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June 05000390-08-301 Exam Final Combined Ro/Sro Written Exam with Kas, Answers, & References (Part 2 of 2)
	ML081900165
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Site:	Watts Bar
Issue date:	06/03/2008
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RO Watts Bar 2008 NRC Initial Lic en se Exam

WRITTEN QUESTION DATA SHEET

Question Number: 51

KIA: 076 K4.03

Service Water

Knowledge of SWS design featur e(s) andlo r interlock(s) which provide for the following: Automatic ope ning

features associated with SWS isolat ion valves to CCW heat exchang ers.

Tier: 2 RO Imp: 2.9 RO Exam: Yes Cognitive Level: LOW

Group: 1 SRO Imp: 3.4 SRO Exam: Yes Source: WBN Bank

Applicable 10CFR55 Section : 41.7

Learn ing Objective: 3-0 T-SYS067A, Objective 13: Given a loss of power, determine the correct respons e

of the ERCW System includ ing : a. "C" CCS Heat Exchanger outlet valves .

References: 3-0 T-SYS067A, Rev. 10. 1-47W611- 67-5 , E-O Append ix A. Rev. 27.

Question:

Reactor trip and safet y injection signals have been manuall y initiated. Wh ich ONE of the following describes

the required posi tions for the listed ERCW valves in accordance with E-O, App endi x A, Equipment

Verifica tion?

O-FCV-67-144, O*FCV-67-152,

"CCS Heat Exchanger 'C' "CCS Heat Exchanger 'C'

Disch to Hdr A" Alt Disch to Hdr B"

A. CLOSED OPEN to Position A

B. THR OTTLED CLOSED

C THROTTLED OPEN to Position A

U. OP EN CLOSED

DISTRACTOR ANALYSIS

a. CORRECT. Under normal conditions O-FCV-6 7-152 is closed with power on the valve . Upon the

receipt of an SI signal 0-FC V-67-152 strokes to the 35% position automa tically. During performan ce of

E-O, Appendi x A the operator places the O-FCV-67-152 handswitch for the valve in the Position A. 0-

FCV-67-144 is normally open and is closed by manual operator action during performan ce of E-O,

Appendix A.

b. Incorrect Plausi ble since the valve positions are reversed.

c. Incorrect. Plausible since the operator may confuse listed valves with others with similar numbers.

d. Incor rect. Plausible since the operator may conf use listed valves with others with simila r numbers .

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RO Watts Bar 2008 NRC Ini ti al Licen se Exam

WRITTEN QUESTION DA TA SHEET

Question Number: 52

KIA : 076 A 1.02

Service Water

Ability to predict and/or monito r changes in parameters (to prevent exceeding design limits) assoc iated with

operating the SWS controls including : Reactor and turbine building closed cooling water temperatures .

Tier: 2 RO Imp: 2.6 RO Exam : Yes Cogn itive Level : HIGH

Group : 1 SRO Imp : 2.6 SRO Exam : Yes Source: NEW

Appli cable 10CFR55 Section: 41.5/45 .5

Learning Objective : 3-0T-SYS067A, Objec tive 8: State the ERCW System normal disch arge path and

given a failure of the path. discuss the alterna te discha rge paths .

References : AOI-13, LOSS OF ESSENTIAL RAW COOLING WATER , Rev. 35.

Question :

Given the following plant conditio ns:

The plant is ope rating at 100% power with 1B CCP in service .

The Control Room Operator shutdowns down the C-A ERCW pump in prepa ration for a test on the

2A 6.9 KV Shutdown Board.

Which ONE of the following describes the impact of shutting down the pump on the listed parameters?

(Assume no other operator action.)

16 CCP Oil Seal Water Return Heat

Temperature Exchanger Temperature

A. Rises Rises

B. Rises Remains constant

C. Remai ns constant Rises

D. Remains constant Remains constan t

DISTRACTOR ANALYSIS

a. Incorrect. The C-A ERCW pump supplies flow to the A header. Whe n the pump is stopped, flow to

both the 1A and 2A ERCW headers is decreased. This causes a dec rease in flow through the seal

water heat exchangers. A reduction in flow to the 1A CCS heat excha nger will cause the A ESF

heade r to heat up as well as the Reactor Building and Miscella neous Equipment Headers to heat up.

The 1A CC P which would normally be in service would show an increase in oil temperature. The Seal

Water Heat Exchanger, supplied off the Miscellaneous Equipment Header wou ld also heat up .

b. Incorrect. Plausible, beca use oil temperat ure does rise, howeve r, Seal Water Return Heat Exchanger

temperatu re also rises.

c. CORR ECT. Seal Wate r Return Heat Exchanger temperatu re does rise, and CCP oil tempera ture

remai ns constan t.

d. Incorrect. Plausible. since candida te incorrectly recalls cooli ng water supp ly to this component.

56 of 81

RO Watts Bar 2008 NRC Initial License Exam

WRITTEN QUESTION DATA SHEET

Qu estion Number: 53

KIA : 078 K1 .01

Instrument Air

Knowledge of the physica l conne ctions and/or cause -effec t relationships betwee n the lAS and the following

systems: Sensor air.

Tier: 2 RO Imp : 2.8 RO Exam: Yes Cogn itive Lev el : HIGH

Group : 1 SRO Imp: 2.7 SRO Exam : Yes Source : NEW

Applicable 10CFR55 Section: 4 1.2 to 41.9/45.7 to 45.8

Learn ing Obj ect ive : 3-0 T-SYS032A, Objecti ve 16: List the even ts and their correspond ing set points that

take place on decre asing control air pressure .

Refer ences : 1-47W611- 32-2 rev 4, SOI-32.02 rev 19 Note on page 12.

Question :

Which ONE of the follow ing identifies both of the following ?

(1) The lowest of the listed containment pressures that resul ts in 1-FCV -32-80 , Aux Air to Rx Bldg

Train B, being automa tically isolated, and

(2) The lowest of the listed air pressure s sensed downstream of the valve that allows the valve to

REMAIN OP EN after the valve was opened and the control switch on 1-M-15 placed to A-Au to after

the isolation signa l was reset.

ill ill

A. 2.0 psid 68 psig

B 2.0 psid 78 psig

C. 3.0 psid 68 psig

D. 3.0 psid 78 psig

DISTRACTOR ANALYSIS

a. Incorre ct. Conta inme nt pressure is not high enough to cause the isolation and the sens or downstream

of the valve will not allow the valve to remain open with the pressure at 68psig , but plaus ible because

with the contai nment pressu re above 1.5, a Phase A isolation wou ld have occu rred and man y paths

would have isolated .

b. Incorrect. Containment press ure is not high enough to cause the isolatio n, but since it is above 1.5,

then a Phase A isolation would have occurred and many paths would have isola ted and the

downs tream pressure is high enough to allow, but plausible because with the containment pressure

above 1.5 then a Phase A isolation wou ld have occurred and many paths would have isolated and 78

psig is high enough for the downstream sensor to allow air to open the valve .

c. Incorrect. Con tainment press ure is high enoug h to cause the isolation , but the pressur e sensed

dow nstream of the valve is not high enough to allow operating air pressu re to maintain the valve open

after the switch is placed in A-Auto, but plausible because with the containment press ure above 3.0 psig

then a Phase B isolation would have occurred causing isolation of the valve.

d. CORR ECT. With containmen t pressure greater than 2.8 psig, a phase B isolation will automatica lly

occur , and the sensor downstre am of the isolation valve must detect grea ter than 75 psig to allow the

valve to remain open after the contro l switch was placed to the A-Auto posi tion ,

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RO Watts Bar 2008 NRC Initial License Exam

WRITTEN QUESTION DA TA SHEET

Question Number: 54

KIA : 103 A2.05

Containment

Ability to (a) predict the impacts of the following malfunct ions or operat ions on the containment system-and

(b) based on those predictions , use procedures to correct, control, or mitigate the cons equen ces of those

malfunct ions or operations: Emergency containme nt entry.

Tier: 2 RO Imp: 2.9 RO Exam : Yes Cognitive Level: HIGH

Group : 1 SRO Imp : 3.9 SRO Exam : Yes Source: NEW

Applicable 10CFR55 Section : 41.5/43.5/45.3/45 .13

Learning Objective: 3-0T -SYS088A , Objective 10: Describe the containment and penetration testing

required and the acceptance criteria.

References: 3-0T -SYS088A , Technical Specification 3.6.1., 1-SI-88-24, rev 7, TI-1207A, Rev. O.

Question :

Given the following plant conditions:

Plant is in Mode 4.

Lower containment air lock is broken and inner door is jammed and will not open .

If conditions were to requ ire an emergency entry into lower containment by opening the sub-h atch, which

ONE of the following is REQUIRED to be contacted prior to openin g the sub-hatch, and what action is

required as a result of the sub-hatch being opened?

A. Shift Manager;

Perform 1-SI-88-24, Containment Divider Barrier Personnel Access Hatches & Equipment Hatches,

within one (1) hour.

B. Shift Manager;

Perform 1-SI-88-2 4, Containment Divider Barrier Personn el Access Hatches & Equipment Hatches,

prior to Mode 3 entry .

C. Work Week Manager;

Perform 1-51-88-24 , Containment Divider Barrier Personnel Access Hatche s & Equipment Hatches ,

within one (1 ) hour .

D. Work Wee k Manag er;

Perform 1-SI-88-24, Containment Divider Barrier Personnel Access Hatches & Equipment Hatches,

prior to Mode 3 entry .

D1STRACTOR ANALYSIS

a. CO RRECT. The sign on the sub-hatch requires the Shift Manager to be notified and the Sl must be

com pleted within one hour , with the plant in Mode 4.

b. Incorrect. While the sign on the sub-hatch requires the Shift Manager to be notified , the SI must also be

completed with the plant is in Mode 4. Plausible because the notificat ion is correct and the candidate

could conclud e the SI is not required until Mode 3.

c. Incorrect. Plaus ible since TI-12 07A, Conta inment Access for Modes 1-4, does ment ion Work Week

Manag er as a cons ultant sour ce only, not as a required notification.

d. Incorrect. Plausible since TI-1207 A, Containmen t Access for Mode s 1-4, does mentio n Work Wee k

Manager as a consultant source only, not as a required notification.

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RO Watts Bar 2008 NRC Ini tial License Exam

WRITTEN QUESTION DA TA SHEET

Qu estion Number: 55

K/A: 103 M .09

Containment

Abil ity to manuall y operate and /or monitor in the con trol room : Containment vacuum system .

Tier : 2 RO Imp: 3.1 RO Exam : Yes Cogn it ive Level : HIGH

Group : 1 SRO Imp : 3.7 SRO Exam: Yes Source: NEW

Applic able 10CF R55 Section : 41.7 /45 .5 to 45 .8

Learni ng Objective : 3-0T-SYS065A. Objective 5: Describe how the EGTS and Annul us Vac uum Sys tems

maintain annulus pressu re .

References : SO I-65 .01, Ann ulus Vacu um System , Rev. 18.

Question :

Whi ch ON E of the following identifies the NORMAL pressure band controlled by the Containment Annulus

Vacuum System and the requi red me thod of contro lling pressu re if 1-M-278 Window 232- 8 , A NN ULUS 6P

LO/DAMPE R SWA POV ER is LIT?

Annulus Pressure Band Method of Controlling Pressure

A. -6.0 to -6.2" WC Dispatch a NAUO to RESET the dampers locally.

8. -6.0 to -6.2" WC Swap the dampers using handswiches on 1-M- 27B.

C. -4.3 to -4.5" WC Dispatch a NAUO to RESET the dampers loca lly.

D. -4.3 to -4.5" WC Swap the dampers using handswiches on 1-M-27B.

DISTRACTOR ANALYSIS

a. CORRECT. The co rrec t pressure band of -6.0 to -6.2 "wc is provided an d the co rrect response to

Wind ow 232-8 is prov ided .

b. Incorrect. Plau sible , since the correct pressure band of -6.0 to -6.2 "wc is provided but incorrect

actio ns to restore alignment are stated .

c. Inco rrect. Plausible since the 4.3 to -4.5 "wc range is associated with the swapover of the dampers

and the co rrect response to Window 232- 8 is provided .

d. Incorrect. Incor rect damper sequence, and the loca l action of the NA UO is de scribed corre ctly .

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RO Watts Bar 2008 NRC Initial License Exam

WRITTEN QUESTION DA TA SHEET

Questio n Number: 56

KIA: 014 A2.02

Rod Position Indication

Ability to (a) predict the impac ts of the following malfun ctions or operations on the RPIS ; and (b) based on

those on those predictions, use procedures to correct , control, or mitigate the consequences of those

malfunctions or operations : Loss of powe r to the RPIS.

Tier : 2 RO Imp: 3.1 RO Exam : Yes Cognitive Level : HIGH

Group: 2 SRO Imp: 3.6 SRO Exam: Yes Source: NEW

Appl icable 10CFR55 Section : 41 .5/43 .5/45 .3/45 .13

Learning Objective: 3-0T-SYS085A, Obje ctive 25: Explain the bases , input, alarms , and oper ator actions

relative to the rod insert ion limits.

Refer ences : 3-0 T-SYS085A, Atta chment 7 Engineering Evaluation of Westinghouse Information

Regardll1g Computer Enhan ced Rod Position Indication (CERPI) Displa ys.

Question :

Given the following plant cond itions :

The unit is at 100% power.

1-M-1B, Window 17-0 , 120 AC VITAL PWR BD 1-1 UV/CKT TRIP , is LIT.

Which ONE of the follow ing describes the impact on the Computer Enhanced Rod Position Indication

(CERPI) System and what is required to ensure that Tech Spec Rod Group Alignm ent Limits are met?

Rod pos ition ind icati on Ac tions Required for Rod Group alignment

A. Available Use the "ALL RODS" function on the ope rating disp lay.

B. NOT Availab le Flux map is required to confirm rod position.

C. Available Flux map is requi red to confi rm rod posit ion .

D. NOT Availab le Use the "ALL RODS" function on the oper ating display.

DISTRACTOR ANALYSIS

a. CORR ECT . The power loss described in the stem results in only the left side CERPI displ ay going

dark. The right side CERPI disp lay remains powered, and the operator can select "All Rods " on the

displa y to determine positions .

b. Incorrect. The right hand display rema ins powe red, with capabil ity to mon itor all rods . No flux map is

required .

c. Incorrect. The power loss described in the stem results in only the left side CERPI display going dark.

The right side CERPI disp lay rema ins powered , and the operator can select "All Rods " on the display

to determine positions. No flux map is required.

d. Incorrect. The right hand display remains powered , with capab ility to moni tor all rods . No flux map is

required .

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RO Watts Bar 2008 NRC Initial License Exam

WRITTEN QUESTION DA TA SHEET

Question Number: 57

KIA: 028 A 1.02

Hydrogen Recom bin er an d Purge Con trol

A bility to pre dict and /or monitor cha nge s in pa rame ter (to prevent exceeding design lim its ) associated with

ope rating the HRPS co ntro ls includ ing: Containment pressure .

Tier: 2 RO Imp:3.4 RO Exam : Yes Cognitive Level: LOW

Group: 2 SRO Imp:3.7 SRO Exam: Yes Source: NEW

Applicable 10CFR55 Section: 41.5 /45.5

Learn ing Objective : 3-0 T-SY S083A , Objective 8: Des cribe the major com ponent s an d op era ting princip le

of the Hydrog en Recombiners.

References: Ref : SOI-83.01 Rev 15, TI- 83.01 Re v 1, and TS 3 .6.7 Basi s.

Question :

Whi ch O NE identifies BOTH of the follo win g for Hydro gen Recombiner ope ration s?

(1) Th e MA XIMUM Hydr og en Recombiner temperature allo we d wh en op erating ?

AND

(2) The HIGHEST of the listed containment hydr ogen co nce ntrations all owed when placi ng the

recombiner in serv ice ?

(1) Maximum Temperature (2) H2 Concentration

A. 6%

B. 4%

C. 6%

D. 4%

DISTRACTOR ANALYSIS

a. Incorrect. Plau sibl e due to 1150 °F is the temperature ab ove wh ich the Hydr ogen Rec ombiners will

rem ove hydr ogen, and greater than 6% is the limit (w ithout taking into ac count instrume nt accuracy)

whe re Hydro gen Recombiners shall NOT be placed in service lAW SOI-83 .01 .

b. Incorrect. Plaus ible due to 1150 °F is the temperatur e above which the Hydro gen Re combi ners wi ll

rem ove hydr ogen . Th e hyd rogen concentration is correc t.

c. Incorr ect: Th e ma ximum temperature value is correct. Plausible du e to great er than 6% is the lim it

(without tak ing into account instrument accura cy ) where Hydrogen Re combiners sha ll NOT be placed in

servic e lA W SOI- 83.01 .

d. CO RREC T. Th e maximum temperature value is correct, and 4% is the co rrec t hydrogen concen tra tion ,

per the gi ven reference .

