place and an ABORTED light will illuminate. Abort will de-energize the two solenoid valves that open on automatic actuation only. A manual electric actuation cannot be aborted.

SLC 16.9.3 (Halon Systems) 16.9 AUXILIARY SYSTEMS 16.9.3 Halon Systems COMMITMENT The following Halon Systems shall be OPERABLE:

a. Elevation 716 ft. Auxiliary Building Room No. Equipment 600B Turbine Driven Aux. FW Pump Unit I 601 B Turbine Driven Aux. FW Pump Unit 2

b. Elevation 733 ft. Auxiliary Building Room No. Equipment 703-704 Diesel Generators Unit 1 714-715 Diesel Generators Unit 2 APPLICABILITY Whenever equipment protected by the Halon System is required to be OPERABLE.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Establish a continuous fire 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Halon Systems watch with backup fire inoperable in an area in suppression equipment.

which redundant systems or components could be damaged.

B. One or more required B.1 Establish fire watch patrol. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Halon Systems inoperable in areas other AND than Condition A.

Once per hour thereafter

TESTING_REQUIREMENTS TEST FREQUENCY TR 16.9.3.1 Verify each manual, power operated, or automatic valve 31 days in flow path is in its correct position.

TR 16.9.3.2 Verify Halon storage tank weight> 95% of full charge 6 months weight and pressure > 90% of full charge pressure.

TR 16.9.3.3 Verify system actuates upon receipt of a simulated 18 months manual and automatic actuation signal and damper closure devices receive an actuation signal upon system operation.

TR 16.9.3.4 Perform a flow test through headers and nozzles to 18 months assure no blockage.

BASES The OPERABILITY of the Fire Suppression Systems ensures that adequate fire suppression capability is available to confine and extinguish fires occurring in any portion of the facility where safety-related equipment is located. The Fire Suppression System consists of the water system, spray, and/or sprinklers, Halon, and fire hose stations. The collective capability of the Fire Suppression Systems is adequate to minimize potential damage to safety-related equipment and is a major element in the facility fire protection program.

In the event that a Halon System becomes inoperable, compensatory actions are required to be taken in the affected areas until the inoperable equipment is restored to service.

For Rooms 704 (lB DIG), 714 (2A DIG) and 715 (2B DIG), Condition B is applied if the Halon system is declared inoperable since there is no impact to redundant systems and components.

For Room 703 (IA DIG), Condition A is applied if the Halon system is declared inoperable since there is impact on redundant equipment (1 CA-42B power cable).

The Testing Requirements provide assurance that the minimum OPERABILITY requirements of the Fire Suppression Systems are met. An allowance is made for ensuring a sufficient volume of Halon in the Halon storage tanks by verifying either the weight or the level of the tanks. Level measurements are made by either a UL or FM approved method.

The main bank (1 cylinder for the TDCA Pump Room, 8 cylinders for the D/G Room) or the reserve bank (1 cylinder for the TDCA Pump Room, 8 cylinders for the DIG Room) provides a sufficient quantity of halon to totally flood any of the TD CA Pump Rooms or Diesel Generator

Rooms with the required design concentrations. Therefore, the Halon System is OPERABLE with the system aligned to either the main or the reserve bank of cylinders. The system is aligned to the main or reserve bank of cylinders by means of a local manual toggle switch.

TR 16.9.3.1 requires that valves in the flow path for the required halon systems be verified to be in their correct position. Although the selector valves and the cylinder valves are in the flow path, these valves are excluded from this testing requirement for the following reasons:

1. There is no visible means of determining valve position,

2. The valves are spring loaded piston actuators which fail closed and require halon discharge header pressure to open (Selector Valves Only),

3. There is no credible means to mis-position these valves other than actual actuation of the halon system,

4. These valves are an integral component of the actuation circuitry for the halon system, which is tested perTR 16.9.3.3, and

5. This exclusion is consistent with fire protection industry practices.

This selected licensee commitment is part of the McGuire Fire Protection Program and therefore subject to the provisions of McGuire Facility Operating License Conditions C.4 (Unit 1) and C.7 (Unit 2).

Parent Question (2009 NRC Exam Q92):

Examination Outline Cross-reference: Level RO SRO x

Tier# 2 Final Group # 2 KIA# 086A2.03 Importance Rating 2.9 Fire Protection:

Ability to (a) predict the impacts of the following on the (Fire Protection) system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: (Inadvertent actuation of the FPS due to circuit failure or welding.)

Proposed Question: SRO 92 1 Pt Given the following:

Unit 1 was operating at 100% RTP

While performing maintenance activities in the 1A DIG Room, one of the heat detectors associated with the Halon system was inadvertently actuated

The Halon actuation was successfully aborted by the fire watch depressing the ABORT/OFF pushbutton prior to actual Halon discharge Which ONE (1) of the following describes how the 1A D/G Halon system is affected AND what actions are required per SLC 16.9.3 (Halon Systems)?

A. Auto actuation ONLY is blocked Establish an Hourly fire watch within one hour B. BOTH Manual and Auto actuation of the system is blocked.

Establish a Hourly fire watch within one hour.

C. Auto actuation ONLY is blocked Establish a Continuous Fire Watch within one hour.

D. BOTH Manual and Auto actuation of the system is blocked.

Establish a Continuous Fire Watch within one hour.

Proposed Answer: C

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21111 MNS SRO NRC Examinatiun QUESTION 84 EPEOO7 EA2.04 Reactor Trip ility to determine or interpret the following as they apply to a reactor trip: (CFR 41.7/45.5 /45.6) reactor should have tripped but has not done so, manually trip the reactor and carry out actions in ATWS EOP Given the following initial conditions on Unit 2:

Unit 2 is operating at 75% RTP with 120 VAC vital instrument bus EKVC de-energized for maintenance

All required Tech Spec actions have been taken Given the following conditions:

Both LOCA Sequencer Actuated status lights (2S1-14) are LIT

PZR Pressure Ch 4 instrument has failed low

The reactor trip breakers have failed in the closed position due to improper maintenance and retesting

1) Which ONE (1) of the folloAng sequences describes the plant response?

2) For the conditions described above what actions are required and why?

A. 1. The reactor protection system generates a trip signal and an ATWS is in progress.

2. Trip the main turbine to ensure SG inventory is maintained.

B. 1. The reactor protection system generates a trip signal and an ATWS is in progress.

2. Trip the main turbine to generate a redundant automatic reactor trip signal.

C. 1. The reactor protection system generates a trip signal and a failure of RPS occurs (however, no ATWS is in progress).

2. Enter TS 3.0.3 and be in Mode 3 within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months due to the plant being in an unanalyzed condition.

D. 1. The reactor protection system generates a trip signal and a failure of RPS occurs (however, no ATWS is in progress).

2. Locally open the Unit 2 RTBs because a valid reactor trip signal has been generated.

Wednesday, April 20, 2011 Page 249 of 299

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2(Ui MNS SRO NRC Examinaffm. QUESTION 84 General Discussion In the scenario given, the applicant is presented with indications that Channel 3 protection bistables are in a tripped condition for both the Low PZR Pressure Reactor Trip and Safety Injection. (EKVC feeds all of the channel 3 protection bistables). When the PZR pressure channel 4 fails low, this results in both a reactor trip signal and a safety injection. Because the reactor trip breakers fail to open and the fact there is a transient in progress (Loss of feedwater from SI) this is an ATWS. The main turbine is tripped immediately in this scenario to preserve S/G inventory because feedwater has been isolated and there isno directtripofthe main turbine generated because there is no P4 signal present.

Answer A Discussion CORRECT: See explanation above.

Answer B Discussion INCORRECT: See explanation above. 1 PLAUSIBLE:

Part (I) is correct and therefore plausible.

Part (2) Plausible because the action is correct but the reason is not. An additional reactor trip signal would be generated and this would seem a reasonable course of action to the applicant.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part (1) is plausible because the reactor trip signal by itself does not constitute an ATWS (2 channels of PZR pressure failed low). The applicant may fail to realize that an SI signal would be generated as well and if so, this would seem a reasonable answer.

(2) Plausible because the action and reason are correct if the first part of the answer is correct. This would be the correct answer if the licant determines that the conditions described do not constitute an ATWS.

iswer D Discussion 4

INCORRECT: See explanation above.

PLAUSIBLE:

Part (I) is plausible because the reactor trip signal by itself does not constitute an ATWS (2 channels of PZR pressure failed low). The applicant may fail to realize that an SI signal would be generated as well and if so, this would seem a reasonable answer.

Part (2) is plausible because the applicant would correctly determine that a real reactor trip signal has been generated and determine that manual actuation is required.

Basis for meeting the KA This K/A is matched because the applicant is being asked to evaluate (interpret) a set of plant conditions and determine whether or not an ATW has occurred. He is then asked about actions contained in the ATWS EOP (Immediately tripping the main turbine) and the reason for that action.

Basis for Hi Cog This question is Hi Cog because the applicant must evaluate a given set of conditions and through a multipart mental process, determine the required actions based on these conditions evaluate the required actions and the reason for that action.

Basis for SRO only Eis 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 10CFR55.43(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.

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 assess given plant conditions and determine whether or not an ATWS has actually occurred and n selecting a given action that is contained in the correct procedure to mitigate the event and the basis for that action.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Thursday, April 21, 2011 Page 250 of 299

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2011 MNS SRO NRC Eamimition QUESTION 84 Dob Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Development References Student References Provided 4.3 (Use of Abnormal and Emergency Procedures) Page 10 of 35 FRP S. 1 (ATWS) Background Document Pg 25 of 69 EPEOO7 EA2.04 Reactor Trip Ability to determine or interpret the following as they apply to a reactor trip: (CFR 41.7 / 45.5 / 45.6)

If reactor should have tripped but has not done so, manually trip the reactor and carry out actions in ATWS EOP 401-9 Comments: RemarkslStatus Thursday, April 14, 2011 Page 251 of 299

Question 84

References:

From Operations Mgmt Procedure (OMP) 4.3 (Use of Abnormal and Emergency Procedures) Page 10 of 35 (Rev 32)

OMP4-3 Page 10 of 35 7.8 ATWS AnATWS (Anticipated Transient Without Scram) is defined in 10 CFR 50.62 as an anticipated operational occurrence followed by the failure of the reactor trip portion of the protective systeni. An anticipated operational occurrence is defined in IC) CFR 50, Appendix A, as those conditions of normal operation which, are expected to occur one or more times during the life of the nuclear power unit and include but are NOT limited to loss ofpower to all NC pumps, tripping of the turbine generator, isolation of the main condenser and loss of all offsite power. Clearly, to have an ATWS there must be transient followed by a failure of the reactor trip breakers.

Instniment failures, by themselves, are NOT necessarily transients. For example, if one channel of Power Range Nuclear Instnment was out of service for preventive maintenance Qaistable in tripped condition) and if another Power Range Nuclear Instrument channel failed, a reactor trip signal would be generated. the reactor failed to trip, this would be a failure of the reactor trip breakers and the automatic trip features of the reactor protection system and NOT an ATWS event. Obviously, the control operators would have to recognize and check that the channel failure was indeed a channel failure by checking the other two channels in this example. This would, however, force OPS to shutdown the affected unit to at least Hot Standby per Tech Specs.

7,) Adverse Containment Setpoints Many setpoints in the EPs are presented in a dual format with a second setpoint enclosed in parentheses. This second setpoint is used to account for the additional error in the setpoint due to the containment environment folloving a high-energy line re The setpoint in parentheses will be used whenever containment pressure has exceeded 3 psig

From FRP S.1 (ATWS) Background Document Pg 25 of 69 (Rev 10)

STEP 2 Check Turbine Trip: (IMMEDIATE ACTION)

PURPOSE: To ensure that the turbine is tripped.

BASIS: For an ATWS event where a loss of normal feedwater has occurred, analyses have shown that a turbine trip is necessary (within 30 seconds) to maintain SIG inventory. For other ATWS events, manual tripping of the turbine may yield a higher system pressure than would otherwise occur. However, this action has been determined to be necessary due to the analytical results discussed earlier. Since there are many initiating ATWS events and some that require immediate mitigating actions, diagnosis of the initiating event would not be feasible and separate guidance for different ATVVS events would complicate training and could delay timely performance of necessary operator actions.

If the turbine will not trip, a turbine runback (manual lowering of load) at maximum rate will also reduce steam flow in a delayed manner. If the turbine stop valves cannot be closed by either trip or runback, the MSIVs and MSIV bypass valves should be closed.

STEP 3 Monitor foldout page.

PURPOSE: Remind the operators to monitor the Foldout Page.

BASIS: The Foldout Page contains three items:

1. Transfer to Cold Leg Recirculation if FWST low level is reached. This operator action is required no matter what EP is in effect to ensure the transfer is accomplished without delay.

2. CA Suction Source Monitoring.

3. Criteria for isolating and unisolating the NV Pump Recirculation Isolation Valves (NV-150 and NV-151).

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2011 MNS SRO NRC Examination - QUESTION 85 L EPEOO9 EA2.1 1 Small Break LOCA bility to determine or interpret the following as they apply to a small break LOCA: (CFR 43.5 I 45.13)

.ontainment temperature, pressure, and humidity Given the following plant conditions:

Unit 2 is operating at 50% RTP

A NC system leak has developed inside Containment

Containment pressure has increased from 0.1 PSIG to 04 PSIG

Lower Containment temperature has increased to 118°F Which ONE (1) of the following describes the concern with continued operation with these conditions per the basis for TS 3.6.4 (Containment Pressure) and TS 3.6.5 (Containment Temperature)?

A. Design peak Containment temperature could be exceeded during a LOCA B. Design peak Containment pressure could be exceeded during a LOCA C. Design peak Containment temperature could be exceeded during a Steam Line Break D. Design peak Containment pressure could be exceeded during a Steam Line Break Thursday, April 14, 2011 Page 252 of 299

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2011 MNS SRO NRC Examination. QUESTION 85 eneral Discussion The applicant is given indications that a NC system leak inside containment (Small LOCA) has resulted in elevated containment temperature and pressure. The upper limit assumed in accident analysis is 0.3 PSIG and per the basis for TS 3.6.4 (Containment Pressure), the DBA assumed in

the analysis is a LOCA. Maintaining the initial pressure below 0.3 PSIG ensure that peak containment pressure will not exceed maximum containment design pressure. The highest containment temperature assumed in accident analysis per the basis for TS 3.6.5 (Containment Temperature) is 135 degrees. The LCO for TS 3.6.5 allowed operation for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months with lower containment temperature up to 135 degrees.

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible if the applicant confuses the Upper and Lower Containment Tech Spec limits for allowable temperature. Since the temperature provided in the stem is higher than normal Containment temperature and outside the Tech Spec limit, it is reasonable for the applicant to conclude that the peak Containment temperature limit could be exceeded on a DBA. If the applicant confuses the basis for the Containment Temperature and Pressure tech specs it would be reasonable for them to conclude that the limit would be exceeded during a design is LOCA event.

Answer B Discussion

[9RRECT: See explanation above.

Answer C Discussion CORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible if the applicant confuses the Upper and Lower Containment Tech Spec limits for allowable temperature. Since the temperature provided in the stem is higher than normal Containment temperature and outside the Tech Spec limit, it is reasonable for the applicant to conclude that the peak Containment temperature limit could be exceeded on a DBA. If that were the case they would conclude that the limit would be exceeded on a design basis Steam Line Break since this is the bounding accident for the Containment temperature Tech Spec nit.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible if the applicant confuses the bases for the Containment Temperature and Pressure Tech Spec limits. Because of the mass and energy associated with a SLB inside containment, it would be reasonable for the applicant to believe that this would be the bounding DBA for the analysis associated with the Containment pressure LCO. It would also be reasonable for the applicant to believe that even though the given pressure in greater than the LCO limit, that safety analysis would assume a higher initial value, we have a containment high pressure annunciator which alarms a 0.5 PSIG and places maximum containment cooling in operation.

Basis for meeting the KA The applicant is given indications that a NC system leak inside containment (Small LOCA) has resulted in elevated containment temperature pressure. He must then interpret this information and determine why continued operation would not be possible.

Basis for Hi Cog The applicant is required to evaluate a given set of indications, recall the correct limit and precaution and determine an outcome. This is a part mental process and an application of a rule.

Basis for 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 03/1 1/20 10) under the Screening Criteria for question linked to IOCFR55.43(b)(2) (Tech Specs):

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

2) It can NOT be answered solely by knowing the LCO/TRM information listed above-the-line

3) It can NOT be answered by knowing the Tech Spec Safety Limits or their bases

4) It DOES require the applicant to have detailed knowledge of Tech Spec basis information to determine the correct answer.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Thursday, April 14, 2011 Page 253 of 299

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2011 MNS SRO NRCExamination. QUESTION 85 as Development References Student References Provided 3.6.4 (Containment Pressure)

- S 3.6.4 Basis (Containment Pressure)

TS 3.6.5 (Containment Temperature)

TS 3.6.5 Basis (Containment Temperature)

EPEOO9 EA2. 11 Small Break LOCA Ability to determine or interpret the following as they apply to a small break LOCA: (CFR 43.5 /45.13)

Containment temperature, pressure, and humidity 401-9 Comments: ;RemarkslStatus Thursday, April 14, 2011 Page 254 of 299

Question 85

References:

From TS 3.6.4 (Containment Pressure) 3.6.4 Containment Pressure LCO 3.6.4 Containment pressure shall be -0.3 psig and +0.3 psig.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment pressure A.1 Restore containment 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months not within limits, pressure to within limits.

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. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE_REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1 Verify containment pressure is within limits. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

From TS 36.4 Basis (Containment Pressure)

B 3.6.4 Containment Pressure BASES BACKGROUND The containment pressure is limited during normal operation to preserve the initial conditions assumed in the accident analyses for a loss of coolant accident (LOCA) or steam line break (SLB). These limits also prevent the containment pressure from exceeding the containment design negative pressure differential with respect to the outside atmosphere following an event which has the potential to result in a net external pressure on the containment.

Containment pressure is a process variable that is monitored and controlled. The containment pressure limits are derived from the input conditions used in the containment functional analyses and the containment structure external pressure analysis. Should operation occur outside these limits coincident with a Design Basis Accident (DBA), post accident containment pressures could exceed calculated values.

APPLICABLE Containment internal pressure is an initial condition used in the DBA SAFETY ANALYSES analyses to establish the maximum peak containment internal pressure.

The limiting DBAs considered, relative to containment pressure, are the LOCA and SLB, which are analyzed using computer codes designed to predict the resultant containment pressure transients. The worst case LOCA generates larger mass and energy release than the worst case SLB. Thus, the LOCA event bounds the SLB event from the containment peak pressure standpoint (Ref. I).

The initial pressure condition used in the containment analysis was 15.0 psia (0.3 psig). The containment analysis (Ref. 1) shows that the maximum peak calculated containment pressure, Pa, results from the limiting LOCA. The maximum containment pressure resulting from the worst case LOCA does not exceed the containment design pressure 15.0 psig.

The containment was also designed for an external pressure load equivalent to -1.5 psig.

There are four conditions which have a potential for resulting in a net external pressure on the containment:

From TS 3.6.5 (Containment Temperature) 3.6 CONTAINMENT SYSTEMS 3.6.5 Containment Air Temperature LCO 3.6.5 Containment average air temperature shall be:

a. 75°F and 100°F for the containment upper compartment, and

b. 100°F and 120°F for the containment lower compartment.

NOTES

1. The minimum containment average air temperature in MODES 2, 3, and 4 may be reduced to 60°F.

2. Containment lower compartment temperature may be between 120°F and 125°F for up to 90 cumulative days per calendar year provided lower compartment temperature average over the previous 365 days is less than 120°F. Within this 90 cumulative day period, lower compartment temperature may be between 125°F and 135°F for 72 cumulative hours.

APPLICABILITY: MODES 1, 2, 3, and 4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment average air A.1 Restore containment 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months temperature not within average air temperature to limits, within limits.

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. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

From TS 3.6.5 Basis (Containment Temperature)

APPLICABLE SAFETY ANALYSES (continued)

The limiting DBA for the maximum peak containment air temperature is an SLB. For the upper compartment, the initial containment average air temperature assumed in the design basis analyses (Ref. 1) is 100°F. For the lower compartment, the initial average containment air temperature assumed in the design basis analyses is 135°F. This resulted in a maximum containment air temperature of 317°F. The current environmental qualification temperature limit is 341°F.

The temperature upper limits are used to establish the environmental qualification operating envelope for both containment compartments. The maximum peak containment air temperature for both containment compartments was calculated to be within the current environmental qualification temperature limit during the transient. The basis of the containment environmental qualification temperature is to ensure the performance of safety related equipment inside containment (Ref. 2).

