Text

Initial Exam 2009-301 Draft SRO Written Exam
	ML092850007
Person / Time
Site:	Summer
Issue date:	10/10/2009
From:	; NRC/RGN-II
To:	; South Carolina Electric & Gas Co
References
	50-395/09-301
	Download: ML092850007 (72)
	v • d • e

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

1. 001 AA2.03 AA1.03 001lNEW/IHIGHERlISROISUMMERl2/2009INO

1. OOIINEWIIHIGHERIISROISUMMERI2I2009INO Given the following plant conditions:

Following a Refueling Outage, Outage. the crew is performing a power ascension from 80% power.

Control rods began moving OUT at 72 SPM. SPM.

Control rods were placed in MAN, MAN. and rod motion is stopped.

Control bank '0D' rods stepped out 14 steps.

Tavg is 580°F 580"F and rising.

If Tavg continues to increase.

increase, which ONE (1) of the following statements identifies the FIRST condition that will occur and the basis for the limitltrip limitllrip setpoint associated with condition?

condnion?

A . A reduction in the OT,AT A. n trip setpoint will occur; To prevent exceeding peak linear power density.

B~ A reduction in the OT&

B:' ATT trip setpoint will occur; To prevent exceeding ONB. DNB.

C. The DNB C. ONB technical specification limit for Tavg will be exceeded; exceeded;

To prevent exceeding peak linear power density.

D. The ONB DNB technical specification limit for Tavg will be exceeded; To prevent exceeding DNB ONB..

Friday, November 21,2008 Friday, 21, 2008 9:52:57 AM 1

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Feedback

A. Plausible because the 1st part Is is correct - the setpoint setpolnt for OT&~T will be penalized a.

soon as Tavg starts increasing. 2nd part is plausible because peak linear power density (Kw/ft) is the bases for the OPL1Ttrip.

OPLlTtrip.

as Incorrect because the OTAT OT~T setpeint setpoint is not calculated to maintain peak linear power density, it~ is calculated to prevent DNB.

B. CORRECT. The OT&T OT~T setpoint setpeint is calculated to prevent exceeding ONBDNB and the setpoint will be penalized as soon as Tavg starts increasing. (see IC-6, pages 16-21)

C. Plausible because, at BOL (following RF outage), the DNB lim~ of 589.2' ONB Parameter limit 589.2°FF for Tavg may be exceeded. 2nd part is plausible because peak linear power density (Kw/ft) is the bases for the OPL1Ttrip.

OPLlTtrip.

Incorrect because the OT OTAT~T setpoint setpeint is will be penalized as soon as Tavg starts increasing. Tavg must increase to 589.2°F from the given value of 580°F. Also incorrect because the OTAT OT~T setpoint is not calculated to maintain peak linear power

~ is calculated to prevent DNB.

density, it ONB.

D.

O. Plausible because, because, at BOL (following RF outage), the DNB Parameter limit lim~ of 589.2°F 589.2'F for Tavg may be exceeded. A small MTC at BOl BOL would mean that Tavg would have to significantly change to compensate for the given rod movement (14 steps).

Significantly steps). Also plausible because, for situations where the OT~ AT setpoint is penalized by large changes in .11;t1/; there could potentially be enough time lag for Tavg to exceed the ONS DNB Parameter limitlim~ before the OT&T~T setpoint is actually penalized by &1

~1..

Incorrect because the OTAT OTL\T setpoint is will be penalized as soon as Tavg starts increasing. Tavg must increase to 589.2°F from the given value of 580°F. 580°F.

Friday, Friday, November 21,20089:52:57 21, 2008 9:52:57 AM 2

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Nota Notes

Ability to determine and interpret the following as they apply to the Continuous Rep actions to be taken if automatic safety functions have not taken place.

Tier:

Group:

1 2

place.

Withdrawal: Proper RO,d Withdrawal:

Importance Rating: SR04.8 Technical

Reference:

IC-S, pp 16-21 1S-21 Proposed references to be provided to applicants during examination:

None Learning Objective: IC-S-18 IC-S-1S Question History: NEW 10 CFR Part 55 Content: 43(b)(2)

Comments:

Not an exact match to the KIA because any action taken as a result of a failure of an automatic action would be an Immediate Operator Action; Action ; therefore, would not be an SRO Only question.

Question was written to the impact that a Continuous Rod Withdrawal has on the automatic calculation of a protection setpoints and the bases for that that calculation.

SRO Only because the question tests knowledge of T.S. bases that is not a Safety Limit.

question.

Friday.

Friday, November 21, 2008 9:52:57 AM 20089:52:57 3

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

2. 001 G2.4.34 2.001 02.4.34oolIMODlFIEO//LOWERlISROISUMMERI2I2009INO 001IMODIFIED//LOWERlISRO/SUMMERl2/2009INO

Given the following plant conditions:

100% power AOP-600.1, Control Room Evacuation, has been implemented.

Because of the immediate nature of the evacuation, NO operator actions were taken prior to leaving the Control Room.

Which ONE (1) of the following describes the strategy used within the procedure to ensure that there is a secondary heat sink?

A. The NROATC will open all Reactor Trip Breakers (RTBs); THEN proceed to locally trip all of the Main Feed Pumps (MFPs).

B. The BOP will open all RTBs; RTBs; THEN report this action to the NROATC, who will THEN locally trip all of the MFPs.

C~ The NROATC will open all RTBs; THEN report this to the BOP, who will THEN C,.

locally trip all of the MFPs.

D. The BOP will open all the RTBs; RTBs; THEN proceed to locally trip all the MFPs.

Feedback

A. Plausible because the 1st part is correct.

consistent with AOP-600.1.

AOP-600.1.

correct. Also plausible because the sequence is Incorrect because the BOP locally trips the MFPs. MFPs.

B.

B. Plausible because the sequence is consistent with AOP-600.1.

Incorrect because the NROATC opens the RTBs, not the BOP. Also incorrect since the BOP trips the MFPs, not the NROATC.

C. CORRECT. According to Att. AOP-600.1 (Rev. 2), the NROATC will open all RTBs Alt. 1 of AOP-600.1 (Alt. 1, Alt Action for Step 2.a); THEN report this action to the BOP (Alt.

(Att. (Att. 1, Step 10).

According to Att.Alt. 2, the BOP will locally trip all MFPs only after verifying from the NROATC that the reactor has been tripped (Att. (Alt. 2, Caution*

Caution - Step 10, & Step 10).

D. Plausible because the sequence is consistent with AOP-600.1 AOP-600.1 and because the 2nd part is correct - the BOP will trip the MFPs.

Incorrect because the NROATC opens the RTBs, not the BOP.

Friday, November 21, Friday, 21 , 2008 9:52:57 AM 4

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Notes

(Control Rod Drive) Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.

Tiar:

Tier:

Group:

2 2

effects .

Importance Rating: SR04.1 SR04.1 Technical

Reference:

AOP-600.1 AOP-600.1 Proposed references to be provided to applicants during examination:

None Learning Objective: AOP-600.1-2258 Question History:

MODIFIED (Although written "from scratch",

scratch", this question is similar enough to Closed Reference questions AOPS 69, 395, 395, & 402 AND Open Reference question AOPS 42 to be classified as MODIFIED)

10 CFR Part 55 66 Content~

Content: 43(b)(5)

Comments:

The KIA is matched because the operator must demonstrate knowledge of the RO tasks (actions by NROATC and the BOP - open RTBs & trip MFPs), performed outside the Control Room, associated with the CRDS; CRDS; during an emergency (CR evacuation), and within the context of resultant operational effects (actions to prevent loss of heat sink).

The question is SRO-Only because it ~ requires the operator to recall detailed procedural steps from attachments within the procedure(Att. 1, Alt All Action, Step 2.a & Alt.

Att. 2, Step 10)

21,20089:52:57 Friday, November 21, 2008 9:52:57 AM 5

QUESTIONS REPORT forVCS for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION) 3.

3 . 003 AA2.03 OOIINEW/ILOWER/ISRO/SUMMER!2I2009INO 001INEW//LOWERlISRO/SUMMERl2/2009INO Given the following plant conditions:

85% power.

power.

Power Range channel N-41 is failed and has been removed from service.

One Control Bank 'C' rod bottom light energized and Control Bank 'D' moved out 10 steps.

steps.

A Quadrant Power Tilt Ratio (QPTR) indicates a QPTR of 1.06.

The dropped Control Bank 'C' rod can be recovered in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Which ONE(1) of the following completes the statement below?

Reactor Engineering must confirm the QPTR using . In accordance with the bases for Technical Specification 3/4.2.4, the power distribution limtt(s) limit(s) of _ __

is/are affected.

affected.

A. BEACON ONLY;ONLY; Fa(z). Heat Flux Hot Channel Factor, ONLY B. BEACON ONLY; B.

Fa(z). Heat Flux Hot Channel Factor AND FN Fa(z). FNMI, MI, Nuclear Enthalpy Rise Hot

C.

Channel Factor C. BEACON or In-Core moveable detectors; Fa(z). Heat Flux Hot Channel Factor, ONLY D~ BEACON or In-Core moveable detectors; D"

Fa(z).

Fa(z). Heat Flux Hot Channel Factor AND FN MI, Nuclear Enthalpy Rise Hot FNMI, Channel Factor Feedback A. 1 st part plausible because BEACON is a software system which obtains IPCS 1st data and uses ittt to update a real time analytical core model. This core model may be used for Tech Spec surveillances, and is also referred to as the Power Distribution Monitoring System (PDMS). (Ref. (Ref. STP 212.001 Core Power Distribution Measurement rev.rev. 12)

Also plausible because the 2nd half is partially correct - (Ref T.S. Bases 3/4.2.4).

limit of 1.02, at which corrective acition is required, The limtt required , provides DNB and linear heat rate generation rate protection with x-y plane power tilts. A limiting tilt of 1.025 can be tolerated before the margin for uncertainty in FQ is depleted.

depleted .

Incorrect because BEACON is not the only method used per T.S. 3/4.2.4.

Surveillance requirement 4.2.4.2 requires that the QPTR be determined within Friday, November 21.

Friday. 21,2008 9:52:57 AM 20089:52:57 6

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION) the limit when above 75 percent rate thermal power with one Power Range

Channel inoperable at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by using the PDMS (aka BEACON) Q! 2! Movable Incore Detectors to confirm that the normalized symmetric power distribution is consistent with the indicated QPTR. Also incorrect because both FNaH the quadrant power tilt FN~H and Fa(z) are affected affected.. lAW T.S. Bases 3/4.2.4, titt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation. The limit of 1.02, at which corrective acition is required, required, provides DNB and linear heat rate generation rate protection with x-y plane tilts. A limiting tilt of 1.025 can be tolerated before the margin for power titts.

uncertainty in FQ is depleted.

B. st part plausible because BEACON is a software system which obtains IPCS 1st 1

data and uses it to update a real time analytical core model. This core model may be used for Tech Spec surveillances, and is also referred to as the Power Distribution Monitoring System (PDMS). (PDMS). (Ref. STP 212.001 Core Power Distribution Measurement rev. 12) .

2nd part is plausible because it is correct. - FNdH, MI, Nuclear Enthalpy Rise Hot Channel Factor, is also a power distribution limit which is associated with DNBR vs DNB and the PDMS and Incore Thermocouple systems are also used in

determination of F N6.H. FNm. limits are also associated with the bases of the FN MI. . FNMI.

specification. The quadrant power titt rod misalignment specification. tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically. The limit of 1.02, at which corrective acition is required, required ,.provides DNB and linear heat rate generation rate protection with x-y plane power tilts. titt of 1.025 can be tolerated before the margin for titts. A limiting tilt uncertainty in FQ is depleted.

1st part incorrect because BEACON is not the only method used pert because T.S. 3/4.2.4 surveillance requirement 4.2.4.2 requires that the QPTR shall be determined within the limit when above 75 percent rate thermal power with one Power Range Channel inoperable at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by using the PDMS or Movable Incore Detectors to confirm that the normalized symmetric power Q!

distribution is consistent with the indicated QPTR.

C. Plausible because the 1st part is correct. correct. T.S. 3/4.2.4 surveillance requirement 4.2.4.2 requires that the QPTR shall be determined within the limit when above 75 percent rate thermal power with one Power Range Channel inoperable at

Friday, least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by using the PDMS 2! Q! Movable Incore Detectors to confirm that the normalized symmetric power distribution is consistent with the indicated QPTR. Also plausible because the 2nd half is partially correct - (Ref Friday, November 21 21,, 2008 9:52:57 AM 7

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION) provides DNB and linear heat rate generation rate protection with x-y plane e power tilts. A limiting tilt of 1.025 can be tolerated before the margin for uncertainty in Fa is depleted.

FN~H and FQ(z) because, both FNIIH Incorrect because, the aPTR limit Fa(z) are affected. lAW T.S. T.S. Bases 3/4.2.4, limij assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation. The limit of 1.02, at which corrective acijion acition is required, provides DNB and linear heat rate generation rate protection with x-y plane power tilts. tilts. A limiting tm limijing tilt of 1.025 can be tolerated before the margin for uncertainty in Fa is depleted.

D.

D. CORRECT. T.S. 3/4.2.4 surveillance requirement 4.2.4.2 requires that the aPTR shall be determined wijhin within the limij limit when above 75 percent rate thermal power with one Power Range Channel inoperable at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by using the PDMS !!! or Movable Incore Detectors to confirm that the normalized symmetric power distribution is consistent with the indicated aPTR.

FN~H, Nuclear Enthalpy Rise Hot Channel Factor, is a power distribution limij FNIIH, limit which is associated with DNBR vs DNB and the PDMS and Incore

e. Thermocouple systems are also used in determination of FN lAW T.S.

T.S. Bases 3/4.2.4, the quadrant power till FN~H.

are also associated with the bases of the rod misalignment specification.

FN~H. limits t1 H. FNAH.

speCification.

tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup capabilijy testing and periodically during power operation. The limij limit of 1.02, at which corrective acijion acition is required, provides DNB and linear heat rate generation rate limiting tilt of 1.025 can be tolerated protection with x-y plane power tilts. A limijing before the margin for uncertainty in Fa is depleted.

depleted .

e*

Friday, November 21,2008 Friday. 21, 2008 9:52:57 AM 8

QUESTIONS REPORT forVCS for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Notes

Ability to (a) predict the impacts of the following on the Dropped Control Rod and (b) based on predictions, predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:

operations: Dropped rod, measurements.

measurements.

rod , using in-core/ex-core instrumentation instru mentation in-core or loop temperature on those operation: Facility conditions and selection of appropriate procedures during abnormal and emergency Tier: 1 Group: 2 Importance Rating: SR03.8 SRO 3.8 Technical

Reference:

T.S. 3/4.2.4

T.S. Bases pp B 3/4 2 2-5 T.S.

T.S. 6.9.1.11 pp pp6*16, 6-16, 16a

STP-212.001,. p 3 of 17 STP-212.001 Proposed references to be provided to applicants during examination: None Learning Objective: S8+18 SB-4-18 Question History: NEW

10 CFR Part 55 Content:

Comments:

Matches the KIA 43(b)(2) 43(b}(2}

KJA in that the question tests the methodology for conducting a T after a dropped rod.

.S. surveillance T.S.

SRO Only in that it tests T.S. requirements below the double line and tests bases that are not Lim~s .

Safety Limits

Friday, November 21,21 , 2008 9:52:57 AM 9

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTERVALIDATION) AFTER' VALIDATION) 4 . 008 G2.4.6 OOllNEW/IIllGHERIISROISUMMERl2/2009INO 4.008 ool/NEW/IHIGHERIISRO/SUMMERI2I2009INO

Given the following plant conditions:

100% power

Following a plant transient a PZR Code Safety Valve lifts and remains stuck partially open causing a Small Break lOCA. LOCA.

An automatic Reactor Trip and Safety Injection occur.

All RCPs are manually tripped.

tripped.

The crew has entered the appropriate plant procedures.

The crew has been directed to check if an RCP should be started.

cond~ions :

The NROATC notes the following conditions:

RCS Subcooling 36°F 36*F

Level 100%

PZR level

PZR Pressure 1000 psig Which ONE (1) ofthe of the following identifies whether or not an RCP should be started AND the basis for this decision?

A~

A'! An RCP should be started;started; To mix the reactor coolant to a uniform temperature.

B. An RCP should be started; To ensure that PZR Spray is available for RCS pressure control.

C. All RCPs should remain secured; C. secured; To prevent an undesirable RCS pressure surge. surge.

D. All RCPs should remain secured; D. secured; To prevent excessive thermal stresses on the RCP seals. seals .

21 , 2008 9:52:57 AM Friday, November 21, 10

QUESTIONS REPORT for forVCS VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Feedback

A.