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RO Watts Bar 2008 NRC Initial License Exam

WRITTEN QUESTION DATA SHEET

Que stion Number: 58

KIA: 033 G2 ,2,36

Spent Fuel Poo l Cooling

Ability to ana lyze the effec t of maintenance acti vities , such as degraded power sources, on the status of

limiting cond itions for operations ,

Tier: 2 RO Imp: 3.1 RO Ex am : Yes Cogn itive Level : HIGH

Group : 2 SRO Imp: 3.2 SRO Exam : Yes Source: NEW

Appl icable 1OCFR55 Section : 4 1.10/43.2/45 ,13

l earnin g Obje ctive : 3-0 T-SYS078A , Object ive 8: Descr ibe the Spen t Fuel Pit pumps including: c. Power

suppl ies ; Object ive 4: Given plan t condition s and par ameters, correctly determine the Conditions for

Ope ration or Technical Requi rements for various co mponents listed in Section 7 of Tech . Specs,

References : SOI-78.01, Spen t Fuel Poo l Cool ing and Cleaning System, Rev. 52 .

Que stion :

Following a refue ling outage, the followi ng conditions exist:

Unit 1 is in Mode 3 preparing for Mode 2 ent ry.

Fuel shuffles are being conducted in the Spen t Fuel Pit.

Spent Fue l Pool Cooling pump A is the only Spent Fuel Poo l Cooling pump in service ,

The 2A-A Shutdo wn Boa rd normal feeder breaker is inadvertently open ed during testing.

The DG starts and reenergizes the shutdown board ,

Which ON E of the following describes the initial effect on the Spent Fuel Pool Coo ling system, and the

required action, if any, per Spen t Fuel Pool Tech Specs?

A. The Spent Fuel Coo ling Pump stri ps from the board , and then sequences bac k on to the shutd own

board .

Spent Fuel Pool Tech Spe cs requires that mo vement of irradiated fuel ass emblies in the fuel storage

poo l be immediately suspended.

B. The Spent Fuel Cooling Pump strips from the board , and then sequences back on to the shutdow n

boa rd,

Spent Fuel Pool Te ch Spe cs does NOT requ ire that movement of irradia ted fuel assemblies in the fuel

storage poo l be suspended,

C, The Spen t Fuel Coo ling Pump str ips from the boa rd, and rema ins off.

Spent Fuel Pool Tech Spe cs requires tha t movem ent of irradiated fuel assemblies in the fuel storag e

pool be immediately suspended .

D. The Spent Fuel Cooling Pump strips from the boa rd, and remains off .

Spent Fuel Pool Tech Specs do es NOT requi re that movement of irradiated fuel assemb lies in the fuel

storage poo l be suspend ed.

DISTRACTOR ANALYSIS

a, Incorrect. While it ;s correct that the pum p strips from the board due to the blackout rela ys , the pump is

not seq uenced back on afte r the diesel generator recovers the boa rd . It is also correct that no action is

requ ired to immediately suspend the movement of radi ation fuel. Plausible because the pump could be

confused with other pumps that do sequence bac k on .

b. Incorrect. While it is correct that the pump strips from the board due to the blackout rela ys, the pump is

no t sequ enced back on after the diesel gene rato r recovers the board and the re is no requ irement to

suspen d the movement of radiation fuel. Plausible becau se the pump could be co nfused with other

pump s that do sequence back on and because the immediate suspension of irradiated fuel movemen t

is required for othe r con ditions associa ted wit h the spent fuel piUcooling syst em,

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RO Watts Bar 2008 NRC Initial License Exam

WRITTEN QUESTION DA TA SHEET

c. Incorrect. While it is correct that the pump strips from the board due to the blackout relays and not

sequenced back on after the diesel generator recovers the board , there is no actio n require d to

immedia tely suspend the movement of radiation fuel. Plausible because the immedi ate suspensio n of

irradiated fuel movement is required for other conditio ns associated with the spent fuel pit/coo ling

system.

d. COR RECT. The SFP pump A normal suppl y is from 480V Shutdown Board 2A2- A. The pump strips

from the board due to the blackout relays and does not sequen ce back on whe n the diesel generator

recovers the board . It is also correct that no action is required to immediately suspend the moveme nt of

radiation fuel.

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RO Watts Bar 2008 NRC Initial License Exam

WRITTEN QUESTION DA TA SHEET

Question Number: 59

KIA: 045 K5.17

Main Turbine Genera tor

Knowle dge of the operational implications of the following concepts as the(y) apply to the MT/B(G) System:

Relationship between mode rator temp erature coefficient and boron concentration in RCS as T/G load

Increases.

Tier: 2 RO Imp: 2.5 RO Exam: Yes Cognitive Level : HIGH

Group : 2 SRO Imp: 2.7 SRO Exam : Yes Source: SQN BANK

Applicable 10CFR55 Section : 41.5/45.7

Learning Objective : 3-0T-G00400 , Objective 6: Explain the average coolant tempe rature (Tavg) program

utilized during power increase or decr ease. Objective 7: Expla in why reactor would "follow" the turbine up

in power dur ing a load increase.

References : Nuclear Parameter and Operation s Package (NuPOP) Cycle 9.

Question:

Which ONE of the follow ing identifie s (a) how main steam header pressu re resp onds as turb ine load is

raised from 25% to 65%, and (b) which method of maintaining Tavg matched with Tref results in the value

for MTC being the MOST negative as turbine load is raised?

A. (a) Main steam header pressure lowers .

(b) Rods are withdra wn to maintain Tavg on program, with Boron concentration held constant.

B. (a) Main steam header pressure rises.

(b) Rods are withdrawn to maint ain Tavg on progr am , with Boron concentration held constant.

C (a) Main stearn header pressure lowers.

(b) Rod positio n is held cons tant, while Boron concen tration is lowered to maintain Tavg on

program .

D. (a) Main steam header pressure rises.

(b) Rod position is held constant, while Boron concentration is lowered to mainta in Tavg on

program .

DISTRACTOR ANALYSIS

a. Incorrect. Main Steam Header pressure lowers as turbine load is raised. Plausible since

candidate could conclude that withdrawing rods makes MTC more negati ve. It actually makes

it less negative . However a competing effect of Tavg rising has a negative effec t on MTC .

b. Incorr ect. Main Steam Header pressure lowers as turb ine load is raised . Plausible since

candidate could conclude that withd rawing rods makes MTC more negative . It actually makes

it less negative. However, a comp eting effect of Tavg rising has a negative effect on MTC .

c. CORRECT. Main Steam Header pressure lowers as turb ine load is raised. Redu ction of

boron concentration results in more negati ve MTC . Tavg rising has a negati ve effect on MTC.

This additiv e negative effect is the most negati ve of all choices given .

d. Incorre ct. Main Steam Head er pressure lower s as turbin e load is raised. Redu ction of boron

concentration results in more negat ive MTC . Tavg rising has a negative effect on MTC . This

additive negati ve effect is the most negati ve of all choices given .

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RO Wat ts Bar 2008 NRC Init ial License Ex am

WRITTEN QUESTION DA TA SHEET

Question Number: 60

KIA : 055 G2.4.3

Condenser Air Remo val

Ability to identify post-accident instrumentation .

Tier : 2 RO Imp : 3.7 RO Exam : Yes Cogn it ive Level : LOW

Group: 2 SRO Imp: 3.9 SRO Exam: Yes Source: NEW

Applicable 10CFR55 Section : 41 .6/45.4

Learning Objective : 3-0 T-SYS090A, Objective 16: Identify where Post Acc ident Monitors are used & read

out.

References : SOI-90 .05, POST-ACCIDENT RAD MONITORS , Rev 12; 1-47W610-90-5 R40 .

Question:

Which ON E of the following identifies monitors associated with Condenser Vacuum Pump disc harge which

are Post Accid ent Monitors (PAM) , in accordance with SOI-90 .05, Post Acci dent Radiation Monitors?

A. Both 1-RM-90-1 19 and 1-RM- 90-404 .

B. Neith er 1-RM-90 -1 19 nor 1-RM-90 -404 .

C. 1-RM-90-119 is a PAM, but 1-RM-90-404 is NOT .

D. 1-RM-90-404 is a PAM , but 1-RM-90-119 is NOT .

DISTRACTOR ANALYSIS

a. Incorrect. SOI-90.05 identifies1-RM-90-40 4 as a Post Accident Rad Monitor, but 1-RM-90-119 is NOT

identified as a Post Accide nt Rad Moni tor. Both being Post Accident Rad Monito rs is plausible because

1-RM-90-11 9 is used during a SGTR event as an indication of the accident.

b. Incorrect. SOI-90 .05 iden tifies1-RM-90-404 as a Post Accident Rad Monitor, but 1-RM-90-119 is NOT

identified as a Post Accident Rad Monitor. Neither being Post Accident Rad Monitors is plausible

beca use neither is listed in the Tech Spec for Accident Monitoring Instrumen tation.

c. Incorrect. SOI-90 .05 identifies1-RM-90 -404 as a Post Accident Rad Monitor, but 1-RM- 90- 119 is NOT

identified as a Post Accide nt Rad Monitor. Plausible beca use the 1-RM-90-119 is used during a SGTR

event as an indicatio n of the accident and the candidate may know one of the mon itoring is a Post

Accide nt Rad Monitor.

d. CO RRECT. 1-RM-90-404 is identified in SOI-90.05 as a Post Accident Rad Monitor, but 1-RM-90-119

is NOT identified as a Post Accident Rad Monitor.

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RO Watts Bar 2008 NRC Initial Licens e Exam

WRITTEN QUESTION DA TA SHEET

Question Number : 61

K/A : 056 K1.03

Condensa te

Knowledge of the physical connections and/or cause-effect relationships between the Cond ensate System

and the following system s: MFW .

Tier : 2 RO Imp: 2.6 RO Exam : Yes Cogn itive Level : LOW

Group : 2 SRO Imp: 2.6 SRO Exam: Yes Source : NEW

Applicable 10CFR55 Section : 41.2 to 41.9/45.7 to 45.8

Learning Objective: 3-0 T-SYS002A Objective 16: List the conditions which will caus e the main feed

pump turbine condense r valves to automatically close .

References : 1-47W6 11-2-1; AOI-16, Rev. 30.

Question :

Which ONE of the follow ing occurs automa tically if the "B" MFP trips due to thrust bearing wea r with the

plant initially at 100% powe r? (Assume no othe r equipmen t failures .)

A. The motor driven AFW Pumps start .

B. The Condensate DI pumps trip if feedwat er flow drop s to <80%.

C. The "B" MFPT condenser condensate inlet and outlet valves go closed .

D. The short cycle valve, 1-FCV-2-35, modulates open to dump excessive condensate flow.

DISTRACTOR ANALYSIS

a. Incorrect. The AFW pumps start on the loss of both MFPs and other cond itions that are not descr ibed

in the stem of the question. Loss of one MFP at the stated power level starts the standb y MFP.

Further plausibility is added due to some Westinghouse plants are design ed this way.

b. Incorrect. Suction pressure <20 psig causes the Conden sate DI pumps to trip.

c. CORR ECT. Since MFW flow is greater than 40 %, the condensa te inlet and outlet valves for the MFP

turbine condensers go closed .

d. Incorrect. Plausible since a differential pressure is deve loped across a flow elem ent (FE-2- 35 ). This

L\ P is converted to a flow signa l which is used to control Short Cycle Valve (FCV -2-35) to ensure a

minim um of 5500 gpm condensate flow.

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WRITTEN QUESTION DA TA SHEET

Question Numb er : 62

K/A : 068 K6.10

Liquid Radwas te

Knowledge of the effect of a loss or malfunction on the following will have on the Liquid Radwaste System :

Radiation monitors.

Tie r : 2 RO lmp: 2.5 RO Exam : Yes Cogn itive Level : HIGH

Group : 2 SRO Imp : 2.9 SRO Exam : Yes So urce: NEW

Applic able 10CFR55 Section : 41.7 /45 .7

Learning Objective : 3-0T-SYS0 90A . Objective 7: Determine Interlocks and/or cau se-effect relatio nships

between the Rad Monitoring Systems (ARM & Process) and the areas they mon itor. Include HVAC systems

and area isolations.

References: 1-47W611-77-2 R5. ARI-1 80-187 Rev 30.

Question :

Which ONE of the following identifies a condition that causes an instrument malfunction alarm on 0-RM-90-

122, WDS Liquid releas e radia tion monitor, and the effect the instrument malfu nction alar m has on valve 0-

RCV-77 -43. CT BLDN LN RAD RELEASE CNTL?

Cause of the alarm Effect on O-RCV-77-43

A. Loss of signal from detector Auto closes 0-RCV -77-43 if the valve was open .

B. Loss of signal from detector Prevents 0-RCV-77-43 from opening if the valve's

local control handswitch was placed to OPEN.

C Loss of flow through the monitor Auto closes 0-RCV-77-43 if the valve was open.

D. Loss of flow through the monitor Prevents 0-RCV-77-43 from opening if the valve's

local control handswitch was placed to OPEN.

DISTRACTOR ANALYSIS

a. CORR ECT. As identified in ARI-181-C , a loss of signal from the detector would cause an Instrument

Malfuncti on alarm and if 0-RCV -77-43 was open it would be automa tically closed as iden tified on 1-

47W611-77-2.

b. Incorrect. As identified in ARI-181-C. a loss of signal from the detector would cause an Instrument

Malfunc tion alarm and if 0-RCV -77-4 3 handswitch was placed to open the valve would ope n but would

reclose when the switch was released as identified on 1-47W6 11-77-2.

c. Incorrect. A loss of flow through the detector is not identified in ARI-181-C as a condition that would

cause an Instrumen t Malf unction alarm but if 0-RCV-77-43 handswi tch was open it would be

automa tically closed due to the low flow cond ition as identified on 1-4 7W61 1-77-2.

d. Incorrect. A loss of flow through the detector is not identified in AR I-181-C as a condi tion that would

cause an Instrument Malfunction alarm and if 0-RCV-77-43 handswitch was placed to open the valve

would ope n but would reclose when the switch was released as identified on 1-47W61 1-77-2.

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WRITTEN QUESTION DA TA SHEET

Question Number: 63

KIA: 075 K1 .02

Circulating Water

Knowledg e of the physical connect ions andl or cause -effect relationships between the circulat ing water

system and the following systems: Liquid radwast e discharge.

Tie r : 2 RO Imp : 2.9 RO Exam : Yes Cog nit ive Le vel : LOW

Group : 2 SRO Imp: 3.1 SRO Ex am : Yes Source: WBN BANK

Appl icable 10CFR55 Section : 41.2 to 4 1.9/45 .7 to 45.8

Learning Objective : 3-0T-SYS0077A, Object ive 20: Correctl y locate the local and MCR controls for the

Liquid Radw aste System ; 3-0T-SYS027A, Object ive 16: Explain the minimu m cooling tower blowdown flow

rate interlock with Radwaste, S /G Slowdown , and Cond Demin discharge valves.

References : SOI-77 .01, Liquid Waste Disposal, Rev 57.

Qu est ion :

A release of the Monitor Tank is in progr ess through the Liquid Radwaste System . Which ONE of the

following cond itions directly results in automatic closure of 0-FCV -77-43, Liquid Radwa ste Release Flow

Control Valve?

A. High radiation signal on 0-RM -90-225, Condensate Demineralizer Release Liquid Radiation Monitor.

S. River flow drop s to less than 3500 cfs after a 1 minute time delay.

C. Cooling Tower Slowdown flow drops below 25,000 gpm .

D. SG Slowdown flow exceeds 150 gpm .

DISTRACTOR ANALYSIS

a. Incorrect. Plausible. since the candidate may believe that this high radiation condition causes

termination of a release . If aligned for a release from the Condensate Demineralizer , this radiat ion

Monitor would in fact termina te the release.

b. Incorrect. Plausible, since river flow dropp ing to less than 3500 cfs after a 1 minute time delay closes

the diffuser valves , causing the release to be discha rged to the 35 acre pond, but will not close.

c. CORRECT. Per 501-77 .01, this condit ion causes automatic closur e of 0-FC V-77-43.

d. Incorrect. Plausible, since SG Slowdown flow exceeding 150 gpm extends the time by 1.5 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

for a release due to increased backp ressure on the release line.

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WRITTEN QUESTION DATA SHEET

Question Number : 64

KIA: 079 A4.01

Station Air

Ability to manually ope rate and /or monitor in the control room : Cross- tie valves with lAS .

Tie r: 2 RO Imp : 2.7 RO Exam : Yes Cogn itive Level : HIGH

Group: 2 SRO Imp: 27 SRO Exam: Yes Source: WBN BANK

Appl icable 10CFR55 Sectio n : 41.7 / 45.5 to 45.8

Learning Objective: 3-0T-SYS0 32A , Objec tive 2: Describe Auto Actions for Loss of Control Air per AOI-

10.