The temperature upper limits are also used in the depressurization analyses to ensure that the minimum pressure limit is maintained for both containment compartments following an event which has the potential to result in a net external pressure on the containment.

The containment pressure transient is sensitive to the initial air mass in containment and, therefore, to the initial containment air temperature.

The limiting DBA for establishing the maximum peak containment internal pressure is a LOCA. The temperature lower limits, 75°F for the upper compartment and 100°F for the lower compartment, are used in this analysis to ensure that, in the event of an accident, the maximum containment internal pressure will not be exceeded in either containment compartment.

Containment average air temperature satisfies Criterion 2 of 10 CFR 50.36 (Ref. 3).

LCO During a DBA, with an initial containment average air temperature within the LCO temperature limits, the resultant peak accident temperature is maintained below the containment environmental qualification temperature. As a result, the ability of containment to perform its design function is ensured. Two Notes to the LCO provide containment air temperature flexibility. Note 1 establishes that in MODES 2, 3, and 4, containment air temperature may be as low as 60°F because the resultant calculated peak containment accident pressure would not

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2011 MNS SRO NRC £xamination QUESTION 86 EPEO1 1 2.4.31 Large Break LOCA EOll GENERIC nowledge of annunciator alarms, indications, or response procedures. (CFR: 41.10/45.3)

Given the following conditions on Unit 2:

A Large-Break LOCA has occurred

The FWST switchover criteria has been reached

The crew has just implemented ES-I .3 (Transfer to Cold Leg Recirc)

Annunciator Windows D-2 (CONT SUMP LEVEL GREATER THAN 3 FT), on both 1AD-14 and 1AD-15 are DARK

Annunciator Windows D-l (CONT SUMP LEVEL GREATER THAN 2.5 FT), on both 1AD-14 and IAD-l5areUT

Both NS Pumps are RUNNING

NEITHER Containment Sump to ND/NS Isolation Valves (2N1-184B and 2N1-I 85A) can be opened Based on the conditions described above:

1) What would be the NEXT procedure/enclosure that would be implemented?

2) How would this transition affect monitoring of the Critical Safety Function Status Trees?

A. 1. ECA-I .1 (Loss of Emergency Coolant Recirc)

2. Implement the CSF Status Trees as required.

B. 1. ECA-I .1 (Loss of Emergency Coolant Recirc)

2. Monitor the CSF Status Trees for information only.

C. I. Enclosure 2 (Low Containment Sump Level), of ES-i .3

2. Implement the CSF Status Trees as required.

D. 1. Enclosure 2 (Low Containment Sump Level), of ES-i .3

2. Monitor the CSF Status Trees for information only.

Wednesday, April 20, 2011 Page 255 of 299

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2011 MNS SRO NRC Examination QUESTION 86 eneraI Discussion According to Step 2 of EP/2/A15000/ES-1 .3 (p2; Rev 23), CSF procedures should not be implemented until directed by this procedure.

According to EP-El (p154; Rev 18), this is done to call attention to the fact that the operator actions to realign the S/I system must be done in a rapid manner. This is because the amount of water in the FWST between the switchover setpoint and the empty point is limited, and the realignment of the S/I systems to cold leg recirculation must be done as quickly as possible. According to ECC-NI (p13; Rev 24), in the event of a major LOCA, the containment sump will receive water from the NC system break, ECCS injection flow through the break, and the ice condenser during ice melt. Annunciators inform the operator when containment sump level has reached greater than 2.5 and greater than 3 feet.

This lets the operator know that adequate volume is contained in the sump for the recirculation phase. At Step 3 of EPI2/AJ5000IES- 1.3 (p2; Rev 23), the operator is directed to check the CONT SUMP LEVEL GREATER THAN 3 FT annunciators on 2AD-14 or 2AD-l5, lit. If they are NOT lit the RNO directs the operator to check, if the NS Pumps are running, and if so, check if either alarm CONT SUMP LEVEL GREATER THAN 2.5 FT is lit on 2AD-14 or 2AD-15. If NS is running, the RNO will allow continuing with transfer to Cold Leg Recirc if the 2.5 FT alarm is lit. 2.5 ft is adequate sump level to operate pumps from sump and remaining FWST inventory will be quickly added to the sump by NS pump operations. According to EP-El (p158, Rev 18), Step 6 checks that the suction valves to the sump have been opened and that isolation of the ND pumps from the FWST is performed since these pumps may still be drawing from the FWST. If a sump suction does not open and can not be opened the associated ND pump is stopped to prevent depleting the FWST. If both ND pumps are off then Recirculation flow will not be achievable and Transition to ECA- 1.1 Loss of Emergency Coolant Recirculation is required. If neither, RB Sump to ND/NS Suction can be opened, according to EPI2/AJ5000/ES-l.3, (PS; Rev 25), Step 6.f RNO will direct the operator to implement EP/2/A15000/F-O (Critical Safety Function Status Trees), and to transition to EP/2/AJ5000IECA 1.1.

Jherefore the operator will monitor the CSFSTs and transition to the appropriate FRP as required upon transition to ECA 1.1.

Answer A Discussion CORRECT: See explanation above.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part (1) is correct and therefore plausible.

art (2) Plausible because the operator may incorrectly believe that they retain the ES-i .3 direction to monitor the CSFST for information only.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part (1) is plausible because the operator may incorrectly believe that with lower than expected Sump levels the transition to Enclosure 2 must be made. This would be correct if the NS Pumps are OFF, or if both sets of level lights were DARK.

Part (2) is correct and therefore plausible.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part (1) is plausible because the operator may incorrectly believe that with lower than expected Sump levels the transition to Enclosure 2 must be made. This would be correct if the NS Pumps are OFF, or if both sets of level lights were DARK.

Part (2) is plausible because if the operator believes that the transition to Enclosure 2 or ES 1.3 is correct then this would be correct as well and refore plausible.

Basis for meeting the KA The KA is matched because the operator must demonstrate the knowledge to asses annunciator alarms (i.e. low Containment Sump Level) as they apply to the Loss of Emergency Coolant Recirculation and the appropriate procedure.

Basis for Hi Cog This is a higher cognitive level question because the applicant must perform a level of analysis concerning the given indications and predict the mpact and determine the correct procedural course of action.

isis for SRO only ihe question is SRO-Only because the question cannot be answered by knowing system knowledge, immediate operator actions, or EOP Entry conditions, but rather requires that the operator assess plant conditions, and then prescribe a procedure or a portion of the procedure, including ecalling of a strategy within the procedure (i.e. monitoring of CSFs).

Thursday, April 14, 2011 Page 256 of 299

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2Afll MNS SRO NRCExamination QUESTION 86 Job Level Cognitive Level QuestionType Question Source SRO Comprehension BANK Bank 2989 Development References Student References Provided EPI2IAI5000/ES-l.3 (p2, 5; Rev 26)

EP-El (p 154-158; Rev 25)

ECC-NI (p13; Rev 25)

EPEOII 2.4.31 -LargeBreakLOCA EPEO 11 GENERIC Knowledge of annunciator alarms, indications, or response procedures. (CFR: 41.10 / 45.3)

[iiZ9 Comments: RemarkslStatus I_______________

Thursday, April 14, 2011 Page 257 of 299

Question 86

References:

From EOP ES 1.3 (Transfer to Cold Leg Recirc) Pg 2 of 43 (Rev 26)

MNS TRANSFER TO COLD LEG RECIRC PAGE NO.

EPIILAJ5000IES- I .3 2 of 43 7%ffj Rev. 28

ow/:::ED s;oi.s :os uo csa:D C. jfr Actions

1. Have STA monitor Foldout page.

2 Perform this EP without delay:

CSF procedures should not be implemented until directed by this procedure.

Double 3-way communicatien is not required..

3. Check containment surup level by jj both alarms are dark THEN perform checking at least one of the following the foIowing:

alarms LIT:

a. IF both NS pumps are off, THEN GO

. CONT SUMP LEVEL GREATER THAN IQ Enclosure 2 (Low Containment 3 FT on 1AD-14 LIT- Sunip Level)..

OR b. If either alarm CONT SUMP LEVEL GREATER TI-iAN is lit on

. CONT SUMP LEVEL GREATER THAN 1AD-14 or 1AD-15, TI-LEN GO TO 3FTon lAD-iS- LIT - St&p4.

This step provIdes estructor plausbihty. L

& Check KC flow to each ND Hx Perform the following:

GREATER THAN 5)OO GPM.

a. CLOSE the following v3lves:

1 KC-50A (Tm A Aux Bldg Non Ess Sup Isol)

IKC-535 (Tm 6 Aux Bldg Non Ess Sup isol).

b. OPEN the following valves:

1KC-56A (1A ND H KC Inlet Isol)

IKC-8l5 (lB ND Hx KG Inlet Isol).

From EOP ES 1.3 (Transfer to Cold Leg Recirc) Pg 3 of 43 (Rev 26)

MNS TRANSFER TO COLD LEG RECIRC PAGE NO.

EPIIIAI5fJOOIES-1.3 3 of 43 Rev. 26 UNIT 1 E8PON5 CN5E 1OT C3:A:UD

5. Reset the following:

a. S/I. a. Reset S/I PER EPII!AJ5000JG-i (Generic Enclosures), Enclosure 23 (Local Reset of S/I Signal) wfiuie continuing in this procedure

b. Sequencers. b. Dispatch operator to open affected sequencer control power brealec A Train - I EVA breaker 6 5 Train - IEVDD reaRer 6.

6. Align ND System for recirc as follows:

a. Check 1NI-185A (1A ND Pump Suction a. Perform the following:

From Cont Sunip Isol) OPEIL

1) Ploce control permissive in 5ypas&

and OPEN lNl.185A.

2) Jr 1NI-165A is opening, ILIEN wait up to 30 seconds to allow valve to open.

3) Jr lNl-1854 is full open, TFjN GO

] Step 6.b.

4) j 1NI-185A is closed OR interrned1ate, IHEJI perform the following:

a) Stop IA ND peump b) GO TO Step 6..

From EOP ES 1.3 (Transfer to Cold Leg Recirc) Pg 4 of 43 (Rev 26)

MNS TRANSFER TO COLD LEG RECIRC PAGE N0 EPII/AI5000IES-1.3 4 of 43 Rev. 26 urr 1 T:oN/E::EcE RCN5E cEPONE

6. (Continued)

c. Check lNl-1 845 (18 ND Pump Suction c. Perform the following:

From Cont Sump Isol) OPEN.

1) Place control permissive in Bypass and OPEN INI-1 848.

2) ffl INI-1 846 is opening, THEN wait up to 30 seconds to allow valve to open.

_3) IF1NI-1845isfullopenTI-IENGO mstep 6.d.

4) IE INI-184B is closed QR intermediate, TUEJ4 perform the following:

_a) StoplBNDpump.

_b) GOTOStep6.e.

d. Check 15 ND pump- ON. d. Start 18 ND pump.

e. Enable power disconnect and CLOSE 1FW-27A (Unft I FWST to ND Pumps Isol).

From EOP ES 1.3 (Transfer to Cold Leg Recirc) Pg 2 of 43 (Rev 26)

MNS TRANSFER TO COLD LEG RECIRC PAGE NO.

EP/1/AI5000IES-1.3 5 of 43 26 RCV ou/x?EcED Rspos so;s ;oT cz:;

6. (Continued)

f. Check any ND pump - 0N f. j both ND pumps are off, THEN perform the following:

1) IEbothl1Spumpsareon,Th1EL stop one NS pump as follows:

a) On train to be stopped, reset Containment S pray.

b) Stop one NS pump.

IOTE If EPI1INOOO!fR-Z1 (Response To High CQrtainment Pressure) is required, it my 1 on,plte as time allows.

2) EPIIIA!50001F-O (Critical Safety Function Status Trees) may now be implemented.

3) GOTOEPII/A15000!ECA-1.1 (Loss Of Emergency Coolant Recirc).

From Lesson Plan OP..MC-EP-E1 Pg 155 (Rev 25) 0.1. Detailed Description of Procedural Steps STEP I Have STA monitor foldout page.

PURPOSE: To instruct the STA that the foldout page should be monitored.

BASIS: The foldout page provides a list of important items that should be continuously monitored. If any of the parameters exceed their limits, the appropriate operations should be initiated.

Due to the time criticality of performing the swapover evolutions, as described in the UFSAR, the operating crew does not have time to monitor these parameters.

Giving the foldout page to STA to monitor saves the crew approximately 30 seconds by not reviewing foldout page. There is nothing on the foldout page that the crew must be aware of to successfully transfer to cold leg recirc for design base event.

STEP 2 Perform this EP without delay. CSF procedures should not be implemented until directed by this procedure. Double 3-way communication is not required.

PURPOSE: To call attention to the fact that the operator actions to realign the S/I system must be done in a rapid manner.

BASIS: Since the amount of water in the FWST between the switchover setpoint and the empty point is limited, the realignment of the S/I systems to cold leg recirculation must be done as quickly as possible.

From Lesson Plan OP-MC-EP-E1 Pg 157 (Rev 25)

STEP 3 Check containment sump level by checking containment sumps alarms lit on either AD-14 or AD-15 CONT SUMP LEVEL GREATER THAN 3 FT.

PURPOSE: To ensure there is sufficient level in the sump to support the transfer to Cold Leg Recirc.

BASIS: Step 3 specifies to use the sump level alarm when checking for minimum sump level. Only one alarm has to be lit for step to confirm adequate level.

The 3 ft alarm will be checked first. If either 3 FT alarm is lit, operator will continue with transfer to Cold Leg Recirc. This should be expected response.

Actions in step 3 RNO depend on whether NS is running.

If NS is off, the RNO will send you to Enclosure 2 (with sump level less than 3 ft). Note that sump level may be below 2.5 ft alarm (inadequate for swappover) or may be between 2.5 and 3 ft (marginal sump level). Enclosure 2 will ensure protection of pumps and restore margin in the sump.

If NS is running, the RNO will allow continuing with transfer to Cold Leg Recirc if the 2.5 FT alarm is lit. 2.5 ft is adequate sump level to operate pumps from sump and remaining FWST inventory will be quickly added to the sump by NS pump operations. If level is below 2.5 ft though, the RNO will send you to Enclosure 2 (Low Sump Level). will direct operator to take actions depending on event (tornado, LOCA outside containment, or LOCA inside containment with low sump level). Actions for low sump level remain similar, including use of Alternate Transfer to Cold Leg Recirc if required. We will ensure adequate sump level before we allow continued swappover to the sump.

Note that if an Orange or Red path Procedure is in effect upon transition out of ES-i .3, it takes priority over any other procedure including ECA-i .1.

STEP 4 Check KC flow to ND heat exchangers GREATER THAN 5000 GPM.

PURPOSE: To ensure KC flow to the ND heat exchangers.

BASIS: This step assumes that the ND heat exchangers are used for heat removal during the post accident recirculation phase and that either KC flow has been automatically provided to the heat exchangers or the operator has manually established KC flow prior to the switchover. If KC flow had not previously been established, then it should be established at this time.

Parent Question (Bank 2989):

Given the following conditions on Unit 2:

A Large-Break LOCA has occurred

The FWST switchover criteria has been reached

The crew has just implemented ES-I .3 (Transfer to Cold Leg Recirc)

Annunciator Windows D-2 (CONT SUMP LEVEL GREATER THAN 3 FT), on both IAD-14 and IAD-15 are DARK

Annunciator Windows D-1 (CONT SUMP LEVEL GREATER THAN 2.5 FT), on both IAD-14 and IAD-15 are LIT

Both NS Pumps are RUNNING

NEITHER Containment Sump to ND/NS Isolation Valves (2N1-184B and 2N1-185A) can be opened Based on the conditions described above:

1. What would be the NEXT procedure/enclosure that would be implemented?

2. How would this transition affect monitoring of the Critical Safety Function Status Trees?

A. 1. ECA-1.I (Loss of Emergency Coolant Recirc)

2. Implement the CSF Status Trees as required.

B. 1. ECA-tl (Loss of Emergency Coolant Recirc)

2. Monitor the CSF Status Trees for information only.

C. 1. Enclosure 2 (Low Containment Sump Level), of ES-I .3

2. Implement the CSF Status Trees as required.

D. 1. Enclosure 2 (Low Containment Sump Level), of ES-l .3

2. Monitor the CSF Status Trees for information only.

ANSWER: A

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2011 MNS SRO NRC Examination QUESTION 87 APEO56 AA2.72 Loss of Offsite Power ility to determine and interpret the following as they apply to the Loss of Offsite Power: (CFR: 43.5 /45.13) jixiliary feed flow Given the following events and conditions on Unit 1:

Unit was operating at 100% RTP and experienced a LOOP at 0200

The resulting transient caused a I B feedwater line break inside Containment

lB DIG failed to start and the TD CA pump is unavailable

1A CA pump tripped on overcurrent

FR H.1 (Response to Loss of Secondary Heat Sink) has been entered and feed and bleed of the NC system was initiated at 0230

The lB CA pump has been returned to service and available as a source of feedwater Given the following conditions at 2:45 AM:

Containment pressure = 3.5 PSIG

Core exit T/Cs are stable at an average value = 560 °F Indication A Loop B Loop C Loop D Loop S/GWRlevel(%) 19 0 16 8 Into which Steam Generator(s) should CA flow be restored and what limitation, if any is required on CA flow rates?

A. S/GC ONLY No limitation on CAflow rate B. S/GsAORS/GC No limitation on CA flow rate C. S/GC ONLY 100 gpm limitation on CA flow rate D. S/GsCANDS/GD 100 gpm limitation on CA flow rate Wednesday, April 20, 2011 Page 258 of 299

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2011 MNS-SRO NRC Examination - QUESTION 87-eneral Discussion In this question, a LOOP has resulted in a loss of heat sink and entry into FRP H. 1 (Response to loss of secondary heat sink). The applicant is told that feed and bleed has been established and that one of the aux feedwater pumps has been recovered. This sets up a procedural flowpath to where the crew would be performing an evaluation of hot dry S/Gs prior to reestablishing feedwater flow. He is then given a set of indications that show that 3 of 4 of the S/Os have reached dry out conditions per step 7h in FR H.1. (17% ACC). From the indications given, the 1A S/G is available to receive full flow but the lB CA pump can only be aligned to the C and D S/Os. Per the procedure, flow should be established to the CS/G (Highest level) at a rate less than 100 GPM.

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part (1) is correct and therefore plausible.

Part (2) Plausible because the applicant may miss the fact that ACC values are in effect and consider this generator to not be hot and dry (12% is the non ACC value). If containment pressure had remained below 3 PSIG this answer would be correct.

Answer B Discussion iNCORRECT: See explanation above.

PLAUSIBLE:

Part (I) is correct and therefore plausible.

Part (2) Plausible because the applicant may miss the fact that ACC values are in effect and consider this generator to not be hot and dry (12% is the non ACC value). If containment pressure had remained below 3 PSIG this answer would be correct. This is correct for the 1A S/G but the applicant may overlook the alignment restrictions associated with only having the lB CA pump available.

.nswer C Discussion 9RRECT: See explanation above.

Answer D Discussion fNCORRECT: See explanation above.

PLAUSIBLE:

Part (1) is correct and therefore plausible.

Part (2) Plausible because the applicant would be correct in his assessment that both of these 5/Os are hot and dry and feeding them both is possible but overlook the procedural requirement to feed the 5/0 with the highest level.

Basis for meeting the KA The K/A is matched because the applicant must evaluate conditions resulting from a loss of offsite power. Specifically, based on interpretation of*

given indications he must determine both where and how much aux feed flow would be established.

Basis for Hi Cog This is a higher cognitive level question because the applicant must perform a level of analysis concerning the given indications and predict the impact and determine the correct procedural course of action.

Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev ldated 03/11/2010 for screening questions linked to IOCFR55.43(b)(5) (Assessment and selection of procedures):

1) This question can NOT be answered by knowing systems knowledge alone. This is strict procedure knowledge, This is not covered during systems training or discussed in a systems lesson plan.

2) This question can NOT be answered by knowing immediate operator actions. There are no immediate actions in FR-H. 1.

3) This question can NOT be answered by knowing the entry conditions for AOPs. The steps to be taken by the crew are not based on the entry conditions provided.

This question can NOT be answered by knowing the purpose, overall sequence of events, or overall mitigative strategy of the AOPs. The stion is based on knowledge of specific procedure content.