A. CORRECT. The plant has experienced a Small Break LOCA and will enter EOP-1 Reactor Trip/Safety Injection Actuation, Actuation, transition to EOP*2.0.

Coolant, and then to EOP-2.1, Secondary Coolant, At step 9 of EOP-2.1 EOP-2.0, Loss of Reactor or EOP-2.1, Post-LOCA Cooldown and Depressurization.

.0, EOP-1.0, Depressurization.

EOP-2.1 the crew will evaluate whether or not an RCP should be started.

EOP-2.1 would have already initiated inijiated an RCS cooldown to rEHlStablish In this event, EOP-2.1 re-establish RCS Subcooling, and since a PZR Steam Space Accident has occurred, PZR level will be off-scale high within minutes of the event. The ES-1.2 ES-1 .2 WOG Background Document for Step 12 (HES12BG.doc, HP-Rev 2,4/30/05, 2,4130/05, p94) states that forced circulation is the preferred mode of operation to allow for normal RCS ReS cooldown and provide PZR spray.spray.

A knowledge item associated with wijh this step (p95) states that even n if plant specific procedures require that a steam bubble be present in the PZR (which they do

[SOP-101 , Precaution 2.a.6.d,

[SOP-101, 2.a.6.d, Rev 26]),

26]). if an RCS leak path is certain (which is the case in the stated conditions), RCP Rep restart should be permitted since the leak ensures that there will NOT be a significant surge when the RCP Rep is started. This is restated on page 44,44. in a section analyzing a Steam Space Small Break LOCA. This section also indicates that the Rep ~needed for normal spray capability but served to mix the RCP is not "needed RCS coolant to nearly uniform temperature." EOP-2.1, EOP-2.1 , Step 9 (Rev 13), reflects the requirements of the Background Document. When checking to see if an Rep RCP should be started, the crew is first directed to check that all RCPs are secured (which they are).

Then, Then, the crew is directed to check RCS Subcooling > 30°F and PZR level> 30%

[50%J. When these are satisfied (as is the case in the established conditions), the

[50%].

operator is directed to (1) establish normal conditions for starting an RCP, Rep, refer to SOP-101 SOP-1 01,, and (2) start the A RCP RCP..

B.

B. Plausible because there is a large range of Small Break LOCA lOeA sizes and locations for which the PZR will NOT go water solid, and the RCP will be started to regain PZR Spray control.

Incorrect because PZR spray is not needed, needed, nor would it be effective, for this event with the PZR level at 100%.

C. Plausible because normally, an RCP Rep should not be started unless there is a bubble in the PZR <<1400 ft3). fl3) . Also plausible because a undesirable pressure surge could occur if the RCS were water solid without wijhout a large break/hole in the PZR steam space.

Incorrect because, ReS leak path is certain, because, if an RCS Rep restart should be permitted since certain, RCP the leak ensures that there will NOT be a significant pressure surge when the Rep RCP is started.

D. Plausible because normally, Rep should not be started unless there is a bubble in normally, an RCP the PZR <<1400 ft3). ft3) . 2nd part is plausible since CAUTION - Step 9 of EOP-2.1EOP-2.1 warns against RCP restart (to prevent seal failure) n if seal cooling had previously been lost.

Incorrect because, if an RCS leak path is certain, RCP restart should be permitted since the leak ensures that there will NOT be a Significant significant pressure surge when the RepRCP is started started..

21 . 2008 9:52:57 AM Friday, November 21, Friday, 11

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Notes

(Pressurizer Vapor space Accident) Knowledge symptom based EOP mitigation strategies.

TIer:

Tier: 1 Group: 1 Importance Rating: SR04.7 Technical

References:

ES-1.2 ES-l .2 WOG Background Document for Step 12 (HES12BG.doc, HP-Rev 2, 4/30105, p44, 94-95) 4/30105,

SOP-l0l , Precaution 2.a.6.d, SOP-101, 2.a.6.d, Rev 26

AB-4 (p39, Rev 12)

EOP-2.1, Step 9 (Rev 13)

Proposed references to be provided to applicants during examination:

None Learning Objective:

learning EOP-2.1-05 & 07 Question History: NEW 10 CFR Part 55 Content:

Content 43(b)(5)

Comments:

The KA is matched because the operator must demonstrate knowledge of the EOP mitigation strategies during a PZR Vapor Space Accident (Le. permit Rep (i.e. it is acceptable to permrt RCP restart even PZR, and the reason for starting an Rep) without a steam bubble in the PZR, RCP)..

The question is SRO-Only because the operator must assess plant conditions,conditions, and know the content of mitigation procedures, including the basis for this content.

Friday, November 21 21,, 2008 20089:52:57 9:52:57 AM 12

QUESTIONS REPORT forVCS for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION) 5.

5 .013 013 G2.2.42 OOI 00IlNEW/IHIGHERIISRO/SUMMERl2/2009/NO INEWfIHlG HER/fSROfSUMMERI2I2009INO

Given the following plant conditions:

100% power RM-A4, conditions:

RM-A4, Reactor Building Purge Exhaust Monitor, is Out-of-Service.

RB pressure has increased and it~ becomes necessary to place the Alternate Purge Supply and Exhaust System in operation.

following identifies the Technical Specification and/or ODCM Which ONE (1) ofthe foliowing Specification of Controls which will prevent placing the Attemate Alternate Purge Supply and Exhaust System in operation?

REFERENCES PROVIDED A. lCO A. 3.3.2, ESFAS Instrumentation; AND LCO 3.3.2, lCO Mon~oring Instrumentation.

3.3.3.1, Radiation Monitoring LCO 3.3.3.1, B~ lCO B!" 3.3.2, ESFAS Instrumentation; AND LCO 3.3.2, ODCM 1.2.1, Radioactive Gaseous Effluent Monitoring Instrumentation.

C. lCO LCO 3.3.3.1, Radiation Monttoring Monitoring Instrumentation; AND Monttoring Instrumentation.

ODCM 1.2.1, Radioactive Gaseous Effluent Monitoring Mon~oring Instrumentation ONLY.

D. ODCM 1.2.1, Radioactive Gaseous Effluent Monitoring ONLY.

Friday. 21,2008 Friday, November 21, 2008 9:52:57 AM 13

QUESTIONS REPORT forVCS for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Feedback

A. Plausible because the 1st to radiation monitoring.

1st part is correct.

correct. Also plausible because LCO 3.3.3.1 Incorrect. According to T.S. LCO 3.3,3.1, 3.3.3.1 is related 3.3.3.1, the Radiation Monitoring Instrumentation channels shown in Table 3.3-6 shall be OPERABLE. Table 3.3-6, Instrument 2.b.i, Containment - Gaseous Activity - Purge & Exhaust Isolation (RM-A4), a minimum of one channel must be OPERABLE in Mode 6. Therefore, the LCO does NOT apply at this time.

B. CORRECT. According to T.S. LCO 3.3.2, the ESFAS Instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE. Table 3.3-3, 3.3-3, Functional Unij Unit 3C2, Purge and Exhaust Isolation, a minimum of two channels required to trip must be OPERABLE in Modes 1-<1, 1-4, whenever purge exhaust is OPEN. According to G5-9, GS-9, Table GS9.4, these two channels (Le. (i.e. monijors) monitors) are RM-A2, RB Sample Line, and RM-A4, RB Purge Exhaust Monitor, both of which will automatically close the Alternate RM-A4, Purge Supply and Exhaust System Isolation Valves on high radiation.radiation. If RM-A4 is inoperable, According to TS Table 3.3-3, ACTION 17 must be addressed. This ACTION statement requires that with less than the minimum channels OPERABLE, operation may continue provided the containment supply and exhaust valves are maintained closed. According to ODCM aDeM Requirement 1.2.1, the Radioactive Gaseous Effluent Monitoring Instrumentation channels shown in Table 1.2-1 shall be OPERABLE. Table 1.2-1, Functional Unit Unij 3, Reactor Building Purge System, a minimum of one channel shall be OPERABLE at all times during releases from this pathway. According to Table 1.2-11.2-1,, this channel is RM-A4, RM*A4, RB Purge Exhaust Monitor, which has separate requirements for the gas sampler including alarm and automatic termination of the release, sampler, the particulate sampler, the flow release, the iodine sampler, measurement devices, and the sampler flow rate measuring device device.. If the gas sampler (RM-A4) is inoperable, ACTION 10 must be addressed. This ACTION statement requires that, with less than the minimum channels OPERABLE, OPERABLE, immediately suspend purging of radioactive effluents via this pathway.

pathway.

C. Plausible because the 2nd half is correct.

correct. Also plausible because LCO 3.3.3.1 3.3.3.1 is related to radiation monitoring.

monitoring.

Incorrect. According to TS LCO 3.3.3.1, 3.3.3.1, the Radiation Monitoring Instrumentation Ta~le 3.3-6 shall be OPERABLE. Table 3.3-6, Instrument 2.b.i, channels shown in Table Containment - Gaseous Activity - Purge & Exhaust Isolation (RM-A4), (RM*A4), a minimum of one channel must be OPERABLE in Mode 6. Therefore, the LCO does NOT apply at this time.

time.

D. Plausible because ODCM 1.2.1 1.2.1 is part of the correct answer.

answer.

Incorrect since both LCO 3.3.2 and ODCM 1.2.1 prohibij prohibit placing the Alternate Purge Supply and Exhaust System in operation.

operation .

Friday, November 21, 2008 9:52:57 AM 14

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Noles Notes

(Engineered Safety Features Actuation) Ability to recognize system parameters that are entry-level conditions for Tech Tier:

Group:

nical Specifications Technical Specifications..

2 1

Importance Rating: 5R04.6 SR04.6 Technical Referencea:

References:

T5 TS LCO 3.3.2, Table 3.3-3, Action 17

GS-9, Table GS9.4

ODCM Requirement 1.2.1, Table 1.2-1, 1.2-1 , Action 10

TS LCO 3.3.3.1, Table 3.3-6 refetences to be provided to applicants during examination:

Proposed references None Learning Objective: S8-4-19 56-4-19 Question History: NEW 10 CFR Part 55 Content: 43(b)(2)

Comments:

The KA is matched because the operator must demonstrate the ability to recognize system parameters (RM-A4 inoperable) that are entry-level conditions for Technical Specifications associated with the ESFAS.

The question is SRO-Only because not only does the operator need to khow the requirements of the LCO, but the operator needs to know action associated with w~h the inoperablity of RM-A4, RM-M, specifically that with RM-A4 inoperable, radioactive releases via this pathway can NOT continue .

continue.

Friday, November 21,20089:52:57 21, 2008 9:52:57 AM 15

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION) 66.. 015 AA2.10 AA2. IO 001IMODIFIEDIIIDGHERlISRO/SUMMERl2/2009INO OOI IMODlFlED/IIflGHERIISRO/Sl)MMER!2I2009INO Given the following plant conditions:

condttions:

8% power

CCW flow to the RCP Bearing Oil Coolers is lost.

The crew is addressing the appropriate Annunciator Response Procedures.

During the recovery the following RCP parameters are observed EIGHT (8) minutes after the loss of flow:

A B C

Highest Motor Bearing Temperature 190°F 190'F 191°F 196' 191'F 196°F F

Lower Seal Water Bearing Temperature 102' 102°FF 105°F 103'F 105'F 103°F

Seal Water Outlet Temperature 100°F 100'F 104°F 102'F 104'F 102°F

I&C reports that the valve can be opened in ONE (1) minute.

Which ONE (1) of the following identifies the RCP(s) that have met their trip criteria AND the procedure(s) that are REQUIRED to be addressed?

Art ONLY RCP C; MONLYRCPC; SOP-101, Reactor Coolant System; GOP-4B, GOP-4B, Pow.erOperation Power Operation (Mode 1 Descending); and GOP-5,GOP-5, Reactor Shutdown from Startup to Hot Standby (Mode 2 to Mode 3). 3) .