References : ARI-36-4 2, Heaters, Turb Seal, & Air, Rev 16.

Qu estion :

Given the follow ing plant conditions :

Unit 1 is operati ng at power when control air press ure starts to drop .

Annunciator 42-F , SERV ICE AIR PCV-33 -4 CLOSED , alarms .

The CRO responds in accordance with the Annunciator Respon se Instruction (ARI).

Which ONE of the following ident ifies the decreasing Control Air system pressu re that cause s this alarm to

occur and whether the Auxiliary Air comp ressors wou ld have started if the air pressu re dropp ed low enough

to cause the alarm, but then recovered without dropping any lower?

Pressure to Caus e Al arm Aux Air Compresso rs

A. 83 psig Will have started

B 83 psig Will NOT have started

C. 80 psig Will have started

D. 80 psig Will NOT have started

DISTRACTOR ANALYSIS

a. Incorrect. The isolation pressure is 80 psig not 83 psig. Plausible because 83 psig is the pressure

whe re the aux air compressors start, which makes the second part of the distr actor correc t. This

pressure setpoint could be misapplied to the station service air isolation pressu re.

b. Incorrect. The isolation pressure is 80 psig not 83 psig. Plausible beca use the 83 psig setpoint could

be misapp lied to the station service air isolation pressure and for the aux air comp ressor start, the

candidate recalls the pressu re at which the aux air compresso rs load (79.5 psig) and applies this

pressure as the starting press ure .

c. CORR ECT . The serv ice air isolates at 80 psig and the aux air compressors would have started at 83

psig .

d. Incorrect. The isolation press ure is correct , but the aux air comp ressors would have started. Plausible

if the candidate recalls the pressure at which the aux air compressors load which is 79.5 psig and

applies this pressu re as the starting pressure .

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WRITTEN QUESTION DATA SHEET

Qu est ion Number : 65

KIA: 086 K3.01

Fire Protection

Knowledge of the effect that a loss or malfunction of the Fire Protection System will have on the following :

Shutdown capabil ity with redundant equ ipment.

Tier : 2 RO Imp : 2.7 RO Exam : Yes Cognitive Level: High

Gro up : 2 SRO Imp : 3.2 SRO Exa m : Yes Sou rce : Ba nk , mod ified

Applicable 10CFR55 Section : 41 .7 /45 .6

Learning Objective : 3-0T-AO I3000 Objecti ve 10: State the two primar y limiting safety conditions which

must be maintained following a postulated Appendi x R fire as specified in AOI-30.2 .

Ref erenc es: AOI- 30.2, Safe Shutdown, rev 27.

Que st ion:

Given the following plant cond itions:

Unit 1 is currently at 100%.

A fire occur s in the Cable Spreading Room .

The crew was unable to start the HPFP pumps.

The incident Commander also reports that due to multiple fire damper failu res the fire is spreading

quick ly.

The crew has entered AOI- 30.2, Fire Safe Shutdown .

In accordance with AOI -30.2, which ONE of the follo wing failures results in a loss of a Cont rol Function

required to place the Plant in Hot Standb y?

REFERENCE PROVIDED

A. Motor Driven AFW Pumps will not start .

B. RCS Therm al Barrier Booster Pumps trip.

C. One Main Steam Isolation Valve fails to close.

D. Letdown Isolation Valve 1-FCV- 62-69 fails closed .

DISTRACTOR ANALYSIS

a. Incorrect. The TO AFW Pump is required for the Safe Shutdown . Plausible since the candidate may

incorrectly ass ume that ALL AFW pumps are required .

b. Incorrect. The trip of the thermal barrier booster pumps does not impact the Safe Shutdown capabilities

of the plant. Plausible if the candidate assumes that forced circulation is a concern .

c. Incorrect. Per the Safe Shutdown Logic Diagram, the failure of the MSIV does not impact the ability to

place the plant in Hot Standby. Plausible. if the candid ate misapp lied the diagram .

d. COR RECT . Letdown is required by the Safe Shutdown Logic Diagram to place the plant in Hot

Standby.

NOTE: This question requires the cand idate to use the following refere nce:

AOI -30.2 Section 4.5, Safe Shutd own Logic Diagram

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WRITTEN QUESTION DA TA SHEET

Qu estion Number: 66

KIA: G2 .1.5

Ability to use pro cedures related to shift staffing , such as minimum crew complement, ove rtime limitations,

etc .

Tier: 3 RO Imp : 2.9 RO Ex am : Yes Cognitive Level : HIGH

Group : SRO Imp: 3.9 SRO Exam : Yes Sou rce : Bank Mod

Applicable 10CFR55 Section : 41 .10 /43 .5/45 .12

Learn ing Objective: 3-0T-SPP1000 , Objective 6: Describe shift staffing requirements.

References : OPDP-1 CONDUCT OF OPERATIONS, Sec tion 3.1.3; Tech Spec 5.2 .2 Shift Staffing .

Question:

For Mode 1 ope ration . wh ich ONE of the followin g describes the MINIMUM number of Licensed Operator

positions to man a shift , AN D the MAX IMUM time requirement to restore if the minim um shift manning for

Lice nsed Ope rators is not me t per OPDP-1, Conduct of Opera tions and Tech Spec 5.2.2, Unit Staf f?

Minimum Shift Manning Time Requirement

A. 2 ROs , 2 SROs Restore within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

B. 2 ROs , 1 SRO Restore within 1 hou r

C 2 ROs, 1 SRO Restore wit hin 2 hou rs

D. 2 ROs , 2 SROs Restore within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

DISTRACTOR ANALYSIS

a. CO RRECT. Per OPDP-1 and TS 5.2 .2, 2 ROs are requi red with a 2 Hour time limit per TS 5 .2.2.

b Inco rrect. 2 ROs req uired is correct. 1 hou r time limit is incorrect. Plausible since TS doe s have one

hour time requirem ents and if minimum shi ft man ning is not met student may concl ude this is impo rtant

enough for a 1 hou r time requ irement.

c. Incorrect. 1 RO is incorrect. Plausible due to in modes 5 and 6 the minimu m req uiremen t is only 1 RO.

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months time requirement is correct.

d. Incorrect. 1 RO is inco rrec t. Plausible due to in mode s 5 and 6 the minim um req uirement is only 1 RO.

Second part is also inco rrect. Plausib le due to TS does have one hour time requi rements and if

minimum shift manning is not met student may conclude this is important enough for a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months time

requ irement.

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WRITTEN QUESTION DATA SHEET

Question Number: 67

KIA: G2.1.28

Know ledge of the purpose and function of major system components and controls.

Tier : 3 RO Imp: 4.1 RO Exam: Yes Cognitive Level: LOW

Group: SRO Imp: 4.1 SRO Exam : Yes Source : WBN Bank

Applicable 10CFR55 Section: 41.7

Learning Objective: 3-0 T-SYS00 1B, Obje ctive 9: Identify which controller is in servi ce when Tavg is

selected with the unit at powe r.

References : 1-47W611-1-2.

Question :

With the Steam Dump Mode switc h in the Tavg mode, what determin es whether the load rejection controller

or the Rx trip controller will be in service?

A. Loss of Load (C-7) Interlo ck.

B. LO-LO Tavg Interlock (550°F).

C. "A" Train Reactor Trip brea ker position .

D. "B" Train Reactor Trip breaker position .

DISTRACTOR ANALYSIS

a. Incorrect. Plausible . since the Loss of Load interlock will arm the steam dump s, but doe s not

determine which controller output will position the dump valves.

b. Incorre ct. Plausi ble , since the LO-LO Tavg Interlock will close all of the steam dumps if temperature

drops to below 550 °F

c. Incorrect. The A train Reactor Trip breaker is used to ARM the steam dumps on a react or trip.

d. CORRECT. The B train reactor trip breaker is used to select the controller in servic e. With NO reactor

trip, the Load Rejection Controller is select ed. Once the B train reactor trip break er open s, the Reac tor

Trip controller is selected.

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WRITTEN QUESTION DA TA SHEET

Question Number: 68

KIA: G2.1.36

Knowledge of procedur es and limitation s involved in core alterations.

Tier: 3 RO Imp : 3.0 RO Exam: Yes Cognitive Level : LOW

Group: SRO Imp : 4.1 SRO Exam: Yes Source: SQN BANK

SIG MOD

Applicable 10CFR55 Section: 41.10/43.7

Learning Objective: 3-0T-G00700, Objective 5: Describe the major steps that the operator must take

whe n unloadin g fuel per this instructio n.

References: FHI-7 , Rev. 31.

Question:

Given the following plant conditions:

Unit is in Mod e 6.

15 fuel assem blies have been reloaded after a com plete core off-load .

Source Range N-131 indic ates 10 cps and is selected for audible count rate indication.

Source Range N-132 indicates 5 cps .

In acco rdance with FHI-7 , "Fuel Handl ing and Movement", which ONE of the following unanticipated

changes in count rate requir es suspension of core alterations?

A. N131 indicates 25 cps and N132 indicates 8 cps.

B. N131 indicates 15 cps and N132 indicates 20 cps .

C. N 131 indica tes 40 cps and N132 indicates 8 cps .

D. N131 indicates 20 cps and N132 indicates 15 cps .

DISTRACTOR ANALYSIS

a. Incorrect. Plausible since N 131 has increased by greater than a factor of 2, but N132 has not changed

by a factor of 2.

b. Incorrect. Plausible since N131 has not increased by greater than a factor of 2, but N132 has changed

by a factor of 4 but has not exceeded the factor of 5 which would require suspens ion of core

alterations .

c. Incorrect. Plausible, since N131 has increased by a factor of 4, but has not exce eded the factor of 5

whic h would require suspension of core alterations.

d. CORRECT Both sourc e range chann els have doub led, which would requi re suspension of core

alterations.

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WRITTEN QUESTION DA TA SHEET

Questio n Number: 69

KIA : G2 .2.12

Know ledg e of surveillance pro cedu res .

Tier: 3 RO Imp : 3.7 RO Exam : Yes Cognitive Level : LOW

Group : SRO Im p : 4.1 SRO Exam: Yes Source : WBN Ban k

App licable 10CFR55 Sec tio n : 41 .10 /45 .13

Learn ing Objective : 3-0T-S PP0802, Obj ective 2: Explain the diffe rence between the surveillance due

date and the WBN extens ion date for both Tech Spec and Non Tech Spec surveillances.

References: SPP-8.2, Rev. 3.

Question :

A Power Range channel has failed requiring the crew to impleme nt AO I-4, NIS Malfu nc tions . The US has

ente red the appropriate Technical Specifications and states that a flux map will be requ ired per Surve illanc e

Requi rement 3.2.4.2.

SR 3.2.4.2 directs the operators to verify OP TR is withi n limits using moveable incore detectors onc e within

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months the reafter.

What is the maxim um time for the initia l performance of the flux map and the ma ximum time for subsequent

perfo rmances ?

Initial Performance Subsequent Performances

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

B. 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

c. 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s 15 hour1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months s

D. 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months 15 hou rs

DISTRACTOR ANALYSIS

a. Incorrect. Plausible sinc e the first performance must be acc omplished wi thin the specified time with

NO extension . An extension is allowed for subsequent performances.

b. Inco rrect. Plausible since an exten sion is allowed for SUbsequent performances, but the first

performa nce must be acco mplis hed within the spec ified time wit h NO extension .

c. CORR ECT . There is no extension allowed for the first perfo rmance, and an extension of 25% of the

allow ed time may be granted for subsequent perf ormances .

d. Incorrect. Plausibl e since an exten sion is allowed for subsequent perfo rmances, but the first

performance must be accomplished within the spec ified time with NO extension .

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WRITTEN QUESTION DA TA SHEET

Question Number: 70

K/A : G2.2.44

Ability to interpret control room indications to verify the status and ope ration of a system . and unde rstand

how ope rato r actions and directives affect plant and system conditions.

Tier: 3 RO Imp: 4.2 RO Exam: Yes Cognitive Level: HIGH

Group: SRO Imp: 4.4 SRO Exam: Yes Source : NEW

Applicable 10CFR55 Section : 41.5/43.5/45.12

Learning Objective : 3-0T-SYS003B, Objec tive 23: Using plant drawi ngs . deter mine the effect of a loss of

instrument air/control power on the follow ing valves/components: a. MDAFWP regul ating valve (main and

bypas s). b TDAFWP regulating valve, c. AFW pumps .

References : AOI- 10, LOSS OF CONTROL AIR . Rev. 38; SOI-3.02, Auxiliary Feedwater System . Section

8.5.1.

Question:

Given the following conditions:

Unit 1 shutdown is on progr ess.

Reactor power is 14% and decreasing .

Interm ediate Range NI-36 fails HIGH .

Which ONE of the following ident ifies how the failure of the NI will effect the reactor trip system and the

effect the failure will have on the Source Range Nl s?

Reactor Tr ip System Effect on Source Range Nls

A. Reactor trip will occu r at the time of failure. Source Range Nls will have to be

MAN UALLY reinstated.

B, Reactor trip will occur at the time of failure, Source Range Nls will AUTO MATICALLY

reinstate.

C. Reactor trip will occu r if powe r reduction is Source Range Nls will have to be

continued . MANUALLY reinstat ed.

D, Reactor trip will occ ur if power reduc tion is Source Range Nls will AUT OMATIC ALL Y

continued, reinstate.

DISTRACTOR ANALYSIS

a, Incorrect. Plaus ible, since power is greater than 10%. the P-10 blocks are active. This would prevent

an immediate reactor trip at the time of the failure, The SR Nl s WOULD have to be manually

reinstated since the P-6 perm issive requires both IR channels to be below the setpoint to automatically

reinstat e.

b. Incorrect. Plausibl e, since power is greater than 10%. the P-10 block s are active. This would prevent

an immedia te reactor trip at the time of the failure, The SR Nl s WOULD have to be manually

reinstated since the P-6 permissi ve requires both IR channe ls to be below the setpoint to automatically

reinstate,

c. CORRE CT . Power is great er than 10%, initiall y and the P-10 blocks are active . This would prevent an

immediate reactor trip at the time of the failure . If the power reduction cont inues to a point less than

10% on 3/4 PR channels, P-10 would be unblocke d and the 1/2 IR trip would occu r. The SR Nls

WOULD have to be manuall y reinst ated since the P-6 permi ssive requires both IR channels to be

below the setpoint to automatically reinstate.

d. Incorrect. Plaus ible since power is greater than 10% initially. and the P-10 blocks are active, This

would prevent an immediat e reactor trip at the time of the failure, If the powe r reduc tion con tinues to a

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WRITTEN QUESTION DATA SHEET

point less than 10% on 3/4 PR channels . P-10 woul d be unblocked and the 1/2 IR trip woul d occur.

The SR Nls would have to be manually reins tated since the P-6 permiss ive requires both IR channels

to be below the setpoi nt to automa tically reinsta te.

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WRITTEN QUESTION DA TA SHEET

Questio n Number : 71

KIA: G2.3.7

Ability to comply with radiation work permit requirements during normal or abnorma l conditions .

Ti er : 3 RO Imp : 3.5 RO Ex am: Yes Cognitive Level: LOW

Group: SRO Imp: 3.6 SRO Exam : Yes Source: NEW

Applicable 10CFR55 Section : 4 1.12 /45 .10

Learn ing Obj ect ive: 3-0 T-RAD0003, Objecti ve 8: Identify the responsibil ities of the followin g concerning

the ALARA program : a. Radi ation Protection Manager/Radiation Safet y Officer , b. TVA NPG Organization ,

c. Employee.

References : RCI-153 , Radiation Work Permits (RWPs) Rev 0000 ; RCI-100 , Control of Radiological Work,

Rev 32.

Questio n :

An individual enters a Radiological Controlled Area (RCA) covered by a Genera l RWP to perform equipment

inspec tions .

Which ONE of the following identi fies an area within the RCA where a Job Specific RWP is requir ed before

entry is allowed , in accordance with RCI-153. Radiation Work Permits?

A. Area where whole bod y dose rates exceeds 100 mrem /hr .

B. Area posted as Hot Part icle Area.

2

C. Area with general contamination levels greater than 200 dpm/100cm .

D. Area where total expec ted dose is greater than 5 mrem .

DISTRACTOR ANALYSIS

a. Incorrect. Required if >1,000mrem/hr.

b. CORR ECT. Required for area posted as Hot Particle Area .

c. Incorrect. Required if> 50,000 dpm/ tuucrn".

d Incorrect. Required if total expected dose exceeds 50 mrem .

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WRITTEN QUESTION DA TA SHEET

Question Number: 72

KIA: G2.3.14

Knowledge of radiation or contamination hazards that may arise during normal , abno rmal, or emergency

conditions or activities .