) The question requires the applicant to have in-depth knowledge of specific steps within FR H. 1. Specifically, it requires the applicant to recall that the Step 7h is in effect requiring evaluating requirements both identification and feeding a hot dry S/G. These requirements are further nded by a procedural note which the applicant must recall to successful answer the question. For these reasons, this is SRO level knowlej Thursday, April 14,2011 Page 259 of 299

Question 87

References:

From FR-H.1 (Response To Loss of Secondary Heat Sink) Pg 5 MNS RESPONSE TO LOSS OF SECONDARY HEAT SINK PAGE NO.

EP/1/N5000/FR-H.1 5 of 129 UNIT 1 Rev. 15 Ac:cN/ExPE:E RESPONSE RESPONSE flOE DETAINED 7 (Continued)

e. Check total flow to SIG(s) GREATER

- e. Perfomi the following:

THAN 450 GPM.

1) lEonlyone MD CA pump is on, AND its discharge path cannot be reopened to its associated S/Gs, THEN evaluate aligning flow to another S/G through MD CA train N cross-tie PER Enclosure 5 (MD CA Pump Train NB Cross-tie Alignment).

In the scenario given step 38 would have been implemented after 2) iF any CA pump is started, D establishing teed and bleed, therefore Lein1ftrnented this step would be followed wten the I B CA pump was restored

3) iF any feed flow to at least one SJG is indicated, THEN perform the following:

a) Maintain flow to restore NIR level to greater than 11%

(32% ACC).

b) WHEN N/R level is greater than 11% (32% ACC), THEN RETURN JO procedure and step in affect.

c) Q J Step 8.

4) IF no feed fow indicated, THEN perform the following:

From FR-H.1 (Response To Loss of Secondary Heat Sink) Pg 6 MNS RESPONSE TO LOSS OF SECONDARY HEAT SINK PAGE NO.

EP/1IN50001FR-I-i1 6 of 129 UNIT 1 Rev. 15 AC:CN/EXEC:ED ci*sz scxs: IOT DSAINED

7. (Continued)

f. Check feed and bleed -

f. RET URN TQ procedure and step in ESTABLISHED PER STEPS 23 effect.

through 27.

_g. QIQStep40.

h. Check any S/C WIR level LESS THAN

- h. Perform the following:

12% (17% ACC).

1) THROTTlE OPEN CA control valves to establish CA flow to S/Gs.

_2) GOTOStep4O.

NOTE

It may be preferable to feed 18 or IC SJG first, to maintain steam supply form CA pump Selecting S/G with highest level will reduce risk of thermal shock to S!G when reestablishing feed flow.

i. Check core exit TICs STABLE OR

i. Perform the following:

GOING DOWN.

1) THROTTLE OPEN CA control valve to one S!G to establish flow rate required to lower core exit T/Cs.

2) W core exit TiCs continue to go up, THEN THROTTLE OPEN CA control valve to feed another SIG as required to lower core exit TIC&

3) GO TO Step 7.m.

j. Slo4y ThROTTLE OPEN CA control valve to one SIG to establish feed flow less than or equal to 100 GPM.

k. Maintain feed flow rate less than or equal to 100 GPM until S!G WR level is greater than 12% (17% ACC).

I. WHEN S1G WIR level is creater than 12% (17% ACC), THEN feed flow may be raised greater than 100 GPM.

m. Check SIG WJR levels on intact S!Gs m. Q 12 Step 7.o.

with feed flow isolated ANY GREATER THAN 12% (17%ACC).

Parent Question (2005 NRC Exam Q15):

Unit 1 has experienced a feedwater line break inside containment and a total loss of feedwater at 2:00 AM. FR-Hi, Response to Loss of Secondary Heat Sink has been entered and feed and bleed of the NC system was initiated at 2:30 AM. Shortly after opening the PORVs, the 1 B CA pump is returned to service and a source of feedwater is available. The operators are directed to restore steam generator level for a heat sink per FR-H.1.

Given the following conditions at 2:45 AM:

Indication A Loop B Loop C Loop D Loop S/G WR level (%) 19 9 16 11

Containment pressure = 3.5 psig

Core exit T/Cs are stable at an average value = 560 °F Into which steam generator(s) should CA flow be restored and what limitations are required on CA flow rates?

A. S/GD No limitation on CA flow rate B. Either S/Gs A or S/G C No limitation on CA flow rate C. S/GC 100 gpm limitation on CA flow rate D. Both S/Gs C and D 100 gpm limitation on CA flow rate ANSWER: C

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2011 MNS SRO NRC Examination. QUESTION 88 APEO57 2.1.23 Loss of Vital AC Electrical Instrument Bus

\PE057 GENERIC oility to perform specific system and integrated plant procedures during all modes of plant operation. (CFR: 41.10 / 43.5 / 45.2 /45.6)

Given the following conditions on Unit 1:

Unit 1 is operating at 100% RTP

Power Panel Board EKVA feeder breaker opens due to a breaker fault

AP-1 5 (Loss of Vital or Aux Control Power) has been implemented

1) Per AP-15, the crew will restore power to the affected bus per

2) The indication used to verify that those actions have been successful is the Which ONE (1) of the following completes the statement above?

A. 1. actionsdirectedinAP-15

2. top row of channel status lights - NORMAL B. 1. actions directed in AP-15

2. switch indication on any pump powered from 1 ETA LIT -

C. 1. OP/1/A/6350/OO1A (125VDC/12OVAC Vital I&C Power System)

2. top row of channel status lights NORMAL -

D. 1. OP/1/A/6350/OO1A (125VDC/12OVAC Vital l&C Power System)

2. switch indication on any pump powered from I ETA LIT -

Thursday, April 14, 2011 Page 261 of 299

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2011 MNSSRO NRC Examination- QUESTION 88

_eneral Discussion In this question, the applicant is given a situation where power has been lost to Vital AC instrument bus 1EKVA. AOP AP-l 5 (Loss of Vital or Aux Control Power) has been implemented. Per AP-15, RHR is checked first (Not in service in this scenario), a check is them made of L/D which should still be service. MSIVs are checked open and then the vital AC panel boards are checked energized by checking the top row of channel status lights- NORMAL, it they are not, which for this question they would not be, then L/D is addressed and VCT level control. The procedure then checks other AC and DC panels energized and if no others are found then the reader is directed step 25 where if one of the vital AC panel boards is deenergized (EKVA) then the reader is sent to step 27. Per Step 27, the crew is sent to Enclosure 10 of AP-15 to Restore power to 1EKVA.

Answer A Discussion LC011ECT See explanation above.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part (1) is correct and therefore plausible.

Part (2) is plausible because this is correct for checking whether IEVDA is energized and the applicant may confuse these two because there are a number of different checks listed throughout the AP and it would be possible for the applicant to get them mixed up.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part (1) is plausible because this is correct for almost any other situation dealing with this bus. Annunciator lAD-i 1 GI (120 VAC ESS PWR Channel A TRBL) would be in alarm for this situation and the ARP for this alarm refers the operator to this procedure. Additionally, this

-ocedure would provide guidance to reenergize this bus but would not be utilized in the scenario given. It would be reasonable for the licant to believe that this would be the correct choice.

(2) is correct and therefore plausible.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part (1) is plausible because this is correct for almost any other situation dealing with this bus. Annunciator lAD-i 1 Gi (120 VAC ESS PWR Channel A TRBL) would be in alarm for this situation and the ARP for this alarm refers the operator to this procedure. Additionally, this procedure would provide guidance to reenergize this bus but would not be utilized in the scenario given. It would be reasonable for the applicant to believe that this would be the correct choice.

Part (2) is plausible because this is correct for checking whether 1 EVDA is energized and the applicant may confuse these two because there are a number of different checks listed throughout the AP and it would be possible for the applicant to get them mixed up. I Basis for meeting the KA This K/A is matched because the applicant is presented with set of conditions where a Vital VC Electrical Bus (1 EKVA) has been lost and must demonstrate specific procedural knowledge (ability to perform) of the procedure required to restore the bus. Additionally, he must choose between two possible procedures, both of which would address the problem. (Integrated plant procedures).

Basis for Hi Cog Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev ldated 03/1112010 for screening questions linked to 10CFR55.43(b)(5) (Assessment and selection of procedures):

1) This question can NOT be answered by knowing systems knowledge alone. This is strict procedure knowledge. This is not covered during vstems training or discussed in a systems lesson plan.

This question can NOT be answered by knowing immediate operator actions. The immediate actions from AP- 11 address attempting to isolate stuck open Pressurizer PORV. However, the actions to be taken by the crew as pressure continues to decrease are not part of the immediate actions.

3) This question can NOT be answered by knowing the entry conditions for AOPs. The steps to be taken by the crew are not based on the entry conditions provided.

4) This question can NOT be answered by knowing the purpose, overall sequence of events, or overall mitigative strategy of the AOPs. The Wednesday, April 20, 2011 Page 262 of 299

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2011 MNS SRO NRC Examination QUESTION 88 question is based on knowledge of specific procedure content.

1 The question requires the applicant to have in-depth knowledge of specific steps within AP-1 5. Specifically, it requires the applicant to recall at the Step 27 directs restoration of power from Enc 10 of this procedure verses using the Normal OP which would accomplish the needed alignment. Additionally, the applicant must recall what indications would be checked to determine that this alignment has been successful.

Job Level Cognitive Level QuestionType Question Source SRO Memory NEW Development References Student References Provided

-15 (Loss of Vital or Aux Control Power) Pg 15 of 279 AP- 15 (Loss of Vital or Aux Control Power) Pg 16 of 279 APEO57 2.1.23 Loss of Vital AC Electrical Instrument Bus APEO57 GENERIC Ability to perform specific system and integrated plant procedures during all modes of plant operation. (CFR: 41.10/43.5 / 45.2/45.6) 401-9 Comments: RemarkslStatus Thursday, April 14, 2011 Page 263 of 299

Question 88

References:

From AP-1 5 (Loss of Vital or Aux Control Power) Pg 15 of 279 (Rev 22)

MNS LOSS OF VITAL OR AUX CONTROL POWER PAGE NO.

AP/1/N5500/15 15 of 279 ReV. 22 u:crr 1 Ill liii liii 11111111 liii l iiii pill 11111111 III I

23. Check all Vital AC panelboards Step 2.

energized as follows OFor IEK/A:

= Top row of channel status lights -

NORMAL Fcr1EKVB;

_: Second row of channel tatu lights -

NORMAL For 1EKVC:

Third row of channel status lights -

NORMAL

For IEKVD:

Bottom row of channel status lights.

NORMAL.

_24. GOTOStep39.

25. Check all Vital AC panelboards Perform the following:

energized as follows:

a. Announce toss of affected bus on 2 For 1EKVA: paging system.

_: Top row of ahGnnel etatus lights - b. IQ.Step27 NORMAL

For IEKVB:

Second row of chaainel status lights -

NORMAL

For 1EKIC:

Third row of channel atatu liahts

From AP-15 (Loss of Vital orAux Control Power) Pg 16 of 279 (Rev 22)

M NS LOSS OF VITAL OR AUX CONTROL POWER PAGE NO.

API1JN5500I1 5 16 of 279 Rev. 22 UNIT 1 IIIIIIiIIIIIIIIIIlIIII IIltII4ltIHIIIlIlI1t

26. jQStep39.

27. Dispatch operator to perform the foHowinj while continuing with this procedure:

a. Detem,ine exact cause of alami(s).

b. Restore power to the affected pQnelbard using the following enclosures:

Enclosure 10 (Restoring Power To 1 EKVA)

OR

_: Enc[osure 11 (Restoring Power To 1 EKVB)

OR

_: EncLosure 12 (Restoring Power T01EKVC)

OR

_: EncLosure 13 (Restoring Power To 1 EKVD).

28. WHEI4 power is restoreI to all affected bus(s) THEN GO TO Step 72.

29. Cheek both of the following Vital DC TO Step 34.

panelboards energized as follows:

For IEVDA:

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2011MNS SRO NRC Examination QUESTION -89 -

SYS062 2.1.20 AC Electrical Distribution System YS062 GENERIC ability to interpret and execute procedure steps. (CFR: 41.10 I 43.5 I 45.12)

Given the following plant conditions:

A complete Loss of Off-Site power has occurred

Unit 2 DIGs started and sequenced normally

Unit 1 DIGs did NOT automatically start and attempts to start them have been unsuccessful

Subsequently, a SGTR occurs on the IA SIG Which ONE (1) of the following describes the procedure that will be used to isolate the ruptured S/G in this situation AND the procedural guidance regarding WHEN the ruptured S/G will be isolated?

A. E-3 (Steam Generator Tube Rupture) is used to isolate the ruptured SIG AS SOON AS it is identified.

B. E-3 (Steam Generator Tube Rupture) is used to isolate the ruptured SIG ONLY AFTER ruptured SIG NR level is greater than 11 %.

C. ECA-O.O (Loss of All AC Power) is used to isolate the ruptured SIG AS SOON AS it is identified.

D. ECA-O.O (Loss of All AC Power) is used to isolate the ruptured SIG ONLY AFTER ruptured SIG NR level is greater than 11 %.

Thursday, April 14, 2011 Page 264 of 299

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2011 MNS SRO NRC Examination QUESTION 89 eneral Discussion If at least one DIG had started on Unit 1, ECA-0.O would have been exited and the ruptured SIG would have ultimately been isolated in E-3 when ruptured SIG level> 11% was established.

For this particular situation, since neither DIG can be started on Unit 1, the crew remains in ECA-O.0 and the ruptured SIG is isolated in accordance with Enclosure 14 (Faulted or Ruptured SIG Isolation). Unlike E-3, the body of ECA-0.0 and Enclosure 14 do NOT direct restoring ruptured SIG level to> 11% prior to isolating the ruptured SIG.

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible if the applicant does not recall that there is guidance in ECA-0.0 for isolating a ruptured SIG and concludes that the isolation is not performed until ECA-0.0 is exited and transition is made to E-3. Isolating the SIG as soon as it is identified is the course of action taken for isolation in ECA-O.0 and hence it would be plausible for the applicant to choose this action.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible if the applicant does not recall that there is guidance in ECA-O.0 for isolating a ruptured SIG and concludes that the isolation is not performed until ECA-O.0 is exited and transition is made to E-3. lAW E-3, the ruptured SIG is NOT isolated until ruptured SIG NR level Answer C Discussion CORRECT: See explanation above.

Answer D Discussion CORRECT: See explanation above.

LAUSIBLE:

ECA-0.0 is the correct procedure since at least one DIG has not been restored. Restoring ruptured SIG level >11% prior to isolation is plausible since the is the normal method for isolating a ruptured SIG as directed in E-3.

Basis for meeting the KA The K/A is matched since a loss of all AC (AC Electrical Distribution) has occurred and the applicant demonstrates the ability to execute procedure steps associated with the AC Electrical Distribution System malfunction by demonstrating detailed procedure step knowledge of ECA 0.0.

Basis for Hi Cog This is a higher cognitive level question because it requires multiple mental steps. The SRO applicant must first analyze the given conditions to determine which procedure is appropriate for isolating the ruptured SIG. The applicant must then recall from memory the difference between ECA-0.0 and E-3 regarding when the SIG must be isolated (i.e. whether or not ruptured SIG NR level must be >11% prior to commencing the isolation).

Basis for 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 03111/2010 for screening questions linked to 10CFR55.43(b)(5) (Assessment and selection of procedures):

1) The question can NOT be answered solely by knowing systems knowledge. The applicant must have detailed knowledge of procedural steps in ECA-0.0 to arrive at the correct answer.

2) The question can NOT be answered by knowing immediate operator actions. There are NO immediate actions associated with ECA-0.0 or E-3.

3) The question can NOT be answered solely by knowing entry conditions for AOP or direct entry conditions for EOPs. The knowledge required to correctly answer this question is not associated with the entry conditions for ECA-0.0 or E-3.

4) The question can NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of the procedure.

This question requires detailed knowledge of when ECA-0.0 or E-3 would be applicable for the ruptured SIG isolation and how the procedure lifer with regards to the isolation requirements.

This question requires the SRO applicant to assess plant conditions and determine which procedure (ECA-0.0 or E-3) will ultimately be used

)erform the SIG isolation. It also requires the applicant to have detailed knowledge of how the steps in ECA-0.0 and E-3 differ regarding the

[isolation.

Wednesday, April 20, 2011 Page 265 of 299

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2011 MNS SRO NRC Examination___QUESTION_89 Job Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Development References Student References Provided

References:

EP/1/A15000/E-3, Steam Generator Tube Rupture, EP/l/A15000/ECA-O.0, Loss of All AC Power Learning Objectives: OP-MC-EP-ECA-0.0 Objective 6, OP-MC-EP-E3 Objective 6 SYSO62 2.1.20 AC Electrical Distribution System SYSO62 GENERIC Ability to interpret and execute procedure steps. (CFR: 41.10 / 43.5/45.12) 401-9 Comments: RemarkslStatus Thursday, April 14, 2011 Page 266 of 299

Question 89

References:

From ECA-O.O, Loss of All AC Power:

MNS LOSS OF ALL AC POWER PAGE NO.

ED!1/Ar50Q0EA0.0 19 of 190 Rev. 28 UNIT I kcTIow/E:PcTEo SPONSE ESPDNSE NO O3TANED

22. Check if SIG tubes intact as follows: IF SIG levels qoincj up in an uncontrolled

h. manneranv EMF normal, THEN

The following secc>ndary EMFs - perform the following:

NORMAL r

a. Attempt to identify the rup:ured SIG(s).

b. WULN ruptured 51(3(s) identified, THEN isolate the ruptured S1G(s) as a *l EMF-34(L) (s:G Sample Ito Range)) follows:

1) Isolale ruptured S/C(s) PER Enclosure 14 (Faulted or 1EMF-25(SIGO) Ruptured SIG Isolation).

1EMF-26(SIGC) /2) )&1ffi14 each ruptured S/G(s) pressure less than 1092 PSG, 1EMF-27(S/GD). THEN perform the following:

-.---z--

a) &isure wptared S!G(s) SM PflRVl C)Ffl b) IE ruptured S/G(s) SM PORV cannot be dosed, ifiEN perform the following:

(1) CLOSE SM PORVisolatkri vlvG.

(2) jf SM FORVisolation vve cannot be closed. THEN dispatch operator to CLOSE SM PORV isuiatk.ni valve.

3) Ensure ruptured SIG(s) remains isolated throughout this procedure.

CAUTION Restoring vital batterj chargers in the next step is a time critical action.

NOTE The enclosure in next step has control roorr and local actions. It should be rdintcd frnm ctnntrnl rcknm h/ Confrcil Room Siprvisnr cr other lictnd operator.

23. Align DC Busses LEE abnormal procedure APJIIAJS500!07 (Loss of Electrical Power), Enclosure 7 (DC Bus Alignment).

MNS LOSS OF ALL AC POWER PAGE NO.

EP/1IN50O0IEA-0.0 ncIosure 14- Page 1 of 4 UNIT 1 Faulted or Ruptured SIG Isolation ev.

1. Maintain Steam flow to TD CA pump from at least one 51G.

2. Isolate CA flow to faulted or ruptured SIG(s) as follows:

For 1A S/G, perform the following:

CLOSE the following valves:

1CA-64A6 (Ui TD CA Pump Disch To 1A SIG Control)

. 1 CA-60A (1A CA Pump Disch To 1A SIG Control).

Dispatch operator to CLOSE the following valves:

. 1CA-62A (1A CA Pump Disch To 1A SIG Isol) (Unit 1 exterior doghouse, 750+12, DD-44, southeast corner)

. 1CA-66AC (Ui TD CA Pump Disch To IA S/G lsol) (Unit 1 exterior doghouse, 750+6, FF-44, 4 ft from inner wall, 8 ft from column DD-44).

IF exterior docjhouse not accessible, QR CA cannot be isolated, THEN dispatch operator to unlock and CLOSE the following valves:

. 1 CA-63 (Unit 1 TD CA Pump Disch To IA SIG Control Inlet IsoL) (Unit I CA pump rm, 716+9, BB-51, above door to TD CA pump, 4 ft south of 1A CA pump)

. 1CA-59 (1A CA Pump Disch To 1A S!G Control inlet Isol) (Unit 1 CA pump rm, 71 6+10, CC-SO, above 1 B CA Pump).

From E-3, Steam Generator Tube Rupture:

MNS STEAM GENERATOR TUBE RUPTURE PAGE NO.

EPI1IN5000IE-3 3 of 76 Rev. 21 UNIT 1 ACTION/EXPECTED RESPONSE RESPONSE NO CBTA:NED 4.

a. a. WHEN ruptured S!G pressure is less than 1092 PS IC, THEN perform the following on affected SM PORV:

1) Check SM PORV closed.