B. ALL RCPs; SOP-1 01, Reactor Coolant System; SOP-101, System; GOP-4B, Power Operation (Mode 1 Descending); and GOP-5,GOP-5, Reactor Shutdown from Startup to Hot Standby (Mode 2

~~~

to Mode 3) .

C. ONLY RCP C; C. C; EOP-1 EOP-1.0, .0, Reactor Trip/Safety TriplSafety Injection Actuation; and SOP-1 01, Reactor Coolant System, ONLY.

ONLY. .

D. ALL RCPs; EOP-1.0, TriplSafety Injection Actuation; and SOP-101 EOP-1 .0, Reactor Trip/Safety SOP-101,, Reactor Coolant ONLY.

System, ONLY.

Friday. 2008 9:52:51 AM 21 , 20089:52:57 Friday, November 21, 16

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Feedback

A. ARP'()()l, Rev 5, XCP-602, 2-3, Supplemental Action 1, if CORRECT. According to ARP-001, Motor Bearing Temperature reaches, affected RCPs must be stopped. If the RCPs .

must be stopped while in Mode 4 or above, and if power is > than 10%, trip the Reactor and secure the RCPs per SOP-101.

SOP-l0l . If power is < 10%, the operator is directed to SOP-l0l . When this is done, T.S.

secure the RCPs per SOP-101. T .S. LCO 3.4.1 is NOT satisfied.

The LCO requires that when in Modes 1 and 2, all RCS loops be in operation. The ACTION states that with < all RCS Loops in operation, be in at least Hot Standby within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . In order to place the plant in Hot Standby, GOPs 4B & 5 must be used.

B.

8. Plausible because the second part is correct. Plausible because the 1st 1st part would be correct if the 10 minute criteria had been exceeded. According to ARP-001, Rev 5, XCP-602, 2-3, Supplemental Action 1, if flow is lost to the bearing coolers, the RCPs must be stopped within 10 minutes or before motor bearing temperature reaches 195°F.

Incorrect because, according to the given conditions, flow will be restored within the 10 minutes. Also incorrect because only RCP 'C' has exceeded the 195°F limit lim~ on Motor Bearing Temperature.

C. Plausible because the 1st1st part is correct.

correct. Also plausible because SOP-l SOP-101 01 will will be used to shut down the RCP. EOP-1.0 would be used if power RCP. 2nd part is also plausible since EOP-l.0 conseNatively used if 10 were greater than 38% (for 1 RCP) and could possibly be conservatively minutes had been exceeded and all 3 RCPs required tripping. Additionally, initiatingin~iating a Reactor Trip would place the unit in Hot Standby, which is the desired end state state..

D.

2nd part is incorrect because the ARP specifically addresses the use of a Reactor Trip ONLY when power is > 10%.

Plausible because the 1st part would be correct if Wthe 10 minute criteria had been exceeded. According to ARP-001, Rev 5, XCP-602, 2-3, Supplemental Action 1, if flow is lost to the bearing coolers, the RCPs must be stopped ~hin within 10 minutes or before motor bearing temperature reaches 195°F. Also plausible because SOP-101 will be used to shut down the RCP. 2nd part is also plausible since EOP-l EOP-1.0 .0 would be used if power were greater than 38% (for 1 RCP) and could possibly be conservatively conseNBtively used if Add~ionally, initiating 10 minutes had been exceeded and all 3 RCPs required tripping. Additionally, a Reactor Trip would place the unit in Hot Standby, which is the desired end state.

Incorrect because only RCP 'C' has exceeded the 195°F limit lim~ on Motor Bearing Temperature.

conditions, flow will be restored within the 10 Incorrect because, according to the given conditions, minutes. 2nd part is incorrect because the ARP specifically addresses the use of a Reactor Trip ONLY when power is > 10% 10%..

Friday, November 21 21,, 2008 9:52:57 AM 17

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION VALIDATION))

Notes

Ability to (a) predict the impacts of the following on the Rep predictions, operation:

Tier:

Group:

RCP Malfunctions and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: W hen to secure Reps RCPs on loss of cooling or seal injection.

1 1

Importance Rating: SRO 3.7 Technical

References:

ARP-001 ARP-001,, Rev 5. 5, XCP-602 AP2-3.

AP2-3, Supplemental Action 1 and 6

TS LCO 3.4.1 3.4.1 and its Basis Proposed references to be provided to applicants during examination:

None Learning Objective: AOP-11B.1-09 AOP-118.1-09 Question History:

QUestion MODIFIED (Although written "from '1rom scratch", this question is similar enough to Closed Reference questions AOPS 379,360, 359, 312,71 379,360,359,312, 71 to be classified as MODIFIED)

10 CFR Part 55 Content Comments:

Content: 43(b)(2),(5)

The KA is matched because the operator must demonstrate the ability to determine when an Rep RCP must be tripped, by interpreting a set of given conditions, during a loss of cooling COOling to the RCPs.

The question is SRO-Only because it requires the operator to recall a strategy or action that is written in a plant procedure (adion(action to take when all Reps RCPs must be tripped),

tripped) , and when to take this strategy << 10% power), AND the question involves application of required TS ACTION (TS 3.4.1) .

LCO 3.4.1).

Friday, November 21, 2008 9:52:57 AM 18

QUESTIONS REPORT forVCS for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION) 7.

7 . 026 G2.2.44 G22.44 00 IIMODlFlED/lHIGHERil1ll OOllMODIFIED//IllGHERllll1 conditions:

Given the following plant conditions:

Mode 1 82 Maintenance Week is in progress Annunciator CCW SRG TK LVL HIILO/LO-LO (XCP-601(XCP-601,, 1-1) has just actuated.

The NROATC has determined that ttit is a LOW level.

CCW Surge Tank 'A' level is decreasing slowly.

Given the following equipment nomenclature:

LCV-7088, MU TO SRG TK (8 SIDE)

LCV-70BB,

LCV-709B, LCV-7098, MU TO SRG TK (A SIDE)

PVG-9627A, SW TO CC LOOP A PVG-9627A, Which ONE (1) of the following describes the response ofthe of the CCW System makeup valves and the actions that must be taken?

LCV-7088 and LCV-709B A. ONLY LCV-70BB LCV-7098 will be open; Swap the running charging pump per SOP-102 SOP-1 02 and apply actions forT.S.

for T.S. 3.1.2.2,

8~

Flow Paths - Operating ONLY LCV-7088 LCV-70BB and LCV-7098 LCV-709B will be open; Swap CCW active loops per SOP-11B Cooling Water System open; SOP-118 and apply actions for T.S. 3.7.3, Component C. LCV-70BB, LCV-708B, LCV-709B, PVG-9627A will be open; LCV-709B, and PVG-9627Awill open; Swap the running charging pump per SOP-1 SOP-10202 and apply actions for T.S.

T.S. 3.1.2.2, Flow Paths - Operating D.

LCV-70BB, LCV-709B, and PVG-9627A will be open; Swap CCW active loops per SOP-11B SOP-118 and apply actions for T.S. 3.7.3, Component Cooling Water System

Friday, November 21, 20089:52:57 2008 9:52:57 AM 19

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Feedback

A. Plausible because the 1st Action 1.b., lCV*70BB 1st part is correct. According to XCP-601, LCV-7088 and lCV*7090 LCV-7090 open on a lOW plausible because it is partially correct*

XCP-601 , 1-1, 1.1 , Automatic LOW level. 2nd part is correct - Supplemental Action 6 of XCP-601, requires the inservice Charging Pump to be transferred to the active CCW loop XCP-601 , 1-1, 1* 1, SOP*l02.

per SOP-102.

The 2nd part is incorrect because the Action of 1.S. T.S. 3.1 .2.2 does not apply.

3.1.2.2 While the original inservice Charging Pump will not be available as a Boration Path,, the LCO requires only two of the three identified flowpaths to be Flow Path OPERABLE. Since the gravity drain path and the other charging pump are available, the LCO is met, and no actions apply.

B. CORRECT. According to XCP-601, 1-1, 1*1 , Automatic Action 1.b., LCV*70BB LCV-7088 and LCV-7090 open on a lOW LCV*7090 LOW level. Supplemental Action 5 of this ARP states: "On low-low level, if level continues to decrease, establish Train B as the low level or low*low active loop per SOP* SOP-118. W~h one train inoperable, refer to Tech Spec 3.7.3.

llB. With C. Plausible because the 1st 1st part of the 1st 1st part is correct. According to XCP-601 ,

1-1, Automatic Action 1.b., LCV*70BB 1*1, LCV-7088 and LCV-7090 LCV*7090 open on a LOW level. Also PVG-9627A will eventually open below the low*low plausible because PVG*9627A low-low level if the leak is not isolated. Incorrect since PVG-9627A will not open on a lOW level..

LOW level 2nd part is plausible because it ~ is partially correct*

correct - Supplemental Action 6 of XCP-601, 1*1 1-1,, requires the inservice Charging Pump to be transferred to the active CCW loop per SOP*l SOP-102.02.

The 2nd part is incorrect because the Action ofT.5. of T.S. 3.1.2.2 does not apply.

While the original inservice Charging Pump will not be available as a Boration Flow Path, the LCO requires only two of the three identified flowpaths to be OPERABLE. Since the gravity drain path and the other charging pump are available, the lCO LCO is met, and no actions apply.

D. Plausible because the 1st part of the 1st part is correct. According to XCP-601, 1-1, Automatic Action 1.b., LCV*

1*1, LCV-7088 LCV-7090 open on a lOW 70BB and LCV*7090 LOW level. Also plausible because PVG-9627 PVG*9627A will eventually open below the low*low low-low level if the leak is not isolated.

PVG*9627A will not open on a LOW level. Supplemental Action 5 Incorrect since PVG-9627 of this ARP states: "On low level or low*lowlow-low level, if level continues to decrease, decrease, establish Train B as the active loop per SOP*llB.SOP-118. With one train inoperable, inoperable, refer to Tech Spec 3.7.3.3.7.3 .

Friday, November 21, Friday, 21 , 2008 9:52:57 AM 20

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION) forVCS Notes

(Loss of Component Cooling Water) Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

Tier:

Group:

1 1

Importance Rating: SR04.4 Technical

References:

XCP-601, 1-1 (Rev 5)

T.S. 3.7.3 T.S.3.7.3

T.S . 3.1.2.2 T.S.3.1.2.2 Proposed references to be provided to applicantB applicants during examination:

None Learning Objective: 1B-2-21 IB-2-21 Question Hiatory:

History:

MODIFIED (Although written "from scratch", this question is similar enough to Closed

Reference question CCW SYSTEM 136 to be classified as MODIFIED) 10 CFR Part 55 Content:

Comments:

43(b)(5)

The KA is matched it involves interpretation of CR indications to verify status of the CCW system (CCW Surge Tank alarm)and understanding of how actions affect the plant (application of ARP Supplental Actions and INOPERABLE CCW Train)

The question is SRO-Only because the operator must assess the stated conditions, conditions, which are abnormal, and apply a detailed step from the ARP Supplemental Actions (specific mitigation m~igation strategies).

strategies) .

Friday, November 21,2008 21, 2008 9:52:57 AM 21 21

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION) 88.. 034 A2.03 OOI 001INEW//lllGHERI/SRO/SUMMERl2/2009INO INEWIIIflGHERIISROISUMMERI2I2009INO Given the following plant condnions:

conditions:

A core reload is in progress per REP-107.013, Core Reload.

The Control Room crew notes the following:

Source Range NI N-31 is reading 5 cps and stable.

Source Range NI N-32 is reading 5 cps and stable.stable.

Fuel assembly U13 is being moved to the 3X3 array near N-31. N-31 .

The operator on the Manipulator Crane reports that the assembly has been lowered into the core at location J-2.

The Control Room crew then notes the following:

Source Range NI N-31 is reading 10 cps and stable.stable.

Source Range NI N-32 is reading 6 cps and stable.stable.

Which ONE (1) of the choices below completes the following statement?

Given these conditions, the fuel assembly is mispositioned based on _ _ _ and the refueling team should _ _ ___ '

REFERENCE PROVIDED

A. the response of N-31 ;

unlatch the assembly and perfonn B. being in the wrong location; reliabiltty check of N-31 perform a new statistical reliability A~erations unlatch the assembly and suspend Core Alterations C. the response of N-31 N-31;;

NOT unlatch the assembly and perfonn perform a new statistical reliability check of N-31 N-31, D~ being in the wrong location; NOT unlatch the assembly and suspend Core Alterations

Friday. November 21 Friday, 21,, 2008 9:52:57 AM 22

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Feedback e* A. REP-107.013, Core Reload, Section 3.3.7, count rate Plausible for 1st part because per REP-l07.013, doubling is expected as the three fuel assemblies are loaded nearest each NI. NJ. In the stated conditions, a fuel assembly is being placed in the 3X3 array near NI-31 and therefore count rate doubling on Source Range Instrument NI-31 would be expected. expected.

Also plausible because it is possible that a statistical reliability check would be performed if ~ SR counts behaved unexpectedly.

infonnation given in the stem of the question and Incorrect because, based upon the information the provided Material Transfer Form, the fuel assembly has been placed in the wrong location. Also incorrect because the N-31 response is acceptable and a statistical reliability check is NOT required.

B. Plausible because the 1st part is correct correct. Also plausible because Core Alterations would be suspended.

Incorrect because, if ~ the assembly is placed in the wrong posaion, position, the Reactor Engineer, or his designee, should NOT grant permission to unlatch, which would, would. in effect, suspend Core Alterations.

C. Plausible for 1st part because per REP-l07.013, REP-107.013, Core Reload, Section 3.3.7 count rate doubling is expected as the three fuel assemblies are loaded nearest each NI. NJ. In the stated conditions, a fuel assembly is being placed in the 3X3 array near NI NI-31 and,

-31and.

therefore, count rate doubling on Source Range Instrument NI-31 would be expected.

e* Also plausible because it is possible that a statistical reliability check would be performed if SR counts behaved unexpectedly.

because. based upon the information given in the stem of the question and Incorrect because, the provided Material Transfer Form, the fuel assembly has been placed in the wrong location. Also incorrect because the N-31 response is acceptable and a statistical reliabilay check is NOT required reliability required..

D. CORRECT: eased Based upon the information given in the stem of the question and the provided Material Transfer Form, the fuel assembly has been placed in the wrong location. U13 should have been placed in H-1, H-1 , not J-2. If the assembly is placed in the position, the Reactor Engineer, or his designee, should NOT grant permission to wrong pOSition, would,, in effect, suspend Core Alterations unlatch, which WOUld Atterations..

Friday, November 21 Friday, 21,, 2008 9:52:57 AM 23

QUESTIONS REPORT forVCS for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Notes

Ability to (a) predict the impacts of the following on the Fuel Handling Equipment and (b) based on those correct, control, or mitigate the consequences of those abnormal predictions, use procedures to correct, operation: Mispositioned fuel element.

Tier:

Group:

2 2

Importance Rating: SR04.0 Technical

Reference:

REP-100.001,, Attachment I, Material Transfer Form; and Attachment II, Reactor REP-l00.00l Core Inventory

REP-107.013, 3.3.7 (p 5 of 17); 43.4.1.M (p 6 of 17)

REP-l07.013,

Figure ICB.1 ICB.l Proposed references to be provided to applicants during examination:

Partially completed REP-l00 .00l , Attachment I REP-100.001, Objective:

Learning ObJactlv.: REFUELING SRO-12, 14 Question History: NEW

10 CFR Part 55 Content:

Comments:

Matches the KiA 43(b)(6)

KIA in that it tests ability to determine that an assembly has been mispositioned (per given information and the Matenal Material Transfer Form) and the actions taken (do not unlatch) for the mispositioned element.

SRO Only in that ij it tests *knowledge knowledge of SRO-specific responsibilities (verifying proper movement of fuel as the Special Nuclear Material Checker). Checker) .

Friday, November 21 21,, 20089:52:57 2008 9:52:57 AM 24

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION) forVCS 9.037 9 . 037 G2.4.8 OO IINEW/IHIGHERIISRO/SUMMERlb2009INO 001lNEW/IHIGHERIISRO/SUMMERl2/2009INO

Given the following plant cond~ions:

conditions:

100% power

The crew has implemented AOP-1 12.2, Steam Generator Tube Leak NOT AOP-112.2, Requiring SI....

A plant shutdown has been in~iatedinitiated..

A Reactor Trip occurs and the crew has just trans~ioned transitioned to EOP-1.1, Reactor Trip Recovery.

Recovery.

Which ONE (1) of the following describes the procedure flowpath?

A.

A. Remain in EOP-1 EOP-1.1 .1 and REFER TO and perform the applicable steps of AOP-112.2 as time allows.

B. RETURN TO AOP-112.2, starting at Step 7, and perform the actions of EOP-1.1 EOP-1.1 as time allows.

C~ Remain in EOP-1EOP-1.1 .1 to check Feedwater System status and RCS Temperature, Temperature, and then RETURN TO AOP-112.2, starting at Step 7, and perform the actions of EOP-1 EOP-1.1 .1 as time allows.

D. RETURN TO AOP-112.2, starting at Step 7, and identify and isolate the leaking

Steam Generator, then RETURN TO EOP-1.1 recovery.

EOP-1.1 and continue with the Reactor Trip

Friday, November 2121,, 2008 9:52:57 AM 25

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Feedback