Tier: 3 RO Imp : 3.4 RO Exam : Yes Cognitive Level : LOW

Group: SRO Imp: 3.8 SRO Exam : Yes Source: NEW

App licable 10CFR55 Section: 4 1.12 /4 3.4 / 45.10

Learn ing Objective: 3-0 T-SYS031A, Objective 3: Describe the ventilation flow path provided by the

control building ventilation system during normal and emergency ope ration .

Referen ces : ARI-1 80-187 rev 30, 1-47W866-4 R39.

Question :

Given the following plant conditions :

Following an accident , both Trains of Control Room Isolation have been initiated .

Several Auxiliary Building Area Radiation Monito rs rise to the alarm setpoint.

Which ONE of the following MCR air intake radiation moni tors will detect and alert the crew of a radiation

hazard entering the control room and what actions the ARI will direct the crew to perform ?

A. 0-RM-90-1 25, Stop MCR Emergency Pressurization Fans.

B. 0-RM-90-125 , Align Emergenc y Pressurization Fan suction to alternate source.

C. 0-RM-90-205 , Stop MCR Emergency Pressur ization Fans.

D. 0-RM-90-205 , Align Emergency Pressurization Fan suction to alternate source .

DISTRACTOR ANALYSIS

a. Incorrect. 0-RM -90-125 is in an isolated flow path due to the CRt. Plausible because the rad monitor is

the incorr ect monitor. It would be in the flow path after the suppl y was realigned and stopp ing the fans

would be a way of stopping the intake of radiat ion .

b. Incorrect. 0-RM-90-125 is in an isolated flow path due to the CR t. Plausib le because the rad monitor is

the incorrect monitor. It would be in the flow path afte r the supp ly was realigned and Annunciator 187-A

directs the action to align the Emergency Pressurization Fan suction to alternate source.

c. Incorrect. 0-RM-90-205 will cause annunciator 187-A to alarm and the ARI directs the action to ALIGN

Emerge ncy Pressurization Fan suction to alterna te source, not to stop the Emergency Pressurization

Fan. Plausible because the rad monitor is the correct monitor and stopping the fans would be a way of

stopping the intake of radiation.

d. CORR ECT. O-RM-90 -205 will cause annun ciator 187-A to alarm. 0-RM-90-125 is in an isolated flow

path due to the CRI. Annunciator 187-A directs the action to align the Emergency Pressurization Fan

suction to alternate source.

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WRITTEN QUESTION DA TA SHEET

Question Number: 73

KIA: G2.4 .13

Knowledge of crew roles and responsibilities during EOP usage.

Tier: 3 RO Imp: 4.0 RO Exam : Yes Cognitive Level : LOW

Group: SRO Imp : 4.6 SRO Exam : Yes Source : SQN BANK

Applicable 10CFR55 Section : 4 1.10/4 5 .12

Learn ing Objective: 3-0T-TI1204, Objective 13: Appl y the rules of usage which relate to performing steps

of an EOP in a specified sequ ence to determ ine when steps may/must be performed.

References : TI-12 .04, Rev. 7, Page 31 thru 34.

Question:

The Operator-at-the-Controls (OAC) is responding to an accident. He recog nizes that he mus t take action s

which are NOT contained in the emergency operating procedure in effect and are NOT cove red by prudent

operator actions. Which ONE of the following describe s the proper action to be taken?

A. The OAC shall take no action until a procedure is developed or revised.

B. The OAC shall obtain approval from a licensed SRO prior to taking action.

C. The OAC should obtain concurr ence from another licensed RO prior to taking action.

D. The OAC should immediately take appropriate actions necessary and inform the SRO when time

permit s.

DISTRACTOR ANALYSIS

a. Incorrect. Plausible, since under normal circumstances , the operator would stop a task in progress and

wait until a new proced ure was written before taking any additional actions . The operator must stop

actions long enough to get appro val from the SRO . The SRO would have to address the situation using

1OCFR50.54(x) criteria.

b. CO RRECT. The operator must stop actions long enough to get appro val from the SRO . The SRO

would have to address the situation using 10CFR50.54(x) criteria.

c. Incorrect. Plausible if the candidate confuses the actions required by a PEER CHECK with actions in b.

d. Incorrect. Plausible, since Prudent Operator Actions do allow the RO to take manu al compensatory

actions which are with in the guidel ines of an existin g procedu re.

79 of 81

RO Watts Bar 2008 NRC Initial License Exam

WRITTEN QUESTION DATA SHEET

Question Number: 74

KIA: G2.4 .37

Knowledge of the lines of authority during implementatio n of the emer gency plan.

Tier: 3 RO Imp: 3.0 RO Exam : Yes Cognitive Level: LOW

Group : SRO Imp: 4.1 SRO Exam: Yes Source: WBN Bank

Applicable 10CFR55 Section : 41 .10 /45 .13

Learning Objective: 3-0T-RAD0003, Objec tive 6: List the extreme eme rgency expos ure guide lines.

References : EPIP1 5, Emergency Exposure Guide lines , Rev 13.

Question:

Given the following:

A Site Area Emergency (SAE) has been declared on Unit 1.

Two hours after the SAE declaration , an indiv idual is to be authorized to receive an Emergency

Exposure radiation dose above the TVA whol e body dose limit during the mitig ation of the

emergency situation .

In acco rdance with EPIP-15 . Emergency Exposure Guidelines. whose approva l is require d for the individual

to receive the dose?

A. TSC Radcon Manager.

B. Onshift Shift Manager.

C. Site Emergency Director .

D. Site Vice President.

DISTRACTOR ANALYSIS

a. Incorrect. Per EPIP-15, the Radiat ion Protection group is responsible for completing necessary

paperwork and obtaining SED's approval

b. Incorrect. Plausible, if the Shift Manager was in the role of the SED. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months time frame stated in

the stem allows for the TSC to be manned and therefore the SED duties wou ld have been assumed

from the Shift Manager.

c. CORRECT. The SED is the ONL Y individual responsible for authori zing emergency dose limits.

d. Incorrect. Plausible. since the Site VP may be acting as the SED . Howe ver the Site VP by title does

not have responsibility for authorizing emergency dose limits.

80 of 81

RO Watts Bar 2008 NRC Initial License Exam

WRITTEN QUESTION DA TA SHEET

Question Number: 75

KIA : G2 .4.42

Knowledge of eme rgenc y response facilities.

Tier: 3 RO Imp: 2.6 RO Exam: Yes Cognitive Level: LOW

Group : SRO Imp : 3.8 SRO Exam : Yes Source : NEW

Applicable 10CFR55 Section: 41.10 /45 .11

Learning Objective: 3-0 T-PDC-048C, Objective 20: Use the Satellite Phone to make calls during

emergencies .

References: SOI- 100 .01 , rev 22.

Question:

Which ON E of the follow ing identifies where a Portable Satellite Telephone , available for use during an

emergency , is located ?

A. Main Control Room

B. Technical Support Center

C. Joint Informat ion Center

D. Operations Support Center

DISTRACTOR ANALYSIS

a. Incorrect. Plausible because the Main Control Room does have Satellite phon e capabilities, however

the capability is to selected phones via the Stationary Satellite Telep hone (SST) system .

b. CORRECT. The Portable Satellite Tele phone (PST ) is located in a cab inet in the Techn ical Support

Center.

c. Incorrect. Plaus ible becau se the other 3 locations do have selected phones that can be connected to

the Stationary Satell ite Telephone (SST) system, and the JIC does not that capabili ty.

d. Incorrect. Plausible because the Operations support Center does have Satellite phon e capa bilities,

however, the capability is to selected phones via the Stationary Satell ite Telephon e (SST) system.

81 of 81

Watts Bar Nuclear Plant

NRC Initial License Written Examination - 2008

Master Examination

Please note: The following 29 pages are the Master Examination copy for

the SRO portion of the examination , including the answer key and

distractor analysis data.

SRO Walts Ba r 2008 NRC In itial Lic ense Exam

WRITTEN QUESTION DA TA SHEET

Qu estion Number : Y 7/;;;

KIA: 00000 7 EA2.05

Reactor Trip - Stabilizat ion - Recovery

Ability to det ermin e or interpret the following as they apply to a reactor trip: Reactor trip first-out indication.

Tier: 1 RO Imp : RO Exam : Cognitive Level : High

Group : 1 SRO Imp : 3.9 SRO Ex am : Yes Source : New

Applicable 10CFR55 Sectio n : 41 .7 /45 .5 /45.6

Learn ing Objective : 3-0 T-SYS099A, Objecti ve 13: Describe the causes of "General Warning " on SSPS.

Objecti ve 14: Identify where "General Warning" indication s can be found.

Referen ces : SPP- 3.5, Rev. 19: 1-SI-99-10B , Rev. 42 .

Qu estio n :

Given the following conditions:

Unit 1 is at 100% power. Solid State Protection System (SS PS) Trai n 'B' Actuation Log ic testing is being

perform ed using 1-SI-99-1 OB with :

Train 'B' SSP S Mode Selector switch in the 'TE ST' positio n.

Train 'B' SS PS Input Error Inhibit switch in the "INHIBI T" position.

A unit trip occurs du e to the loss of one of the two 48v DC power suppl ies on TRAIN 'A' SSP S. The

followlllg "First Out" annunci ators are lit:

1-XA-55-4C , Turbin e Tr ip First 1-XA-55-4D Reactor Trip First Out

Window 73C - "RX TRIP BKRS RTA & BYA OPEN" Window 76B "TUR BIN E TRIP "

Window 74C - "RX TRIP BKRS RTB & BYB OPEN"

Window 74B - "M FPT A&B TRIPPED"

Which ONE of the following identifies both the sequence of events of the unit trip , and the time allowed to

make the requ ired NRC 50.72 notification ?

Sequence of Events NRC Notification Required With in

A. Turbine trip caused the Reactor trip . Four Hours

B Turbine trip caused the Reactor trip. Eight Hour s

C. Reactor trip caused the Turbine trip . Four Hours

D. Reactor trip caused the Turbine trip . Eight Hours

DISTRACTOR ANALYSIS

a. Incorrect. The cond itions stated in the stem result in a general warning on both train s of the SSPS

which causes the reacto r trip and bypa ss breake rs to open (but no reactor first out annunc iator will be

lit). Thus the turbine trips as a result of the reactor trip . The four hour notification to NRC for a reactor

trip is correct. Plau sible due to the notification time being correct and no othe r reactor trip first out

annunc iator will be lit except the turbine first out annun ciator .

b. Incorrect. The cond itions stated in the stem result in a general warnin g on both trains of the SSPS

which causes the reactor trip and bypass breakers to open (but no reac tor first out annunciator will be

lit), thus the turbine trips as a result of the reactor tripping. The notification to NRC for a reacto r trip is

a fou r hour notification , not an eight hour notification. Plausible due to the notification time being

correct and no other reactor trip firs t out annunciator lit except the turbine first out annu nciator; and that

the notification tim e could be misappl ied since the ESF actuation required time is 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and an ESF

actu ation of AFW does occur on a reac tor trip.

c. CO RRECT. The conditions stated in the stem result in a general warn ing on both trains of the SSPS

which causes the reactor trip and bypass breakers to open (but no reactor first out annun ciator will be

lit), thus the reactor trip cau sing the turbine trip is correct. The four hour notification to NRC for a

reac tor trip is correc t.

1 of 29

SRO Watts Bar 2008 NRC Initia l License Exam

WRITTEN QUESTION DA TA SHEET

d. Incorrect. The conditions stated in the ste m result in a ge neral warning on bo th trains of the SSPS

which causes the reactor trip and bypass breake rs to ope n (but no reac tor first out annunciator will be

lit), thus the reactor trip caus ing the tur bine trip is correct. The notification to NRC for a reactor trip is a

four hour notification , not an eight hour notificatio n. Plausible beca use the notificat ion tim e could be

misapplied since an ESF actuation requ ires an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months notific ation and an ESF actuatio n of A FW does

occur on the reactor trip .

2 of 29

SRO Watts Bar 2008 NRC Initial License Exam

WRITTEN QUESTION DA TA SHEET

Qu estion Number : ,z"77

KIA : 000022 G2.1.7

Loss of Reacto r Coo lant Makeup

Ability to evaluate plant perform ance and mak e operat ional judgments based on ope rating characteristics,

reactor beha vior, and instrume nt interpretation.

Tier : 1 RO Imp: RO Exam : Cogn it ive Level : High

Group : 1 SRO Imp : 4.7 SRO Exam : Yes Source: NEW

Applicable 10CFR55 Section : 41.5, 43.5

Learn ing Objective: 3-0 T-SYS062A: Explain the automa tic actuation logic and interlocks assoc iated with

the VCT outlet valves, FCV-62-132 and 133 and the CCP suction valves from the RWST , FCV-62-135 and

136.

Referenc es: SOI-62.02, Rev. 47; ARI-109-115, rev. 16, SPP-10A , 5 0, rev. 5.

Qu estion :

Given the following plant conditions :

Core burnup is 1200 MWD/MTU.

Indicated reactor power is stable at 100%.

VCT low level alarm ann unciates .

Auto makeup has failed.

Actual VCT level had lowered to 5% before the crew comp leted the appropriate corrective action.

Reactor power has stab ilized at approximately 97% powe r.

If this even t had occurred with core burnup at 16600 MWD/M TU, the magni tude of the change in reactor

power would be (1) , and the Significan ce Level of the Reac tivity Management Event

classificatio n would be reco rded on the PER (Problem Evaluation Report) by (2) _

(1) (2)

A. less Reactor Engineering.

B greater Reactor Engineering.

C. less the Management Review Comm ittee.

D. greater the Management Review Comm ittee .

DISTRACTOR ANALYSIS

a. Incorrect. Plausible if candida te confuses the change in boron worth with time in core life to be less,

instead of more. In this case, candidate incorrectly conc ludes that the power change response is at a

lower magnitude. Correctly recogni zes that Reactor Engineering records the significance level of the

event on the PER.

b. CORRECT. Since boron worth is greater at EOL, injection of RWST inventor y results in a highe r

magnitude of change in reactor power . Per the appropriate references, Reactor Engineering records

the significance level of the reactivity event on the PER.

c. Incorrect. Plausible if candidate confuses the change in boron worth with time in core life to be less ,

instead of more. In this case, candidate incorrectly concludes that the power change response is at a

lowe r magnitude. Further plausible, since Management Review Committee (MRC) is involved with plant

PERs, but assig ning the reactivity significan ce level is not a function of the MRC.

d. Incorrect. Plausible, since the cand idate correctly recognizes EOL conditions, and applies the

know ledge that boron worth is higher. Further plausible, since Manag emen t Review Committee (MRC)

is involved with plant PERs, but assigning the reactivity signif icance level is not a function of the MRC.

3 of 29

SRO Watts Ba r 2008 NRC Initial License Exam

WRITTEN QUESTION DA TA SHEET

Qu estion Number : .2 7~,

K/A : 0000 25 AA2.01

Loss of RHR System

Ab ility to de termine and interpret the following as they app ly to the Loss of Residual Heat Removal System:

Proper amperage of runn ing LP ll decay heat removal/RHR pum p(s) .

Tier: RO Imp: RO Exam: Cognitive Level : High

Group: 1 SRO Imp: 2.9 SRO Exam: Yes Source: NEW

Applicable 10CFR55 Section : 43.5 /4 5.13

Learning Objective: 3-0 T-AOI1400 , Rev. 6, Objective 7: Demonstrate ability/knowl edge of AOI, to

correctly: a. Recognize entry conditions, b. Resp ond to Action steps, c. Respond to Contingencies , d.

Respo nd to Notes and Ca utions. 3-0T-G01000 , Rev. 5, Objec tive 5: Identify the proced ure to which the

opera tor is refe rred if Residual Heat Removal cooling is lost while in during Reduc ed Invent ory/M id-Lo op

oper ations. (SO ER 88-3 & SOE R 85-4 )

References: GO- 10, Rev. 37.

Question:

Given the following:

Unit 1 is in Mode 5 follow ing a refueling outage .

RHR pump 1A-A is in serv ice.

The opera ting crew is drawing vacuum on the Reactor Coolant System.

The RHR pump begins to show signs of cavitatio n.

Which ONE of the following identifies both how the RHR pum p motor amps are affec ted when the pump is

cav itating , and the mitiga ting strategy that will be implemented if the cav itation cannot be terminated?

Motor Amps Mitigating Strategy

A. Unstable and fluctuating. Break vac uum per GO-10. Reactor Coolant System Drain

and Fill Operations, then enter AO I-14, Loss of RHR

Shutdown Cooling.

B. Unstable and fluctuatin g. Immediately enter AOI-14 . Loss of RHR Shut down

Cooling . Break vacuum as directed by the AOI.