2) IE SM PORV is still open. THEN CLOSE its manual loader.

3) IE SM PORV is still open, THEN perform the following:

a) CLOSE SM PORV isolation valve.

b) IF SM PORV isolation valve cannot be closed, ThEN dispatch operator to CLOSE SM PORV isolation valve.

MNS STEAM GENERATOR TUBE RUPTURE PAGE NO.

EP11IAJ5000JE-3 4 of 76 UNIT 1 Rev. 21 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

4. (Continued)

b. b. Isolate TO CA pump steam supply from ruptured SIG as follows:

1) IE TO CA pump is the only source of feedwater, IUEt maintain steam flowto it from at least one SJG.

2) Ensure operators dispatched in next step imm1i1y notify Control Room SupeMsor when valves are closed.

3) Immediately dispatch two operators to concurrently venfy (CV), unlock and CLOSE valves on ruptured S1G(s):

ForlBS/G:

a 1SA-78(1BS/GSMSupplyto Unit 1 TO CA Pump Turb Loop Seal Isol) (Unit 1 interior doghouse, 767+10, FF53)

. 1SA-2(1BSIG SM Supply to Unit 1 10 CA Pump Turb Maint Isol) (Unit I interior doghouse, 767+12, FF-53).

For1CS!G:

ISA-77 (lCSIGSMSupplyto Unit I U) CA Pump Turb Loop Seal Isol) (Unit 1 interior doghouse, 767+1 0, FF-53)

. 1SA-t (1C SIG SM Supply to Unit 1 10 CA Pump Turb v1aint Isol) (Unit I interior doghouse, 767+10, FF-53, above ladder).

4) IFATANYTIME local closure of SA valves takes over 8 minutes, THEN isolate TO CA pump steam supp[y EJR Enclosure 2 (Tripping TO CA Pump Stop Valve or Alternate Steam Isolation).

MNS STEAM GENERATOR TUBE RUPTURE PAGE NO.

EP1IIA15000IE-3 5 of 76 Rev. 21 UNIT 1

?LCTION/EXPECTED RESPONSE RESPONSE N0 oBrA:NED

4. (Continued)

1) 1) Perform the following:

a) CLOSE valve(s).

b) CLOSE 186-123 (1A SIG

. B lowdown Throttle Control)

2) For I B S/G: 2) Perform the following:

. 1 B8-2B (I B S/G Slowdown ont a) CLOSE valve(s).

Outside Isol Control) b) CLOSE 188-124 (18 SIG

. 1 BB-6A (B S/G 88 Cont Inside Slowdown Throttle Control).

lsol).

3) For IC S!G: 3) Perform the following:

a 1 BB-3B (1 C S/G Blowdown Cont a) CLOSE valve(s).

Outside Isol Control) b) CLOSE 188-125 (IC SIG

. 1 BB-7A (C S/G 88 Cont Inside B lowdown Throttle Control).

lsol).

4) For 1 D S!G: 4) Perform the following:

. 1 BB-4B (1 D SIG Slowdown Cont a) CLOSE valve(s).

Outside Isol Control) b) CLOSE 188-126 (1D S/G a 1 B8-8A (D S/G 88 Cont Inside Slowdown Throttle Control).

Isol).

d._

. 1SM-89 (B SM Line Drain Isol)

. 1SM-95 (C SM Line Drain Isol)

ISM-lOl (D SM Line Drain lsol).

MNS STEAM GENERATOR TUBE RUPTURE PAGE NO.

EP/1/AI5000IE-3 S of 76 UNIT 1 Rev. 21 ACTION/EXPECTED RESPONSE RESPCNSE NOT OBTAINED

4. (Continued)

e. Perform the following:

1) Place STM PRESS CONTROLLER in manual.

. 2) Adjust STM PRESS CONTROLLER output to 0%.

3) Place STEAM DUMP SELECT in steam pressure mode.

4) Initiate Main Steam isolation signal.

5) IE all S/G pressures are above 775 PS1G, THEN reset the following to allow automatic SM PORV operation:

. Main Steamline Isolation

SMPORVs.

6) IF ruptured SIG(s) MSIV and bypass valve are closed. THEN GOTO Step 6.

7) CLOSE the following valves on remaining S/Gs:

MSIV

MSIV bypass valve.

8) IF at least one intact S/G cannot be isolated from ruptured SIG(s),

ThEN GO TO EP/1JA15000IECA-3. 1 (SGTR With Loss Of Reactor Coolant -

Subcooled Recovery Desired).

9) Select 11 OFF RESET on steam dump intetiock bypass switches.

10) Dispatch operator to immediately CLOSE valves EER Enclosure 3 (Local Isolation SP Valves and Steam Drain Bypass Valves).

(RNO continued on next page)

MNS STEAM GENERATOR TUBE RUPTURE PAGE NO.

EPIIIAJE,000)E-3 8 of 76 UNf 1 Rev. 21 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

4. (Continued)

15) CLOSE from Control Room or dispatch operator to CLOSE the following waives:

ISM-14(U1 Main StearnTo CSAE Isol) (Unit 1 turbine bldg, 760+20, 1 H-31, west side of column)

. 1TL-3 (SM To Steam Seal Isol)

(Unit 1 turbine bldg, 760+7, 1 D-33, northeast of DEH skid).

16) WHEN cooldown is initiated in subsequent steps, THEN use intact SIG(s) SM PORV for steam dump.

a. a.

2

3) .QStep6

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201.1 MNS SRO NRC. Examination. QUESTION 90 APEOO3 AK3.04 Dropped Control Rod owledge of the reasons for the following responses as they apply to the Dropped Control Rod: (CFR 41.5,41.10 / 45.6/ 45.13) ctions contained in EOP for dropped control rod Given the following:

UnitlisatlOO%RTP

Control Bank D Rods are at 220 steps

A single rod in Control Bank D drops to 115 steps In accordance with AP-14 (Rod Control Malfunction) thermal power must be reduced to less than (1) power within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

The basis for restricting thermal power to less than this limit is to (2)

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

A. 1. 50%

2. limit core peaking factors due to the misaligned rod B. 1. 50%

2. minimize AFD swings due to the induced Xenon oscillation C. 1. 75%

2. limit core peaking factors due to the misaligned rod D. 1. 75%

2. minimize AFD swings due to the induced Xenon oscillation Wednesday, April 20, 2011 Page 267 of 299

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2011. MNS SRO NRC Examination QUESTION 90 -

.neraI Discussion In the first part of this question, the applicant is asked to recall the requirement for restricting reactor thermal power per AP-14 (Rod Control Malfunction) for a dropped rod. Per Step 12b, the procedure requires that reactor power is less than 75% within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to comply with TS 3.1.4.

The second part of the question addresses the basis for limiting power to less than 75% rated thermal power for continued operation. The basis is to ensure that local LHR (Linear Heat Rate) increases to the misaligned RCCA will not cause core design criteria to be exceeded (i.e. limit local peaking factors due to the misaligned rod).

Answer A Discussion iNCORRECT: See explanation above.

PLAUSIBLE:

Part (1) is plausible since, lAW AP-14 power must be reduced to less than 50% power to retrieve the dropped rod.

Part (2) is correct.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part (1) is plausible since, JAW AP-l4 power must be reduced to less than 50% power to retrieve the dropped rod.

Part (2) is plausible because the dropped rod will result in the buildup of Xenon in the assembly with the dropped rod causing a shift in AFD.

Answer C Discussion COCT: See explanation above.

qswer D Discussion ORRECT: See explanation above.

PLAUSIBLE:

Part (1) is correct.

Part (2) is plausible because the dropped rod will result in the buildup of Xenon in the assembly with the dropped rod causing a shift in AFD.

Basis for meeting the KA This KJA is matched because the applicant must have knowledge of the actions contained in the EOP (in this case AP) for a dropped rod (including basis).

Basis for Hi Cog Basis for 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 10CFR55.43(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-14.

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 power reduction requirements (including the basis).

5) The question is detailed procedure step knowledge and the basis for that procedure step.

Job Level Cognitive Level QuestionType Question Source SRO Memory NEW Development References Student References Provided TS 3.1.4 (Rod Group Alignment Limits)

AP- 14 (Rod Control Malfunction) Page 9 of 36 Thursday, April 14, 2011 Page 268 of 299

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2fl1MNSSRO NRC Examination QUESTION 90 Tech Spec 3.1.4 Basis E003 AK3.04 Dropped Control Rod is.nowledge of the reasons for the following responses as they apply to the Dropped Control Rod: (CFR 41.5,41.10/45.6/45.13)

Actions contained in EOP for dropped control rod 401-9 Comments: Remarks!Status Thursday, April 14, 2011 Page 269 of 299

Question 90

References:

From TS 3.1.4 (Rod Group Alignment Limits) -

3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Rod Group Alignment Limits LCO 3.1.4 All shutdown and control rods shall be OPERABLE, with all individual indicated rod positions within 12 steps of their group step counter demand position.

APPLICABILITY: MODES I and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more rod(s) A. 1.1 Verify SDM is within the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months untrippable. limit specified in the COLR.

OR A.1.2 Initiate boration to restore 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 8DM to within limit.

AND A.2 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One rod not within B.1 Restore rod to within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months alignment limits. alignment limits.

OR B.2. 1.1 Verify SDM is within the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months limit specified in the COLR.

OR B.2.1.2 Initiate boration to restore 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months SDM to within limit.

AND B.2.2 Reduce THERMAL POWER to 75% RTP. 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months AND B.2.3 Verify SDM is within the limit specified in the COLR. Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND B.2.4 Perform SR 3.2.1.1.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months AND B.2.5 Perform SR 3.2.2.1.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months AND B.2.6 Re-evaluate safety analyses and confirm 5 days results remain valid for duration of operation under these conditions.

(continued)

From AP-14 (Rod Control Malfunction) Page 9 of 36 (Rev 13)

MNS ROD CONTROL MALFUNCTION PAGE NO.

API1IA!5500114 6 3

of Enclosure 1 Page 5 of 23 Rev.

UNIT 1 Response To Dropped or Misaligned Rod A.D::CN7EXPEC:Er RESPONSE RESPONsE NOT OEAINED

12. Reduce reactor power below 50% prior to rod realignment as follows:

a. Check only one rod MISALIGNED.

- a. j Step 12.c.

b. Ensure reactor power is less than 76%

within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of rod misalignment to comply with Tech Spec 11.4.

c Reduce load as directed n subsequent steps until reactor power is less than 50% to comply with Reactor Engineering requirements.

d. Observe the following limitations during power reduction:

1) Do not move rods until IAE determines rod movement is available.

2 Borate as required during power reduction to maintain T-Avg at T-Ref.

3) Monitor AFD during load reduction.

4) IEAT XJJM Aft) reachea Tech Spec limit NL reactor power is greater than 50%, ThEN perform the following:

a) Trip Reactor.

b) GO TO EPII/N5000/E-O (Reactor Trip or Safety Injection>.

From TS 3.1.4 Basis:

Rod Group Alignment Limits 83:1.4 BASES APPLICABLE Control rod misalignment accidents are analyzed in the safety analysis SAFETY ANALYSES (Ref. 3). The acceptance criteria for addressing control rod inoperability or misalignment are that:

a. There be no violations of:

1. specified acceptable fuel design limits, or

2. Reactor Coolant System (RCS) pressure boundary integrity and

b. The core remains subcritical after accident transients.

Two types of misalignment are distinguished. During movement of a control rod group, one rod may stop moving, while the other rods in the group continue. This condition may cause excessive power peaking.

The second type of misalignment occurs if one rod fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition requires an evaluation to determine that sufficient reactivity worth is held in the control rods to meet the SDM requirement, with the maximum worth rod stuck fully withdrawn.

Analyses are performed in regard to static rod misalignment, single rod withdrawal, dropped rod, and dropped group of rods (Ref. 4). With control banks at their insertion limits, one type of analysis considers the case when any one rod is completely inserted into the core. The second type of analysis considers the case of a completely withdrawn single rod from a bank inserted to its insertion limit Satisfying limits on departure from nucleate boiling ratio in both of these cases bounds the situation when a rod is misaligned from its group by 12 steps. Another type of misalignment occurs if one RCCA fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition is assumed in the evaluation to determine that the required SDM is met with the maximum worth RCCA also fully withdrawn (Ref. 5).

The Required Actions in this LCO ensure that either deviations from he aliqnment limits will be corrected or that THERMAL POWER will be I adjusted so that excessive local linear heat rates (LHR5) will not occfir, and that the requirements on SDM and ejected rod worth are preseived.

Continued operation of the reactor with a misaligned control rod is allowed if the heat flux hot channel factor (Fo(X,Y,Z)) and the nuclear v hot channel factor (F(X,Y)) are verified to be within their

...I:

is

...-j *- .

1.

McGuire Units 1 and 2 B 3.1.4-3 Revision No. 0

Rod Group Alignment LImits B 3.1.4 BASES APPLICABLE SAFETY ANALYSES (continued) ion3.2(I owen I Limits contains more complete discussions of the relation of F(X,Y,Z) and F (X,Y) to the operating limits.

Shutdown and control rod OPERABILITY and alignment are directly related to power distributions and SDM, which are initial conditions assumed in the safety analyses. Therefore they satisfy Criterion 2 of 10 CFR 50.36 (Ref. 6>.

LCO The requirements on rod OPERABILITY ensure that upon reactor trip, the assumed reactivity will be available and will be inserted The limits on shutdown and control rod alignments ensure that the assumptions in the safety analysis will remain valid, and that the RCCAs and banks maintain the correct power distribution and rod alignments.

The requirement to rnaintah the alignment of any one rod to within plus or minus 12 steps is conservative. The minimum misalignment assumed in safety analysis is 24 steps (15 inches), and in some cases a total misalignment from fully withdrawn to fully inserted is assumed.

Failure to meet the requirements of this LCO may produce unacceptable power peaking factors and LHRs, or unacceptable SDMs, all of which may constitute initial conditions inconsistent with the safety analysis.

APPLICABILITY The requirements on RCCA OPERABILITY and alignment are applicable in MODES 1 and 2 because these are the only MODES in which neutron (or fission) power is generated, and the OPERAOILITY (Le., tippability) and alignment of rods have the potential to affect the safety of the plant.

In MODES 3,4,5, and 6, the alignment limits do not apply because the control rods are normally bottomed and the reactor is shut down and not piuduciriy fisioii puwet. In Die shulduwn MODES, [lie OPERABILIT ul the shutdown and control rods has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the RCS. See LCO 3.1.1, SHUTDOWN MARGIN (5DM), for SDM in MODES 3,4, and Sand LCO 3.9.1, Boron Concentration, for boron concentration requirements during refueling.

ACTIONS A.1.1 and A.1.2 When one or more rods are untrippable, there is a possibility that the required 5DM nay be adversely affected. Under these conditions, it McGuire Units I and 2 B 3.1.4-4 Revision No. 0

Rod Group Alignment Limits 83.1.4 BASES AC11ONS (continued>

In many cases, realigning the remainder of the group to the misaligned rod may not be desirable. For example, realigning control bank B to a rod that is misaligned 15 steps from the top of the core would require a significant power reduction, since control bank D must be moved fully in and control bank C must be moved in to approximately 100 to 115 steps.

Power operation may continue with one RCCA trippable but misaligned.

provided that SDM is verified within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

The Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months represents the time necessary for determining the actual unit SOM and, if necessary, aligning and starting the necessary systems and components to initiate boration.

8.2.2, 8.2.3, 8.2.4, 8.2.5. and 8.2.6 For continued operation with a misaligned rod, RTP must be reduced, 5DM must periodically be verified within limits, hot channel factors (X,Y,Z) and F(X,Y) must be verified within limits, and the safety 0

F analyses must be re-evaluated to confirm continued operation is permissible.

Reduction of power to 75% RTP ensures that local LHR increases die to a misaligned RCCA will not cause the core design criteria to be I exceeded (Ref. 7). The Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months gives the operaor sufficient time to accomplish an orderly power reduction without challenging the Reactor Protection System.

When a rod is known to be misaligned, there is a potential to impact the 5DM. Since the core conditions can change with time, periodic verification of SDM is required. A Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient to ensure this requirement continues to be met.

Verifying that F(X,Y,Z) and F(X,Y) are within the required limits 0

ensures that current operation at 75% RTP with a rod misaligned is not resufting in power distributions that may invalidate safety analysis assumptions at full power. The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months allows sufficient time to obtain flux maps of the core power distribution using the incore flux mapping system and to calculate FoGX,Y,Z) and FN.H(X,Y).

Once current conditions have been verified acceptable, time is available to perform evaluations of accident analysis to determine that core limits will not be exceeded during a Design Basis Event for the duration of McGuire Units I and 2 B 3.1.4-6 Revision No. 0

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2111 MNS SRO.NRCExaminatkin. QUESTION 91 iL APEO36 2.2.12 Fuel Handling Incidents PE036 GENERIC now1edge of surveillance procedures. (CFR: 41.10/45.13)

Given the following conditions on Unit 1:

The unit is in MODE 6 with core reload in progress

NC system boron concentration is 2705 PPM

The following surveillances are being performed:

o PT/i IA/4600/1 00 (Surveillance Requirements For Shutdown Conditions) o PTI1IAI4600/003 C (Weekly Surveillance Items Checklist)

The surveillance for NC system boron concentration performed during PTIIIAI4600I1 00 (SR 3.9.1 .1) ensures that keff during MODE 6 remains less than a MAXIMUM of (1)

The MINIMUM Boric Acid Tank level required to meet the surveillance requirements of PT/I /A/4600/003 C (TR 16.9.14.3) is (2)

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

REFERENCE PROVIDED A. 1. 0.95

2. 8.7%

B. 1. 0.95

2. 13.6%

C. 1. 0.99

2. 8.7%

D. 1. 0.99

2. 13.6%

Thursday, April 14, 2011 Page 270 of 299

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loll MNS SRO NRC Examination QUESTION 91. 9 neraI Discussion In accordance with TS 3.9.1 (Boron Concentration), in MODE 6 boron concentration shall be maintained within the limits specified in the COLR (Core Operating Limits Report). In accordance with the Basis Document for TS 3.9.1 the basis for maintaining boron concentration within limits is to ensure that Keff remains less than 0.95 to ensure that a recriticality does not occur while in MODE 6 (i.e. during fuel movement).

In accordance with SLC 16.9.14, Borated Water Sources (Shutdown) for the BAT to be considered OPERABLE it must meet the minimum volume requirements specified in the COLR. There are two conditions that would require different BAT levels. If the unit was in MODE 6 at the end of a cycle after 455 EFPD and the core had not yet been off loaded, Figure 6 of the COLR would be used to determine the minimum volume and the required minimum level would be 8.7%. However, for the conditions given, since core reload is in progress (i.e. the full cycle core is no longer loaded), the requirements of COLR 2.16 would apply (i.e. 13.6%).

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part 1 is correct Part 2 is plausible because this would be the correct answer if the core had not yet been off loaded.

Answer B Discussion CORRECT: See explanation above.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE:

DaJ.t 1 is plausible because < .99 is the maximum Keff for Cold Shutdown.

art 2 is plausible because this would be the correct answer if the core had not yet been off loaded.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part 1 is plausible because < .99 is the maximum Keff for Cold Shutdown.

Part 2 is correct.

Basis for meeting the KA This question requires the applicant to have knowledge of the basis for NC system boron concentration surveillance requirements during refueling. If the limits of this surveillance requirement are not met, it could result in an inadvertent criticality which would constitute a Fuel Handling Incident.

Basis for Hi Cog This question is higher cognitive level because it requires more than one mental step. The first part of the question requires the applicant to recall from memory the basis for NC system boron surveillance requirement during refueling. The second part requires the applicant to analyze plant conditions and determine which BAT level limit from the COLR applies to the given conditions.

Basis for SRO only This question is SRO-only knowledge linked to 1 OCFR55.43(b)(2) (Tech Specs) as described in the Clarification Guidance for SRO-only Question Rev 1 (dated 03/11/2010):

1) The question can NOT be answered solely by knowing 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months TS/TRM action statements. It requires knowledge of TS basis.

2) The question can NOT be answered solely by knowing the LCO/TRM information listed above-the-line. Again, TS basis knowledge.

The question can NOT be answered solely by knowing the TS Safety Limits. TS 3.9.1 (Boron Concentration) Basis.

4) The question requires the applicant to have detailed knowledge of the TS basis to analyze TS requirements (i.e. basis for the NC system boron concentration surveillance). As such this constitutes SRO-level knowledge.