A. Plausible because ij it is similar to the stipulations in NOTE - Step 4 of EOP-1 Incorrect because AOP-112.2 would NOT be "referred to" and because it is the non-conflicting steps of EOP-1EOP-1.1 .1 that are conducted as time allows.

EOP-1..1.

1.

allows, not the steps of AOP-112.2. Use of "return to" in Step 4 of EOP-1.1 implies that the AOP now becomes the primary procedure.

B. Plausible because this would be true if the crew had completed Steps 1-3 of EOP-1 .1.

EOP-1.1.

Incorrect because the crew is not directed to return to AOP-112.2 until Step 4 and the givenconditions in the stem clearly place them at Step 1 of the EOP.

C.

C. CORRECT. According to OAP-103.4,OAP-103.4. EOP/AOP User's Guide, 6.4.e, Rev O.

Guide. Step 6.4.e, 0, the use of the phrases MGo To* and -Refer "Go To" "Refer To" are used to determine which procedure constitutes the primary flowpath to target the correct recovery action. action. The phrase -Go"Go To* is used to leave the procedure or step in progress and transition to the required To" procedure or step for continued recovery actions, with the procedure transitioned to becoming the new primary procedure in effect. The phrase "Refer To* To" is used to continue with the procedure in progress using the referenced procedure as a guideline to accomplish a specific action. In this case, the referenced procedure is used concurrently with wijh the primary procedure in progress. Upon exiting EOP-1.0,EOP-1 .0. Reactor Trip/Safely Injection Actuation.

Trip/Safety Actuation, the operator is directed to Go To EOP-1 .1. Step 1.

EOP-1.1, 1, at the Step 5 Alternative Action (Rev 22), making EOP-1.1 EOP-1 .1 the primary procedure in progress.

progress.

The first three steps of EOP-1 EOP-1.1 .1 (Rev 15) direct the operator to (1) announce plant condijions conditions over the page system, system. (2) check FW status.

status, and (3) check RCS Temperature. Then the operator is provided with a Note prior to step 4 and the step itself. The Note states that if a transition is made to AOP-112.2, AOP-112 .2. the steps of EOP-1 EOP-1.1 .1 which do NOT conflict with AOP-112.2 should be completed as time allows. Then step

4 directs the operator to return to AOP-112.2,"

AOP-112.2,*step step 7.

7, if EOP-1.0 EOP-1 .0 was entered from AOP-112.2 (which ij was). making AOP-112.2 the primary procedure in progress.

it was),

D. This is plausible because the 1st 1st part is consistent with the requirements of Step 4 of EOP-1 EOP-1.1..1.

Incorrect because the crew will remain in AOP-112.2, AOP 112.2, even after SG identification and 4

isolation.

isolation .

Friday, November 21.

21, 2008 9:52:57 AM 20089:52:57 26

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Notes

(Steam Generator Tube leak) with EOPs.

Tier:

. Group:

Leak) Knowledge of how abnormal operating procedures are used in conjunction 1

2 Importance Rating: SR04.5 Technical

References:

OAP-103.4, EOP/AOP User's Guide, Step 6.4.e, Rev 0

EOP-1 .0, Step 5, Rev 22 EOP-1.0,

EOP-1 .1, Step 1-4, Rev 15 EOP-1.1,

AOP-112.02, Rev 4 Proposed references to be provided to applicants during examination:

None Learning Objective: AOP-112.2-06; EOP-1.1-05 EOP-1 .1-05 Question History: NEW 10 CFR Part 55 Content: 43(b)(5)

Comments:

The KA is matched because the operator must demonstrate knowledge of how abnormal (AOP-112.2) are used in conjunction operating procedures (AOP-i12.2) identified rules of usage.

conjunction. with EOPs (EOP-1.i)

(EOP-1.1) with the The question is SRO-Only because the operator must recall that the strategy for mitigating a Steam Generator Tube Leak makes AOP-1i2.02AOP-112.02 the primary procedure in progress after the afterthe EOP-1.1 first three steps of EOP-1 .i are performed, and renders the WOG based EOP to a secondary or referenced procedure status.

status.

Friday, November 21, 2008 9:52:57 AM 21 ,2008 27

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION) 10.

10.054 AA2.04 001lNEW/IHIGHERIISRO/SUMMERl2/2009INO OO IINEWtIH1GHERltSRO/SUMMERI2J2009INO 054 AA2.04 Given the following plant conditions:

conditions:

100% power

MDEFP 'B' has been tagged out for the past 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months for bearing replacement.

A loss of ALL Main Feedwater occurs.

A manual Reactor Trip was initiated.

The MDEFP 'A' starts, but then fails TWO (2) minutes later.

The crew is throttling EFW flow in EOP-1.1,EOP-1 .1, Reactor Trip Recovery.

Recovery.

IFV-3556, TDEFP Flow Control Valve to SG 'C', fails OPEN.

Which ONE (1) of the following identifies how the level in SG 'C' 'c' will be controlled AND the action required?

A. EOP-1S.2, Response to Steam Generator High Level; A . Transition to EOP-15.2, Immediately take actions to restore an Emergency Feedwater Pump to OPERABLE status.

B~

B ~ Remain in EOP-1EOP-1.1, .1, Reactor Trip Recovery, and locally control SG 'C' floW; flow; Immediately take actions to restore an Emergency Feedwater Pump to OPERABLE

status.

C. Transition to EOP-15.2, Response to Steam Generator High Level; Be in Hot Shutdown with 38 hours4.398148e-4 days 0.0106 hours 6.283069e-5 weeks 1.4459e-5 months . hours.

EOP-1 .1, Reactor Trip Recovery, and locally control SG 'C' flow; D. Remain in EOP-1.1, flow; Be in Hot Shutdown with 38 hours4.398148e-4 days 0.0106 hours 6.283069e-5 weeks 1.4459e-5 months .

Feedback A. Plausible because, w~h IFV-3551 failed open, because, with open, without local operator action, level would eventually rise to the high level and a Yellow Path to EOP-15.2 would exist. Also plausible because the 2nd part is correct.

correct. Prior to the reactor trip (and entry into Mode 3), the crew is operating within LCO LeO 3.7.1.2, 3.7.1.2. Action Statement Stalement a.

a, with 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months m

remaining to complete the work on the TO EFW Pump. However, after Mode 3 is entered, entered, the A MD EFW Pump also is discovered to be inoperable, and Action Statement b is now applicable. According Acoording to T.S.

T.S. 3.0.1 3.0.1 compliance with the Leo LCO is modes, except that upon failure to meet the LeO.

required during the operational modes, LCO, the associated ACTION requirements shall be met. With this in mind, mind, the crew has 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to achieve Hot Shutdown or Mode 4. 4.

IncOrrect because the entry condition for EOP-15.2 Incorrect Friday. November 21, Friday, EOP-1S.2 is 87% Narrow Range Level (see EOP-12.0, Rev 12, Attachment 3), which would NOT exist at the time of the given EOP-12.0, conditions in in the stem. Also incorrect because the operator will apply the Continuous Action to maintain SG level between 40-60% at step 9 of EOP-1 ..1, 21 , 2008 9:52:57 AM 1, well before control 28

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION) would be established in EOP-15.2 EOP*15.2..

B.

B. CORRECT. With W~h MO MD EFW Flow Control Valve to the C SG, IFV*3551 the SG level cannot be controlled from the MCB. According to OAP*103.4, IFV-3551,, failed OPEN, OAP-103.4, EOP/AOP User's Guide, Step 6.14.d, Rev 0, if an EOP directs an operation of a valve from the Room,, and the valve will not operate, it is expected that an operator would be Control Room dispatched to investigate, or attempt to locally operate the valve. valve. According to IB3 (p32*33, (p32-33, Rev 17), the MO MD EFW Flow Control Valves have the .bil~ ability to be controlled w~h. a handwheel override. At Step 9 of EOP*1 locally with. EOP-1.1 .1 (Rev 15), the operator is directed to verify SG levels between 40-60%. This is a Continuous Action step (Le. (i.e.

marked with an asterisk), and therefore when the C SG level rises to to>> 60%, Step 9b required.. This step directs the operator to control EFW Flow to maintain SG will be required Level between 40-60%. Since the direction cannot be completed from the MCB with the Flow Control Valve failed OPEN, an operator will be dispatched to control the valve locally, before the C SG Narrow Range Level rises to 87% (the entry condition for EOP-15.2). Prior to the reactor trip (and entry into Mode 3), the crew is operating within

EOP*15.2).

LCO 3.7.1.2, Action Statement a, w~ with 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months remaining to complete the work on the TO TD EFW Pump. However, after Mode 3 is entered, MOEFP MDEFP 'A' is also becomes inoperable, and Action Statement b. is now applicable. According to TS 3.0.1 3.0.1 compliance with the LeO LCO is required during the operational modes,modes, except that upon LCO, the failure to meet the LCO, associated ACTION requiremenls Ihe associaled requirements shall be mel. met. Wrth With this in mind.

mind, the crew has 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to achieve Hot Shutdown or Mode 4. 4.

C. Plausible because, with IFV-35S1 IFV-3551 failed open, without local operator action, level would

eventually rise to the high level and a Yellow Path to EOP-15.2EOP-1S.2 would exist. 2nd part is plausible because it il a common misapplication of TSR 1003, T.S. T.S. 3.0.3, Applicability Shutdown Time Allowance.

Allowance. For example, adding the remaining 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months (72-40=32) from Action Statement a. to the allowed time to reach Hot Shutdown in Action Slalemenl Statement b. (6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; 32+6=38 hours).

1st part is incorrect because Incorrect because Ihe lsI the entry condition for EOP* 15.2 is 87%

EOP-15.2 Narrow Rang Range e Level (see EOP*12.0, EOP-12.0, Rev 12, Attachment 3), which would NOT exisl exist al at the time of the given conditions in the stem. Also incorrect because the operator will apply the Continuous Action to maintain SG level between 40-60% at step 9 of EOP-1.1, EOP* 1.1, well before control would be established in EOP-15.2.EOP*15.2. Also incorrect because the 2nd part is wrong - application of the time remaining to restore OPERABILITY (or the lime time 10 un~ in HoI to place the unit Hot Slandby)

Standby) would be inappropriate and is specifically prohibited in TSR 1003, T.S. 3.0.3 Applicabil~

prohibiled Applicability Shutdown Time Allowance.

Allowance.

O.

D. Plausible because the 1 1s1 st part is correct correct*- With Wilh MO MD EFW Flow Control Valve 10 to Ihe the C IFV-3551,, failed OPEN, the SG level cannol SG, IFV*3551 cannot be controlled from the MeB. MCB.

According to 10 OAP-103.4, OAP* 103.4, EOP/AOP User's Guide, Slep Step 6.14.d, Rev 0, 0, if an EOP directs an operation of a valve from the Control Room, and the valve will not operate, operate, it is expected that an operator would be dispatched to investigate, or attempt to locally operale operate Ihe the valve. According to IB3 (p32*33, (p32-33, Rev 17), the MO MD EFW Flow Control Valves have the abil~

ability to be controlled locally with a handwheel override. At AI Step 9 of EOP-1.1 EOP*1 the operalor

.1 (Rev 15), Ihe operator is directed 10to verify SG levels between 40-60%. 2nd part is plausible because it ~ a common misapplication of ofTSR TSR 1003, T.S.T.S. 3.0.3, Applicability Shutdown Time Allowance.

Allowance. For example, adding the remaining 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months (72-40=32) from Action Stalemenl Statement a. to Ihe the allowed lime time 10 to reach Hot Shuldown Shutdown in Slalemenl b. (6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; 32+6=38 hours).

Action Statement Friday, November 21.

21, 2008 9:52:57 AM 29

QUESTIONS REPORT for VCS 2009 NRC NRC. SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Incorrect because the 2nd part is wrong*wrong - application of the time remaining to restore

. Notes OPERABILITY (or the time to place the unij Allowance .

Allowance.

unit in Hot Standby) would be inappropriate and is specifically prohibited in TSR 1003, T.S. 3.0.3 Applicability Applicabilijy Shutdown Time Ability to (a) predict the impacts of the following on the Loss of Main Feedwater and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Proper operation of AFW pumps and regulating valves. valves.

Tier: 1 Group: 1

. Importance Ratlng:Rating: SR04.3 Technical

References:

LCO 3.7.1.2, Action Statement alb

EOP-12.0, Rev 12, Attachment 3

OAp-103.4, EOP/AOP User's Guide, Step 6.14.d, OAP-103.4, 6.14.d, Rev 0

1B-3 IB-3 (p32-33, Rev 17)

Step 9 of EOP-I .1 (Rev 15)

EOP-1.1 Proposed references to be provided to applicants during examination:

None Learning Objective:

Objectlve: EOp-I .1-06 EOP-1.1-06 Question History: NEW 10 CFR Part 55 Content: 43(b)(2),

43(b)(2) , (5)

Comments:

The KA is matched because the operator must demonstrate the ability abilijy to determine the proper operation of AFW pumps (two OOS in Mode 3) and regulating valves (One failed OPEN), and determine actions (procedural actions for valve control, control, TS actions for Pumps OOS) to take for both.

both.

The question is an SRO-Only question because it requires the operator to assess plant conditions and select a recovery procedure within the procedural rules of usage for the failed open Flow Control valve, and apply the required actions of Technical Specifications.

Specifications.

Friday, November 21, 21 . 2008 9:52:57 AM 30

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION) 11 . 062 G2.2.36 11.062 G22.36 001INEW/IHIGHERlISRO/SUMMER/2/2009INO OOIINEWIIlUGHERIISROISUMMERI2f.!OO9INO cond~ions :

Given the following plant conditions:

100% power A routine surveillance reveals that Train 'B' of SSPS will NOT automatically actuate RB Spray.Spray.

Train 'B' of ESFAS Instrumentation, Instrumentation, Functional Unit (Item) 2.b, Reactor Building Spray - Automatic Actuation Logic and Associated Relays, is declared INOPERABLE. .

LCO 3.3.2, ACTION 14 is entered, placing the plant on a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months clock.

Which ONE (1) of the following planned maintenance activities should BOTH be operabilitY of SSPS Train 'B' has been restored?

rescheduled until the operability A.

A. Cycle the PZR Block Valve MVG-8000B; AND Swap the Train 'B' Charging Pump to Charging Pump 'C'.

B~

B!' Cycle the PZR Block Valve MVG-8000B; AND Swap the Train 'A' Service Water Pump to Service Water Pump 'C'.

C. Perform Perfonn the monthly Surveillance on the B Diesel Generator; AND

Swap the Train 'B' Charging Pump to Charging Pump 'C'.

Perfonn the monthly Surveillance on the B Diesel Generator; AND D. Perform Swap the Train 'A' Service Water Pump to Service Water Pump 'C'.

Friday, November 21, Friday, !WJ 21 , 2008 9:52:57 AM 31

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Feedback

A. Plausible because the 1st part is correct.

correct. 2nd part is plausible because swapping pumps is a routine maintenance activity under normal conditions.

The 2nd part is incorrect because it would be allowed.

tt does NOT affect the OPERABILITY of Train 'A' and B.

B. CORRECT. According to the T.S. LCO 3.3.2 Basis, entry into ACTION statements 12, 14, 21 or 25 is NOT a typical, pre-planned evolution during power operation, other than 14,21 for surveillance testing; testing; and generally are entered due to an equipment failure. If these state~ents are entered some restrictions will apply. For instance, to preserve ACTION statements activ~ies that degrade the availability of the EFW System, A TWS mitigation capability, activities ATWS RCS Pressure Relief System, AMSAC or Turbine Trip should NOT be scheduled when a logic train is inoperable for maintenance. Additionally, to preserve the LOCA lOCA mitigation capabiUty, capability, one complete ECCS train that can be actuated automatically must be maintained when a logic train is inoperable for maintenance. Furthermore.

Furthermore, to preserve reactor trip and safeguards actuation capability, activities that cause master relays or slave relays in the available train to be unavailable and activities that cause analog channels to be unavailable should not be scheduled when a logic train is maintenance. Finally, inoperable for maintenance. Finally, any activities on electrical systems (e.g. AC and DC power) and cooling systems (e.g. (e.g. service water and CCW) that support the systems or functions listed in the first three conditions should NOT be scheduled when a logic train is inoperable for maintenance.

maintenance. That is, is, one complete train of a function that supports a complete train of a function noted above must be available. With this in

mind, the cycling of the PORV Block Valve can NOT be perfonned performed because ~it is an activity that degrades the availability of the RCS Pressure Relief System. Secondly, the swap of the A Train Service Water Pump to the C Service Water Pump can NOT be performed because it is an activity on electrical systems (e.g. AC power) and cooling systems (e.g.

(e.g. service water) that supports the systems or functions listed in the first cond~ions , such EFW, and ECCS.

three conditions, ECCS. On the other hand, the swap of the B Train Charging Pump to the C Charging Pump may be performed because it~ does NOT affect the A Train of ECCS, aulomatically actuated. Add~ionally, ECCS, which can still be automatically Additionally, the monthly Surveillance on the B B Diesel Generator can be performed because one complete Train (e.g. A Train) that supports a complete train of the function noted above available.