C. Stable but reduced. Break vacuum per GO-1 0. Reactor Coolant System Drain

and Fill Operations, then ent er AOI-14, Loss of RHR

Shutd own Cooling.

D. Stable but reduced. Immediately enter AOI-14 , Loss of RHR Shutdown

Cooling. Break vac uum as directed by the AO I.

DISTRACTOR ANALYSIS

a. CORRECT. GO-10 describes cavitation and amps being unsteady and directs the break ing of

vacuum prior to implementing AOI-14.

b. Incorrect. While GO- 10 describes cavitation and amp s as stated in the distractor , the GO requires

the vacuum break prior to the tran sition to AO I-14.

c. Incorre ct. Amps will not be "stable but reduced" if the pump was cavitating ; however the mitigatin g

strateg y is corre ct.

d. Incorrect. Am ps will not be "stable but reduced" if the pump was cavi tating. The GO requires the

vac uum break prior to the trans ition to AOI-14.

4 of 29

SRO Watts Bar 2008 NRC Initial License Exam

WRITTEN QUESTION DA TA SHEET

Qu est ion Number : ~ 1 q

KIA: 000027 G2.4.6

Pressurizer Pressure Control System Malfunction

Knowledge of EOP mitigation strategi es.

Tier : RO Imp : RO Ex am: Co gn itiv e Level : High

Group : SRO Imp : 4.7 SRO Exam : Yes Source : BANK

Applic abl e 1OC FR55 Sect io n : 41 .10 1 4 3. 5 / 4 5. 13

Learn ing Objective: 3-0 T-AOI1800; Object ive 5: Explain the operator actions for drop ping RCS pressu re.

References : E-O, Rev.27 , Drawing 47W8 13-1 , AOI- 18, Rev . 21.

Qu estion :

Given the following conditions :

The plant is ope rating at 100% powe r steady state conditions.

All systems are aligned normally.

A failure of the Pressurizer Spray Valve PCV-68-340D causes it to go full OPEN .

The OAC has attemp ted to take manua l control of the Spray Valve, but is unable to close it.

Pressurizer pressure continues to LOWER.

What is the appropriate mitigation strategy and which procedure{s) will be used to implement the strategy?

A. Enter AOI-18, Malfun ction of Pressurizer Pressure Control System, initiate a Reacto r Trip , trip RCP #1,

then enter E-O, Reactor Trip or Safety Injection.

B. Refer to ARI-90-B, PZR PRESS LO DEVN BACKU P HTRS ON, isolate Train B Essential Air to

Containment to fail the Pressurizer Spray Valve closed , and then enter AOI-1 8.

C. Refer to ARI 90-B, PZR PRE SS LO-DE VN BACKUP HTR S ON , isolate Train A Essential Air to

Containment to fail the Pressurizer Spray Valve closed , and then enter AO I-18.

S' ~

D. Enter AO I-18. Malfunction of Pressu rizer Pressure Control~, initiate a Reacto r Trip AND SI,

trip RCP #1, then enter E-O, Reactor Trip or Safety Injection.

DISTRACTOR ANALYSIS

a. Incorrect. Guidanc e is given in AOI -18 for tripping the reactor if the Pressurizer spray valves canno t be

closed. This requires entry into E-O . Candidate must recognize that a Reacto r trip AND an SI is

required.

b. Incorrect. Plausible, since isolation of Essential Air to Containment causes the spray valve to fail

closed .

c. Incorrect. Plausible, since isolat ion of Essential Air to Containment causes the spray valve to fail

closed.

d. CORRECT. Guidan ce is given in AOI-18 for tripping the reactor AND an SI if the Pressurizer spray

valves cannot be clos ed. This requir es entry into E-O.

5 of 29

SRO Watts Bar 2008 NRC In itial Licens e Exam

WRITTEN QUESTION DA TA SHEET

Qu estion Number : KW

KIA: 000058 AA2 .01

Loss of DC Power

Ability to determine and interpret the follow ing as they apply to the Loss of DC Power : That a loss of dc

power has occ urred; verification that substitute power sources have come on line

Tie r: ROlmp : RO Exam : Cognitive Level: High

Group : SRO Im p : 3.1 SRO Exam: Yes Source: New

Applicable 10CFR55 Section: 43.5

Learning Objective: 3-0 T-SYS057P, Objective 6: Explain how the operator can tell if the 125v Vital

Charger or the 125v Vital Batter y is supplying power to the 125v Vital Battery Boards.

References : Tech Spec 3.8.4 Bases , page B 3.8-57, Rev 87

Question :

Given the following :

The plant is oper ating at 100% power .

The 125 V DC VITAL CHGR III fails and its output breake r opens .

A report is received that there is arcing occurring on Vital Battery III.

The Shift Manager directs that 0-BKR-236-3/10 9 125V Vital Batter y III Brea ker , betw een Vital

Batter y III and Vital Batte ry Board III be opened.

The Unit Supervisor has determined that power to Vital Battery Board III will be restored using Vital

Batter y V, in conjunction with the Spare Battery Charger.

Vital Battery V is curren tly in service to Battery Board V.

Which ONE of the following describ es the expected indication on 1-EI-57-94, Vital Batt BD III AMPS which

will confirm that power has been restored to Battery Board III, AND what is the ope rabi lity status of Vital

Battery Board III?

1-EI-57 -94 (Batt BO III Amps) Indication Operability Status of Vital Batt. BO III

A. Indicating UPSCALE from zero. Can NOT be considered operable with

both Vital Battery V and the spare charger

connected concurrently.

B. Indicating DOWNSCALE from zero. Can NOT be cons idered ope rable with

both Vital Batter y V and the spare charger

conne cted concurrently.

C. Indicating UPSCALE from zero . Operable with the spare charge r and Vital

Batter y V connected to the Battery Board and with all

applicable surveillances on Vital Battery V satisfied.

D. Indicating DOWNSCALE from zero . Operab le with the spare charge r and Vital

Battery V connected to the Batt ery Board and with all

appl icable surveillances on Vital Battery V satisfied.

OISTRACTOR ANALYSIS

a. Incorre ct. Upscale is the incorrect indication if a battery charger is in service providin g power to the

loads. The battery board is operable with the spare charger and Battery V connected as identified in

SOI-236.03, therefore using them concurrently is not an operab ility concern .

b. Incorrect. Downscale is the correct indication since a battery charger is providing power to the loads,

even though a different battery is connect ed to the Vital Board III. The battery boa rd is operable with

the spare charger and Battery V connected as identified in SOI-236.03, therefore using them

concurrently is not an oper ability concern.

6 of 29

SRO Watts Bar 2008 NRC Initial Licens e Exam

WRITTEN QUESTION DA TA SHEET

c. Incorrect. Upscale is the incorre ct indica tion if a battery charger is in service providing powe r to the

loads . The Battery Board is oper able with the spare charg er and Battery V connec ted provided all

app licable surveillances on Vital Battery V are satisfied.

d. CORRECT. Downscale is the correct Indication when a battery charger is providing power to the

loads, even though a different battery is connected to the Vital Board III. The SOl has the spa re

charger place d in serv ice prior to Battery V being connected. The battery board is operable with the

spare charg er and Battery V (if operable) connected as identi fied in SOI-236.05 . Per Tech. Spec

Basis 3.8.4, page B 3.8-57 , the Vital Battery V can be considered ope rable after it is connected to a

board and all app licab le SRs have been verified satisfactorily.

7 of 29

SRO Watts B ar 2008 NRC In i tial Lice nse Exa m

WRITTEN QUESTION DATA SHEET

Question Number: ~} I

KIA : W/E1 1 G2.2.22

Loss of Emergency Coola nt Recirc.

Knowledge of limiting conditions for oper ation s and safety limit s.

Tier : RO Imp : RO Exam: Cogn itive Level: High

Group: SRO Imp: 4.7 SRO Exam : Yes Source: NEW

Ap plicabl e 10CFR55 Section : 41 .5. 43.2

Learning Objective : 3-0T-SYS0 63A . Rev. 10, Objective 30: Given the cond ition/statu s of the Emergency

Core Cooling system/component and the appropriate sections of Tech Specs . determ ine if operabi lity

requirem ents are met and what actions, if any. are required.

References : LCO 3.3.2, Action K, includi ng basis.

Qu estion :

With the plant at full power , and during the Shift Turnover for the 1900 shift, the Unit Supe rvisor is informed

of the following:

1-LS-63-50A (RWST Low Level) was declared inoperable at 1000 that day .

It was placed in the configuration required by Technical Spec ifications at 1400, and is expected to

remain inopera ble until 2300.

A required surveillance instruction on 1-LS-63-51A (RWST Low Level) must be completed by 2330

toda y to preven t being out of frequenc y due to exceeding the NRC late date.

The surveillance involves having 1-LS-63-51A in the required config uration for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

If the surveillance is completed by 2330, which ONE of the following describes the expec ted effe ct on the

automatic switchover to containm ent sump function while 1~6 3-51A is out of service?

LS ~_

A. Functional. Even though two level switc hes are TRIPP ED and are inoperable, the remaining

operable level switches are sufficient for switchover to be functional.

B. Functional. Even though two level switches are BYPASS ED, the remaining level switches are

sufficient for switch over to be functional.

C. Not functional, becaus e two level switches are TRIPPED and at least three level switches are

requir ed for switchover to be functional.

D. Not functional , because two level switche s are BYPASSED, and at least three level switches are

requir ed for switchover to be func tional.

DISTRACTOR ANALYSIS

a. Incorrect. Plausible, since there are other compone nts in the plant that are placed to trip when

discovered inoperab le. However, candidate fails to recognize that whe n one of these level switches

is inoperable, it is bypas sed, not tripped . Further plausibili ty is added by the condit ions given; i.e.,

there TWO switches affected, and the remaining two are sufficient for actua tion capabi lity.

b. CORRECT . Per LCO 3.3.2, Action K, one inope rable channel of "Automatic Switchover to

Containme nt Sump" must be bypassed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Anothe r channel may also be bypass ed for

surve illance testing, leavinq only two chan nel s. However, the remaining two channels are sufficient

for functional ity of actuation.

c. Incorrect. Cand idate incorrect ly believes that these condit ions require tripping of the two channels,

and uses that incorrect informat ion to conclude that actuat ion capab ility is rendered not functional.

Plausible , since candidate may believe at least three channels are requ ired to be operable at all

times, for Mode 1.

d. Incorr ect. Candidate correctly recognizes that these cond itions require bypassing the two channels .

However , candidate fails to realize that this does not disable emergency coolan t recirculation

actuat ion capabilit y. Plausib le, since some plant equipmen t requires at least three of fou r channe ls to

be operable ; however, the candida te misapplie s that concept for these cond itions .

8 of 29

SRO Watts B ar 2008 NRC In itia l Licen se Ex am

WRITTEN QUESTION DA TA SHEET

Question Number: '~2

KIA: 000059 AA2 .03

Accident al Liquid RadWaste Rei.

Ability to dete rmine and interpret the following as they apply to the Accidental Liquid Radwaste Release:

Failure modes. their symptoms . and the causes of misleading indications on a radioactive-liquid monitor.

Ti er: 1 RO Imp: RO Exam: Cogn it ive Leve l: Low

Gro up : 2 SRO Imp : 3.6 SRO Exam : Yes Source : NEW

Appl icable 1OCFR55 Section : 43.5.45.13

Learn ing Objective: 3-0T-SYS07 7A, Objective 19: Discuss how processed wate r is released .

Referen ces : SOI-77 .01, Rev. 0058; AR11 80-187, Rev. 30;

0-0 0 1-90-1, liquid Radwa ste Tank Relea se, Rev. 0028 .

Question :

Given the following :

The unit is at 100% power and all equipm ent is available.

A planned Cask Decontamination Collector Tank (CDCT ) release is in progress when the follow ing

occurs :

Annunciator 181-A "WDS RELEASE LINE 0-RM-90-122 L1 Q RAD HI" alarms .

The Monitor Tank level is 70% and lowering.

Which ON E of the following ident ifies whether the release permit was violated and the release permit

requirements to allow the CDCT release to continue, in accordance with SOI-77 .01, Liquid Wast e Dispo sal?

Permit Violated Release of COCT

The release permit would be violated

A. becaus e the liquid releas ed was not The same release permi t can be used following

sampl ed prior to the release. independent verification of correct valve lineup.

The release permit would be violated

B. because the liquid release d was not A new release perm it must be gene rated.

sampled prior to the release.

The release permit was NOT

C violated because the release was The same release perm it can be used following

termina ted by the high rad signal. independent verification of correct valve lineup .

The release permit was NOT

D. violated because the releas e was A new release permit must be generated.

term inated by the high rad signal.

OISTRACTOR ANALYSIS

a. Incorrect. The release perm it would be violated because liquid released from the Monitor Tank was not

sampled and the same release permit cannot be used to continue the release . Plausible because the

release permit did sample the CDCT liquid intended to be released. and independent verification of the

valve line-up is something required if the rad moni tor is inoperable.

b. CORRECT. The release permit would be violated because liquid released from the Monitor Tank was

not samp led and a new release permit is required to continue the release due to the release being

terminated by a High Rad signal as identified in SOI-77 .01. The release permit did sample the CDCT

liquid intended to be releas ed. and a new release permit is required in ord er to continue the relea se.

c. Incorrect. The terminat ion of the release by the radiation monitor does not prevent a violation of the

permit. The release perm it would be violated because liquid released from the Monitor Tank was not

samp led and the same release perm it cannot be used to continue the release. Plausible because the

9 of 29

SRO Watts Bar 2008 NRC Initial License Exam

WRITTEN QUESTION DATA SHEET

release permit did sam ple the CDCT liquid intended to be released; the rad monit or did terminate the

release and independent verification of valve line-up is som ethin g required if the rad mo nitor is

inoperable.

d. Incorrect. The termination of release by the radiation monitor does not prev ent a violation of the permit.

The release permi t would be violated beca use liquid released from the Monit or Tank was not samp led.

A new release permit is requi red in orde r to continue the release . Plausible because the release permit

did sam ple the CDCT liquid intended to be released ; the rad monit or did termin ate the release and a

new release permit is requ ired .

10 of 29

SRO Watts Bar 2008 NRC In itial License Exam

WRITTEN QUESTION DA TA SHEET

Que stion Number: ..... 0 .-2.,

Il' L" ___

KIA : 000068 G2.4.8

Control Room Evac.

Knowledge of how abnormal operating procedu res are used in conjunction with EOP s.

Tier: 1 RO Imp : RO Exam : Cogn itive Level : High

Gr ou p: 2 SRO Imp : 4.5 SRO Exam: Yes Source: SQN Bank

MODIFIED

Applicable 1 OCFR55 Section : 43 .b (5)

Learning Objective: 3-0T-AO I3000, Objective 12: Demonstrate abi lity/k nowledge of AOI -30 .1 and 30.2

by: a. Recognizing entry co ndit ions, b. Responding to required actions of the AOI , c. Responding to

contingencies (RNO) , d. Responding to Note s/Caution s.

References : AOI -30.2, Fire Safe Shutdown , Rev 27; Note at beginning of Section 3.0, page 5.

Question :

Given the following plant con ditions :

Unit 1 was at 100% power when a Main Con trol Room (MC R) evacuation was required.

The crew entered AOI-27 , Mai n Control Room Inaccessibility.

The crew performed ES-0.1, Reactor Trip Response , prior to leaving the MCR.

Wh ile perfo rming actions from the ACR a Safe ty Inject ion occurs .

Which ON E of the follow ing will be the status of the MS IVs when the crew establish es co ntrol from the Aux

Contro l Room and describes the cor rect procedure usage ?

MSIV Status Procedure Us age

A. Open AO I-27 will be the controlling proced ure beca use it is wrillen with

mitigating actions to respond to a Safe ty Injection.

B. Open E-O, Reactor Trip or Sa fety Injection, would be used because AOI-27

is wrillen assuming no other accident is occurring .

C. Closed AOI -27 will be the cont rolling proc edure because it is wrillen with

mitigating actio ns to respond to a Safety Injection.

D. Closed E-O, Reactor Trip or Safety Inje ction, would be used becau se AO I-27

is wrillen assum ing no other acciden t is occ urring .