Thursday, April 14, 2011 Page 271 of 299

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2011 MNS SRO NRC Examination QUESTION 91 Job Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Development References Student References Provided

References:

COLR 2.16 (Borated Water Sources Shutdown)

TS 3.9.1 (Boron Concentration) Basis Core Operating Limits Report (COLR) 2.16 and Figure 6 APEO36 2.2.12 Fuel Handling Incidents APEO36 GENERIC Knowledge of surveillance procedures. (CFR: 41.10 I 45.13) 1-9 Comments: RemarkslStatus Thursday, April 14, 2011 Page 272 of 299

Question 91

References:

From TS 3.9.1 (Boron Concentration) Basis:

Boron Concentration 83.9.1 BASES BACKGROUND (continued) cavity mix the added concentrated boric acid with the water in the refueling canal. The RHR System is in operation during refueling (see LCO 3.9.5, Residual Heat Removal (RHR) and Coolant CirculationHigh Water Level, and LCO 3.9.6, Residual Heat Removal (RHR) and Coolant Circulation Low Water Level) to provide forced circulation in the RCS and assist in maintaining the boron concentrations in the RCS, the refueling canal, and the refueling cavity above the COLR limit.

APPLI CABLE During refueling operations, the reactivity condition of the core is SAFETY ANALYSES consistent with the initial conditions assumed for the boron dilution accident in the accident analysis and is conservative for MODE 6. The boron concentration limit specified in the COLR is based on the core reactivity at the beginning of each fuel cycle (the end of refueling) and includes an uncertainty allowance.

The required boron concentration and the plant refueling procedures that verify the correct fuel loading plan (including full core mapping) ensure that the k of the core will remain 0.95 during the refueling operation.

Hence, at least a 5% lc/k margin of safety is established during refueling.

During refueling, the water volume in the spent fuel pool, the transfer canal, the refueling canal, the refueling cavity, and the reactor vessel form a single mass. As a result, the soluble boron concentration is relatively the same in each of these volumes.

The RCS boron concentration satisfies Criterion 2 of 10 CFR 50.36 (Ref.

2).

LCO The LCO requires that a minimum boron concentration be maintained in the RCS, the refueling canal, and the refueling cavity while in MODE 6.

The boron concentration limit specified in the COLR ensures that a core k of 0.95 is maintained during fuel handling operations. Violation of the LCO could lead to an inadvertent criticality during MODE 6.

APPLICABILITY This LCO is applicable in MODE 6 to ensure that the fuel in the reactor vessel will remain subcritical. The required boron concentration ensures a kr 0.95. Above MODE 6, LCO 3.1.1, SHUTDOWN MARGIN (5DM), ensure that an adequate amount of negative reactivity is available to shut down the reactor and maintain it subcritical.

The Applicability is modified by a Note. The Note states that the limits on boron concentration are only applicable to the refueling canal and the MeGuire Units 1 and 2 83.9.1-2 Revision No.68

MCEL-040u-232 PageZ9of32 Revision 0 McGuirt 1 Cytle 21 Ce Operating Limits Report 2.16 Borated Water Sources Sbutdawu (SLC 16.9,14)

2. i&l V1ume and boroe c centrations ibr the Bozic Acid Tank (BAT) and the Refueling Water Storage Tank (RWST) during MODE 4 with any RCS cold leg temperature 300 °F and MODES 5 and 6.

BAT minimum contained borated watar voiunm 10,599 gnllon.s 13.6% Level Note; WIic-n cycle burnup is> 455 EFPD Figure 6 may be used to determine the rcquired SAT niinirnurn level BAT minimum boron ccncentratin 74000 ppm BAT mininruin water volume required to 2,00.gaUons,,

matntain SOM at 7)0130 ppm RWST minimum contained borated water 47,700 gallons volurnc 41. inches RWST mini mum boron concentration 29675 ppm RWST minimum water volunac required to 8,200 gallons nzaErrtaiu 5DM a 2,675 ppm

Me.EU4U-Z Pan31 of 32 Revisioi 0 MGuire 1. Cycle 21 Core Opsrating Limits Report FIgure 6 Boric Acid Storage Tank Indicated Level Yerss RCS Boron Cancenraton (Valid When Cycle Burxrnp is> 455 EFPD)

This figure inelndes additionaL volumes listed in SIX 16.9.14 and 16.9.11 RCS Bat-on ConnLratiori AT Lcvcl (pii) rktev0

<300 37X3 500 0G 2-LO 70<1DOO 23.O

.1OOOi3OO 13.6.

AcceptBbJe 100

[ ecptae opcto I C 2t0 4G E0 1000 1200 1400 1q00 1q00 20cG 2200 2400 3G00 20G RGS Boron Concentraliori (ppmb)

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2011 MNS SRO NRCExamination QUESTION 92 APEO37 AA2.1O Steam Generator (S/G) Tube Leak ility to determine and interpret the following as they apply to the Steam Generator Tube Leak: (CFR: 43.5 /45.13) ech-Spec limits for RCS leakage Unit I is operating at 100% RTP. Given the following:

1EMF-33 (Condenser Air Ejector Exhaust) is in Trip 2 alarm

1EMF-71 (S/GA Leakage) is in Trip 2 alarm

Pressurizer level has been stabilized using AP-lO (Reactor Coolant Leak)

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 Wednesday, April20, 2011 Page 273 of 299

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2011 MNS SRO NRC Examination QUESTION 92 neral Discussion With the indications given, the crew would be required to enter AP-lO (NC System Leakage), Case 1 (S/G Tube leakage). This procedure would direct the crew to stabilize PZR level and determine leak size.

Leakage rate is 78-45-12=21 gpm making the Standby Makeup Pump INOPERABLE in accordance with SLC 16.9.7. Step 6 of AP-lO Case 1, directs an SRO to evaluate if leakage exceeds SLC 16.9.7 Condition C limits. The limit is defined as >20 GPM. Per TS 3.4.13 (NC Operational Leakage), the limit for a individual 5/0 tube leakage of 135 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 7 of AP-lO, Case 1, if the leakage in one S/G is greater than 125 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 125 GPD.

Answer A Discussion CORRECT: See explanation above.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

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 of2l 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.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE:

(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 in Mode 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . It would reasonable for the applicant to believe this would also be the required time specified in AP-lO.

Part (2) correct.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

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 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.

Basis for meeting the KA This K/A is matched because the applicant is first required to demonstrate the ability to determine the actual S/G tube leakage. He is then required to interpret this information as it applies to procedural direction from AP- 10 for leakage being greater than tech specs and the application of SLC 16.9.7 limit on leakage. I Basis for Hi Cog 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.

Basis for 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 10CFR55.43(b)(5) (Assessment and selection of procedures):

The question can NOT be answered solely by knowing systems knowledge.

le question can NOT be answered by knowing immediate operator actions. Neither of the actions described are immediate actions.

i) 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 Wednesday, April 20, 2011 Page 274 of 299

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2011 MNS SRO NRC Examination. QJJFSTION 92 4 Job Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Development References Student References Provided From AP-lO Pages 5 & 7 SLC 16.9.7 (Standby Shutdown System)

APEO37 AA2.1O Steam Generator (S/G) Tube Leak Ability to determine and interpret the following as they apply to the Steam Generator Tube Leak: (CFR: 43.5 /45.13)

Tech-Spec limits for RCS leakage 401-9 Comments: RemarkslStatus Thursday, April 14, 2011 Page 275 of 299

Question 92

References:

From AP-lO (NC Leakage within the Capacity of Both NV Pumps) Case I (SIG Tube Leakage) Page 7 of 129 (Rev 22)

NC SYSTEM LEAKAGE WITHIN THE CAPACITY OF MNS PAGE NO 0TH AP/IJA!550W10 7 of 129 Rev. 22 UNfl I Steirn GenrtrTub Lkge r-::ot;i;E2ED oNs 5?ONSE KOT oTA:K:

7. (Continued)

d. Leakage in one S!G GREATER THAN

- d. Perform the following:

12(ALLON PER DAY).

1) IF unit shutdown required per PTIltN41S001C (Primary to Secondary Leakage Monitoring),

THEN observe Note prior to StepS and Q.IOStepe.

2) IF station management desires to exit procedures THEN exit procedure at this time.

3) Do not continue unless load reduction desired.

4) Obserie Note prior to Step 8 and

.QIQ.Step 8.

e. Observe the following limits while reducing load in Step 8:

. Ensure reactor power is. less than 50% with in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of exceeding 125 GPO.

Be in Mode 3 within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months of exceedirg 125 GPO.

From AP-lO (NC Leakage within the Capacity of Both NV Pumps) Case I (SIG Tube Leakage) Page 5 of 129 (Rev 22)

NC SYSTEM LEAKAGE WITHIN THE CAPACIW OF MNS BOTH NV PUMPS PAGE NO.

AP/11550OI1 0 Case 5 of 129 TJ%fl 1 Stearni Generator Tube Leakage Rv. 22 oN/EcE:E tE5;oNe e;ozs NOT REFER ic. RPlQIN57OIQQO (Classification of Emergency).

6. IF AT ANY TIME NC leakage exceecl Tech Spec linilts. fllEt4 perform the following:

Ensure Outside Air Pressure Filter train in service PER 0 P1O!A16450/O1 I (Control Area VentilationlChillect Water System), En closure 4.4 (Control Room Atmosphere Pressurization During Abiromal Conditions>.

. Have another SRO evaluate if leakage exceeds SLC 16.9.7 condition C limits and immediately notify security if SSF is inoperable.

From SLC 16.9.7 (Standby Shutdown System) 16.9 AUXILIARY SYSTEMS FIRE PROTECTION SYSTEMS 16.9.7 Standby Shutdown System COMMITMENT The Standby Shutdown System (SSS) shall be operable.

APPLICABILITY MODES 1, 2, and 3.

REMEDIAL ACTIONS NOTE

1. The SRO should ensure that security is notified 10 minutes prior to declaring the SSS inoperable. Immediately upon discovery of the SSS inoperability, Security must be notified to implement compensatory measures within 10 minutes of the discovery.

2. If inoperable SSS component is located inside containment, repairs shall be made at the first outage which permits containment access.

CONDITION REQUIRED ACTION COMPLETION TIME NOTE A.1 Verify the OPERABILITY of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Not applicable to the SSS fire detection and Diesel Generator or 24 V suppression systems in the Battery Bank and Charger. associated areas identified Tablel6.9.7-1.

A. One or more required AND SSS components identified in Table A.2 Restore the component to 16.9.7-1 inoperable. 7 da OPERABLE status.

B. SSS Diesel Generator or B.1 Verify the OPERABILITY of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 24 V Battery Bank and fire detection and Charger inoperable, suppression systems in the associated areas identified in Tablel6.9.7-1.

AND (continued)

REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. (continued). B.2 Verify offsite power and 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months one emergency diesel generator OPERABLE.

AND B.3 Restore the component to 7days OPERABLE_status.

C. Total Unidentified C.1 Declare the Standby Immediately LEAKAGE, Identified Makeup Pump inoperable.

LEAKAGE, and reactor coolant pump seal AND leakoff> 20 gpm.

C.2 Enter Condition A.

OR Total reactor coolant pump seal leakoff> 16.3 gpm.

OR Any reactor coolant pump No. 1 seal leakoff

>_4.0_gpm.

D. Lake Norman level D.1 Verify the C Fire 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months below 746 feet. Suppression Pump is OPERABLE_(Unit_1_only).

E. Required Action A.2 and E.1 Prepare and submit a 30 days its associated Special Report to the NRC Completion Time not outlining the cause of the met. inoperability, corrective actions taken, and plans for restoring the SSS to OPERABLE status.

F. Required Action B.3 and F.1 Prepare and submit a 14 days its associated Special Report to the NRC Completion Time not outlining the extent of met. repairs required, schedule for completing repairs, and basis for continued operation.

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2011 MNS SRO NRC Eximinathn QUESTION 93 APEO67 2.2.25 Plant Fire On Site EO67 GENERIC owledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. (CFR: 41.5 / 41.7 / 43.2)

Given the following plant conditions:

Both units are operating at 100% RTP

A fire occurs in the Auxiliary Building on the 733 Elevation in the Battery Room

AP-45 (Plant Fire) has been implemented

The crew has just reached the steps in Enclosure 13 (AB 733 Battery Room Fire Unit I and 2 Actions) which directs closing the PZR PORV Block valves on both units In accordance with the AP-45 Background document, these valves must be closed within (1) minutes of the start of the fire.

In accordance with Tech Spec 3.4.11 (Pressurizer PORVs) Basis ONLY, after the PZR PORV Block valves are closed, the PZR PORVs are (2)

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

A. 1.10

2. INOPERABLE B. 1.30

2. INOPERABLE C. 1.10

2. OPERABLE D. 1.30

2. OPERABLE Wednesday, April 20, 2011 Page 276 of 299

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2011 MNS SRO NRC Examination QUESTION 93 93

..neraI Discussion AY-45 (Plant Fire). Enclosure 13 (AB 733 Battery Room Fire Unit I and 2 Actions) directs the operators to close the PZR PORV Block Valves for both units.

AP-45 directs that the action to close the PORV Block Valves is time critical. However, the time critical time is only listed in the AP-45 Background document.

In accordance with AP-45:

Without operator action what could happen after 10 mm is a Pzr PORV Isol could be open and the power supply cable to the valve motor could be burnt through. Per Appendix R assumptions, all three isolations may have lost power. Also, the one assumed smart short could be sending an open signal to one PORV. To prevent this scenario, the PORV block valves are closed within 10 minutes, before these failures could occur.

With regards to the OPERABILITY of the PORVs after the PORV Block Valves are closed AP-45 goes on to state:

Note this step just closes the block valve; it does not remove power from the valve. We do not have to assume a 2nd smart short would fail the block open. For Modes 1, 2, or 3, just closing the block valve does not make the PORV inoperable per Tech Spec. Of course, if the fire really does burn through the power supply cable to the isolation or sends a fail open signal to the PORV, then the PORV would be inoperable at that time.

Answer A Discussion iCORRECT: See explanation above.

The first part is correct.

The second part is plausible because normally the only reason for closing a PORV Block Valve is under conditions where the PORV is already

-onsidered INOPERABLE.

iswer B Discussion iNCORRECT: See explanation above.

PLAUSIBLE:

The first part is plausible because 30 minutes is a time-critical time that is used extensively throughout the APs.

The second part is plausible because normally the only reason for closing a PORV Block Valve is under conditions where the PORV is already sidered INOPERABLE.

Answer C Discussion RRECT: See explanation above.

Answer 0 Discussion

[iCORRECT: See explanation above.

PLAUSIBLE:

The first part is plausible because 30 minutes is a time-critical time that is used extensively throughout the APs.

The second part is correct.

Basis for meeting the KA The KJA is matched because it requires the applicant to have knowledge of TS basis to be able to determine if the PORV is inoperable with the PORV Block valve closed.

Basis for Hi Cog Basis for SRO only rt 1 of this question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions

V I dated 03/11/2010 for screening questions linked to 10CFR55.43(b)(5) (Assessment and selection of procedures)

1) This question can NOT be answered by knowing systems knowledge alone. This is strict procedure knowledge. This is not covered during systems training or discussed in a systems lesson plan.

2) This question can NOT be answered by knowing immediate operator actions.

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2011 MNS SRO NRC Examination QUESTION 93 [ 93L

3) This question can NOT be answered by knowing the entry conditions for AOPs. This is based on knowledge of the AP-45 Background (Basis) ocument.

1 This question can NOT be answered by knowing the purpose, overall sequence of events, or overall mitigative strategy of the AOPs. The question is based on knowledge of the AP-45 Background document.

5) The question requires the applicant to have in-depth knowledge of the AP-45 Background document. Therefore, this is SRO level knowledge The second part of this question is SRO-only knowledge linked to IOCFR55.43(b)(2) (Tech Specs) as described in the Clarification Guidance for SRO-only Question Rev 1 (dated 03/11/2010):

1) The question can NOT be answered solely by knowing 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months TS/TRM action statements. It requires knowledge of TS basis.

2) The question can NOT be answered solely by knowing the LCO/TRM information listed above-the-line. Again, TS basis knowledge.

3) The question can NOT be answered solely by knowing the TS Safety Limits.

4) The question requires the applicant to have detailed knowledge of the TS basis to analyze TS requirements (i.e. TS 3.4.11). As such this constitutes SRO-level knowledge.

Job Level Cognitive Level QuestionType Question Source SRO Memory NEW Development References Student References Provided

References:

AP-45 (Plant Fire)

AP-45 Background Document Lesson Plan OP-MC-AP-45 arning Objectives: OP-MC-AP-45 Objective 2 APEO67 2.2.25 Plant Fire On Site APEO67 GENERIC Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. (CFR: 41.5 /41.7/43.2) 401-9 Comments: Remarks/Status Thursday, April 14, 2011 Page 278 of 299

Question 93

References:

From Lesson Plan OP-MC-AP-45:

From AP-45 Background Document for Enclosure 13 Step 3:

APIW!55Q0I045 (Plant FIFE)

Cue the operator of the time c.ritial nature of the following step, which must be compted witrln 10 minutes otaetermlningrne fire Is ACTIVE. Note Itmaytake a couple mnutes to determine the tire is active.

DISCUSSION:

One rule of usae for Appendbc R fire assumptors is there can be multiple faiLires asscciated with power supply cables burning through. This v/ill cause a loss of power to those valves or imp rnntors However, another nile of usage is there is onlV one assumec smart fre short This smart short could cause a valve to misjxsition for example. These failures are assumed to not occur within the first 10 minutes of the fire. It takes some amount of tie (40 mm) to bum thrcugh the insulation. Thats the basis fo the 10 mm setpcmnt.

ENCLI: STEP 3:

FURPOSE:

Freent an unisolable Pzr PORV LDCA.

DISCUSSION:

Without operator action what could happen 10 mm is a Pzr PORV isol could be open and the power suppy cable to the valve motor could be burnt throuu. Per Appendix R assumptions, aN three sclations may have lost power. Aso, the one assumed smart short could be sending an open signal to one PORV. To prevent iruls scenarIo, tne PORV blocK valves are closed within lU minutes, betcce these Tailures cculd occur Note this step just closes the block valve; t does not remove power from the va1e. We do not have to assume a 2nd smart short would fail the block open. For Mcdes 1, 2, or 3, just closinq the block valve does not make the PORV noperable per Tech Spec. Of course, if the fire really does hum through the power supply cable to the solation or sends a fail open signaL In the PORV, then tue PDRV would be inoperable at that time.

ENCLl: STEP 4:

PURPOSE:

Prevent water hamrrer on the GB System.

11of256 Rev9

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2011 MNS SRO-NRC Examination QUESTION 94 GEN2. 1 2.1.20 GENERIC Conduct of Operations nduct of Operations bi1i to interpret and execute procedure steps. (CFR: 41.10 I 43.5 / 45.12)

With Unit I at 100% RTP the following conditions exist:

The 1C SIG develops a 10 GPM tube leak

AP-1 0 (NC System leakage Within the Capacfty of Both NV Pumps) has been implemented

Plant load is reduced using AP-04 (Rapid Down power), and manually tripped at 15% power

The crew has just transitioned to ES-0.1 (Reactor Trip Response)

Subsequently:

NCS Pressure is 2210 PSIG, and slowly DECREASING

The crew arrives at Step 23.a of ES-0.1 and is directed to Maintain Pzr pressure AT 2235 PSIG

Simultaneously, the crew arrives at Step 18.a of AP-lO and is directed to Depressurize to between 1900-1 955 PSIG using normal Pzr spray.

Which ONE (1) of the following describes how the implementation of AP-lO and ES-0.1 is coordinated?

A. Suspend actions in AP-lO until ES-0.1 is complete; THEN Return to AP-lO and complete all required actions.

B. Suspend actions in ES-0.1 until AP-lO is complete; THEN Return to ES-0.1 and complete all required actions.

C. Continue simultaneous implementation of ES-0.1 and AP-1 0; If conflicting guidance is provided, AP-lO actions will have priority.

D. Continue simultaneous implementation of ES-0.1 and AP-1 0; If conflicting guidance is provided, ES-0.1 actions will have priority.