is still available.

C.

C. Plausible because both are routine evolutions that could be performed under normal conditions.

conditions.

Incorrect because the monthly Surveillance on Diesel Generator 'B' can be performed because one complete Train (e.g. (e.g. AA Train) that supports a complete train of the function available.

noted above is still available.

D. 11st st part is plausible because both are routine evolutions that could be performed under normal conditions.

conditions. Also credible because the 2nd part is correct.

Incorrect because the 1st part is wrong - the monthly Surveillance on DIG 'B' can be pertormed because one complete Train (e.g.

performed (e.g. AA Train) that supports a complete train of

the function noted above is still available Friday, November 21 Friday, 21,2008 9:52:51 AM

, 2008 9:52:57 available..

32

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Notes

(Loss of Nuclear Svc Water)

Tier:

Group:

W ater) Ability to analyze the effect of maintenance activities, such as deg operations.

sources, on the status of limiting conditions of operations.

power sources, 1

1 raded degraded Importance Rating: SR04.2 Technical

Reference:

TS 3.3.2 & Basis.

reference. to be provided to applicants during examination:

Proposed references None Learning Objective: S8+19,21 SB-4-19,21 Question History: NEW 10 CFR Part 55 Content: 43(b)(2)

Comments:

The KA is matched because the operator must LOSS demonstrate the ability to analyze the effect of maintenance activities, such as degraded power sources, and others, on the status of lim~ing conditions limiting cond~ions for operations (LCO 3.3.2).3.3.2).

The question is SRO-Only because it requires that the operator have knowtedge TS LCO 3.3.2 necessary to analyze the TS action required. required.

knowledge of the basis of

Friday, Friday, November 21,21. 2008 9:52:57 AM 33

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

12. 062 G2.4.6 OOIINEWIIHlGHERIISROISUMMERI2!2009INO 001lNEW/IlllGHERlISRO/SUMMERl2/2009INO

Given the following plant conditions:

o o

o*

100% power A sustained total loss of offsite power has occurred.

Both ESF Diesel Generators have failed to start automatically and manually.

manually.

o* The crew has implemented EOP-6.0, Loss of All ESF AC Power.

o ESF equipment has been placed in PULL-TO-LOCK.

o* Continued attempts to start the DGs have been unsuccessful.

o* Subsequently, at Step 11 of EOP-6.0, the crew successfully energizes Bus 1DA via XTF-5052, ALTERNATE AC POWER SUPPLY TRANSFORMER.

Which ONE (1) of the following identifies the breaker that, when closed, will restore power to the bus, AND identifies the action required when Bus 1DA is restored?

A~ The 1DA NORM Feed Breaker; GO TO Step 30 and stabilize SG pressure.

B. The 1 DA AL T Feed Breaker; GO TO Step 30 and stabilize SG pressure.

C. The 1DA NORM Feed Breaker; Proceed to Step 12 and establish IA D. The 1DA1DA ALT ALT Feed Breaker; I.A. to the RB.

Proceed to Step 12 and establish I.A. to the RB .

FOday.

Friday, NOIJember November 21, 2008 9:52:57 AM 34

QUESTIONS REPORT ves 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION) for VCS Feedba,k Feedback

A. CORRECT. The crew is operating at Step 11 Alternate A~emate Action of EOP-6.0 EOP-S.O (Rev 21),

oper.tor to refer to SOP-304,Section V.A, (Rev 11) to supply offs~e which directs the operator power from the Alternate AC Power Supply. SOP-304 provides a Caution prior to step 2.8 indicating that only one AC ESF Bus can be supplied using that Alternate AC Power offsite Supply at anyone time. A series of steps is provided in SOP-304, Step 2.8.a.4, to energize the 1DA ESF Bus. This section includes a step to close the 1DA NORM Feeder Breaker which will re-energize the bus. bus. According to Step 2.B.b.4, 2.8.b.4, in order to re-energize the 1 1DB DB ESF Bus the lOB 1DB ALALTT Feeder Breaker must be closed closed., According to Caution - Step 8 just prior to Step 8 of EOP-S.O, EOP-6.0, when power is restored to either ESF Bus, Bus, recovery should continue with Step 30 to minimize the deterioration of plant conditions.

B.

B. Plausible because the 2nd part is correct.

correct. Also plausible because the lOB 1DB ALT Feeder Breaker must be closed to Re-energize the 1DB ESF Bus.

Incorrect because the 1st part is wrong - SOP-304 provides a Caution prior to step 2.8 indicating that only one AC ESF Bus can be supplied using that Altemate Alternate AC Power Supply at anyone time. time. A series of steps is provided in SOP-304, Step 2.8.a.4, 2,8.a.4, to energize the lOA 1DA ESF Bus, Bus. This section includes a step to close the 1DA NORM Feeder Breaker which will re-energize the bus.. ..

C..

C Plausible because the 1 1st st part is correct - a series of steps is provided in SOP-304, Step 2.8.a.4, to energize the 1DA ESF Bus. Bus. This section includes a step to close the 1DA NORM Feeder Breaker which will re-energize the bus, bus. 2nd part is plausible because it is the next step in the procedure if the Caution is misapplied.

Incorrect because, per CAUTION - Step 8, the crew is directed to Step 30, which stabilizes SG pressures, NOT to Step 12, establish RB I.A.

D.

D. 1st 1 st part is plausible because the AL AlT feeder breaker would be used to restore Bus 1DB 1DB from the Alternate AC source. 2nd part is plausible because it is the next step in the misapplied.

procedure if the Caution is misapplied.

Incorrect because the first part is wrong - a series of steps is provided in SOP-304,SOP-304, 2.8 .* .4, to energize the 1DA ESF Bus.

Step 2.8.a.4, Bus. This section includes a step to close the 1DA NORM Feeder Breaker which will re-energize the bus. Also incorrect because, per CAUTION - Step 8, the crew is directed to Step 30, which stabilizes SG pressures, NOT to Step 12, establish RB I.A.

Friday. November 21, Friday, 21 ,20089:52:57 2008 9:52:57 AM 35

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Notes

(AC Electrical Distribution) Knowledge symptom based EOP mitigation strategies.

Tier: 2 Group: 1 Importance Rating: SR04.7 Technical

References:

EOP-6.0 Caution 6, Step 8

SOP-304,Section V.A, Step 2.8, Rev 11

TS Basis for LCO 3.8.1 3.8.1 Proposed references to be provided to applicants during examination:

None Learning Objective: EOP-6.0-05 Question History: NEW 10 CFR Part 55 Content: 43(b)(5)

Comments:

The KA is matched because the operator must demonstrate Qemonstrate have system based knowledge of the AC Distribution System (Breaker nomenclature), and knowledge of the EOP mijigation mitigation strategies (specific procedure f1owpath).

flowpath) .

The question is SRO-Only because the operator must recall that the specific strategy (and the specific action taken) for moving forward to step 30 is written into EOP-6.0, when past Step 10.

specifiC 10, and an ESF Bus is restored wijh with the Alternate AC Power Supply.

Supply.

Friday, November 21 21,. 2008 9:52:57 AM 36

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

13. 064 A2.13 OOI INEWIIHIGHERIISROISUMMERI2I2009INO 001lNEW/IHIGHERlISRO/SUMMERl2/2009INO

Given the following plant condttions:

100% power Diesel Generator 'A' conditions:

A routine surveillance of Diesel Generator 'A' IAI is in progress.

IAI is operating in parallel with wtth offsite offstte power.

Service Water Pump 'C' ICI is aligned mechanically to Train IAI.

'A'.

Subsequently, the feeder breaker for Bus 1EA trips OPEN.

The following DG IN conditions are observed:

'A' condttions

The DG 'A' 'AI Lube Oil Temperature is 160*F 160°F and rising at 1°F/minute.

1*F/minute.

The DG 'A' Coolant Temperature is 178*F 178°F and rising at 1.S*F/minute.

1.SoF/minute.

Which ONE (1) of the following identifies the approximate time frame that Diesel Generator 'A' 'AI will trip within if NO action is laken, taken, AND the actions that should be lakentaken wtthin AOP-117.1, Loss orServics within AOP-117.1, of Service Water?

A. 4-8 minutes; Refer to Annunciator Response Procedures on the Service Water System and transfer Service Water System loads to Train 'B'. IBI.

B. 4-8 minutes; Ct Start SWP 'C' Slart IC' and open PVG-310SA, FS TO DG A, to provide water from Fire Service to DG 'A'.

C~ 10-12 minutes; Refer to Annunciator Response Procedures on the Service Water System and transfer Service Water System loads to Train 'BI. 'B'.

D. 10-12 minutes; Start SWP 'c' lei and open PVG-310SA, FS TO DG A, to provide waterfrom water from Fire Service to DG 'A' 'A'..

Friday, November21 November 21,, 2008 9:52:57 AM 20089:52:57 37

QUESTIONS REPORT forVCS for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Feedbac:k Feedback

A. Plausible because the 2nd part is correct.

correct. Also plausible because the Lube Oil High Temp Alarm will come in within 7 minutes [167-160=7/1=7] and the High Jacket Water Temp Alarm will come in in 4.7 minutes [185-178=7/1.5=4.67].

Incorrect because the 1 1st

[185-178=711.5=4.67].

st part is wrong - The heat rate on the DG will NOT resuK result in a trip within 7 minutes - See trip calculations in Choice C.

B.

B. Plausible because the Lube Oil High Temp Alarm will come in within 7 minutes

[167-160=711=7]

[167-160=7/1=7] and the High Jacket Water Temp Alarm will come in in 4.7 minutes

[185-178=711 .5=4.67]. Also plausible because ARP-604, 3-4 Supplemenatl Action I1

[185-178=7/1.5=4.67].

restores cooling to the DG via PVG-3105 and because Corrective Action 2 of the same ARP directs the crew to place the spare SWP in service per SOP-I SOP-117.17.

Incorrect because the 1st I st part is wrong - The heat rate on the DG will NOT result in a trip within 7 minutes - See trip calculations in Choice C. C. Also incorrect because SWP

'C' cannot be started under the given conditions in the stem.

C.

C. CORRECT. According to GS-2, GS-2, Table GS2.2 (p81 (pSI,, Rev 15), lhe the A Service Water Pump is powered from 7.2 KV Stub Slub Bus 1EA lEA and therefore, will be de-energized when Ihe the feeder breaker for ESF Stub Slub Bus 1EAlEA lrips trips OPEN. According 10 to IB-5 (p49, Rev 21)

(and ARP XCX-5201, 1-2), 1-2) , the Diesel will wililrip trip on high lube oil temperature of 175*F 175°F and jackel coolant high jacket coolanl temperature temperalure of 195°F, I 95*F, when Ihe the DG is operaled operated in Ihe the TEST Mode il is when conducting a routine surveillance). At the rate indicated, the High (Which it Lube Oil Temperalure Temperature will be reached in 15 minutes [(175*F -160°F) x minutell"F

[(175°F -160*F) =

minute/1°F = 15 minutes), and Ihe minutes], the High Jacket Coolant Trip will be reached in 11 .3 minutes [(195*F 11.3 [(195°F

-178°F) xx 2 minutes/3°F

-178*F) minutesl3*F = 11.311 .3 minutes].

minutes). According to AOP-117.1 AOP-117.1 (Rev 3), Ihe the enlry entry cond~ions for Ihe conditions operalor will refer to the appropriate ARPs 10 the AOP are met. The operator to reslore Service Water, one of which will be ARP-001, restore ARP-OOI, XCP-604, AP3-4 AP3-4,, DG A CLR SW FLO LO TEMP HI. This ARP has a Supplementary Action (I) (1) thaI that will restore SW 10 to the operating DG by opening PVG-3105A to 10 provide water from the Fire System. This will end Ihe the immediate threat to Ihe the loaded DG which is operaling operating without cooling waler.

water.

Add~ionally, Steps Additionally, Sleps 5 and 6 of AOP-117.1 AOP-I17.1 will direct the operalor operator to transfer Service Water System loads to Train B.

D. Plausible because Ihe the 151 1st part is correct - see lrip calculalions in Choice C. Also trip calculations plausible because ARP-604, 3-4 Supplemenatl Action I1 restores cooling to the DG via PVG-3105 and because Corrective Action 2 of the same ARP directs the crew to place the spare SWP in service per SOP-117.SOP-I 17.

Incorrect because SWP 'C' cannot be started under the given conditions in the stem. stem .

Friday, November 21, 2008 9:52:57 AM Friday, IW. 38

QUESTIONS REPORT for VCS 2009 NRC SRO WOR WORKSHEETKSHEET (AFTER CHANGES AFTER VALIDATION)

Notes

Ability to (a) predict the impacts of the following on the Emergency Diesel Generator and (b) based on those predictions predictions,, use procedures to correct, control.

operation:

Tier:

Group:

2 1

control, or mitigate the consequences of those abnormal operation: Consequences of opening auxiliary feeder bus (ED/G sub supply). supply).

Importance Rating: SR02.8 SRO 2.8 Technical

Reference:

GS-2, Table GS2.2 (p81 (p81,, Rev 15)

IB-5 (p49, Rev 21) 18-5

AOP-117.1 (Rev 3)

AOP-117.1 Proposed references to be provided to applicants during examination:

None Learning Objective:

Loaming Objectlvo: IB-5-21;; AOP-117.1-06 18-5-21 Question History: NEW 10 CFR Part 55 Content: 43(b)(5)

Comments:

The KA is matched because the operator must demonstrate the ability to predict the impact opening auxiliary feeder bus (ED/G sub-supply or ESF Stub Bus 1EA) on the ED/G system (Requires a DG trip if no action taken); and (b) based on those predictions, identify the sleps steps AOP-117.1 to mijigate taken within AOP-117.1 mitigate and control the event.

The question is SRO~Only SRO-Only because the operator must recall how the strategy is written into AOP-117 .1 for dealing with opening the backup cooling water supply to an operating ESF DG, AOP-117.1 when one train of of SW is lost (i.e. Use ARPs and shift SW loads to the operating train) .

Friday, November 21, 2008 9:52:57 AM 39

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

14. 068 A2.04 001IMODIFIED/IHIGHERIISRO/SUMMER/2/2009INO 00 IIMODIFIED/IHIGHERlISROISUMMER/2I20Q9INO cond~ions :

Given the following plant conditions:

A Waste Monitor Tank #2 release is in progress.

The following annunciators have actuated:

LlQ WST DISCH RM-L9 HI RAD (XCP-644, 2-5)

LIQ

LlQ WST DISCH RM-L9 TRBL (XCP-644, 2-6)

Release flow rate remains at 55 gpm.

Which ONE (1) of ofthe the following completes the statement below?

The operator must close as required by procedure. Failure to take this action may result s~e boundary from resun in the dose commiment to an individual at the site radioactive materials in liquid effluents to exceed the legal limit of to the total body during any calendar quarter.

A. Close RCV-018, A. RCV-018, Liquid Waste Control Valve;Valve; 1.5 mrem

B. Close RCV-018, Liquid Waste Control Valve; C,..

1 Rem Valve; C~ Close PVD-6910, Liquid Effluents to Fairfield Penstocks;Penstocks; 1.5 mrem PVD-6910, D. Close PVD-691 0, Liquid Effluents to Fairfield Penstocks; 1 Rem

Friday. November 21 Friday, 21,, 2008 9:52:58 AM 40

QUESTIONS REPORT forVCS for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Feedback

A. Plausible because the 2nd part is correct. Also plausible because RCV-018 monitors sense a HI RAD.RAD.

RCV*OI6 is one of the two discharge isolation valves that may close when their respective radiation Incorrect because RCV-018 RCV-016 shuts when RM-L5 RM*L5 senses a HI RAD, RAD. not RM-L9.

RM*L9. The procedure that must be addressed is the RM-L9 RM*L9 HI RAD and TRBL Annunciator Response Procedures. Each of these procedures direct that PVD-6910 PVD-691 0 be closed.

closed, and NOT RCV*OI6.

RCV-018.

B. Plausible because RCV-018 is one of the two discharge isolation valves that may close when their respective radiation monitors sense a HI RAD. RAO. The dose is plausible because it is the admin limitation for annual Whole Body exposure.exposure. It is also the TEDE value in Initiating Condmon Condition 461 for declaration of a General Emergency (see EPP-001, EPP-001 ,

Att II, All II. Page 13 of 25).

Incorrect because RCV-018 RCV-016 shuts when RM-L5 RM*L5 senses a HI RAD, RAD. not RM*L9.

RM-L9. The procedure that must be addressed is the RM-L9 RM*L9 HI RAD and TRBL Annunciator Response Procedures. Each of these procedures direct that PVD-6910 PVD-691 0 be closed, closed. and NOT RCV-018.

RCV* 016. Also incorrect because 2nd part is wrong lim~ per Spec 1.1.3.1 wrong*- the limit 1.1.3.1 is 1.Smrem.

1.5 mrem.

C. CORRECT. According to GS-9, GS*9. (p27, (p27. Rev 10) high radiation on RM-L9RM*L9 automatically PVD-6910. Liquid Effluents to Fairfield Penstocks. According to ARP-019 (Rev closes PVD-6910, (Rev..

1) for XCP-644, AP 2-5, the operator must verify that the automatic action occurs, which with an indication of 55 gpm still sbll flowing, flowing. it did not. Therefore.

Therefore, the operator will be required to manually close cloSe PVD-6910, PVD-6910. Liquid Effluents to Fairfield Penstocks.

According to Specification 1.1.1.1 of the Offsite Oftsite Dose Calculation Manual (ODeM).(ODCM), the minimum number of radioactive liquid effluent monitoring instrumentation channels must be OPERABLE to ensure that the limits of ODeM ODCM Specification 1.1.2.1 1.1.2.1 are not exceeded. According to specification 1.1.2.1, this limitation provides assurance that exceeded.