DISTRACTOR ANALYSIS

a. Incorrect. AO I-27 directs the closu re of the MS IVs prior to leaving MC R and is wrillen to assume no

othe r acc ide nt is occ urring . Plau sible becaus e if the MSIVs were ope n, steam dump valve s would be

available and the AO I used for abando ning the MC R during an Appendi x R fire is wrillen for assuming a

spu rious SI occu rs . (The Appendi x R AOI is the controlling procedu re during Appendi x R fire) .

b. Incorrect. AO I-27 directs the closure of the MS IVs prior to leaving MCR and is wr illen to assume no

other acc ide nt is occ urring . If an acc ident is occu rring then the eme rge ncy procedure network would be

used . Plausible because if the MS IVs were open, steam dum p valves would be availa ble and the

eme rgenc y procedure network would be used because the AOI is wrillen assuming no othe r acci de nts .

c. Inco rrect. AOI- 27 direct s the closure of the MSI Vs prior to leaving MCR however and the AOI is wrille n

to assum e no othe r accident is occu rring . Plau sible because the MSIVS are dire cted to be closed prior

to leaving the MCR and the AO I used for abandoning the MC R during an Appendi x R fire is wrillen

ass uming a spurio us SI occ urs . (The Appendix R AO I is the controlling proced ure du ring Appendi x R

fire).

d. CORR ECT . AO I-27 directs the MS IVs to be closed prior to leaving the MCR. Th e discussion sect ion of

til e AOI states the MCR inaccessibility is no t considered to occur with or subsequently with anoth er

accident.

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SRO Watts B ar 2008 NRC Initial Lic ens e Exam

WRITTEN QUESTION DATA SHEET

Qu estion Number: % 81

KIA: W/E13 EA2. 1

Ste am Gener ator Over-pressure

Ability to determine and interpret the following as they apply to the (Ste am Gene rator Ove rpressu re)

Facility cond itions and selection of appropriate procedu res during abno rmal and emergency operations .

Tier : 1 RO Imp : RO Exam: Cogn it iv e Lev el : High

Group : 2 SRO Imp : 3.4 SRO Ex am : Yes Source: NEW

Ap plicabl e 10CFR55 Sec tion : 4 3 .5 /4 5 .13

Learn ing Objective: 3-0T-T11 204, Rev . 1, Obje ctive 25 : Describe when a Function Restoration Instruction

can be exited or transition ed out of.

Ref er ences : FR- H.2, Steam Gene rator Overpress ure , Re v 5 page 3.

FR-H.3, Steam Generator High Level, Rev 6 page 3.

Question :

Given the following:

The crew is performing FR-H .2, Steam Generator Overpressure, for an ove rpressure cond itio n on

SG #2.

When the step is address ed to check affected SGs NR leve l it is noted that the SG #2 level is

indicating 94% narrow range.

Which ONE of the following identifies the correct crew actions as a result of the SG level ind icating 94% ?

A. Continue in FR-H.2, but do not initiate any steam release un til TSC eva lua tion is com plete.

B. Cont inue in FR-H .2, steam release may conti nue until NR level indi cates 100% .

C Transition to FR-H.3, Steam Generator High Level , but do not initiate any steam release until TSC

evaluation is comp lete.

D. Transition to FR- H.3, Steam Generator High Level , steam relea se ma y continue until NR level

indicates 100% .

DISTRACTOR ANALYSIS

a. Incorrect. Plausible sin ce the release of steam is prohib ited with a high level (> 93 %) condition until

after a TSC evaluation is comple te. Trans ition to FR-H .3 is directed from FR-H .2 at Step 3.

b. Incorrect. The transit ion to FR-H.3 is requir ed , howe ver candidate may corre ctly conclude that FR-H .3

is lower in priority on the FR-H status tree and not recall the transition . Plausible since eve n w hen NR

level indicates 100%, there is still significa nt volum e before the stea m gen erator fills with wat er.

c. CO RRECT. The RNO for the ste p dire cts the transition to FR-H.3 and FR -H.3 restr icts the releas e of

steam until a TSC evaluation is complete .

d. Incorrect. The RNO for the step dire cts the transition to FR-H. 3 and the release of steam is restr icted

until a TSC evaluation is complete if the level exceed s 93% , therefore with the level at 100% , the

release will be restricted . Plausible since even when narrow range SG level indi cates 100% , ther e is

still significant volume be fore the steam generator fills with water.

12 of 29

SRO Watts Bar 2008 NRC In itial Lice nse Ex am

WRITTEN QUESTION DA TA SHEE T

Question Number : j.(f t,s

KIA : W/E08 G2.4 .18

RCS Ove rcoo ling - PTS

Knowle dge of the specific base s for EOPs.

Tie r: 1 RO Imp : RO Exam : Cogn iti ve Le vel : High

Group : 2 SRO Im p : 4.0 SRO Exam: Yes Source: NEW

Applicable 1OCFR55 Sectio n : 41.10 /43 .1 145 .13

Learning Objective: 3-0 T-FRP000 1, Rev 10, Objective 9: Explain the basis for returning to the instruction

in effect after identifying that RCS pressure 5. 150 psig and RHR is deliver ing flow when perfo rming step 1 of

FR-P.1.

References : FR-P .1, Pressurized Thermal Shock , Rev 14; TI-12.04, Users Guide for Abn ormal and

Emergency Instru ctions, Rev.0007 .

Quest ion :

Given the following :

Unit 1 experienced a Reactor Trip and Safety Injection .

The crew tran sitioned to FR-Z .1. High Conta inment Pressure, from E-1, Loss of Reactor or

Second ary Coolant.

While FR-Z.1 was being perfo rmed , the crew transitioned to FR-P.1. Pressurized Thermal

Shock .

The STA reports the containment pressure has dropped and the containment status tree is

GREEN and that no other RED or ORA NGE paths exist.

Which ON E of the following identifies the basis of the FR-P.1 step for checking RCS pressure great er than

150 psig, and the proced ure the crew will implement if a transition is made from FR-P.1 during perfor mance

of the step?

A. To preclude the need to perform FR-P.1 action s, since pressuri zed thermal shock is not a serious

concern for a large-b reak LOCA ;

Transi tion is made back to E-1.

B. To preclude the need to perform FR-P .1 actions, since pressurized therm al shock is not a serious

concern for a large-break LOCA ;

Transition is made back to FR-Z.1.

C. To avoid delays caused by unnece ssary soak periods required by FR-P.1.

Transition is made back to E-1.

D. To avoid delays caused by unne cessary soak periods required by FR-P.1.

Transition is made back to FR-Z.1.

DISTRACTOR ANALYSIS

a. Incorrect. While a pressurized therma l shock is not a serious concern for a large -break LOCA , the

transition will be back to FR-Z. 1, not to E-1. Plausib le, since basis is correct, and with the FR-Z.1

status green the candidate could conclude the return to E-1 will be required.

b. CORRECT. A pressur ized thermal shock is not a serious concern for a large -break LOCA because

the system cannot repressurize with a large break LOCA. The step RNO will transition back to

instruction in effect, which is FR-Z.1 even though the status tree for it is now green .

c Incorrect The bases is not to prevent the soak periods , but the transition to E-1 is correct. Plausible

because the procedure does contai n modified Sl termination criteria and SI will be terminated in the

proce dure whic h does reduce coo ling to the core and the transit ion is correct

d. Incorrect. The bases is not to prevent the soak periods , and the transition to FR-Z. 1 is not correct.

Plausible because the procedure does contain modified SI termination criteria and SI will be

termin ated in the procedure which does reduce cooling to the core and with the FR-Z.1 status green

the candidate could conclude the return to FR-Z.1 will be required .

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SRO Watts Ba r 2008 NRC In it ial Licens e Exam

WRITTEN QUESTION DA TA SHEET

Question Number : .-11' Eb

KIA : 003 A205

Reactor Coolant Pump

Abilit y to (a) predict the impacts of the following malfunctions or operations on the RCP S; and (b) based on

those predict ions . use procedures to correct. control. or mitigate the conseque nces of those malfunc tions or

operat ions: Effects of VCT pressure on RCP sealleakoff flows .

Tier : 2 RO Imp: RO Exam : Cognitive Level : Hig h

Group : 1 SRO Imp : 2.8 SRO Ex am : Yes So urce: NEW

Appli cabl e 1OCFR55 Section : 41.5 /4 3.5/45 .3 /4 5/13

Learning Objective: 3-0T-EOPOOOO, Objective 15: Explain the purpose for and basis of each step in E-O .

ES-O.O. ES-O.1. ES-0.2. ES-0.3, and ES-O.4.

Refere nces: ES-0.2, Natural Circulation Cooldown. Rev 20; SOI-68.02. Reacto r Coolant Pump s. Rev 33.

Qu est ion :

Given the following:

Unit 1 was in Mode 2 with startup in progress when a loss of off-site powe r occurred .

The decision was made to place the plant in Mode 5.

The crew implemented ES-0.2, Natural Circulat ion Cooldown, and started cooling the plant down .

Four (4) hours after the cooldown was initiated both trains of offsite power were restored to the

plant.

The crew determines all criteria to restart the RCPs are met excep t for the #1 seal leakoff flow on

RCP #2 which is lower than the norma l operating band .

Which ONE of the follo wing iden tifies a change that causes an increase in the # 2 RCP sealleakoff flow and

the actions to be taken and procedure to be used if the seal leakoff flow cann ot be established within the

normal operating band ?

A. Lower PRT pressu re;

Start RCP #1, cont inue perfo rming ES-0.2 until all RCS temperatu res are less than 200°F. then

transition to GO-6 , Unit Shutdown From Hot Stand by To Cold Shutdown.

B. Lower PRT pressure;

Start the other 3 RCPs and immediately transition from ES-0.2 to GO-6 , Unit Shutdown From Hot

Standb y To Cold Shutdown.

C. Lower Ve T pressure;

Start RCP #1. continue perform ing ES-O.2 until all RCS temperatures are less than 200°F. then

transiti on to GO-6 . Unit Shutdown From Hot Standby To Cold Shutdown .

D. Lower VCT pressu re;

Start the other 3 RCPs and immediately transition from ES-0.2 to GO-6 , Unit Shutdown From Hot

Standby To Cold Shutdown.

DISTRACTOR ANALYSIS

a. Incorrect. Lowering the PRT pressu re affects the #1 sealleakoff flow. but only if the leakoff flow path ,

which is routed to the VCT. is isolated . Plausible because starting an RCP provides forced circulation,

and some spray flow correct and if ES-0.2 was continued until the end of the procedu re. a transiti on to

GO-6 would be made .

b. Incorrect. Lowering the PRT pressure affects the #1 sealleakoff flow. but only if the leakoff flow path.

which is routed to the VCT, is isolated . Plausible beca use starti ng the other 3 RCPs and transitioning to

GO-6 after starting the RCPS is correct.

c. Incorrect. Lowering the VCT pressure wou ld allow increased sealleakoff flow. but ES-0.2 wou ld not be

continued after the RCP was started (a transition would be made to GO-6 whe n the pump was started .)

Plausible because lowering the VCT pressure is correct and if ES-0.2 was con tinued until the end of the

proced ure. a transition to GO-6 would be made.

d. CORR ECT . Lowering the VCT pressure would allow increased seallea koff flow and the other 3 RCP s

are directed to be started if RCP #2 cannot be started.

14 of 29

SRO Watts Ba r 2008 NRC Initial License Exam

WRITTEN QUESTION DA TA SHEET

Question Number: j~8 -7

KiA: 0 12 G2.2.44

Reactor Protection

Abi lity to interpret con trol room indica tions to verify the status and operation of a syste m , and und erstand

how operator actions and directives affect plant and system co nditions .

Tier : 2 RO Imp: RO Exam: Cognitive Level : High

Group : 1 SRO Imp : 4.4 SRO Exam : Yes Source: NEW

Applicable 10CFR55 Section : 41.5 /435 /45 .12

Learning Objective: 3-0T-A0 12100 , Obje ctive 7: Describe sig nifica nce of loss of de to Protection and

Co ntrol sys tems (SO ER 83- 5, Rec 9)].

References: AOI-21 .01, Loss of 125v DC Vital Batt ery Bd I, Rev 2 1; 45W 600 -99-1 R6; LCO 3.3.2.

Question:

Given the follow ing :

Unit 1 is at 100 % power.

1-SI-99- 10-B, 31 Day Functional Test of SSPS Train B and Reac tor Trip Breaker B, is in progress

with Reactor Trip bypa ss brea ker (BYB) closed .

Appendix F, Reactor Breaker Rep lace ment , of 1-S I-99-1O-B is in prog ress with Rea ctor Trip

Breaker B (RTB) racked out.

A feed water tran sient occu rs due to an electri cal power loss, res ulting in the MC R ope rator

MANUALLY tripping the reacto r.

Wh en Tavg, Pressurizer Pressure , and SG levels are stabilized post trip, the follow ing co nditions are

observed:

All 4 Diesel Ge nerators running but NOT co nnected to the shutdown boards .

All Reactor Trip Breaker and Reactor Trip Bypass breaker indicating light s on 1-M-4 are DAR K

except for the GREEN light on the Reactor Tri p Bypass Breaker B (B YB ) which is LIT.

Which ONE of the following ide ntifies both the position of Reactor Trip Breaker A (RTA ) and the corr ect SRO

deci sio n relative to com pleting 1-SI-99-1O-B followi ng the stabil ization of the plant and the elec trica l powe r

restoration?

RTA Position 1-SI-99-10-B Status

A. Closed The surveillance is requ ired to be complet ed .

B. Closed Th e surveillan ce is NOT required to be co mpleted with the plant in

this Mode .

C Tripped The survei llance is requ ired to be co mpleted .

D. Tripped The survei llance is NOT required to be completed wi th the plant in

this Mode.

DISTRACTOR ANALYSIS

a. Incorrect. RTA will be tripp ed by the UV coil. Plausible, if candidate doe s not co rrectly apply the

func tion of the reac tor trip bre aker UV coil trip function but does realize the need complete the SI which

is requir ed eve n with til e plan t in MO DE 3.

b. Incorrect. RTA will be tripped by the UV coil and the SI is still required with the plant in Mode 3.

Plausible, if candida te does not cor rectly ap ply function of the reac tor trip brea ker UV coi l trip functio n

and det ermines the SI wou ld not be required to be completed becau se the plant is now in Mode 3.

c. CORRECT. RTB and BYA will have no indicating light s lit because the breakers are racked out of the

cubicles, BYB will be open and have gree n light lit; RTA will hav e no light lit due to a loss of 125v DC

15 of 29

SRO Watt s Bar 2008 NRC Initial Licens e Exam

WRITTEN QUESTION DA TA SHEET

control power , however, the UV coil on RTA will cause the brea ke r to open . The SI is requir ed with the

plant in MOD E 3.

d. Incorre ct. The UV coil on RTA will cause the breaker to open , however will be isolated by closin g 1-

FCV-6 2-90 and 91 in accord ance with the AOI. Plausible if the candidate correctly det ermin es

correctly the RTA position is correct but incorrectly determines SI wou ld not be required to be

com pleted because the plant is now in Mod e 3.

16 of 29

SRO Watt s Bar 2008 NRC Initial License Exam

WRITTEN QUESTION DA TA SHEET

Question Number: 4-3' gf)

K/A : 061 A2.09

Auxil iary/Emergency Feedwater

Ability to (a) predict the impacts of the following malfunctions or ope rations on the AF W; and (b) based on

those predic tion s, use proce dures to correct, control, or mitigate the cons equences of those malfunctions or

operations: Total loss of feedwater.

Tier: 2 RO Imp: RO Exam: Cognitive Level: High

Group : 1 SRO Imp: TBD SRO Exam : Yes Source: NEW

Applicable 1OCFR55 Section: 41 .5. 43.5

Learn ing Objective: 3-0T-SYS003B Auxiliary Feedwater System, Objec tive 26. Identify the steps to gain

local control of the Tur bine-D riven Auxiliary Feedwater pump and SG levels.

References: FRH.1, rev. 17.

Question:

Given the following:

Following a reactor trip the plant experienced a total loss of feedwater.

The condition caused the crew to initiate RCS bleed and feed.

Subsequently the TDAFW pump was restored to service and the crew is read y to establish AFW

flow to the selec ted steam generator.

Other plant conditions include:

o Selected SG Wide Range level is 4%.

o RCS loop hot leg temperature at 558°F.

(I Core exit thermoc ouple temp eratur es are RISING .

In accordance with FR-H.1, Loss of Second ary Heat Sink, feedwat er flow will be re-estab lished to the

selec ted SG at ' "

A. the minimum detectable flow gpm.

B. less than 100 gpm until WR level >15%.

C. a rate that causes wide range level to rise and RCS hot leg to drop.

D. a maximum rate.

DISTRACTOR ANALYSIS

a. Incorrect. Plausible, since the feedwa ter flow wou ld be correct for the conditions where a steam

generator was being fed to prevent dryout.

b. Incorrect. Plausible, since the feedwater flow would be correct for the conditio n which meets the "hot,

dry" cond itions and core exit thermo couples are not rising.

c. Incorrect. Plausibl e, since the feedwat er flow would be correct for the condition where Core Exit

thermocouples were stab le or loweri ng.

d. CORRECT. The maximum flow rate is only used if the steam gene rator meets the wide range level is

<15% and RCS temperature is >550°Fand core exit thermocouples are rising .