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2011 MNS SRO NRC Examination QUESTION 94

.eneral Discussion According to the AP1O Background Document (p7-8; Rev 6), during the performance ofAPlO, If the reactor is tripped, the crew would end up running ES-U. 1 concurrently with the remainder of this AP. For the potential conflicts in guidance given in the two procedures, the actions that affect the strategy of this AP would take priority. This is because the actions in AP-1U were written to minimize primary-to-secondary leakage by quickly depressurizing and cooling down the NC System. ES-U. 1 was written to stabilize the plant following a reactor trip. Some examples of conflicting actions are: 1) ES-U. 1 would turn Pzr heaters on if NC pressure is low whereas AP/1 U instructs the operator to turn off Pzr heaters prior to the NC depressurization; 2) ES-U. 1 would establish letdown when Pzr level is greater than the letdown isolation setpoint where as AP/lO operates letdown on the ability to maintain Pzr level and adequate boration flow; 3) ES-U.1 has a step to verif SI actuation on the low Pzr pressure SI setpoint whereas AP/lO purposely depressurizes the NC System and blocks the SI signal: 4) ES-U. 1 stabilizes NC temperature whereas AP/l U performs an NC cooldown. This guidance is consistent with the guidance provided in OMP 4-3 (p22; Rev 27), which states that generally the use of APs in conjunction with an EP should be avoided. However, it goes on to say that in some cases it would be proper to use an AP during a major accident that is addressed in the EPs.

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE:

It is plausible because the operator may incorrectly believe that since the EP network has been entered, that the AP no longer takes precedence.

Generally this is true so it would be reasonable for the applicant to believe that an EP would take precedence over an AP.

Answer B Discussion iNCORRECT: See explanation above.

PLAUSIBLE:

This is plausible because the operator may incorrectly believe that the AP takes precedence to the point to which the EP must be set aside.

Considering that the remaining actions in ES U. 1 are just clean up and will not address a major problem that has yet to be dealt with, this would seem a reasonable course of action.

nswer C Discussion ORRECT: See explanation above.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

This is plausible because the operator may incorrectly believe that the actions of an EP always take precedence over the actions of an EP. In almost every scenario, this would be true so it would not be unreasonable for the applicant to believe that it would in this scenario as well.

Basis for meeting the KA The KA is matched because the operator must demonstrate the ability to interpret and execute procedure steps during a SGTL. Specifically he must demonstrate the ability to decide between conflicting procedural guidance.

Basis for Hi Cog This is a higher cognitive level question because the applicant must perform a level of analysis concerning the given indications and determine correct procedural course of action.

Basis for SRO only The question is SRO-Only because the question cannot be answered by knowing system knowledge, immediate operator actions, or EOP Entry conditions, but rather requires that the operator know how to implement two procedures concurrently during a specific event (i.e. SGTL).

This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev 1 dated U3/l 1/2U 1 U for screening questions linked to 1 UCFR55.43(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.

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 SRO level procedural selection with conflicting guidance between an higher and lower level procedure.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension BANK Bank 32U3 Thursday, April 14, 2011 Page 280 of 299

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2011 MNS SRO NRC Examination QUESTION 94 Development References Student References Provided S 0.1 page 13 & 20 of 57 P- 1 (NC System Leakage) Page 11 of 129 AP/lO Background document. Pg 7 of 67 GEN2. 1 2.1.20 GENERIC Conduct of Operations Conduct of Operations Ability to interpret and execute procedure steps. (CFR: 41.10/43.5/45.12) 401-9 Comments: RemarkslStatus ii Thursday, April 14,2011 Page 281 of 299

Question 94

References:

From ES 0.1 (Reactor Trip Response) page 13 of 57) (Rev 31)

MNS REACTOR TRIP RESPONSE PAGE NO.

EPI1/A15090/ES-0.1 13 of 57 UNIT 1 Rev. 31 1111111111 H I I 1I 11111 11111)1111111111111

16. (Continued)

b. LE Pzr preasure greater than 2235 P810 going up, THEN perform the following:

1) Ensure all Pzr heaters off.

2) Control Pzr pressure using normal Pzr spray.

3 IF normal spray not availabe, THEN perform tle following:

a) if letdown in service, THEN use NV aux spray PER EPJ1JAJSOOQ!G-i (Gnric Endosures), Enclosure 3 (Establishing NV Aux Spray).

b) IF letdown isolated OR W aux spray not effective, THEN ensure proper operation of Pzr PORV()

17. Control SIG levels as follows:

a Check N/R level in any SIG - a. Maintain total feed flow greater than GREATER TF-t.N 11%. 450 GPM unt at least one S!G N/R level greater than 11%.

b. THROTTLE feed flow to maintain S/G b. IF N1R level in any SIG continues to go N/R levels between 11% and 50%. up, THEN stoz feed flow to that 51G.

18. Evaluate initiating actions in abnormal procedures:

a. Check 1 ETA and 1 ETB ENERGIZED

- a. REFER TO A11IN5500IO7 (Loss of BY OFFSITE POWER. Electrical Power), while continuing with this procedure.

b. Evaluate implementing applicable abnormal procedures, while continuing with this procedure.

From ES 0.1 (Reactor Trip Response) page 20 of 57) (Rev 31)

MNS REACTOR TRIP RESPONSE PAGE NO.

EPI1/AI5000IES-O. 1 20 of 57 Rev. 31 UNIT 1 II 1111111 11111 I III I I 11111 IllIllIllIll I

23. Maintain stable plant conditions:

a. Maintain Pzr pressure AT 2235 P51G.

b. Maintain Pzr level at program PZR LEVEL SETPOINT.

c. Maintain SIG NIR levels BE1WEEN 11%AND5O%.

d. Check condenser available as follows: d. IF AT ANY TIME S/G PORVs are cycling in automatic, THEN perfomi the ECheck C-9 COND AVAILABLE FOR following to stabilize SIG pressure and STEAM DUMP status light (1SI-18) - NC temperature:

1) Perform the following substeps in x MSIVs on all intact SIGs OPEN

- rapid succession to hmit risk of opening SM safeties:

a) CLOSE SM PORV manual loaders.

b) Select MANUAL on SM PORV MODE SELECT.

c) ThRO1TLE intact SIG SM PORVs as necessary to stabilize S/G pressure at approximately 1100 PSIG.

2) Request STA to assist in monitoring SIG pressures and NC

From AP-1 (NC System Leakage) Page 11 of 129 (Rev 22)

NC SYSTEM LEAKAGE WITHIN THE CAPACITY OF N1N BOTH NV PUMPS PAGE NO AP/l/AJ55001i0 fi of 129 CaeI UNIT 1 Steam Generator Tube Leakage -=

E5ONaE cr o: ::

CAUTION l1 Pzr pressure goes above 1955 PSIG after perfonning blacks then Pzr Sf1 actuation circuit and Low Pressure Steamline Isolation circuit will a utornatically unblock.

HOTE Depressu rization to perform blocks should be performed as quickly as possible while maintaining control of pressure.

18. Depressurize Pzr to perform blo cks as follows:

a. Depressurize te between 1900- a. Depressurize to 1900 19.55 PSIG 1955 PSIG using normal Pzr spray. using one of the following:

IF Ietdown in service, AD time allows, use NV aux spray PER EPIIIAI5000IG-1 (Genelic Enclosures), Enclosure 3 (Establishing NV Aux Spray).

Use one Pzr PORV.

b. Cheak P-il PRESSURIZER Sf1 b. Perform the following:

BLOCK PERMISSIVE status light (IS 1-18) LIT.

- 1) Noti RO to be prepared to blo.k Pzr 511 as soon as Pzr pressure goes below P-Il.

2) Do not oonnue trntil P-I I

From the APIIO Background document. Pg 7 of 67 and some shutdown banks withdrawn. In this plant alignment (initial conditions plant-shutdown below P-Il), entry to E-0 for a reactor trip does not apply. Reference OMP 4-3 (Use of Abnormal and Emergency Procedures) and its associated deviation document.

If the leak rate does not require a reactor trip but a unit shutdown is required/desired, the time limits for accomplishing the unit shutdown are listed. If the leak rate is> 125 GPD, then it is required to be below 50% power in one hour and in MODE 3 within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months . If the leak rate is < 125 GPD but above the limits requiring shutdown per PTIIIAI415OC (Primary to Secondary Leakage Monitoring), the PT will list the time requirements. Ref.: NSD 513 (Primary to Secondary Leak Monitoring Program) and EPRI PWR Primary-to-Secondary Leak Guidelines, Revision 2, April 2000.

If the leakage is less than the requirements of the PT for shutdown, this AP can be exited or a hold point established here, unless it is desired to shutdown. Note that if leakage is> 135 GPD, Tech Spec 3.4.13 requires the unit to be in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

If the reactor is tripped, the crew would end up running ESO.1 concurrently with the remainder of this AP. For the potential conflicts in guidance given in the two procedures, the actions that affect the strategy of this AP would take priority. This is because the actions in this AP were written to minimize primary-to-secondary leakage by quickly depressurizing and cooling down the NC System. ES-0.1 was written to stabilize the plant following a reactor trip. Some examples of conflicting actions are:

1) ES-0.l would turn Pzr heaters on if NC pressure is low whereas AP/lO instructs the operator to turn off Pzr heaters prior to the NC depressurization; 2)

ES-0.I would establish letdown when Pzr level is greater than the letdown isolation setpoint where as AP/1 0 operates letdown on the ability to maintain Pzr level and adequate boration flow; 3) ES-0.1 has a step to verify SI actuation on the low Pzr pressure SI setpoint whereas AP/1 0 purposely depressurizes the NC System and blocks the SI signal: 4) ES-0.l stabilizes NC temperature whereas AP/I0 performs an NC cooldown.

Most conflicts occur because of timing issues when performing concurrent guidelines. If ES-0.l was completed prior to entry into AP/lO, then most conflicts would disappear.

REFERENCES:

Tech Spec 3.4.13, PT/I /A/4150/OO1C (Primary to Secondary Leakage Monitoring), NSD 513 (Primary To Secondary Leakage Monitoring Program), and EPRI PWR Primary-to Secondary Leak Guidelines

Parent Question (Bank 3203):

With Unit 1 at 100% RTP the following conditions exist:

The IC SIG develops a 10 GPM tube leak

AP-1 0 (NC System leakage Within the Capacity of Both NV Pumps) has been implemented

Plant load is reduced using AP-04 (Rapid Downpower), and manually tripped at 15% power

The crew has just transitioned to ES-0.1 (Reactor Trip Response)

Subsequently:

NCS Pressure is 2210 PSIG, and slowly DECREASING

The crew arrives at Step 22.a of ES-0.1 and is directed to Maintain Pzr pressure AT 2235 PSIG

Simultaneously, the crew arrives at Step 18.a of AP-lO and is directed to Depressurize to between 1900-1955 PSIG using normal Pzr spray.

Which ONE (1) of the following describes how the conflicting procedural guidance must be resolved?

A. Suspend actions in AP-lO until ES-0.1 is complete; THEN Return to AP-lO and complete all required actions.

B. Suspend actions in ES-0.1 until AP-lO is complete; THEN Return to ES-0.1 and complete all required actions.

C. Continue simultaneous implementation of ES-0.1 and AP-lO; If conflicting guidance is provided, AP-lO actions will have priority.

D. Continue simultaneous implementation of ES-0.1 and AP-lO; If conflicting guidance is provided, ES-0.1 actions will have priority.

ANSWER: C

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2q11 MNS SRO NRC Examination QUESTION 95 (3EN2.1 2.1.35 GENERIC Conduct of Operations

)nduct of Operations

.now1edge of the fuel-handling responsibilities of SROs. (CFR: 41.10/43.7)

Given the following conditions on Unit 1:

The unit is in MODE 6 with core alterations in progress

It is determined that a Fuel Handling interlock must be bypassed to insert the next fuel assembly into the core

The interlock which must be bypassed is NOT specified in a procedure In accordance with NSD-414 (Fuel Handling), which the individuals listed below is allowed to approve bypassing the interlock?

1. Fuel Handling SRO

2. Reactor Engineering

3. Operations Shift Manager A. 1ONLY B. I AND 2 ONLY C. lAND 3ONLY D. l,2,AND3 Wednesday, April 20, 2011 Page 282 of 299

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2(111 MNS SRO NRC Examiiiation QUESTION 95 95-eneraI Discussion In accordance with NSD 414 (Fuel Handling) AND OMP 2-2 (Conduct of Operations):

Fuel Handling SRO may approve use of fuel handling bypass interlocks as necessary when NOT specified by an approved procedure.

In accordance with NSD 414, the responsibilities of the OSM include:

During fuel movement, fuel receipt, special projects, and dry cask storage:

1. Ensure SROs/ROs are cognizant of all fuel handling activities in progress or planned.

2. Maintain awareness of any activities that could impact fuel handling activities and ensure appropriate fuel handling personnel are aware of these activities.

3. Ensure appropriate response and notifications to any abnoinial fuel handling event and verify any Technical Specification implications.

4. Has ultimate responsibility for the safety of the reactor core and fuel stored on site.

5. Ensure the 91-01 Briefing is performed prior to core reload.

Answer A Discussion CORRECT: See explanation above.

Answer B Discussion 1NCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible because the Reactor Engineering has responsibility for the majority of all fuel movement activities as defined in NSD 414. The Fuel Handling SRO is correct.

Answer C Discussion rCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible because OSM has numerous oversight functions during fuel handling as defined in NSD 414. The Fuel Handling SRO is correct.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible because the Reactor Engineering has responsibility for the majority of all fuel movement activities as defined in NSD 414 and the OSM has numerous oversight functions during fuel handling as defined in NSD 414. The Fuel Handling SRO is correct.

Basis for meeting the KA The K/A is matched because the item evaluates a decision by refueling SROs.

Basis for Hi Cog Basis for SRO only SRO level because knowledge of SRO responsibilities during refueling is 1 OCFR55.43 (b) item 6/7 specific Job Level Cognitive Level QuestionType Question Source SRO Memory NEW velopment References -

Student References Provided Keferences:

NSD 414 (Fuel Handling)

OMP 2-2 (Conduct of Operations)

Thursday, April 14, 2011 Page 283 of 299

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2011 MNS SRO NRC Examination QUESTION Lesson Plan OP-MC-FH-FC Section 1.2 and 2.2

.arning Objectives: OP-MC-FH-FC Objectives 2 and 5 Learning Objective:

GEN2.1 2.1.35 GENERIC.- Conduct of Operations Conduct of Operations Knowledge of the fuel-handling responsibilities of SROs. (CFR: 41.10/43.7)

I Comments: RemarkslStatus Thursday, April 14, 2011 Page 284 of 299

Question 95

References:

From Lesson Plan OP-MC-FH-FC Section 1.2:

The following is a specific list of Fuel Handling SRO responsibilities:

1. Ensure all fuel handling activities are performed in a safe and efficient manner.

2. Securing fuel handling operations as required by Tech Specs, Plant conditions, Safety concerns, or during times of uncertainty.

3. Should monitor refueling cavity to insure FME is being maintained.

4. Maintain constant communications with the control room during core alterations.

5. Assist the control room in monitoring refueling canal level, audible count rate and EMF or containment evacuation alarms.

6. Assist fuel handling crew in visually verifying fuel assemblies are lowered and raised safely. Gives hoist operator clearance to engage or disengage on fuel assemblies. Verifies assemblies are aligned properly and down on core plate prior to giving concurrence to disengage gripper.

7. Gives verbal clearance prior to pulling control rods during control rod latching, unlatching, and drag testing activities.

8. During core alterations, approve use of fuel handling bypass interlocks as necessary when not specified by an approved procedure (NSD 414).

From Lesson Plan OP-MC-FH-FC Section 2.2:

2.2 Bypassing Fuel Handling Interlocks Objective #5 Fuel handling procedures direct bypassing an interlock when required by known specific operations. During core alterations, the Licensed SRO for Fuel Handling is tasked with approving the use of bypasses for fuel handling interlocks as necessary when not specified by an approved procedure (NSD 414).

From NSD 414 (Fuel Handling):

VERIFY HARD COPY AGAINST WEB SITE IMMEDIATELY PRIOR TO EACH USE Nuclear Policy 1anual Vc,lunie 2 NSD 414 414. FUEL HANDLING 4

14.1 INTRODUCTION

The purpose of this NSD is fo identify the roles and responsibilities for Fuel Handling activities at the three nuclear sites. iJechanical Maintenan.ce (eactor Services) is the owner of Fuel Handling and performs operation of all the tools and equipment used to move and manipulate fuel and components. eactor Engineering prcrviden the core desigas. connguration control, and astists with the coordination and oversight of fuel handling activities. Operations provide oversight and ensure reactivity management during fuel manipulations.

414.2 ROLES AND RESPONSIBILITIES 414.2.1 REACTOR ENGINEERING (RES)

A. During Fuel Movement:

Determine fuel movement seftuence.

2. Detennine .acceptable stora ge locatrons per Tech Specs.

3, Ensure reactivi monitoring is performed during reuiseling.

4, Ptvtde t .bniaa.! vtrtigkt f anttellin rttedure during unl.nul antd relead.

2. Azrisr with pre-job beug m required.

6. Provide inztuctr ens for alternate moies as requited. (must be approved by FH SkO).

7. Provide technical assistance durinr foreign. object retrievaL S. Perform plant engineering roles (i.e., core rerification, gap alignments, ensure SNM dattbase is updated, eto.) im accorda ewith Equipment Reliability Program (NSD12O)

NOTE; Operations is responsible for perforuring the SOEF..91Ol Brieg for core reload.

9. Assist in providing the following SOER91-Ol overszght før core

Enur Manarements eupectations are met.

Provide periodic oversight rs required.

3epreentatthe1-OL ieng B. During Fuel Remipt:

1. Serve as point contact for scheduling fuel and/or component receipt (i.e. rendsir interface).

The dater are established and ptovided to the FE Supervisor for ftstker notification and fliow up.

I Prepare documentation for new fuel receipt.

.3. Ensure QA inspection. bai been performed.

3. Interlace with GOvendor for evahutoa of any defects found.

5. Responsible for loading patterns in the Spent Fuel Pool.

& Ensure the SNht (Spe.aial NuaiearMatetia1r) databare it updated C. During Special Projects:

REVISION 2 VERIFY HARD COPY AGAINST WEB SITE IMMEDIATELY PRIOR TO EACH USE

VERIFV HARD COPV AGAINST WEB SITE IMMEDIATELY PRIOR TO EACH USE NSD 414 Nuclear Policy !fanual Volume 2

1. Prepare test procedure.

2. Provide technical vendor interface.

3. Provide oversight of inspection (UT ECT, visuals. P Isradiation Examinations, etc.)

4. Formulate repair plan and any necessary procedures.

5. Provide oversightofrepairs (re-cage, reconstitute).

6. Provide Job Sponsor with NSD 105 information required for in-processing ifor badging, RP.

iscellanenus in-processing derails).

7. Review an [process vendorprocedwes in accordance with NSD 703 S. Generate applicable woth requests and schedule changes.

1). Dining Thy Cask Storage

1. Approve specific FAs to be moved.

2. Ensure the SNM (Special Nuclear Materials) database is updated.

3. Fulfill the System Engmeer role for Dry Cask Storage.

a) Procedure development and review b) Perform 5O.59 and 72.48b, c) Equipment review.

d) Eqilpment interface.

e) Vendor interface,

4. Ensure required surveillances are pesfomied

5. Provide technical oversight fuel assembly loading instructions, and ensure final load verifications.

414.2.2 EQUIPMENT ENGINEER (RES)

A. Dining fuel movement, fuel receipt special projects, and dry cask storage:

1. Provtde a qualified reviewer function for procedures relating to Fuel Handling equipment and activities.

2. Develop procedures for non-routine Fuel Handling equipment.

3. Assist in solving Fuel Handling related problems.

4. Incoxporate OEP items into Fuel Handling program.

5. Originate Fuel Handling equipment mod paperwork.

6. Maintain Fuel Handling equipment drawings and manuals (documents).

7. Track and trend Fuel Handling equipment perfonnance to ensure equipment reliability.

S. Pesfonn failure analysis.

9. Assist Fuel Handling coordinator/sponsor as required.

10. Review/revise Fuel Handling PM program as required.

11. Research industry standards for equipment and tooling upgrades.

12. Ensure necessary spare parts are available for fuel handling equipment.

REVISION 2 VERIFV HARD COPY AGAINST WEB SITE IMMEDIATELY PRIOR TO EACH USE

VERJFV HARD COPY AGAINST WEB SITE IMMEDIATELY PRIOR TO EAC:H USE Nurlear Policy Manual Volume 2 NSD 413 9 Resuonsihe for PMs PTs on all Fue1in Handling equipment.