concentrations of radioactive radioadive materials in bodies of water outside the site will not result in exposures in excess of 10CFR20 and 10CFR50 limits. lim~s. According to Specification 1.1.3.1, the dose or dose commitment to an individual from radioactive materials in liquid effluents released from the site must be limited to 1.5 mrem to the total body and 5 mrem to any organ during any calendar quarter, and 3 mrem to the total body and 10 mrem to any organ during any calendar year. Therefore, the means to ensure that these dose limitations are not exceeded is to comply with the concentration limits of Specification 1.1.2.1, and the means to ensure that these concentration limitations are not exceeded is to comply with the equipment operability reqUirements requirements of Specification 1.1.1.1. Since they are not being complied with (Automatic (AutomatiC closure failure of the release flow control valve),

valve). the limits limrts of Specification 1.1.3.1 may be exceeded.

D.

D. Plausible because the 1st part is correct.

correct. The dose is plausible because it is the admin exposure. It is also the TEDE value in Initiating limitation for annual Whole Body exposure.

Condition 461 for declaration of a General Emergency (see EPP-001, EPP*OOI. Att All II, II. Page 13 of 25).

25) .

Incorrect because the 2nd part is wrong 21,2008 Friday, November 21, 2008 9:52:58 AM wrong*- the limrt 1.1 .3.1 is 1.5 mrem.

limit per Spec 1.1.3.1 41

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Notes

Ability to (a) predict the impacts of the following on the Liquid Radwaste and (b) based on those predictions, predictions, use procedures to correct.

operation: Failure of automatic isolation Tier:

Group:

control , or mitigate the consequences of those abnormal correct, control, isolation."

2 2

Importance Rating: SR03.3 SRO 3.3 Technical

References:

OOCM ODCM 1.1.3.1 1.1.3.1 Proposed references to be provided to applicants during examination:

None Learning Objective: HPP-71 0-02 HPP*710-02 Question History:

MODIFIED (Although written "from scratch", this question is similar enough to Closed Reference question LIQUID RAD WASTE 8, R$ference 19, & 42 to be classified as MODIFIED) 8,19, t

10 CFR Part 55 66 Content: 43(b)(5) 43{b){5) f; Comments:

Matches the KIA in that the question tests knowtedge knowledge of what automatic action should occur to

.....J. tEl[minate tecmmate a liquid waste release when a high radiation condition exists.

SRO Only in that ij it tests the bases from the OOCM ODCM .

Friday, November 21,21 , 2008 9:52:58 AM 42

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

15. 073 G2.4.8 001lNEW/IHIGHERIISRO/SUMMERl2/2009INO 15.07302.4.8001INEW/IHIGHERlISRO/SUMMERI2I2009INO

Given the following plant conditions:

o

o

o conditions:

A Reactor Trip occurs due to a locked rotor on RCP IBI.

The crew has just transitioned transnioned to EOP*l.l, Annunciators RC LTDN HI RNG RM*L EOP-1.1, Reactor

'B'.

Reac/or Trip Recovery.

Recovery.

RM-L 1 HI RAD (XCP-642, 1-5) 1*5) and RC LTDN LO RNG RM-L RM*L1 HI RAD (XCP-642, 4-3) 4*3) both actuate.

o Chemistry is directed to sample the RCS.

o HP is directed to survey the area.

area.

While implementing EOP-1.1, EOP*l. l, which ONE (1) of the following identifies other procedures that should be referred to and actions that should be taken while awaiting results from Chemistry and HP?

A'!

Art Increase letdown to 120 gpm in accordance with wnh the Annunciator Response Procedure ONLY.

B. Increase letdown to 120 gpm in accordance with B. wnh the Annunciator Response procedure; AND Cooldown to less than 500"F 500°F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months as required by Technical Specification 3.4.8, Reactor Coolant System Specific SpeCific Activity.

Activity.

C. Determine C. Detennine the capacity capacny status of the liquid and gaseous radwaste systems in

D.

accordance with SAP-154, D. Determine Detennine the capacity accordancewnh accordance SAP*l54, Failed Fuel Action Plan ONLY.

SAP*l54, Failed Fuel Action Plan; with SAP-154, Declare an Unusual Event in accordance mh Implementation of the Emergency Plan. Plan.

ONLY.

capacny status of the liquid and gaseous radwaste systems in Plan; AND with EPP-001, Activation and

Friday, November 21,2' , 2008 9:52:58 AM 20089:52:58 43

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION) forVCS Feedback

A.

A. CORRECT. According toARP-019 XCP~2, 4-3 (low to ARP-019 (Rev 1), XCP-642, (Low Range) the operator will be required to (1) verify that the alarm is valid, (2) notify Chemistry to sample the RCS, and (3) notify HP to conduct radiological surveys in the area. area.

Supplemental Actions in this ARP such as increasing letdown Letdown flow, referring to Technical Specifications and the Failed Fuel Action Plan (SAP-154) (SAP-l54) are dependent upon the results of the sample.sample. According to ARP-019 (Rev 1),

XCP~2, XCP-642, 1-5 (High Range) the operator will be required to (1) verify that the alarm is valid, (2) notify Chemistry to sample the RCS, (3) notify HP to conduct radiological surveys in the area, and (4) increase Letdown letdown flow to 120 gpm.

gpm.

Again, Supplemental Actions in this ARP such as referring to Technical Specifications and the Failed Fuel Action Plan (SAP-154) (SAP-l54) are dependent upon the results of the sample. Therefore, only the ARPs require that letdown Letdown flow be increased at this time.

B.

B. The 1st 1st part is plausible because it is correct - Both ARPs direct increasing letdown flow. The 2nd part is plausible because if sample results indicate that the LCOlCO is exceeded, the Action (cooldown to less than 500°F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ) is correct. Also plausible because Accident Analysis assumes failed fuel on a correct.

locked rotor, so the lCO LCO would probably be exceeded.

C.

Incorrect because, although Letdown (Rev 1), if Action Level letdown flow is required to be increased, the TS ACTION is NOT required until LCO 3.4.8 can be evaluated after the sample.

allerthe Plausible because this is an action taken, as stated in Step 7.3.6 of SAP-154 SAP-l54 level 1 is declared and because the action is a preliminary

~action (similar to increasing letdown flow) if there was failed fuel. Also plausible

--action because Accident Analysis assumes failed fuel on a locked rotor, so SAP-l54 SAP-154 eventuany be implemented.

would probably eventually implemented.

- Incorrect because the implementation of SAP-l54 SAP-154 -is is NOT yet required until sample analysis confirms failed fuel, fuel , and will ONLY be required after the Chemistry Sample is evaluated.

D.

D. Plausible -because, in accordance with Attachment II of EPP-001 Plausible*because, EPP-001,, Unusual Event In~iating Condition 106, Fuel Damage Indication, is met when the RM-l1 Initiating RM-L 1 High Range Alarm is activated (which it ~ is) AND primary coolant dose equivalent iodine activity is greater than 30 ~Cilml

~Ci/ml.. Also plausible because Accident Analysis assumes failed fuel on a locked rotor, rotor, so SAP-154 SAP-l54 would probably eventually be implemented.

Incorrect because the 30 ~Cilml ~Ci/ml has not yet been confirmed by Chemistry; therefore the EAl EAL does not yet apply. AlsoA lso incorrect because the

implementation of SAP-l54 failed fuel, SAP-154 is NOT yet required until sample analysis confirms fuel , and will ONLY be required after the Chemistry sample is evaluated.

Friday, November 21,20089:52:58 21 . 2008 9:52:58 AM 44

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Notes .

(Process Radiation Monitoring) Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

Tier:

Group:

EOPs.

2 1

Importance Rating: SR04.5 Technical

Reference:

ARP*019 (Rev 1), XCP-642, 4-3 (Lo Range)

ARP-019

ARP'()19 ARP-019 (Rev 1), XCP-642, 1-5 (Hi Range)

TS LCO 3.4 .S 3.4.8

Step 7.3.6 of SAP-154 (Rev 1)

.Step 4.1 of EPP-001 & Attachment II, p8

Step pS of 25 (Rev 29) references to be provided to applicants during examination:

Proposed reference.

None Learning Objective: AB-3-25 Question History: NEW

10 CFR Part 55 Content:

Comments:

43(b)(5)

The KA is matched because the operator must demonstrate knowledge of how abnormal operaliml operating procedures (i.e. ARP'()19, ARP-019, Note that VCS uses ARPs for AOP related to the Process Mon~ors) are used in conjunction with EOPs (EOP-1 Rad Monitors) (EOP-1.1)

.1) with the identified rules of usage.

The question is SRO-Only because the operator must recall that the strategy for responding to an alarm condition on both ranges of RM-l1 RM-L 1,, the Primary Coolant Letdown Monitor.

Monitor.

Friday, November 21.

21, 2008 9:52:58 AM 45

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

16. 078 A2.01 001INEW/IHIGHERI/SRO/SUMMERl2/2009INO OOIINEW/IHIGHERIISRO/SUMMERI2I2009INO 100% power cond~ions :

Given the following plant conditions:

Instrument Air Compressor XAC-3B is in operation.

Instrument Air Compressor XAC-3A is in standby.

The Supplemental Breathing Air Compressor XAC-12 is aligned to provide activ~ies in the Aux Building.

Breathing Air to support on-going maintenance activities

The Instrument Air Dryer fails, fails, causing desiccant to clog the after filters.

fiijers.

Instrument Air header pressure decreases rapidly to 35 psig.

The crew has just entered AOP-220.1, AOP-220.1, Loss of Instrument Air.

Which ONE (1) of the following describes the action that must be taken to restore Instrument Air header pressure AND what is the consequence of I.A. header pressure .

dropping below 40 psig?

M Ar:! Suspend maintenance activities, open XVB-2633, IA BACKUP SYSTEM SUP HDR ISOLATION VLV, VLV, and ensure the Supplemental Breathing Air Compressor is running; running; Reactor power will increase.

B. Suspend maintenance activities, activ~ies , open XVB-2633, and ensure the Supplemental

Breathing Air Compressor is running; Turbine blade erosion will increase.

running ;

C. Locally start the Diesel Driven Air Compressor; Turbine blade erosion will increase.

D. Locally start the Diesel Driven Air Compressor; Reactor power will increase.

Feedback A.

A. CORRECT. A similar event occurred at VCS (see TB-12 (p28, Rev 12). While at 100% power, the Instrument Air Dryer failed, causing desiccant to clog the filters.. This, in tum after filters turn caused a low Instrument Air header pressure alarm.

The standby air compressor was started, however, Instrument Air header pressure continued to decrease, and ultimately the Turbine Group A and B Drain Valves failed open. When air pressure continued to drop, the Supplemental Air Compressor was started and air header pressure began to recover, ultimately uijimately stabilizing at 112 psig. The event established in the conditions of the question is similar, and yet different. According to TB-12 (p26, Rev 12), the Breathing Air

System is usually in operation during maintenance outages, but may be required during any mode of plant operation to support maintenance requirements, as is the case here. According to TB-12 (p17), when Breathing Air is required, the Friday, November 21 21,, 2008 9:52:58 AM 46

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

XVB-2633 is closed.closed. This means that this air compressor can NOT function as a

ready backup to the Instrument Air header should an event like an After Filter require. The operator can start the Diesel Air Compressor, clogging would require.

however, this compressor taps into the Instrument Air System upstream of the Air Dryers and therefore, will not aid in raising Instrument Air header pressure.

The Only means to raise Instrument Air header pressure is to suspend pressure.

activ~ies , open XVB-2633, and ensure the Supplemental Breathing maintenance activities, Air Compressor is running.

running . According to ARP-001, ARP-001 , XCP-607, XCP-607, 2-5, 2-5, Rev 5, 5, when IPI05875 senses less than 40 psig Group A and B Turbine Drain Valves will fail OPEN resulting in a reactor power increase.

B. Plausible because the 1 1st st part is correct.

correct. Also plausible because, according to ARP-001, ARP-001 , XCP-607, 2-5, Rev 5, 5, when IPI05875 senses less than 40 psig Group A and 8 B Turbine Drain Valves will fail OPEN. CLOSED, it~ is OPEN. If the valves failed CLOSED, conceivable that moisture content in the steam could increase to the point where turbine blade erosion increases. Also credible because these drain valves are mOisture, which could contribute to blade erosion.

designed to remove moisture, Incorrect because the failed-open valves will simply pass more steam/moisture and, if anything, will reduce the potential for turbine blade erosion.

erosion.

C. Plausible because the Diesel Air Compressor is the source of I.A. after a loss of all AC power. Also plausible because AOP-220.1 AOP-220.1 (Rev 2) identifies starting this

component as action to mitigate a Loss of Instrument Air.

because, because, according to ARP-001, less than 40 psig Group A and 8 valves failed closed, ARP-001 , XCP-607, XCP-607, 2-5, Air. Also plausible 2-5, Rev 5, 5, when IPI05875 senses B Turbine Drain Valves will fail OPEN. If the closed, it is conceivable that moisture content in the steam could increase to the point where turbine blade erosion increases. Also credible because these drain valves are designed to remove moisture, which could contribute to blade erosion.

Incorrect because the 1st part is wrong - The Diesel Air Compressor taps into the Instrument Air System upstream of the air dryers; therefore,therefore, it will not aid in raising Instrument Air header pressure. Also incorrect because the failed-open valves will simply pass more steam/moisture and, if anything, anything, will reduce the potential for turbine blade erosion.

D.

D. Plausible because the 2nd part is correct -~ ARP-001, XCP-607, AP2-5, AP2-5, Rev 5, 5, states that when IPI05875 senses less than 40 psig, Group A and 8 B Turbine resu~ing in a reactor power increase.

Drain Valves wili fail OPEN resulting increase. Plausible because the Diesel Air Compressor is the source of I.A. after a loss of all AC power.

power. Also plausible because AOp-220.1 AOp-220.1 (Rev 2) identifies starting this component as action to mitigate a Loss of Instrument Air. Air.

Incorrect because the 1st 1st part is wrong - The Diesel Air Compressor taps into the Instrument Air System upstream of the air dryers; therefore, it will not aid in raising Instrument Air header pressure.

Friday, 21 , 2008 9:52:58 Friday, November 21, 9 :52:58 AM 47

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Notes

Ability to (a) predict Tier:

Group:

predicl the impacts of the following on the Instrument Air and (b) based on those predictions, use procedures to correct. control, or mitigate the consequences of those abnormal operation:

correct, control, and filter malfunctions.

malfunctions.

2 1

predictions, operation: Air dryer Importance Rating: Rating : SR02.9 SRO 2.9 Technical

Reference:

AOP-220.1, Rev 2

T8-12, p17, 26, 28, Rev 12 TB-12,

ARP-001, XCP-607, XCP-607, AP2-5, AP2-5, Rev 5

B-208-057 8-208-057-IA0008 Proposed reference.

references to be provided to applicants during examination:

None Learning Objective:

ObjectJve: AOP-220.1-5 Question History: NEW

10 CFR Part 55 Comments:

Content:

56 Content 43(b)(5)

The KA is matched because the operator must ability to (a) predict the impacts of Air dryer and filter malfunctions on the lAS (Reactor Power will increase when Turbine Drain Valves Open);

and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions (realign and start the Supplemental breathing Air Compressor).

Compressor).

The Question is SRO-Only SRO*Only because the operator must assess plant conditions during an abnormal situation sijuation (Loss of lAS) and then apply detailed knowledge of specific procedure steps; induding, steps; limijed to: 1) Step 2 of AOP-220.1, including, but not limited AltemativeAction, AOP-220.1 , Alternative SOP-220.1 ,

Action, 2) SOP-220.1, HPP-604, Step 3.12.2,Section IV.B, Step 2.1, and 3) HPP-604, 3.12.2, that enables recovery from the abnormal event.

Friday, Friday, November 21, 21 . 2008 9:52:58 AM 48

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

17. E04 EA2.1 OO2IMODIFIED/IIflGHERIISRO/SUMMERi2I2009INO 002IMODIFIED/IHIGHERIISRO/SUMMERl2/2009INO

Given the following plant conditions:

condttions:

A Reactor Trip and SI have occured.

The crew has entered EOP-1 .0, Reactor Trip/Safety Injection Actuation, and has EOP-1.0, reached the diagnostic steps.

RCS pressure is 1300 psig and DECREASING slowly. slowly.

ALL RB conditions are NORMAL.

ALL SG pressures and levels are stable.

Auxiliary Building Area Radiation Monitors alarm.

Monttors are in alarm.

NO other abnormal radiation monitoring indications exist.

Which ONE (1) ofthe of the following describes the appropriate procedure flowpath?

A. Continue in EOP-1EOP-1.0,.0, Reactor Trip/Safety Injection Actuation; Transition to EOP-2.0, Loss of Reactor or Secondary Coolant. Coolant.

B~ Continue in EOP-1 EOP-1.0,.0, Reactor Trip/Safety Injection Actuation; Transttion Transition to EOP-2.5, LOCA Outside Containment.

Containment.

C. Go to to EOP-2.0.

EOP-2.0, Loss of Reactor or Secondary Coolant;

Transition to EOP-2.5, LOCA Outside Containment.

Transttion D. Go to EOP-2.0, Loss of Reactor or Secondary Coolant; Transition to EOP-2.1, Post LOCA Cooldown and Depressurization Transttion Depressurization..

Friday, November 21, 21 . 2008 9:52:58 AM 49