17 of 29

SRO Watts Bar 200 8 NRC In itial Lic ens e Ex am

WRITTEN QUESTION DATA SHEET

Qu estion Number : .;4[J{

KIA : 064 G2 .2.37

Emergency Diesel Generator

Ability to determine operability and/or ava ilability of safety related equipment.

Tie r: 2 RO Imp: RO Exam: Cognitive Level: High

Group : 1 SRO Imp: 4.6 SRO Exam : Yes Source: NEW

Applicable 10CFR55 Section: 4 1.7,43.5

Learning Objective : 3-0T-SYS082A , Obj . 13: State the Techni cal Specif ication requirements associated

with AC Sources & DG Support Systems .

References : LCO 3.8.4, DC Sou rces Operating , including Basis. LCO 3.8.6 , Table 3.8 .6-1, inc!. Basis.

Question :

Given the followin g conditions :

The plant is shutdown In Mode 4 wit h all safety related equipment operable .

Dur ing the performa nce of surve illance O-S I-215- 21-A, DIESEL GENERATOR 1A-A BATT ERY

QUA RTER LY INSP ECT ION, it is repo rted that the Float Voltage val ues for three (3) connected

cells are as follows :

Cell 17 = 2.09 v

Ce ll 34 = 2.06 v

Cell 39 = 2.12 v

ALL other co nnected cells have Float Vo ltag e values greater than 2.13 v.

A LL other parameters measured during the above surveillance are normal.

Which ONE of the following describes the status of DIG 1A-A batte ry AND of the DIG 1A-A?

REFERENCE PROVIDED

Status of Battery Status of D/G 1A-A

A. Battery is degr aded but it can NO Tech Spec or trac king only entry required for DG.

be consi dered operable for 31 days .

B. Battery is degraded but it can Tech Spec tracking onl y entry required for DG .

be considered operable for 31 days.

C. Battery is inop erabl e. DG is inoperable and Tech Spec tracking on ly

entr y is required.

D. Battery is inop erable . DG is inop era ble and Tech Spec entry

is required.

DISTRACTOR ANALYSIS

a. Incorre ct. Plausible, since if cell voltages had been below Categ ory B limits, but above Category C

allowable limits, battery could be considered operable for 31 days.

b. Incorrect. Plausible, since if cell voltages had been below Category B limit s, but abo ve Categ ory C

allow able Ii~ i ts , battery could be considered operable for 31 days.

I("(C\:12J;*~c ~gGU: Per LCO 3.8.6 , since Cell 34 volt age is less than 2.07v, it does not satisfy Cat ego ry C

requiremen ts, which is Condition B of LCO 3.8.6. The battery is therefore inoperable . It is true the DIG

is also inoperable , however, in Mode 4, onl y one train of DIGs is required . The on e inop erabl e DIG

status woul d be trac ked per Tech Spec track ing on ly.

/" .\ _~ Incorrect. Plaus ible, since first part of distractor is true; however, ca ndida te fails to recognize and

\..Crlee \\ . ~ appl y knowledge that in the cu rrent plant mode , the one ino perab le DIG on ly requires Tech Spec

~ tracking , and NOT

} t i {prOVide only the pages of LCO 3.8 .6.

~V\

Pe 6- PO~ t ~aJ,'.' QC'I\r.eA .

Di~ 5>i iK£G4fGtL b l\t 18 of 29

SRO Watts Bar 2008 NRC Initial License Exam

WRITTEN QUESTION DA TA SHEET

Question Number: .:t.a- CJ"' fLJ

KIA: 076 A2.01

Service Water

Ability to (a) predict the impacts of the following malfunctions or operatio ns on the SW S; and (b) based on

those predictions, use procedures to correct , control, or mitigate the consequ ences of those malfunctions or

operations: Loss of SWS.

Tier : 2 RO Imp : RO Exam: Cognitive Level: High

Group: 1 SRO Imp: 3.7 SRO Exam: Yes Source: NEW

Learn ing Objective : 3-0 T-AOI1300, Objective 4: Identify the genera l location of a rupture give n a Hi

strainer \ P and low flow in the same hdr. 3-0T-T11240, Objective 4: Describe the four "risk" levels used in

the Equipment to Plant Risk Matrix and the signif icance of each level.

References : TI- 124 , Equipment To Plant Risk Matrix. Rev 14.

AOI-13, Loss of Essential Raw Cooling Water, Rev 35 .

Question:

Given the following conditions:

Unit 1 is operati ng at 100% power.

ERCW straine r 2B-B plugs.

It is determined that the strainer must be isolated .

Which ONE of the following identifies the action directed by AOI-1 3, Loss of Essential Raw Cooling Water,

to mitigate the isolation of the 2B strainer and the PSA risk status the plant will be in due to the strainer

isolation?

AOI-13 actions PSA risk

A. Realign cooling water to the A train diesel generators. Orange

B. Crosstie the 2B header with the 1A header until strainer is repaired. Orange

C. Realign cooling water to the A train diesel generators. Red

D. Crosstie the 2B header with the 1A header until strainer is repaired. Red

D1STRACTOR ANALYSIS

a. Incorrect. The PSA risk is Red (high) not Orange for the isolation of the ERCW head er. Plausible,

since the action listed involves realignm ent of cooling water. The cooling water supply for the A train

DGs will not be affected as a result of isolating the strainer and no DG cooling water alignmen t would

be required.

b. Incorrect. The PSA risk is Red (high) not Orange for the isolation of the ERCW header . Plausible,

since cross tie of the headers wou ld be performed in acco rdance with AOI-1 3.

c. Incorrect. Plausible, since PSA risk is correct , and the action listed involves realignment of cooling

water. The cooling water supply for the A train DGs will not be affecte d as a result of isolating the

strainer and no DG cooling water alignment wou ld be required.

d. CORRECT. The 2B ERCW header will be cross tied with the 1A ERCW header during performance of

AOI-13 and the PSA risk would be RED as identified in TI-124.

19 of 29

SRO Watts Bar 2008 NRC In itial Li ce nse Exam

WRITTEN QUESTION DA TA SHEET

Question Number: ~l

KIA: 011 G2 .1.32

Pressurizer Level Control

Ability to explain and appl y system limits and precautions.

Tier : 2 RO Imp: RO Exam : Cognitive Level: High

Gro up: 2 SRO Imp : 4.0 SRO Exam : Yes Source: NEW

Ap pl icable 10CFR55 Section: 41.10,43.2

Learn ing Objective: 3-0 T-T-S0304, Objective 4: Given plant conditions and parameters correctl y

determine the app licable Limiting Condi tions for Operatio ns or Technical Requ irements for the various

components of the RCS .

References : LCO 3.4.9

Question :

Given the following plant conditions:

During a startup, the plant is in Mode 4.

GO-1, Appendix C, Mode 4-to-Mode 3 Review and Appro val, has been comple ted up to the last

step, Operations Supe rintendent Hold Point, for granting appro val to enter Mode 3.

During scaff olding removal activities, a worker contacts the air line to 1-FCV-62-93, Cha rging Flow

Control Valve, result ing in pulling the air line loose from the valve ope rator , such that the air system

rema ins intact, due to crimping of the line on the air supply side of the break .

Which procedures will the Unit Supervisor use to respond to this event and the basis for taking quick action

to limit the effects?

A. The Unit Superviso r will use AOI-10, Loss of Control Air to direct actions to prevent Pressurizer level

from exceeding 92% in order to maintain the presence of a steam bubb le in the pressurizer.

B. The Unit Supervisor will use AO I-10 , Loss of Contro l Air, and direct actions to prevent the level

decrease to less than 17% to ensure subcooli ng margin can be maintai ned.

C. The Unit Superv isor will use SOI-62.01, CVCS- Charging and Letdown and direct actio ns to prevent

the level decrease to less than 17% to ensure subcooling margin can be maintained .

D. The Unit Superv isor will use SOI-62.01, CVCS-Ch arging and Letdown, to direct actions to prevent

Pressurizer level from exceeding 92% in order to maintain the presen ce of a steam bubble in the

pressurizer .

DISTRACTOR ANALYSIS

a. Incorrect. Plausible, if a failed open chargi ng flow control valve condition is not mitigated PZR level

would increase to the point where the desir ed steam bubble would not be maintained.

b. Incorrect. Plausible since it contains the correct proced ure. Candidate inco rrectl y concludes that when

1-FCV-62-93, Charging Flow Control Valve, loses air it fails CLO SED . Using this incorrect conclusion ,

it is logical to see why a cand idate would then conclude that Pressurizer level will lower . Candidate

correctly recalls that the re is a lower limit for Pressurizer level of 17%, and applies it to arrive at the

concern for losing all banks of Pressurizer heaters resulting in a potentia l loss of desired subcooling .

c. Incorrect. Plausible since it contains a procedu re with a title similar to the cond itions given in the stem .

Candida te incor rectly concludes that when 1-FCV-6 2-93, Charging Flow Control Valve loses air it fails

CLOS ED. Using this incor rect conclusion, it is logical to see why a candidate would then conclude that

Pressurizer level will lowe r. Candidate corre ctly recalls that there is a lower limit for Pressu rizer level

of 17%, and applies it to arrive at the concern for losing all banks of Pressu rizer heate rs resulting in a

potential loss of desired subcoo ling.

d. CORR ECT. When 1-FCV- 62-93 loses operating air, it fails OPEN , resulting in signifi cantly more

charging flow . If this is allowed to continue, Pressu rizer level will exceed 92%, which could result PZR

level increasing to the point where the desired steam bubble would not be maintained .. SOI-62 .01

contains detailed action steps for isolating and bypassi ng 1-FCV-62-93, which will restore charging

flow control to the control room, even with the normal flow cont rol valve failed open .

20 of 29

SRO Wat ts Bar 2008 NRC In itial Licens e Exam

WRITTEN QUESTION DATA SHEET

Question Number : ~qL.

KIA : 075 A2.03

Circulating Wate r

Ability to (a) predict the impa cts of the follow ing malfunctions or operations on the circulating water system ;

and (b) based on those predictions, use procedures to correct, contro l, or mitigate the consequences of

those malfu nctions or opera tions : Safety features and relationship between condenser vacuum, turbine trip,

and steam dump .

Tier: 2 RO Imp: RO Exam: Cognitiv e Leve l : Low

Group : 2 SRO Imp: 2.7 SRO Exam: Yes Source: NEW

Applicabl e 10CFR55 Section : 41.5, 43.5

Learn ing Objectiv e: 3-0 T-SYS027A, Objective 11: Describe the main conden ser . 3-0T-A0 11700,

Objective 1: Demonst rate knowledge of the Purpo se/Go al of AOI, Rev. 9.

Refe ren ces : AOI-1 1, Loss of Condenser Vacuum, Rev. 27.

Question :

Given the following :

Unit 1 was operating in Mode 1 at 14% reactor power.

A loss of condenser vacuum occu rs resulting in an automatic turbine trip.

The operating crew stabilizes the plant.

Which ON E of the following identifi es both how the condenser circ ulating water box LlTs will be affected and

the notifications required due to the turbine trip in accordan ce with SPP-3.5, Regulatory Repo rting

Requir ement s?

Water Box LlT Notificat ions required

A. Rises Internal TVA notifications only

B. Rises NRC and Internal TVA notifications

C. Lowers Internal TVA notifications only

D. Lowe rs NRC and Interna l TVA notifications

DISTRACTOR ANALYSIS

a. Incorrect. Steam dump s are prevented from openi ng due to loss of vacuum . Therefore, no steam is

entering the condenser, and water box LlT will drop . The notifications required by SPP-3.5 are intern al

only. Plaus ible if candida te fails to recall the condenser vacuum interlock with steam dump ope ration

but correctly ident ifies the required notifications .

b. Incorrect. Steam dumps are prevented from opening due to loss of vacuum. Therefore, no steam is

entering the condenser and water box LlT will drop . The notification required by SPP- 3.5 are internal

only, no notifi cation to the NRC is required . Plausible if candidate fails to recall the condenser vacuum

interlock with steam dump opera tion and incor rectly iden tifies the required notifications .

c. CORR ECT . With a loss of condenser vacuum, steam dump operation is blocked . Therefore, no

steam is entering the condenser, resulting in a lower LlT across the condenser waterboxes . SPP-3 .5

require s internal TVA notification be made due to the turbine trip.

d. Incorrect. No steam is enteri ng the condenser, resulting in a lower LlT across the condenser

wate rbo xes. The notification required by SPP-3 .5 are internal only, no notification to the NRC is

required . Plausible because the waterbo x LlT response is correct and candidate may incorrectly

identify the required notifications .

2 1 of 29

SRO Watts Bar 2008 NRC Initial License Exam

WRITTEN QUESTION DATA SHEET

Question Number: -+8"1>

K/A : 086 G2.4 .9

Fire Protection

Knowledge of low powe r/shutdown implications in accident (e.q., loss of coolant acc iden t or loss of residual

heat remov al) mitigation strategies .

Tier : 2 RO Imp: RO Exam: Cognitive Level : High

Group : 2 SRO Imp : 4.2 SRO Exam : Yes Source: NEW

Applicable 10CFR55 Section: 45.5

Learning Objective: 3-0T-AOI-700, Rev. 10: Explain the 2 mod es of maintaining the core cool & stable

during flood mode ope ration.

References : AO I-7.01, Rev. 16.

Question:

Given the following:

At0100 Unit 1 entered Mode 3 during shutdow n for a refueling outage .

At 0200 AO I-7.0 1, Maximum Probable Flood, was implem ented due to extremely heavy

rainfall in the upstre am watershed .

At 1100 The plant is in Mode 4 on RHR with cooldow n in prog ress with the Re S press ure 320

psig and temperature 220 F. 0

At 1200 River System Operations (RSO) confirms the flood level at the plant is predicted to

crest at EI. 730' within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

Which ONE of the following identifies both the Flood Stage Preparation level(s ) that is/are required to be

completed, and how the cooling of the core will be maintained in accor danc e with AOI-7 after the

prepa rations are complete ?

A. Only the procedure for Stage 1 Preparations is required to be completed .

Core cooling will be maintained by the steam generators with water being supplied by high pressure fire

protection with the RHR system removed from service.

B. Only the procedure for Stage 1 Preparations is required to be completed.

Core cooling will be maintained by the Spent Fuel Pool cooling system crosstied with the RHR system .

C. Both Stage 1 and Stage II Preparations proced ures are required to be completed .

Core cooling will be maint ained by the steam generator s with water being supplied by high pressure fire

protection with the RHR system removed from service.

D. Both Stage 1 and Stage II Preparation s procedures are required to be completed.

Core cooling will be maintained by the Spent Fuel Pool cooling system cross tied with the RHR system .

DISTRACTOR ANALYSIS

a. Incorrect. Plausible, since me thod of maintaining core cooling is correct.

b. Incorrect. Plausible, since Stage 1 preparation is part of the correct action. AOI- 7.01, Attachment 2

Step 15 refers the operator to Appendix E, which will cross connect the RHR and SFPC systems if in

open mode cooling configura tion.

c. CORRECT. Per given reference.

d. Incorrect. Plausible, since the first part is correct. AOI-7.01, Attac hme nt 2, Step 15 refers the operator

to Appendi x E, which will cross conne ct the RHR and SFPC systems if in ope n mod e cooling

configuration.

22 of 29

SRO Watt s Bar 2008 NRC Initial Licens e Exam

WRITTEN QUESTION DA TA SHEET

Question Number: ~{;71

KIA : G2.1.23

Ability to perform specific system and integrated plant proced ures duri ng all modes of plant ope ration.

Tier: 3 RO Imp: RO Exam : Cognitive Level: Low

Group : SRO Imp: 4.4 SRO Exam: Yes Source: NEW

Applicable 10CFR55 Section : 4 1.10, 43.5

Learning Objective : 3-0 T-SYS079A, Rev . 6, Objective 4: Identify the maximum quantity of fuel that shall

be out of approved storage loca tions during fuel handling ope rations .

References : FHI-7, Fuel Handling and Movement , Rev 0034.

Question :

Which ON E of the following satisfies the requirement of FHI-7 for the maximum number of fuel assemb lies

allowe d outside of approved storage?

A. Two unirradiated assemblies within the fuel-handling area.

B. Two irradiated assemblies within the spent fuel storage pool bounda ry.

C. One assembly in the transfer cart, two assemblies in the RCCA fixture and one asse mbly in the

refueling machine mast over its proper location in the reactor vessel.

D. One assembly in the transfer cart, two assem blies in the RCCA fixture and one assemb ly in the

refueling mac hine mas t over the reactor side upender.