10. Resuonsibie for maintaining and nnubleshoating all Fuel Handling equipment

11. Responsible for preparing, loading, and transporting of Dry Storage Canisters.

12. Qualified certified to Fuel Handling procedures for assisted equipment.

11 Peiform procedures related to SNM (Special Nuclear Material) inventory contol related to fuel.

14. Install and maintain communications systems required for refueling activities (installation and checkout).

15. Maintain underwater lights.

16. Support special projects as needed.

17. Peifonn au fuel handling activities.

18. Operate overhead cranes and hoists as necessary dining fuel handling activities

19. Establish and maintain houzekeeniug, material condition, and Fl4E controls of all fuel handling areas. This includes Upper Containment Refueling Canal Area. Spent Fuel Pools and Fuel Receiving Areas.

414.2.6 FUEL HANDLING ADVISORS (VENDOR)

A. Provide expertise for fuel handling acuvines (cranes, hoists, tooling, including indust-v knowledge, etc.).

B. Participate as an active member of the Fuel Handling Team.

C. Cat peifoim the following:

Review procedures.

Provide hands on work as requested and approved by the Job Sponsor.

4142.7 OPERATIONS SHIFT MANAGER (OPS)

NOTh: Operations is responsible for performing the SOER. 9 1-01 Briefing for core reload.

A. Dining fuel movezient, fuel, receipt. special project, and dry cask storage:

1. Ensure SROs ROs are cognizant of all fuel handling activities in progress tr planned.

2. Maintain awareness of any activities that could impact fuel handlIng activities and ensure appropriate fuel handling personnel are aware of these activities.

3. Ensure appropriate response and notifications to any abnormal fuel handling event and veri any Technical Specification implications.

4. Has ultimate responsibility for the safety of the reactor core and fuel stored on site.

Ensure the 91-01 Briefing is performed prior to core reload.

414.2.8 CONTROL ROOM SRO AND RO (OPS)

A. Dining fuel movement, fuel receipt special projects, and dry cask storage:

1. Monitor the Nuclear Instrumentation during core alterations.

2. Implement any responses required by Abnormal Procedures.

REVISION 2 VERIFY HARD COPY AGAINST WEB SITE IMMEDIATELY PRIOR TO EACH USE

VERIFY HARD COPY AGAINST WEB SITE IMME]>LTELY i>RIOR TO EACH USE NSD 314 Nuclear Policy Manual Vohime 2

3. Log, erifj. and. maintain Technical Specication for Mode 6. Core Alterations, and other Technical Specifications for Spent Fuel Building activities.

4. Maintain awareness of fuel handling and Spent Tuel Building activities (i.e. logging.

tunver et.).

5. Maintain awareness of core congurntion during core alterations.

6. Ensure reactivity nsomtoiing is performed during refueling.

4142.9 REFUELING SRO RESPONSIBLE FOR FUEL HANDLING A. rhunig core alterations:

1. Shall be resent inth eattorEuilding to ctbserre and provide oveatight of fuel handling activities anytime Core Alterations are being perfonaae&

2. Shall have an SRO Li.cense or a SRO license luatited to fnel han.diing.

3. Maintain a winking knowledge ofprocedtues and Teclinical Specifications araccisted with fuel handling and consniand inunedinte action as recuired,

3. Approve use of fuel handling bypass interlocks as iiecessaay when not specified by an uppro red procedure.

5 Approve alternyte fuel assembly moves as recommended byRe.actor Engineering.

& The Refuelina 51W should be stationed on the refuelina bndae anytime Fuel Assemblies are being moved in the Reactor.

7. The !.&eling SRO will ensure the following:

a) Fuel flandling Procedinat are performed as written, is) kll refl.ielinn personnel adhere to STAkSelf-checkin tinhniqier procedure use and adherence, oomnsunication standards and independent verification.

c) Understands the need fur and approve nil contingency actions which may be required. in accordance with. Maintenance procedures for operating the Reactor Building Manipulator Crane.

Direct eactnr building Acctvitiez during performance of Abnorina Procedures.

e) No activities occur that adversely affect reactivity controL fi Foreirn Material Exclusion controls are implemented per NSD l4 in the Refueling Canal area and that all housekeeping standards are maintained.

g) Assure approved saitiv practices are followed during cpeia*ion of the Manipulator Crane.

h) Suspend all refueling operations anytime heishe thialci refueling operations are not being performed correctly or safely.

414.2.10 TRAINING A. Develop and maintain initial tiaining for designated Maintenance, Operations, and contract personnel on fisel handling topics B. Assist in the development of Just in Time (J1TI) on relevant fuel handling topics using the systematic approach to training (SAT) process. Provide this training for the above designated personnel to maintain a wei qualified work force for sate and efficient fuel handling operations 1

and to maintain awareness of NSD 414 and fuel handling related issues.

6 REVISION 2 VERIFY HARD COPY AGAINST WEB SITE IMMEThIATELY PRIOR TO EACH USE

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2011 MNS SRO NRC Examination QUESTION 96 GEN2.2 2.2.18 GENERIC Equipment Control uipment Control nowledge of the process for managing maintenance activities during shutdown operations, such as risk assessments, work prioritization, etc.

(CFR: 41.10/43.5/45.13)

In accordance with SOMP 02-02 (Operations Roles In The Risk Management Process):

In MODE 3, risk assessment during non-core business hours for emergent maintenance activities is performed ONLY by the (1)

Defense in Depth (DID) assessments (2) performed during MODE 3.

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

A. 1.WCCSRO

2. are B. 1. CRS

2. are C. 1.WCCSRO

2. are NOT D. 1. CRS

2. are NOT Wednesday, April 20, 2011 Page 285 of 299

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2011 MNS SRO NRC Examination QUESTION 96 General Discussion In accordance with SOMP 02-02 one of the responsibilities of the WCC SRO is:

During non-core business hours when the Work Window Manager (WWM) or Risk Site Expert is unavailable, it is the responsibility of the WCC SRO to evaluate the current risk when the existing conditions do NOT match those evaluated based on the schedule due to emergent work or schedule carry-overs.

With regards to Defense in Depth assessment:

An on-shift SRO is designated the DID Sheet Coordinator and is responsible for maintaining the current risk status by updating the DID sheet when a condition affecting the key safety functions changes. Details on completing the DID sheet are found in NSD 403 (Shutdown Risk Management (Modes 4, 5, 6, and No-Mode) per 10 CFR 50.65(a)(4)).

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part 1 is correct.

Part 2 is plausible because Defense in Depth (DID) assessments are part of Shutdown Risk Management. The unit is shutdown in MODE 3.

Therefore, it is plausible for the applicant to conclude that DID assessments would be performed in this MODE.

Answer B Discussion NCORRECT: See explanation above.

T LAUSIBLE:

Part 1 is plausible since the CRS has very specific responsibilities described in SOMP 02-02 with regards to the Risk Management Process.

Part 2 is plausible because Defense in Depth (DID) assessments are part ofShutdown Risk Management. The unit is shutdown in MODE 3.

Therefore, it is plausible for the applicant to conclude that DID assessments would be performed in this MODE.

Answer C Discussion CORRECT: See explanation above.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part 1 is plausible since the CRS has very specific responsibilities described in SOMP 02-02 with regards to the Risk Management Process.

Part 2 is correct.

Basis for meeting the KA The K/A is matched because the applicant must possess detailed knowledge of how risk assessments are performed during different MODEs of operation.

Basis for Hi Cog Basis for SRO only This is an SRO-Only question as designated by 10 CFR 55.43(b)(5):

ssessment of facility conditions and selection of appropriate procedures ing normal, abnormal, and emergency situations.

For this particular instance the SRO applicant must have knowledge of the administrative procedural requirement for performing risk assessments for maintenance activities. Since these risk assessments have been put in place to meet the regulatory requirements of 10 CFR 50.65 (a) (4) (Maintenance Rule) it constitutes required SRO knowledge.

Thursday, April 14, 2011 Page 286 of 299

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2011 MNS SRO NRC Examination QUESTION 96 Job Level Cognitive Level QuestionType Question Source SRO Memory NEW Development References Student References Provided

References:

SOMP 02-02 (Operations Roles in the Risk Management Process)

Learning Objectives: OP-MC-ADM-OMP Objective 37 GEN2.2 2.2.18 GENERIC Equipment Control Equipment Control Knowledge of the process for managing maintenance activities during shutdown operations, such as risk assessments, work prioritization, etc.

(CFR: 41.10/43.5/45.13) 401-9 Comments: I RemarksIStatus Thursday, April 14, 2011 Page 287 of 299

Question 96

References:

From SOMP 02-02:

SONLP 02-02 Page 9 of 25 5.4 Operations Work Process Manager (OWPM):

5.4.1 The OWPM is responsible for the day-to-day oversight of OPS interface with the risk management process management.

5.4.2 The OWPM will assist in resolution of scheduling conflicts.

5.4.3 The OWPM is responsible for the day-to-clay inanagemeat of protected Equipment Postings and monitoring coasistency of implementation.

5.5 Operations Shift Manager (OSM):

5.5.1 The OSM maintains overall responsibility for control of the key shutdown safety functions during outages. The primary method of controlling key safety functions is the prior approval by the OSM of any changes in plant configuration that will change the DID sheet.

5.5.2 The OSM maintains an awareness of current Electronic Risk Assessment color conditions for each imit.

5.5.3 The OSM provides guidance and direction during resolution of any scheduling conflicts identified by the Electronic Risk Assessment tool.

5.5.4 WIWN entering an orange or red condition *oni emergent work, the OSM will evaluate the restoration plan and have final authority on whether the plan is implemented. Additionally, at their discretion., the OSM may require development of a written risk management plan for actions to be taken in the event of further degadatioas. Refer to NSD 515 (Operational Decision Making) to determine if a Unit Threat Team is warranted.

5.5.5 The OSM is responsible for communicating the risk assessment results, DID sheet stams, and Protected Equipment status to OPS Shift personnel at the beginning of each shift.

5.6 Work Control Center SRO (WCC SRO):

5.6.1 Prior to releasing on-line work, an SRO assigned to the WCC will verify the work is part of the committed schedule, has the correct PRA code for tile current plant configuration, and is being perfoirned at the scheduled time.

5.6.2 During non-core business hours when the Work Window Manager (WWM) or Risk Site Expert is unavailable, it is the responsibility of the WCC SRO to evaluate the current risk when the existing conditions do NOT match those evaluated based on the schedule doe to emergent work or schedule carw-overs.

8. Outage Risk Management (Modes 4 6 and No Mode)-

8.1 Prior to the start of an outage, the OPG will assist Work Control in a detailed schedule review utilizing the Electronic Risk Assessment tool results to ensure maintenance of adequate defense in depth.

8.2 The OWPG develops outage tagouts and outage plans to assure adequate Defense in Depth (DID) is maintained fr Key Safety Functions.

8.3 The OSM maintains overall responsibility for control of the key shutdown safety functions during outages. The primary method of controlling key safety flunctions is the prior approval by the OSM of any changes in plant configuration that will change the DID sheet, 8.4 An on-shift SRO is desiøiated the DID Sheer Coordinator and is responsible for maintaining the current risk status by updating the DII) sheet when a condition affecting the key saftri ftinctions changes. Details on completing the DID sheet are found in NSD 403 (Shutdown Risk Management (Modes 4, 5.6, and No-Mode) per 10 CFR 50.65(a)(4)).

8.5 The OPS Manager in the 0CC is responsible for communicating DID sheet status and Protected Equipment status to OPS Shift personnel at the beginning of each shift and at 0CC status meetings.

8.6 OPS personnel in the Control Room, Outage Command Center, Outage Execution Groups. and Work Control Center review the status of the DID sheet at the beginning of each shift and notify the on-duty DII) Sheet Coordinator of any discrepancies between actual plant configuration and current DID sheet status and of any upcoming changes that would affect the status.

8.7 All conflg;ration changes that impact the time to boil shall be approved by the OSM and be documented in AutoLog by the DID Sheet Coordinator.

8.8 OPS personnel, when requested by Work Control, will assist in the reassessment of shutdown risk using the Electronic Risk Assessment tool when changes to the outage llan warrant.

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24111 MNS SRO NRC Examination QUESTION 97 GEN2.2 2.2.19 GENERIC Equipment Control iquipment Control nowledge of maintenance work order requirements. (CFR: 41.10 / 43.5 I 45.13)

Given the following conditions on Unit 2:

Several upcoming maintenance activities to be worked concurrently have been evaluated and it is determined that a Risk Management Plan is required

The Risk Management Plan does NOT meet the exemption requirements of NSD 213 (Risk Management Process)

In accordance with SOMP 02-02 (Operations Role In The Risk Management Process):

The Electronic Risk Assessment Tool color associated with the LOWEST risk condition where a Risk Management Plan must be generated is (1)

The Risk Management Plan shall be approved by the (2)

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

A. 1. RED

2. Station Manager B. 1. ORANGE

2. Station Manager C. 1. RED

2. Plant Operations Review Committee D. 1. ORANGE

2. Plant Operations Review Committee Thursday, April 14, 2011 Page 288 of 299

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2011 MNS SRO NRC Examination QUESTION 97 eneraI Discussion In accordance with SOMP 02-02:

For planned work resulting in an Orange condition in the Electronic Risk Assessment tool the following actions are required.

7.5.1 WHEN entering a planned activity which the Electronic Risk Assessment tool has assessed as an orange condition, there must be a written Risk Management Plan approved by the Plant Operations Review Committee (PORC) unless it meets the exemption requirements in NSD 213 (Orange result is caused by a single work activity). The Work Control organization has lead responsibility for ensuring these plans are developed with input from other groups as necessary.

Answer A Discussion fNCORRECT: See explanation above.

PLAUSIBLE:

Part 1 is plausible because a RED condition does require a Risk Management Plan.

Part 2 is plausible because the Station Manager approves Critical Activity Plans in accordance with NSD 213 (Risk Management Process).

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part 1 is correct.

irt 2 is plausible because the Station Manager approves Critical Activity Plans in accordance with NSD 213 (Risk Management Process).

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part 1 is plausible because a RED condition does require a Risk Management Plan.

Part 2 is correct.

Answer D Discussion CORRECT: See explanation above.

Basis for meeting the KA The KIA is matched because the knowledge of the requirements in SOMP 02-02 related to risk assessments and when planned work requires a Risk Management Plan is part of the Work Release Process.

Basis for Hi Cog Basis for SRO only This is an SRO-Only question as designated by 10 CFR 55.43(b)(5):

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

For this particular instance the SRO applicant must have knowledge of the administrative procedural requirement for performing risk assessments for maintenance activities. Since these risk assessments have been put in place to meet the regulatory requirements of 10 CFR 50.65 (a) (4) (Maintenance Rule) it constitutes required SRO knowledge.

Job Level Cognitive Level Questionrype Question Source SRO Memory NEW Thursday, April 14, 2011 Page 289 of 299

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2011 MNS SRO NRC Examination QUESTION 97 veIopment References Student References Provided ferences:

SOMP 02-02 (Operations Role In The Risk Management Process Learning Objectives: OP-MC-ADM-MRA Objectives 4 and 6 GEN2.2 2.2.19 GENERIC Equipment Control Equipment Control Knowledge of maintenance work order requirements. (CFR: 41.10 / 43.5 / 45.13) 401-9 Comments: RemarkslStatus Thursday, April 14, 2011 Page 290 of 299

Question 97

References:

From SOMP 02-02:

so: 02-02 Page 11 of25 7.4 For planned work resulting in a Yellow condition in the Electronic Risk Assessment tool the following actions are required:

7A. 1 Risk management actions for yeflow conditions are focused on providing increased risk awareness.

7.4.2 OPS supervision shall discuss the planned work activity within their organization and with the Maintenance personnel performing the work to increase Operator and Maintenance awareness of the associated risk.

7.5 For planned work resulting in an Orange condition in the Electronic Risk Assessment tool the following actions are reired.

T5. I WREN entering a planned activity vthicb the Electronic. Risk Assessment tool has assessed as an orange condition. there must be a written Risk Management Plan approved by the Plant Operations Review Committee (PORC) unless it meets the exention requirements in NSD 213 (Orange result is caused by a single work activity). The Work Control organization has lead responsibility for ensuring these plans are developed with input fiom other groups as necessary.

7.5.2 The WCC SRO will verify the work schedule and OWPM guidance on the activities then release the work for execution. OPS must ensure that Maintenance understands the risk level of their work and any required actions as designated in a Risk Management Plan.

7.6 Entry into a valid red condition is NOT normally allowed and will be scheduled without PORC approval and a written risk management plan. PORC approved entry into red conditions will conform to the requirements for planned orange conditions.

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2011 MNS SRO-NRC Examination QUESTION 98 GEN2.3 2.3.4 GENERIC Radiation Control

{adiation Control nowledge of radiation exposure limits under normal or emergency conditions. (CFR: 41.12 / 43.4/ 45.10)

Given the following conditions:

A Large-Break LOCA has occurred on Unit 2

The 2A ND pump tripped after starting

The ECCS is in cold leg recirculation using the 2B ND pump

A SITE AREA EMERGENCY (SAE) is in effect

The OSM has determined that an effort must be made to make the 2A ND pump functional

A team will be dispatched from the OSC to repair the 2A ND pump

The maximum dose rate in the area of the 2A ND pump is 20 REM/hour Considering ONLY the dose rate received at ND Pump A, the MAXIMUM stay time is (1) minutes.

In accordance with RP-003 (Site Area Emergency), Enclosure 4.4 (Request for Emergency Exposure) the emergency exposure shall be APPROVED by the (2)

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

A. 1.30

2. Radiation Protection Manager B. 1.30

2. Emergency Coordinator C. 1.75

2. Radiation Protection Manager D. 1.75

2. Emergency Coordinator Thursday, April 14, 2011 Page 291 of 299

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- 2011 MNS SRO NRC Examination QUESTION 98 9g eneral Discussion According to RP/0/A15700/003 (Enclosure 4.4, p1; Rev 26), the TEDE allowed during emergencies to protect valuable property is 10 Rem. If the dose rate at the A ND Pump is 20 Rem/hour, the workers can be authorized to work for 30 minutes (i.e. stay time). According to RP101A157001003 (Enclosure 4.4, p1; Rev 26), the authorization of emergency exposure limits must be acknowledged by the RPM and approved by the EC or EOF Director.

Answer A Discussion NCORRECT: See explanation above.

PLAUSIBLE:

Part I is correct.

Part 2 is plausible because the RPM must sign ACKNOWLEDGING the planned emergency exposure.

Answer B Discussion CORRECT: See explanation above.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part I is plausible because this answer would be correct if the applicant confuses the allowed exposure for lifesaving/protection of the public with the allowed exposure for protection of valuable property.

2 is plausible because the RPM must sign ACKNOWLEDGING the planned emergency exposure.

Answer 0 Discussion INCORRECT: See explanation above.

AUSIBLE:

Part 1 is plausible because this answer would be correct if the applicant confuses the allowed exposure for lifesaving/protection of the public with the allowed exposure for protection of valuable property.

Part 2 is correct.

Basis for meeting the KA The KA is matched because the SRO applicant must demonstrate knowledge of radiation exposure limits under emergency conditions.

Basis for Hi Cog This is a higher cognitive level question because it requires more than one mental step. First the applicant must recall from memory the emergency exposure limit for protection of valuable property. The applicant must then calculate the stay time based on the dose rate provided and compare the result to the exposure limit recalled from memory. The applicant must also recall from memory who is required to approve the emergency exposure.

Basis for SRO only This is an SRO-only question as it applies to 1 OCFR55.43(b)(4) (Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions):

Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures (specifically what are the emergency exposure limits during E-Plan implementation, and who must authorize the use of these limits).

Job Level Cognitive Level QuestionType Question Source SRO Comprehension BANK Bank 3214 velopment References Student References Provided

.,.eferences:

RP-003 (Site Area Emergency)

Learning Objectives: N/A Thursday, April 14, 2011 Page 292 of 299

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2011 MNSSRO NRC Examination - QUESTION 98-GEN2.3 2.3.4 GENERIC Radiation Control adiation Control

.nowledge of radiation exposure limits under normal or emergency conditions. (CFR; 41.12 / 43.4/45.10) 401-9 Comments: RemarkslStatus Thursday, April 14, 2011 Page 293 of 299

Question 98

References:

There are NO MNS Learning Objectives associated with this KIA.

From RP-003 (Site Area Emergency):

Enclosure -L4 RPiO/Ai57OO/OO Request for Eiergeucy Exposuie (a) Page 1 of 1 Acrivirv Trotal Effective Dose Lens of Eve Other Orans (b)

Equivalent (TEDE)

All S rem 15 rem SO rem Protecting Valuable 10 rem 30 rein 100 rem Property Lifesaving or Protection 25 rem 75 rem 250 rem of Large Populations Lifesaving or Protection >25 rem >75 rein 250 rem of Large Populations (c)

(a) Exciudes declare4 pregnant women.