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Feedback Feulback.

A.

A. Plausible Plausible*because the 1st part is correct.

correct. 2nd part is plausible because there are indications of a loss of RCS inventory an indavertent SI or misdiagnosis could resuij (12, 13, & 14). Once past result in proceeding past the diagnostic steps (12,13, those steps, there are other opportunities in EOP-l EOP-1.0.0 which diagnose conditions to determine the proper Optimal Recovery Procedure (Steps 19, 21 &

21,, 22, 23, &

25).

Incorrect because the criteria for transition to EOP-2.0 at Step 25 of EOP-1.0EOP-l .0 are NOT met (RCS pressure is > 250 psig)

B. CORRECT.

CORRECT. According to EOP-l .0, conditions are NOT met for transition at the EOP-1.0, diagnostic steps (RB rad levels are normal."

normal,* RB sump levels are normal, RB pressure is normal, and RBCU drain flow annunciators are not actuated, actuated, SG pressure and levels are normal and the only abnormal rad monitors are those outside the RB and not asociated with the SGs). The crew would then continue in EOP-l EOP-1.0 EOP-l .0 (based

.0 until the transition to EOP-2.5 is met at Step 23 of EOP-1.0 on abnormal rad levels in the AB). AB). The Alternate Aijemate Action for Step 23 requires the crew to GO TO EOP-2.5.

C. Plausible because the 2nd part is correct.

correct. EOP-2.5 would be the next procedure used. The 1 st part is plausible because there are indications of a loss of RCS 1st

inventory.

inventory. Also plausible because there is a transition to EOP-2.0 at the diagnostic steps.

Incorrect because the criteria for transition to EOP-2.0 at Step 14 of EOP-1.0 NOT met (RB rad levels are normal, RB sump levels are normal, RB pressure is EOP-l .0 are normal, and RBCU drain flow annunciators are not actuated).

D. The 1st 1st part is plausible because there are indications of a loss of RCS inventory. Also plausible because there is a transition to EOP-2.0 at the diagnostic steps. 2nd part is plausible because the given indications are consistent with a SBLOCA and EOP-2.1 EOP-2.1 would normally be used to mitigate this event. EOP-2.1 may uttimately ultimately be the optimal recovery procedure selected by the TSC when transitioning from EOP-2.4. For these conditions, the procedure flowpath could be 1.0-2.5-2.4-2.1 1.0-2.5-2.4-2.1 Incorrect because the criteria for transition to EOP-2.0 at Step 14 of EOP-1.0EOP- l .0 are NOT met (RB rad levels are normal, RB sump levels are normal, RB pressure is normal, and RBCU drain flow annunciators are not actuated) actuated)..

Friday, November 21 21,, 2008 9:52:58 AM 50

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION) forVCS Notes

Ability to (a) predict the impacts of the following on the LOCA Outside Containment and (b) based on those predictions, predictions, use procedures to correct.

operation:

correct, control, or mitigate the consequences of those abnormal operation: Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

operations.

Tier: 1 Group: 1 Importance Rating: SR04.3 Technical

Reference:

EOP-2.0, Rev 13

EOP-2.4, Rev 11

EOp-2.5, EOP-2.5, Rev 7 Proposed reference.

references to be provided to applicants during examination:

None Learning Objective: EOP-1.0-07 EOP-1.Q..07 Question History:

MODIFIED (Although written "from scratch",

Reference questions EOPS 439, 10 CFR Part 55 Content:

Comments:

439, 459, 43(b)(5) scratch", this question is similar enough to Closed 471,, 478, & 525 to be classified as MODIFIED) 459, 471 The KA is matched because the operator must demonstrate the ability to interpret a given set of plant conditions and select an appropriate procedure flowpath associated with an unisolable LOCA outside the Containment. .

The question is SRO-Only because the question involves assessing plant conditions of an abnormal/emergency event, and then prescribing a procedure, procedure. and procedure flowpath.

flowpath.

Fridav. November 21, Friday, 21,2008 2008 9:52:58 AM 51

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

18. E08 G2.2.25 003/NEW/NINHlGHERIISROISUMMERI2I20091 003INEW/N/AlHIGHERlISRO/SUMMERI2I20091

When implementing EOP-16.0, Response to Imminent Pressurized Thennal Thermal Shock, Shock, which ONE (1) of the following identifies the required soak time and the bases for the soak?

M A'!I ONE (1) hour; prevent flaw growth B. ONE (1) hour; relieve compressive stress on the inner wall of the Reactor Vessel C. EIGHT (8) hours; prevent flaw growth D. EIGHT (8) hours; relieve compressive stress on the inner wall of the Reactor Vessel Feedback

A.

B.

CORRECT: lAW EOP-16.0, Step 24. According to HFRP1BG, HFRP1 BG, page 2, objective of EOP-16.0 is to prevent the growth of a flaw.

flaw.

2, the Plausible because the 1st part is correct lAW EOP-16.0, Step 24. Also plausible because compressive stress is a stress mechanism on the outer wall due to outerwall thermal stresses.

Incorrect because both pressure and thermal stresses on a cooldown are tensile on the inner wall (see Figure TS15.24).

C. Plausible because the 2nd part is correct. 1st part plausible because this is a soak period required for PZR safety valves during a plant heatup.

Incorrect because the soak period is only 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per Step 24 of EOP-16.0.

D.

D. Plausible because this is a soak period required for PZR safety valves during a plant heatup. Also plausible because compressive stress is a stress mechanism on the outer wall due to thermal stresses.

outerwall stresses.

Incorrect beCause because the soak period is only 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per Step 24 of EOP-16.0. Also incorrect because both pressure and thermal stresses on a cooldown are tensile

Friday, Friday, November 21 on the inner wall (see Figure TS15.24) 21,, 2008 9:52:58 AM TS15.24)..

52

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Notes

(Pressurized (Pressu rized Thermal Shock) Knowledge of the bases in Technical Specifications for limiting conditions for op~rations operations and safety limits Tier:

Group:

limits..

3 1

Importance Rating: SRO SR04.2 4.2 Technical

Reference:

T.S. 3.4.9.1 3.4.9.1

T.S. Figures 3.4-2 & 3.4-3

T.S. Bases 314.4.9.1)b) 3/4.4.9.1)b)

EOP-16.0, Step 24 Proposed references to be provided to applicants during examination:

None Learning Objective: EOP-16.0-07 Question History: NEW 10 CFR Part 55 Content: 43(b)(2)

Comments:

Meets KIA by application of TS RCS Chemistry requirements.

SRO-Ievel because n requirements.

it requires detailed knowledge of EOP-16.0 and the bases for a step that is not an Immediate Operator Action Action..

Friday, Friday, November 21, 21 , 2008 9:52:58 AM 53

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

19. G2.1.27 002fNEWINIAIMEMORYlFUNDAMENTALIISRONCSNS OO2INEWIN/NMEMORY/FUNDAMENTAU/SRONCSNS NRC 200911

Which ONE (1) of the following describes the basis for OPERABILITY limits on accumulator volume and boron to ensure that assumptions used for accumulator injection in the safety analysis are met?

A.

A. All THREE (3) accumulators are required to provide the inijial initial core cooling for a postulated Loss of Coolant Accident and limit the maximum power that may be reached during large secondary pipe ruptures.

ruptures.

B. ONLY TWO (2) accumulators are required to provide the initial inijial core cooling for a postulated Loss of Coolant Accident and limit the maximum power that may be reached during an excessive heat removal event as a result resutt of a feedwater ma~unction .

malfunction.

C. All THREE (3) accumulators are required to provide the initial core cooling for a postulated Loss of Coolant Accident and limit the maximum power that may be reached daring during an excessive heat removal event as a result resutt of a feedwater ma~unction .

malfunction.

D~ ONLY TWO (2) accumulators are required to provide the initial D'!' inijial core cooling for a postulated Loss of Coolant Accident and limit the maximum power that may be reached during large secondary pipe ruptures.

ruptures.

Feedbac::k Feedback

A.

A. Plausible because the During a Large Break Loss of Coolarit described in Chapter 15.4 of the FSAR (Ref.

Coolant accident as (Ref. 15.4.1.1.2 15.4.1.1 .2 VCSNS FSAR) when the RCS depressurizes to 600 psia the accumulators begin to inject water into the reactor coolant loops. From the latter stage of blowdown and then the beginning-of~reflood, the safety injection accumulator tanks rapidly discharge beginning-of-reflood, borated cooling water into the RCS, contributing to the filling of the reactor vessel downcomer. The downcomer water elevation head provides the driving force required for refJooding of the reactor core. Also plausible because the 2nd ha~ is correct. (Ref. T.S. Bases 3/4.5.1) half because, according to VCSNS FSAR (Table 15.4-5), presents the Incorrect because, reflood mass and energy release to the containment and the broken loop accumulator mass and energy flowrate to containment. i.e. i.e. it is postulated that the broken loop accumulator mass is unavailable for reflood.

B. Plausible because the 1st part is correct. During a Large Break Loss of Coolant accident as described in Chapter 15.4 of the FSAR (Ref. 15.4.1.1.2 VCSNS FSAR) when the RCS depressurizes to 600 psia the accumulators begin to inject water into the reactor coolant loops.

loops. From the latter stage of blowdown and then the beginning-of-reflood, the safety injection accumulator tanks rapidly

discharge borated cooling water into the RCS, contributing to the filling of the reactor vessel down downcomer.

comer. The downcomer water elevation head provides the driving force required for reflooding of the reactor core.

Friday, November 21, 21 , 2008 9:52:58 AM core. However, according to 54

QUESTIONS REPORT for far VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION) cantainment and the broken laap the containment loop accumulatar accumulator mass and energy flawrate flowrate to.

to

containment.

cantainment. i.e. n unavailable for removal Heat Removal it is postulated reflood.

far reflaod.

pastulated that the broken 2nd part is remaval transient as described in, (Ref 15.2.10 of reactar power for an increase in reactor because accumulator mass is braken loop accumulatar plausible n it is a significant heat af VCSNS FSAR; Remaval Due To Feedwater Sysytem Malfunction), which would result in FSAR; Excessive far which if accumulators could inject waurd would tenninate terminate the power pawer rise.

Incorrect Incarrect because the T.S. bases for accumulators states specifically the far the accumulatars barated water serves to limn borated limit the maximum power pawer which may be reached during large secondary secandary pipe ruptures ruptures..

C. Plausible because the During a Large Break Loss of Coolant accident as described in Chapter 15.4 of the FSAR (Ref. 15.4.1.1.2 VCSNS FSAR) when the RCS depressurizes to 600 psia the accumulators accumulatars begin to inject water into the reactor coolant loops. From the latter stage of blowdawn blowdown and then the beginning-of-reflood, the safety injection accumulator tanks rapidly discharge borated cooling water into the RCS, contributing to to. the filling afof the reactor vessel downcomer.

downcamer. The downcomerdawncamer water elevatian elevation head provides the driving force required for far reflooding of af the reactar reactor care.

core. 2nd part is also plausible because it is a significant heat removal transient as described in, (Ref 15.2.10 of VCSNS FSAR; Excessive Heat Remaval Removal Due To. To Feedwater Sysytem Malfunction),

Malfunctian), which wauld would result in an increase in reactor reactar power for which if pawer far

accumulators could inject would tenninate broken looplaap accumulator terminate the power rise.

Incorrect because Incorrect because according to.

Incarrect accumulatar mass and energy flawrate flowrate to rise.

to VCSNS FSAR (Table 15.4-5) presents the reflood mass and energy release to the containment and the to. containment.

cantainment. i.e. nis i.e. it postulated that the broken loop loap accumulator accumulatar mass is unavailable for reflood.

far reflood.

Also incorrect because the T.S. bases for Also. far the accumulators states specifically the borated barated water serves to limit limn the maximum pawer power which may be reached during large secondary pipe ruptures. ruptures.

D.

D. CORRECT: During a Large Break Loss Lass afof Caalant Coolant accident as described in Chapter 15.4 oltheaf the FSAR (Ref. 15.4.1.1.2 VCSNS FSAR) when the RCS depressurizes to 600 psia the accumulators accumulatars begin to inject water into into. the reactor coolant loops.

loaps. From Fram the latter stage of af blowdown blawdown and then the beginning-of-reflood, beginning-of-refload, the safety injectioninjectian accumulator tanks rapidly discharge borated barated cooling water into the RCS, contributing to the filling of the reactar reactor vessel downcomer.

downcamer. The downcomer dawncamer water elevation elevatian head provides the driving force required for reflooding af farce of the reactor core. However, Hawever, according to. to VCSNS FSAR (Table 15.4-5) presents the reflood reflaad mass and energy release to to. the containment and the broken loop laap accumulator accumulatar mass and energy flowrate flawrate to cantainment.

containment. i.e. n it is postulated that the brokenbraken loop laap accumulator mass is unavailable far reflaod. Also.

for reflood. Also correct because the T.S. T.S. bases far for the

accumulators states specifically the borated accumulatars power Friday, November 21 Friday, . 20089:52:58 AM 21,20089:52:58 barated water serves to limit the maximum pawer which may be reached during large secondary pipe ruptures. ruptures.

55

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Notes

(Conduct of Operations) Knowledge of system purpose and/or andlor function.

Tier: 3 Group: 1 Importance Rating: SR04.0 Technical

Reference:

Chapter 15.4.1.1 15.4.1.1.2.2 VCSNS FSAR

TS Bases 3/4.5.1 3/4.5.1

Chapter 15.2.10 ofVCSNS of VCSNS FSAR Proposed references to be provided to applicants during examination:

None Learning Objective: AB-10-21 Question History: NEW 10 CFR Part 55 Content:

Contont: 43(b)(2)

Comments:

Meets KJA KIA by application of TS RCS ~CS Chemistry requirements.

SRO-Ievel because ij it requires detailed knowledge of a GOP and the bases for a T.S. that is NOT a Safety Limit. .

Friday, November 21, Friday, 21 , 2008 9:52:58 AM 56

QUESTIONS REPORT for VCS 2009 NRCSRO NRC 'SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

20. G2.1.34 00lIMODIFIEDliLOWERIISROISUMMER!2/l009INO OOlIMODIFIED/ILOWERlISRO/SUMMERI212009INO

Which ONE (1) of the choices below completes the following statement?

In accordance wtth with GOP-2, Plant Startup Starlup and Heatup (Mode 5 to Mode 3), Dissolved Oxygen must be within limits prior to exceeding corrosion.

in order to prevent _

M 200'F; stress B. 200'F; crevice C. 250'F; stress D. 250'F; crevice

Friday, November 21, 2008 9:52:58 AM 21,2008 57

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Feedback

A. CORRECT. GOP-2, Reference Page, Item 2.E, requires RCS chemistry to be in-spec prior to exceeding 200"F.

exceeding 200"F.

200°F.

200°F. Step 3.1 3.1 also contains this requirement.

Additionally, CP-614, Attachment III, requires Oxygen to be < 0.1 0.1 ppm prior to B.

B. Plausible because the 1st part is correct. 2nd part is plausible because crevice corrosion is one form of corrosion and is associated with oxygen.

oxygen.

Incorrect because T.S. bases for LCO 3.4.7 specifically states that RCS lim~s are designed to minimize and reduce the potential for stress Chemistry limits corrosion.

corrosion.

C. Plausible because the 2nd part is correct.

correct. Also plausible because the note in

"*Lim~ is not applicable with Tavg < 250°F."

T.S. Table 3.4-2 states "*Limit 250"F."

Incorrect because GOP-2 is more restrictive in that Oxygen must be within limits lim~s prior to exceeding 200"F.

200°F.

D. Plausible because the note in T.S. Table 3.4-2 states "*Limit is not applicable wnh Tavg < 250"F."

with 250°F." 2nd part is plausible because crevice corrosion is one form of corrosion and is associated with oxygen.

oxygen.

Incorrect because T.S. bases for LCO 3.4.7 specifically states that RCS stJess Chemistry limits are designed to minimize and reduce the potential for stress corrosion. Also incorrect because GOP-2 is more restrictive in that Oxygen must be within limits prior to exceeding 200°F.

200"F.

Friday, Friday, November 21 21,, 2008 9:52:58 AM 20089:52:58 58

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Notes

(Conduct of Operations) Knowledge of primary and secondary chemlstry chemistry limits.

Tier:

Tier: 3 Group: 1 Importance Rating: SRO 3.5 SR03.5 Technical

Reference:

T.S. Table 3.4-2 (Page 255 of 588)

T.S. 3.4.7 (Page 254 of 588)

Proposed references to be provided to applicants during examination:

None Learning Objective: S8-4-15 Question History:

(Allhough written '1rom MODIFIED (Although "from scratch",

scratch", this question is similar enough to Open Reference question NORMAL OPERATIONS 58 to be classified as MODIFIED)

10 CFR Part 55 Content: 43(b)(2),(5)

Comments:

Meets KiA KIA by application of TS RCS Oxygen Chemistry requirements.

requirements.

SRO-Ievel because it ~ requires detailed knowledge of a GOP and the bases for a T.S. that is Lim~ .

NOT a Safety Limit.

Friday, November 21 21,, 2008 9:52:58 9:52 :58 AM 59

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

21. G2.2.38 G2.2.38001INEW//LOWERIISRO/SUMMERl2/2009INO OOIINEW/ILOWERIISRO/SUMMERI2I2009/NO Given the following plant cond~ions conditions::

The unit was at 100% power when a high pressure feedwater heater bypass valve failure caused Reactor power to peak at 2960 MWt on U9003-SM.

The failed valve has been closed and Reactor power is at 100%.

Which ONE (1) of the choices below completes the following statement regarding transient follow-up actions?

Adjust Reactor power as necessary to obtain ____

_ _ less than or equal to 2900 MWI.

MWt.

Notify the NRC within _ __

A. a five minutes rolling average power; 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

B. a five minutes rolling average power; 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

C. an integrated shift average power;

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

hour.

D~ an integrated shift average power; Dt 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

hours .

Friday, November 21 21,, 2008 9:52:58 AM 60

QUESTIONS REPORT forVCS for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Feedbac=k Feedback

A. 102'lA> overpower is evaluated by checking the five Plausible because the 102%

minutes rolling average computer point. 2nd part is plausible because a Limit Violation) if one hour report would be required (per T.S. 6.7, Safety limit lim~ was violated.

Core Power Safety Limit Incorrect because the 2900 MWt limit is based on an 8-hour a-hour rolling average. Also incorrect because the report is a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months requirement, average. requirement, not one hour.

B. Plausible because the 2nd part (reporting requirement) is correct. Also plausible because the 102% overpower is evaluated by checking the five minutes rolling average computer point.

lim~ is based on an a-hour Incorrect because the 2900 MWt limit 8-hour rolling average.

average.

C. Plausible because the 1st part is correct. 2nd part is plausible because a one hour report would be required if Core Power Safety limitLimit was violated.

Incorrect because the report is a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months requirement, not one hour.

D. CORRECT. The action is correct because licensed power limit OAP-l00.6, lim~

limit is required by OAP-100.06 mon~oring an 8 compliance is maintained by monitoring lim~

a hour shift average per OAP-100.6, Section 10.5.a .. A 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months report for exceeding the license OAP-l00.0e (Section 10.5); Nl-122, NL-122, REGULATORY NOTIFICATION AND REPORTING (Enc. (Ene. A, Item P-9).

Friday.

Friday, November 21, 21,2008 2008 9:52:58 AM 61 61

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Notes

(Equipment Control) Knowledge of conditions and limitations in the facility license.

license.

Tier: 3 Group: 2 Importance Rating: SR04.5 Technical

Reference:

OAP-100.e, Section 10.0 (Pages 11 and 12)

OAP-100.6,

NL-122, Enclosure A, P-9 (Page 14 of 32)

VCSNS Operating License, License, Section 2.C.(1)

Proposed references to be provided to applicants during examination:

None Learning Objective: OAP-100.6-7 Question History: NEW 10 CFR Part 55 Content: 43(b)(1)

Comments:

Meels KIA by requiring knowledge of action(s) associated with exceeding the reactor power Meets limits specified in the facility license.

SRO only because it requires the direction of compensatory actions and knowledge of reporting requirements. Because of the importance, requirements. importance. the license limit and actions may be known by RO and SRO candidates but the direction of corrective actions and intemaVextemal internal/external reporting are the responsibility of the Shift Supervisor (see NL-122, Sections 6.1 6.1 and 6.2).

6.2) .

Friday, November 21 21,. 2008 9:52:58 AM 62

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

22. G2.3. 13 OOlINEW//LOWER//SRO/SUMMERI2I2009INO G2.3.13 OOIINEW/ILOWERIISRO/SUMMERI2I2009INO

Given the following plant conditions:

100% power conditions:

A QA Audit has determined that an ECCS valve lineup was not properly documented and may be incorrect.

A Reactor Building entry has been approved to verify the position of the valves in question.

Which ONE (1) of the choices below completes the following statement?

The team will enter the Reactor Building through the Airlock. At the completion of this entry, the operating crew will generate an Action R&R to ensure that _ _ __

is conducted.

A.. ESCAPE; A

STP-109.001, STP-1 09.001, Reactor Building Closeout Inspection, Inspection, B~ ESCAPE; STP-215.001 B, Reactor Building Personnel Escape Airlock Test,Test,

C. PERSONNEL; STP-109.001, STP-1 09.001, Reactor Building Closeout Inspection, D. PERSONNEL; D.

STP-215.001A, Reactor Building Personnel Airlock Test,

Friday, Friday, November 2121,, 20089:52:58 2008 9:52;58 AM 63

QUESTIONS REPORT .

forVCS for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

FeedbBCk Feedback

A.

A. Plausible because the airlock is correct and the closeout inspection is required at the completion of the entry.

Incorrect because, per OAP-IOB.I OAP-10B.1,, Section 6.5.b., the Action R&R is intended to ensure STP-215B is completed within 7 days B. CORRECT. Per OAP-10B.01, OAP-I OB.OI, Control of Reactor Building Entry, the ESCAPE Airlock is used to reduce exposure to neutron streaming. Per OAP-IOB.I, OAP-10B.1, 6.5.b., the Action R&R is intended to ensure STP-215B is completed Section 6.5.b.,

w~hin 7 days within C. Plausible because this is the normal entry point for non-power entries. Also STP-I09.001 must be performed at the completion of the plausible because STP-109.001 entry.

Incorrect because the Action R&R ensures STP-215.001 B is conducted, not STP-I09.001. Also incorrect because the ESCAPE hatch is used to reduce STP-109.001.

exposure to neutron streaming.

D.

D. Plausible because this is the normal entry point for non-power entries. Also plausible because, if the PERSONEL AIRLOCK were used, this would be the

correct STP.

Incorrect because the ESCAPE hatch is used, not the PERSONEL AIRLOCK.

Friday, November 21, Friday, 21 , 2008 9:52:58 AM 64

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Notes

(Radiation Control) Knowledge of radiological safety procedures pertaining to licensed operator duties, duties.

Tie r:

Tier: 3 Group: 3 Importance Rating: SRO 3.8 Technical

Reference:

OAP-108.1, 6.416.5 (Pages 3 and 4 of 5)

OAP-1 08.1, 6.4/6.5 references to be provided to applicants during examination:

Proposed reference.

None Learning Objective: OAP-108.1-03 Question History: NEW 10 CFR Part 55 Content:

Content 43(b)(4), (5)

Comments:

Meets KIA t<JA by requiring detailed knowledge of the administrative requirements (radiological safety practices - use Escape Hatch to reduce exposure to neutron streaming) associated with the RB entry procedure (OAP-108.1).

(OAP-10B.1).

SRO only in that the Duly Duty Shift Supervisor implements OAP-108.1 OAP-108.1 and the CRS must direct the operating crew to initiate the R&R to ensure completion of STP-215B.

STP-215B.

Friday, November 21, 2008 9:52:58 AM AM 65

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER, (AFTER CHANGES AFTER VALIDATION)

23. 02.3.14 G2.3.14 OO IINEW/ILOWERI/SRO/SUMMERI2I2009INO OOIINEW//LOWER/ISRO/SUMMER/2/2009INO 100% power cond~ions:

Given the following plant conditions:

Operators have just completed shifting from' servioe service to GOT '8' 'B' in service.

servioe.

from'Gas Gas Decay Tank (GOT) 'A' in

The relief valve on GOT 'A' failed open and ALL contents were released.

idenijfies both the assumed radioisotope and the Which ONE (1) of the following identifies LIMITING dose that an individual standing at the nearest exclusion area boundary may receive?

A. Krypton-85; A. Krypton-S5; 5.0 Rem B. Krypton-S5; Krypton-85; 0.5 Rem Xenon-133; C. Xenon-133; C.

5.0 Rem

D~

O~ Xenon-133; 0.5 Rem

Friday, November 21,2008 Friday, 21, 2008 9:52:58 AM 66

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Feedback

A. Krypton-85 is plausible because it~ is the 2nd most contributing isotope to doses after a WGDT Rupture (see Chapter 15.3 and Table 15.3-5 of the FSAR).

2nd part is plausible because 5.0 R is the CDE (thyroid) lim~

Condition 461.461 . .

FSAR). The limit in In~iating Initiating Incorrect because lLCO CO 3.11 .2.6 specifically addresses Xe-133. Also incorrect 3.11.2.6 because T.S. Bases 3/4.11.2.6 specifically gives 0.5 R as the exposure limit.

B. Krypton-85 is plausible because it is the 2nd most contributing isotope to doses Krypton-as after a WGDT Rupture (see Chapter 15.3 and Table 15.3-5 ofthe of the FSAR).

FSAR). Also plausible because the 2nd part is correct (see T.S. T.S. page B 3/4 11-2).

3/411-2).

Incorrect because LCO 3.11 .2.6 specifically addresses Xe-133.

3.11.2.6 C. Plausible because the 1st part is correct.

correct. The 2nd part is plausible because 5.0 Cond~ion 461 R is the CDE (thyroid) limit in Initiating Condition 461..

Incorrect because T.S. bases 3/4.11.2.6 3/4.11 .2.6 specifically gives 0.5 R as the exposure lim~.

limit. .

D.

D. CORRECT. LCO 3.11.2.6 3.11 .2.6 specifically addresses Xe-133. T.S.T.S. bases 3/4.11.2.6 3/4.1 1.2.6

gives 0.5 R as the exposure limit.

Friday, Friday, November 21, 21 . 2008 9:52:58 AM 67

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Notes

(Radiation Control) Knowledge of radiological or contamination hazards that may arise during normal, abnormal, abnormal, or emergency conditions or activities.

Tier:

Group:

3 3

activities .

Importance Rating: SR03.8 Technical

Reference:

3.11 .2.6 (Page 453 of 588)

T.S. 3.11.2.6

T.S. Basis 3.11 .2.6 (Page 544 of 588) 3.11.2.6 Proposed references to be provided to applicants during examination:

None Learning Objective: AB-12-19 Question History: NEW 10 CFR Part 55 Content: 43(b)(2),

43(b)(2), (4)

Comments:

KIA by requiring knowledge of an assumed radiation dose caused by a component Meets KiA failure as referenced in TS.

SRO only because it requires the applicant to apply both the TS and the TS Basis in order to correctly answer the question, question.

Friday, Friday, November 21, 21 , 2008 9:52:58 AM 68

QUESTIONS REPORT forVCS for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

24. G2.4.20 00 OOllNEW/IHIGHERIISRO/SUMMERl2/2009INO IINEW/IHIGHERIISRO/SUMMERI2I2009INO cond~ions ;

Given the following plant conditions:

A large break LOCA has occurred.

Component Cooling Water (CCW) Pump 'C' CCW Loop 'A' is the Active Loop.

'c' is not available.

The operating crew is performing EOP-2.0, Loss of Reactor or Secondary Coolant, Step 16 - Verify equipment is available for Cold Leg Recirculation.

Recirculation.

Which ONE (1) of the follOWing following is the REQUIRED action with respect to shifting CCW Loop 'A' to FAST speed?

A. Shift CCW Pump 'A' to FAST and continue in EOP-2.0. EOP-2.0.

6~

B~ Leave CCW Pump 'A' in SLOW and continue in EOP-2.0.

C. Shift CCW Pump '6' 'B' to FAST, swap active loops per SOP-118, Component Cooling Water System, THEN continue in EOP-2.0.

D. Reduce CCW flow to the minimum required for loads per EOP-2.2, Transfer to Cold Leg Recirculation, THEN continue in EOP-2.0. EOP-2.0.

Feedback

A..

A Plausible because this is a potential alignment in SOP-118 if CCW Pump 'c' was running. Also plausible because it would accomplish the desired end state.

running.

Add~ionally Additionally,, "continue in EOP-2.0" is plausible because it is correct (see CAUTION - Step 16.c, 2nd bullet). bullet).

Incorrect because CAUTION - Step 16.c prohibits shifting the running pump to FAST if the sing CCW pump is not available, as in the given conditions.cond~ions .

6B.. CORRECT. EOP-2.0, CAUTION - Step 16.c prohiMs prohibits shifting the running pump to FAST if the sing CCW pump is not available, as in the given conditions.cond~ions.

"Continue in EOP-2.0" is correct (see CAUTION - Step 16.c, 2nd bullet). bullet).

C. Plausible because this is a possible solution and because it is an evolution addressed in SOP-118. Also plausible because it ~ may uijimately ultimately be the correct alignment when time allows in another EOP.

Incorrect because CAUTION - Step 16.c gives specific direction NOT to realign anything at this time and continue in EOP-2.0.

D.

D. Plausible because this will be accomplished in EOP-2.2 per the 2nd bullet of CAUTION - Step 16.c.

Friday, Incorrect because it EOP-2.2.

Friday, November 21, 21 , 2008 9:52:58 AM tt will NOT be accomplished until the crew implements 69

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

Notes

(Emergency Procedures/Plans) Knowledge of the operational implications of EOP warnings notes.

Tier:

Group:

3 4

wamings cautions and Importance Rating: SR04.3 Technical

Reference:

EOP-2.0. 15)

EOP-2.0, Step 16.c (Page 10 of 1S)

SOP-118,Section IV.D (Pages 20 and 21)

SOP-11B.

references to be provided to applicants during examination:

Proposed referencea None Learning Objective: EOP-2.0-05. 06 EOP-2.0-0S, Qu ....tion History:

Question NEW 10 CFR Part 55 Content: 43(b)(5) 43(b)(S)

Comments:

Meets KIAKJA by requiring an operational decision based on an EOP-2.0 NOTE and knowledge of the SOP (SOP-118).

(SOP-118).

SRO only because it requires application of specific mitigating strategies in EOP-2.0.

EOP-2.0.

Selection of procedures per 10CFRSS.43(b)(S) 10CFRS5.43(b)(5) is also involved in the process of eliminating choices C & 0 and in selecting the correct answer.

Friday, Friday, November 21, 21 , 2008 9:52:58 AM 70

QUESTIONS REPORT for VCS 2009 NRC SRO WORKSHEET (AFTER CHANGES AFTER VALIDATION)

25. G2.4.41 002JNEW/IHIGHERIISRO/SUMMERI2I2009INO 002lNEW/IlllGHERlISRO/SUMMERl2/2009INO

Which ONE (1) of Emergency per EPP-001 the following would REQUIRE a declaration of a Site Area ofthe EPP-001,, Activation and Implementation of the Emergency Plan?

A. For TEN (10) minutes, ALL of the followingfollowing::

1) DC bus undervoltage alanns alarms on all ESF buses, AND 2) 480V ESF Channel A OR B Loss of DC Alanns.

Alarms.

B. For TWENTY (20) minutes, ALL of the following following::

1) UV alanns alarms on 1DA and 1DB, lOB, AND 2) Loss of 115KV ESF Potential lights.

C. For TEN (10) minutes, ALL of the followingfollowing::

1) DC bus undervo~age undervoltage alanns alarms on all ESF buses, AND 2) 480V ESF Channel A OR B Loss of DC Alanns Alarms AND 3) DG A OR B Loss of DC Alann. Alarm.

D~ For TWENTY (20) minutes, ALL of the following:

UV alarms on 1DA and 1DB,

1) UValanns 1DB, AND 2) Loss of 115KV ESF Potential lights, lights, AND 3)

Loss of 230KV ESF Potential lights.

Feedback

A. Plausible because the given 10 minutes exceeds the threshold for an ALERT.

Also plausible because 2nd part contains 2 of 3 Detection Methods for a SAE EPP-001,, Alt.

(see EPP-001 Att. II, Page 12 of 25).

ALERT.

Incorrect because 10 minutes does not exceed the SAE threshold of 15 minutes. minutes.

Also incorrect because all of the Detection Methods are NOT met.

B. Plausible because the 1st 1st part is correct - the given 20 minutes exceeds the threshold for a SAE.

Incorrect because 2nd part contains only 2 of 3 Detection Methods for a SAE.

C.

C. Plausible because the given 10 minutes exceeds the threshold for an ALERT. ALERT.

Also plausible because 2nd part contains 3 of 3 Detection Methods for a SAE and would be correct if the given time was >15 minutes.

Incorrect because 10 minutes does not exceed the SAE threshold of 15 minutes.

D. CORRECT. Per EPP-001 EPP-001,, Att.

Alt. II, Page 11 of 25, an SAE is required when ALL ofthe of the Detection Methods are met for more than 15 minutes. minutes .

Friday, November 21, 2008 9:52:58 AM 71

QUESTIONS REPORT QUESTIONS REPORT for VCS for VCS 2009 2009 NRC NRC SRO SRO WORKSHEET WORKSHEET (AFTER (AFTER CHANGES CHANGES AFTER AFTER VALIDATION)

VALIDATION) '"

Notes Notes

- (Emergency Procedures/Plans)

(Emergency Procedures/Plans) Knowledge Knowledge ofof the the emergency emergency action action level level thresholds thresholds and and classifications.

classifications.

Tier:

Tier: 33 Group:

Group: 4 Importance Rating: SR04.6 SRO 4.6 Technical

Reference:

Technical

Reference:

EPP-001,. Attachment II.

EPP-001 II, Page 1 (Pages" (Pages 11 & 12 ofof 25)

Proposed references to be provided to applicants during examination:

None Learning Objective: EPP-001-4095 Question History:

Queetion NEW 10 CFR Part 55 Content: 43{b){5) 43(b)(5)

Comments:

Meets KIA by requiring knowledge of Oetedion Detection Methods for declaration of a SAE and an Alert Alert.

SRO Only because knowledge of the EAL Tables is required .

Friday, Friday, November November 21, 21 . 2008 2008 9:52:58 9:52:58 AM AM 72 72

ML092850007

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ML092850007

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