(b) Includes skin and body exireinities.

(c) Onlr on a volunteer basis to persons filly aware of the risks involvecL All factors being equal, select volunteers above the age of 45 and those who normally encounter little eposnre.

RP Badge No Age Employer Signature of JadMdual My signature indicates my acknowledgement that I have been informed that I may be exposed to the levels ofradiation indicated aboVe. I haive been fully briefed on the task to be accomplished and on the risks of thi exposure.

I._______________________________ acknowlecge this planned Emergency Exposure JSA arr1.,Ir rrnrr,nTr nr rfl,l ,nrnnn,,rrrI - Uat Efle I, approve this planned Emergency Exposure at rgic Ccrrdiatr EOF Drr. a or r.or or rai aithoriaator) Da T Subsequent Radiation Protection Action:

- Determine need of medical evaluation

- Initiate reporting requirements per 10CFR 20

- Copy to Individua1 Exposure History File

Parent Question (Bank 3214):

Given the following conditions:

A large-break LOCA has occurred on Unit 2

The 2A ND pump tripped after starting

The ECCS is in cold leg recirculation using the 2B ND pump

A Site Area Emergency (SAE) is in effect

The OSM has determined that an effort must be made to make the 2A ND pump operable

Two NEOs and an HP Technician will be dispatched from the OSC to check the local conditions at the 2A ND pump

The maximum dose rate in the area of the 2A ND pump is 20 REM/hour

The task has been designated as protection of valuable property Considering only the dose received at ND Pump A, which ONE (1) of the following identifies the maximum stay time, who is required to approve this emergency exposure?

A. 30 minutes; Radiation Protection Manager must APPROVE the planned Emergency Exposure.

B. 30 minutes; Emergency Coordinator must APPROVE the planned Emergency Exposure.

C. 75 minutes; Radiation Protection Manager must APPROVE the planned Emergency Exposure.

D. 75 minutes; Emergency Coordinator must APPROVE the planned Emergency Exposure.

ANSWER: B

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2011 MNS SRO NRC Examination QUESTION 99 GEN2.4 2.4.11 GENERIC Emergency Procedures I Plan mergency Procedures / Plan nowledge of abnormal condition procedures. (CFR: 41.10 / 43.5/45.13)

Given the following:

Unit 1 is at 46% RTP

IC NCP bearing temperatures have been rising and are as follows:

o Thrust bearing temperatures are approximately 195°F o Pump Lower Bearing temperature is approximately 2 10°F o Motor Bearing temperature is approximately 200°F o All temperatures are rising at approximately 5°F/mm

1A, 1B, and 1D NCP bearing temperatures are stable

The RO determines that KC flow to I C NCP was lost TheICNCP (1)

The Tech Spec action statement entered after the IC NCP is stopped may be exited as soon as the unit is in (2)

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

A. 1. must be tripped immediately

2. MODE 2 B. 1. must be tripped immediately

2. MODE 3 C. 1. mayberunforupto3minutes

2. MODE 2 D. 1. may be run for up to 3 minutes

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2011 MNSSRO NRC Examinafion QUESTION 99 eneraI Discussion According to AP/l/A!5500/08, Case II, Step 2, (p14; Rev 11), the temperature limits of the 1C NCP Pump Motor Bearing (195) have already been exceed and the pump needs to be tripped. According to TS 3.4.4, Four RCS loops shall be OPERABLE and in operation. As long as the plant is in Modes 1 or 2, this LCO will be entered. When the plant enters Mode 3, LCO 3.4.4 will no longer be applicable.

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE:

The first part correct.

The second part is plausible because the operator may incorrectly conclude that the LCO is no longer applicable in Mode 2.

Answer B Discussion CORRECT: See explanation above.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE:

The first part is plausible because the operator may incorrectly conclude that the pump is at least 15°F away from a trip criteria (i.e. 3 minutes).

The applicant may incorrectly conclude that the LCO is no longer applicable in Mode 2.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

ie first part is plausible because the operator may incorrectly conclude that the pump is at least 15°F away from a trip criteria (i.e. 3 minutes).

The second part is correct.

Basis for meeting the KA The KA is matched because the operator must demonstrate knowledge of the organization of the operating procedures network for normal (i.e.

when to use OP), abnormal/emergency evolutions (i.e. operating within EP network), associated with the startup of an NCP.

Basis for Hi Cog Basis for SRO only This first part of the answers to 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 for screening questions linked to 10CFR55.43(b)(5) (Assessment and selection of procedures):

1) The question can NOT be answered solely by knowing systems knowledge.

Knowledge of operating limits for NC pumps is not systems level knowledge. This detailed procedure step knowledge contained in AP-08, Malfunction of NC Pump.

2) The question can NOT be answered by knowing immediate operator actions.

There are NO immediate actions in AP-08.

3) The question can NOT be answered solely by knowing entry conditions for AOP or direct entry conditions for EOPs.

Knowledge of NC pump operating limits is not part of the entry conditions for AP-08.

4) The question can NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of the procedure.

L5 is detailed knowledge of specific diagnostic steps within AP-08.

5) The question requires applicant to assess plant conditions and select an appropriate section of the procedure to mitigate the consequences of the malfunction. It also requires the applicant to have knowledge of diagnostic steps in AP-08 to make decision about required actions within AP 08.

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2O11-MNS SRO NRC Examination QUESTION 99 mhe second pai of the answers to this question is RO-level owledge.

Job Level Cognitive Level QuestionType Question Source SRO Memory BANK Bank 3206 Development References Student References Provided erences:

AP-08, Malfunction of NC Pump T.S. 3.4.4, RCS Loops Modes 1 & 2 Learning Objectives: OP-MC-AP-08, Objective 2 GEN2.4 2.4.11 GENERIC Emergency Procedures / Plan Emergency Procedures / Plan Knowledge of abnormal condition procedures. (CFR: 41.10/43.5/45.13) 401-9 Comments: RemarkslStatus Thursday, April 14, 2011 Page 296 of 299

Question 99

References:

From AP-08:

NS MALFUNOTION OF NC PUMP PAGE NO.

API1Lb4itYU8 1 ci 21.

ReV 12 UNIT 1.

A. Purpose The puIcse of this procedure s to ensure proper response in the event of a malfunction of an Nu punip and to identify the appropritc actions for the following caao:

Ca I NC Pump Sa1 or Pimp Lor BG3rng Malfurctbn II MC Pirrp Mtnr r Mtnr Rring Mjnctin Case Ill Excessive Vibra:ion

MNS MALFUNCTION DF NC PUMP PAGE NO.

AP111A15500/08 14 of 24 Case Rev. 12 UNIT 1 NC Pimp Motor or Motor Beating Malfunction zET::::c:z]jDc::cE B. Symptoms I NC pump stator winding temperature going up i NC pump motor bearing tempcraturcs going up I NC pump upperllower oil reservoir level computer alarm.

C. ptcL1iQn

1. Check thnormal NC pump parameter -

]Q Enclosure I (Validation of NC KNOWN TO BE VALID. Pump Piiniietrs).

2 Check NC pump pameters within IE trip criteria valid, THEN QIQ Step 5.

operating limits:

I Jl NC pump statcc winding temperatures

-LESSTHAN3II°F i All NC pump motcr beating temperature&.

- LESS THAN i95F I All NC pump oil reservoir level computer points- INDICATING BE1WEEN

(-)L25 AND (+)1.25.

3. iEAIXj)Manyoperatinglimitin Step 2 exceeded, THEN [QStep 5.

4. QIQStep6

MNS MALFUNCTION OF NC PUMP PAGE NO.

AP!I1NSSGOIO8 Case II of 24 NC Pump Motor or Motor Bearing Malfunction Rev. 12 UNIT 1

5. Stop affected NC pump as follows:

a. IF A or B NC pump is the affected pump, THEN CLOSE associated spray valve:

I 1NC-27C (A NC Loop PZR Spray Control)

I 1 NC-29C (B NC Loop PZR Spray Control).

b. Check unit status IN MODE 1 OR 2.

- b. Perform the following:

1) Stop the affected pump.

I 2) .iiNCpumpsareoffJFlEN perform the following:

I I a) Secure any boron dilution in progress.

b) I.E in Mode 3, fliN immediately open Reactor Trip Breakers.

c) If the step above results in rods dropping Pzr pressure is above P-Il, THEN GO TO EP/IIA/50001E-O (Reactor Trip or Safety Injection>.

3) IQStep 6.

c. Trip reactor.

d. WHEN reactor power less than 5%,

THEN stop affected NC pump.

e. .QIQEP/1/Al5OOO/E-O (Reactor Trip or Safety Injection).

6. Announce occurrence on paging system.

From T.S. 3.4.4, NC Loops - Modes I & 2:

RCS LoopsMODES 1 and2 3.4.4 3.4 REACTOR COOLANT SYSTEM (RCS) 34.4 RCS Loopc MODES 1 and 2 LCO 3.4.4 Four RCS loops shall be OPERABLE and in operation.

APPLICABILITY: MODES I and 2.

ACTIONS CONDITION REQUIRED AC11ON COMPLETION TIME A. Requirements of LCO A.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify each RCS loop isin operation. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months McCuire Units I and 2 3.4.4-1 Amendment Nos. 18.4/166

Parent Question (Bank 3206):

Given the following:

Unit 1 is at 46% power.

1 C NCP bearing temperatures have been rising and are as follows:

Thrust bearing temperatures are approximately 195F.

Pump Lower Bearing temperature is approximately 21OF.

Motor Bearing temperature is approximately 2OOF.

Motor Stator temperature is approximately 28OF.

All temperatures are rising at approximately 5F per minute.

IA, I B, and 1 D NCP bearing temperatures are stable.

The RO determines that KC flow to IC NCP was lost.

Which ONE (1) of the following describes the MAXIMUM amount of time IC NCP may be allowed to run, and the technical specification MODE limitation?

A. 1 C NCP must be tripped immediately; The technical specification action statement entered due to loss of the NCP will be exited as soon as the unit is in Mode 2.

B. 1 C NCP must be tripped immediately; The technical specification action statement entered due to loss of the NCP will be exited as soon as the unit is in Mode 3.

C. IC NCP may be run for up to 3 minutes; The technical specification action statement entered due to loss of the NCP will be exited as soon as the unit is in Mode 2.

D. 1 C NCP may be run for up to 3 minutes; The technical specification action statement entered due to loss of the NCP will be exited as soon as the unit is in Mode 3.

ANSWER: B

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2011 MNS SRO-NRC Examination QUESTION 100 io GEN2.4 2.4.23 GENERIC Emergency Procedures / Plan Emergency Procedures / Plan (nowledge of the bases for prioritizing emergency procedure implementation during emergency operations. (CFR: 41.10/43.5 /45.13)

Given the following conditions:

A Large-Break LOCA has occurred on Unit 1

Containment pressure is 11 PSIG

FWST level is 100 inches

ECA-1 .1 (Loss of Emergency Coolant Recirculation) is implemented Which ONE (1) of the following describes the priority for NS pump operation in this condition and the BASIS for that operation?

A. ONLY one NS pump is required to be running during ECA-1.1 and ECA procedures always have priority over FRP5.

B. Both NS pumps are required to be running because a total loss of ND causes the NS system to become relatively more important in reducing Containment pressure.

C. ONLY one NS pump is required to be running because this conserves FWST water level while providing sufficient NS flow to reduce Containment pressure.

D. Both NS pumps are required to be running because FR-Z.1 (Response to High Containment Pressure) was implemented in response to an orange path and FRPs have priority over ECA procedures.

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2011 MNS SRO NRC Examination QUESTION 100 J 100

.aneraI Discussion In this question the SRO applicant is presented with a set of conditions where a large break LOCA has occurred resulting in both a challenge to Containment Integrity (3 PSIG in containment is a SDSP Orange Path and requires implementation of FR Z. 1) and a loss of ECR capability which requires implementation of ECA 1.1.

In this situation the crew would be given conflicting guidance concerning the number of Containment Spray Pumps required (NS). FRP Z. 1 ensures that both trains of NS are in operation as long as containment pressure is greater than 3 PSIG. A containment pressure of 11 PSIG is provided in the stem of the question therefore per the guidance in FR Z. 1, both NS pumps are required. In ECA 1.1, with containment pressure between 10 and 15 PSIG, only one NS pump is required to be in operation. Normally, guidance contained in a FRP would supersede guidance in a lower priority procedure (ECA 1.1) but guidance contained in FRP Z. 1 specifically asks if ECA 1.1 has been implemented and if so, the guidance from it should be followed.

Answer A Discussion See explanation above.

rNC011ECT PLAUSIBLE:

This is plausible because ECA-1.1 does have priority for this particular instance. After Containment pressure is checked less than 15 PSIG (Red Path criteria) the remainder of FR-Z. I is performed as a Yellow Path EP and ECA-1.1 has priority.

Basis is incorrect but plausible because the applicant may correctly conclude that ECA 1.1 does have priority but misapply this to only require that one NS pump is required should a Red Path be present (15 PSIG). At a containment pressure greater than 15 PSIG, ECA 1.1 requires both NS pumps to be running.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

This is Plausible because Containment pressure is near the Containment design pressure of 15 PSIG and ND availability would be compromised e to the loss of ECR capability. It is conceivable that the plant could experience a loss of the capability to place ND in a recirc alignment and 1 be able to align NS to the containment sump. Therefore the basis is valid but in the scenario given, both NS pumps are not required.

-inswer C Discussion CORRECT: See explanation above.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

This is plausible because both NS pump would be running with Containment pressure greater than 3 psig if ECA-1.l were not in effect. This is the guidance in FR Z. 1 which is a higher priority procedure.

The basis given is incorrect but plausible because in almost all cases, FRPs do have priority over ECA procedures. I Basis for meeting the KA The K/A is matched because the applicant must know which emergency procedure has priority with regards to the operation of the NS pumps based on a given set of conditions (ECA- 1.1 or FRP-Z. 1) and the basis for that priority.

Basis for Hi Cog This question is Hi Cog because the applicant must evaluate a given set of conditions and through a multipart mental process, determine the required actions based on these conditions and recall the basis for that action.

Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev ldated 03/11/2010 for screening questions linked to 10CFR55.43(b)(5) (Assessment and selection of procedures):

1) This question can NOT be answered by knowing systems knowledge alone. This is strict procedure knowledge. This is not covered during systems training or discussed in a systems lesson plan.

2) This question can NOT be answered by knowing immediate operator actions.

This question can NOT be answered by knowing the entry conditions for AOPs. The steps to be taken by the crew are not based on the entry iditions provided.

-) This question can NOT be answered by knowing the purpose, overall sequence of events, or overall mitigative strategy of the AOPs. The question is based on knowledge of specific procedure content.

5) The question requires the applicant to have in-depth knowledge of specific steps within FRP Z. 1 and ECA 1.1 Specifically, it requires the applicant to recall that FR Z. 1 requires the crew to operate the NS pumps as directed in ECA 1.1 even though the crew is implementing a higher priority procedure. Additionally, the applicant must recall the specific guidance contained in ECA 1.1 for the given containment pressure.

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2011 MNS SRO NRC Examination QUESTION 100 io priority procedure. Additionally, the applicant must recall the specific guidance contained in ECA 1.1 for the given containment pressure.

herefore, this is SRO level knowledge.

tm Job Level Cognitive Level QuestionType Question Source SRO Comprehension BANK Bank 3108 (2009 NRC Exam Q93)

Development References OP-MC-EP-FRZ Objective 4 Student References Provided 1

FR-Z. 1, Response to Containment High Pressure rev. 14 page 4.

Lesson Plan OP-MC-EP-FRZ, Containment rev. 20 page 25 I GEN2.4 2.4.23 GENERIC Emergency Procedures / Plan Emergency Procedures / Plan Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations. (CFR: 41.10 / 43.5 / 45.13) 401-9 Comments: I Remarks!Status Thursday, April 14, 2011 Page 299 of 299

Question 100

References:

From Lesson Plan OP-MC-EP-FRZ Objective #4 From Lesson Plan OP-MC-EP-FRZ: Pg 25 of 89 STEP 9 Check containment isolation PURPOSE: To ensure non-essential containment penetrations are isolated.

BASIS: This step instructs the operator to check the Phase A and Phase B containment isolation valves closed. This prevents the release of radioactive materials from containment. It should be noted that operator actions in other procedures may have resulted in deliberate actions to defeat containment isolation of specific fluid lines.

STEP 10 Check NS System in operation as follows:

PURPOSE: To ensure NS pumps are operated as directed in ECA-1.1 instead of this procedure if ECA-1 .1 is in effect.

BASIS: This step instructs the operator to operate NS as indicated in ECA-1.1 if appropriate. This procedure specifies maximum available heat removal in order to reduce containment pressure. ECA-1 .1 permits reduced spray pump operation depending on FWST level and containment pressure. The criteria for containment spray operation is used in ECA-1 .1 since recirculation flow to the NC is not available and it is very important to conserve FWST water, if possible, by stopping the NS pumps.

From EPIIIAI5000IFR-Z-l (Response to High Containment Pressure)

Rev 17 Pg 4 of 25 kc::oNixEzD RE3ONE RE3ON5E lQT CTAND I

9. (Continued)

b. Check the following windows on Group b. Establish containment isolation on 4 of ESF Monitor light Panel LlT energized train(s) as follows:

_* C-3CONTISOLPHASEATRNA 1) IE Phase A or B valve is reqiired VLVS ALIGNED open by another EP, valve may be Ift open in next s4ep, C-6 CONT ISOL PHASE A TR!N B VLVS ALIGNED 2) Check QAC Monitor Light Program (MONL) for associated light, and G-4CONTISOLPI-IASEBTRNA close Phase A and B isolation VLVS AUG NED valves as required

. G-5 CONT ISOL PHASE B TRN B VLVS ALIGNED.

10. Check NS System in operation as follows:

a. Check EPI1LAJ5000IECA-1 I (Loa Of a. Q IQ Step 1 0d.

Emergency Coolant Recirc) IN-EFFECL This step checks to see if ECA-1 .1 is in effect If yes, then the overriding concern is the preservation of FWST inventory and use of Containment

_c. QIQStep1i. Spray is minimized, This step directs the crew to operate the Containment Spray pump per the guidance in ECA-ti instead of this Funonal Restoration Procedure.

From EP/IIAI5000IECA 1.1 (Loss of Emerg Coolant Recirc) Rev 14 Pg 8 of 101 MNS LOSS OF EMERGENCY COOLANT RECRC PAGE NO EPIIJN5000IECA-1i 8o4 101 Rev14 UNIT1

[ ACTIONJEXPECTEO RESPONSE I RESPONSE NOT OBTAINED

9. (Continued)

b. Determifle number of NS pumps required from the following tabe:

rWST LEVEL CONTAINMENT PRESSURE MS PUMPS RECUIRED GEA1ER THAN 1, PSIG 2 GREATER THAN 33 inches (FWST LOLO BETNEEN 10 P516 AND 15 P516 1 LEVEL a]rIn)

UESS THAN 10 PSIG 0 LESS IHAt4 N/A 0 33 inches (FWST LO-LO LEVtL al driN)

c. Check NS pumps running EQUAL TO

- c. Perform the following:

NUMBER REQUIRED.

1) Reset Containment Spray.

2) Operate NS pumps as required by table above.

d. CLOSE th following NS spray ve on pumps that are off;

lANSpurnp:

. INS-32A (IA NS Hx Outlet Cont Outside tso)

. 1N.S-29A (IA NS Hx Outlet Cant Outside IsoL),

lBNSpur9p;

. 1fLSI2B (IS NS Hx Outlet Cant Outside Isol)

INS-i 58 (lB NS Hx Outlet Cant Outside [sd).

Parent Question (2009 NRC Exam Q93):

Given the following conditions:

A large break LOCA has occurred on Unit 1

Containment pressure is 11 psig

FWST level is below the swap over setpoint

ECA-1 .1 (Loss of Emergency Coolant Recirculation) is implemented Which ONE (1) of the following describes the priority for NS pump operation in this condition and the basis for that operation?

A. ONLY one NS pump is required to be running during ECA-1 .1 even if containment pressure red path criteria is exceeded.

B. Both NS pumps are required to be running because FR-Z.1 was implemented in response to an orange path and FRPs have priority over ECA procedures.

C. ONLY one NS pump is required to be running because this conserves FWST water level while providing sufficient NS flow to reduce containment pressure.

D. Both NS pumps are required to be running because a total loss of ND causes the NS system to become relatively more important in reducing containment pressure.

ANSWER: C

ML112650104

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