DISTRACTOR ANALYSIS

a. Incorrect. Only one unirr adiated asse mbly can be outside of approve d storage within the fuel -handli ng

area.

b. Incorre ct. Only one irradiated assembly can be outside of approved storage within the spent fuel

storage poo l boundary.

c. CORRECT. Per FHI-7, the assemb ly in the mast and over its proper location in the reactor vessel is

allowed beyo nd the three allowe d within the refueling canal; i.e., this answer spec ifies that three

assemblies are within the refueling canal, two can be in the RCCA change fixture and one in the

trans fer cart.

d. Incorrect. The assemb ly in the refueling mast over the upender must be includ ed in the 3 assem blies

allowed within the refuel canal which would result in 4 assemb lies outside approved storage in the

refuel canal, thus exceeding the three allowed in the refuel ing canal area .

23 of 29

SRO Watts Bar 2008 NRC Initial License Exam

WRITTEN QUESTION DA TA SHEET

Question Number: ~ 4~

KIA: G2.1.34

Knowledge of primary and secondary plant chemist ry limits.

Tier: 3 RO Imp: RO Exam: Cognitive Level: Low

Group : SRO Imp : 3.5 SRO Exam: Yes Source: NEW

Applicable 1OCFR55 Section : 41.10, 43.5

Learning Objective: 3-0 T-T-S0304, Objective 4: Given plant conditions and parameters correctly

determine the applicable Limit ing Conditi ons for Operations or Technical Requi remen ts for the various

components of the RCS .

References: CM- 3.01, "System Chem istry Speci fications", Appendix A, pp. 1, 2, 25, and 26; rev 72

Tech Requirement 3.4.4, Chemistry.

Question :

W iltl the plant at full power, the Chemistry Lab has just informed the Unit Super visor that the RCS chloride

level is 1650 ppb and that the SG chloride level is 200 ppb. The source of any impuri ty ingress has NOT yet

been identified.

Based on the reported values, (1) what is the MOST restrict ive Action Level that must be entered and (2) the

impact on plant opera tions caused by the Action Level entry?

ill

A. Action Level 3 for RCS chloride level Initiate actions to take the reactor sub- critical as

quickly as practicable in a controlled manner and

reduce RCS temp erature to < 250" F.

B. Actio n Level 3 for SG chloride level Initiate actions to take the reactor sub-critical as

quickly as practicable in a controll ed manner and

reduce RCS temperature to < 250 " F.

C. Action Level 2 for RCS Chloride level Restore parameter to below Act ion Lev el 1 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

or reduce reactor power to less than 5%.

D. Action Level 2 for SG Chloride level Restore parameter to below Action Level 1 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

or reduce reactor power to less than 5%.

DISTRACTOR ANALYSIS

a CORRECT. Action Level 3, which is the most restrictive (requiring plant shutdown ), applies due to

RCS chloride exceeding 1500 ppb .

b. Incorrect. Candida te incorrectly believe s that Action Level 3 applies for SG chloride levels, when it is

actually Action Level 2 (SG chloride exceeds 50 ppb, but does not exceed 250 ppb . Plausible because

the impact on plant operations of a chemistry param eter being aut-of-specification is correct.

c. Incorrect. Candidate fails to recognize Action Level 3 conditions for RCS chlo ride . However, the

distractor has plausibility because there is similar impact on plant oper ations if there were Action Level

2 conditions for RCS chloride.

d. Incorrect. Plausible, since the Action Level for the given SG chloride value is correct. Further

plausibility is added since there is a time value given to restore the parameter. Howev er, it differs from

the correct time (within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months ) by a factor of 4. The requirement to be at less than 5% power is

incorre ct.

24 of 29

SRO Watts Bar 2008 NRC In it ial License Exam

WRITTEN QUESTION DA TA SHEET

Question Number: 7f Cffv

KIA : G2.2.40

Ability to apply Technical Specifications for a system.

Tier: 3 RO Imp: RO Exam: Cognitive Level : High

Group : SRO Imp: 4 .7 SRO Exam: Yes Source: NEW

Applicable 10CFR55 Section : 41 .10/43 .2/43 .5 /45 .3

Learning Objective: 3-0T-TS0307 , Rev 3, Objecti ve 5: Given plant conditions and param eters, determine

applicable Action Condi tions, Requ ired Actions, and Completion Times associated with diffe rent Plant

Systems . Objec tive 2: Determ ine the Bases for each speci fication, as applicabl e, to Plant Systems.

References: TS 3.7 .5 and Bases.

Question :

Given the follow ing

Unit 1 at 100% power.

Diesel Generator (DG) 2A-A is to be rem oved from service for a mainten ance outage with a

planned out of service of 102 hours0.00118 days 0.0283 hours 1.686508e-4 weeks 3.8811e-5 months .

Which ONE of the following identifies Unit 1 equipment that is listed in Tech Spec 3.8.1, AC Sources

Operating, Bases Contingency Actions, as equipment that is to remain in service conc urrently dur ing the DG

2A-A outage maintenance to be in comp liance with the Tech Spec?

A. Reactor trip breake r A (RTA ).

B. TDAFW pump .

C. RHR Pump 18 -B.

D. Any S/G AFW level control valve .

DISTRACTOR ANALYSIS

a. CORRECT. Tech Spec 3.8.1 Bases Table 3.8.1-2, TS Ac tion or Surveillance Table (SR) Contingenc y

Actions, Item 4 states "Do not rem ove reactor trip breakers from service con currently during planned

DG outage maintenance".

b. Incorrect. The Bases Table item 5 states "Do not rem ove the turbine-driven auxiliary feedwater (AFW )

pump from service concurrently with a Unit 1 DG outage ". Plausible if the candidate misapp lies the

item to the Unit 2 DG outage.

c. Incorrect. The Bases Table item 7 states "Do not rem ove the op posite train resid ual heat remova l

(RHR) pump from service concurrently with a Unit 1 DG outage ". Plausibl e if the candidate misapp lies

the item to the Unit 2 DG outage .

d. Incorrect. The Base s Table item 6 states "Do not remove the auxiliary feedwater level control valves to

the steam generators from service concurrently with a Unit 1 DG outage". Plausible if the candidate

misapplie s the item to the Unit 2 DG outage .

25 of 29

SRO Watts Bar 2008 NRC Initial License Exam

WRITTEN QUESTION DATA SHEET

Qu estion Number: ..2'2'1l

KIA : G2.3.6

Ability to approve release perm its.

Tier: 3 RO Imp : RO Exam : Cognitive Lev el : Low

Group: SRO Imp : 3.8 SRO Exam : Yes Source: NEW

Applicable 10CFR55 Section : 41 .13 I 43.4 14 5.10

Learning Objective: 3-0 T-SYS077B, Objec tive 10: Describe the ge neral procedure to make a gaseo us

release.

Ref er en ces : SOI -77 .02 , Waste Gas Disposal System, Rev 0034.

Question:

Which ONE of the follow ing ide ntifies BOTH the minimum decay time requi red to allow the co ntents of a Gas

Decay tan k to decay prior to release and who ca n waive the minimum time in acco rdance with SO I-77 .02 ,

Waste Gas Disposal system ?

Decay Time Required Who Can Waive

A. 60 days Chemistry Duty Manager

B. 60 days Radiation Protection Manage r

C. 8 days Chemis try Duty Manager

D. 8 days Radiation Prot ection Manage r

DISTRACTOR ANALYSIS

a. CO RRECT . The procedur e requ ires a 60 day decay time and doe s provid e for waiving of the time by

the Chemi stry Duty Man ager .

b. Incor rect. The decay time required is 60 days but the Radiation Protect ion Mana ger is not the position

that ca n wa ive the requ irement if earlier release is requir ed. Plausible becau se the req uired time is for

radioactive deca y wh ich cou ld be addres sed by RadCon .

c Incorrect. The decay time requ ired is 60 days, not 8 days, but the wa iving of the requirement by the

Chemistry Duty Manager is allowed by the proc edu re. Plausible because 8 days is identified in the

ODCM for Gaseous Efflu ents as being the half life of ce rtain radionu clides that set dose rate limit s at

and beyond the Unr estri cted Ar ea Bound ary .

d. Incorr ect. The deca y time requ ired is 60 days , not 8 days and the wa iving of the requ irem ent by the

Radiation Prote ction Manager is not allowed by the procedure. Plausible beca use 8 days is ide ntified

in the ODCM for Gaseous Efflu ents as being the half life of certa in radionuclides that set dos e rate

limits at and beyond the Unre str icted Area Boundary and because the required time is for rad ioac tive

decay which could be addressed by RadCon .

26 of 29

SRO Watts Bar 2008 NRC Initial License Exam

WRITTEN QUESTION DA TA SHEET

Question Number: ~ 1g

K/A : G2.3.15

Knowledge of radiation monitoring systems, such as fixed radiation monitor s and alarms, portable survey

instruments , personnel monitoring equip ment, etc .

Tier : 3 RO Imp: RO Exam : Cogn itive Level: High

Group: SRO Imp: 3.1 SRO Exam: Yes Source : NEW

Applicable 10CFR55 Section: 41 .12,43.4

Learning Objective: 3-0 T-SYS090A, Rev. 13, Objecti ve 8: Regarding Technica l Specifications and

Technical Requirements for this system: Explain the Limiting Conditions for Operation, Applicability, and

Bases .

References : SOI-90 .04, Section 3.0.B , and Section 5.1.(3], Rev. 6.

LCO 3.3.8, Table 3.3.8-1, includ ing Basis; Drawing 45N600-30-4.

Question :

Given the following plant conditions:

Unit 1 is in Mode 4 with preparations being made for fuel moves.

As part of perform ing the Channel Opera tional Test (COT) for 0-RM- 90-102, Spent Fuel Pool Pit

Area Monitor , a source check is to be performed.

What is the effect of performing the source check portion of this test , and what is the SRO 's respo nsibility for

Tech Spec/LCO Tracking Sheet entry?

A. This will result in an automatic actua tion of Train A of ABGTS . The SRO will make an entry on the

LCO Tracking Sheet that Train A of ABGTS is inoperable.

B. RM-90-102 outpu t will be blocked during the source check, the SRO will make an entry on the LCO

Tracking Sheet that Train A of ABGTS is inoperable.

C. RM-90-10 2 output will be blocked during the source check , the SRO will make an entry on the LCO

Tracking Sheet that RM-90-102 is inope rable.

D. This will result in an automatic actuation of Train A of ABGTS . The SRO will make an entry on the

LCO Tracking Sheet that RM-90-102 is inoperable.

DISTRACTOR ANALYSIS

a. Incorrect. Both parts are incorrect, but plausible, because of the close link between ABG TS (Auxiliary

Building Gas Treatment System ) operation and RM-90-102 (causes actu ation of AB GT S). However,

cand idate incor rectly believes that the COT causes an automatic actuat ion of ABGTS.

b. Incorrect. Candidate correctly understands that the COT requires blocking of the outpu t of RM-90-102,

which preven ts it from actuat ing ABG TS, and concludes therefore, that the associated train of AB GTS

is inoperable. This adds to the plausibilit y of this distractor, but it is incorrect, since a different LCO

(LCO 3.7.12) governs the operability of the ABGTS train itself.

c. CORRE CT . This test requires blocking of the output of RM-90-102, which prevents it from auto

actuat ing the assoc iated train of ABGTS . This is done on purpo se, to preven t inadvertent/undesired

actuation of ABGTS . Further , this blocking renders the RM-90-102 inoperable.

d. Incorre ct. Candidate correctly concludes that this test renders RM-90-10 2 inoperable, but for the

wrong reason. It is plausi ble, since there are other components in the plant, that if running, or

actuated, are considered inoperable. However, candidate incorrectly believes that this test causes an

auto actuation of the associated train of ABGTS.

27 of 29

SRO Watts Bar 2008 NRC Initial License Exam

WRITTEN QUESTION DA TA SHEET

Question Number: ~<f1

K/A: G2.4 .23

Knowledge of the bases for prioritizing eme rgency procedure implementation during emergency operations.

Tier: 3 RO Imp: RO Exam : Cogn itive Level : Low

Group : SRO Imp :4.4 SRO Exam: Yes Source: Modified Bank

Applicable 10CFR55 Section : 41.10, 43.5, 45. 13

Learning Objective: 3-0T-EOP0 300 Objec tive 5: Given a set of plant cond itions, use E-3, ES-3.1, ES-3.2,

and ES-3.3 to correc tly diagnose and implement: Action Steps, RNOs , Foldout Pages, Notes and Cautions;

Objec tive 6. Explain the basis for cooling the RCS to a target incore temp prior to depressuization of the

RCS .

References: E-3, Steam Generator Tube Ruptur e, Rev. 22.

Question :

Given the following:

Unit 1 exper iences a Safety Injection due to a steam generator tube rupture on SG #1 .

All Reactor Coo lant Pumps were remove d from service due to loss of sup port systems .

RCS cooldown at maximum rate to targe t incore temperature is in progress.

The STA reports that a RED path for FR-P.1 , Pressurized Thermal Shock, exists on RCS Loop 1

on the FR-O, Status Trees.

Which ONE of the following identifies when the transition to FR-P.1should be made?

A. Immediately transition to FR-P .1 from E-3 because FR-P.1 provides actio ns to limit cooldown and

repressurization of the RCS.

B. Remain in E-3 until the cooldow n is complete and then transition to FR-P.1 only if the RED path still

exists because the coo ldow n is needed to allow depressuriza tion of the RCS.

C. Remain in E-3 until the cooldown is complet e and then transition to FR-P .1 even if the RED path no

longer exists because the cooldow n is needed to allow depressurization of the RCS.

D. Remain in E-3 until the safety injection is term inated, then transition to FR-P.1 only if the RED path still

exists because FR-P .1 provides actions to limit cooldown and repressurization of the RCS .

DISTRACTOR ANALYSIS

a. Incorrect. Plausible , since the normal rules of usag e would require implementation of FR-P .1 as soon

as the condition was confi rme d to exist.

b. Incorrect. Plausible since a stagnant loop condition will exist in the loop associated with the ruptured

SG with the RCPs off. CAUTION prior to the cooldown step in E-3 states that the anticipated red or

orange path on FR-P.1 does not requ ire implem entation until after SI termi nation has been

accomplished.

c. Incorrect. Plaus ible since a stagnant loop condition will exist in the loop assoc iated with the ruptur ed

SG with the RCPs off. CAUTION prior to the cooldown step in E-3 states that the antic ipated red or

orange path on FR-P.1 does not require implementation until after SI termin ation has been

accom plished .

d. CORRECT. A stagnant loop condition will exist in the loop associated with the ruptured SG with the

RCPs off. CAUTION prior to the cooldown step in E-3 states that the anticipated red or orange path

on FR-P.1 does not require implementation until after SI termination has been accomplished. E-3

Note prior to Step 43 directs the operator to begin reevaluating for PTS/Cold Overp ressure conditions.

An evaluation of PTS is delayed to this point since the active SI wou ld contribute to the erroneous

red/orange path .

28 of 29

SRO Watts Bar 2008 NRC Initial Li c ense Exam

WRITTEN QUESTION DA TA SHEET

Qu est ion Number: ~ iDO

KIA: G2.4.29

Knowledge of the emergency plan.

Ti er : 3 RO Imp : RO Exam : Cognitive Level : High

Group : SRO Imp : 4.4 SRO Exam : Yes Source: NEW

Applicab le 10CFR55 Section : 43.5, 45.11

Learning Objective : 3-0T-PCD -048C, Objecti ve 7: Identify Opera tion's responsibilities for the following

emergency response posit ions : Site Emergency Director (who is initially the SM).

References: EPIP-1, 3.3.7 , rev. 29 , SPP-3.5, Appendi x A, 3.1, Rev . 19.

Question :

Given the following condi tions :

The plant is at full power.

A report was received of a tornado being sighted over the Watts Bar Training Center, moving in a

northwest directio n. The tornado continue d to move across Highway 68 and then dissipa ted

without touching dow n onsite.

The MCR crew has just entered AO I-8, Tornado Watch or Warning.

Confirm ation was received that no visible damage had been received to any struc tures or

equipment on site.

The Shift Manager evaluates the Radiologica l Emergency Plan (REP) and determines the

conditions for an NOUE were initially met but are now fully resolved.

Whic h ONE of the followi ng identi fies the ODS and NRC notification requirements in accordance with the

REP?

ODS notif ication NRC Notification required wi th in :

A. Repo rt but not declare 15 minutes.

the event.

B. Repo rt but not declare 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

the event.

C. Declare and terminate the 15 minutes .

event at the same clock time.

D. Declare and term inate the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

event at the same clock time .

DISTRACTOR ANALYSIS

a. Incorrect. Plausible , since a declaration is not made , and a report is required, but candid ate

incorrectly believes that it needs to be made within fifteen minutes.

b. CORR ECT. Per EPIP-1, for events that are totally resolved prior to declaration, no declaration shall

be made : howeve r, a report to the NRC within one hour is required.

c. Incorrect. Plausible, since the NRC must be notified, and since cand idate may fail to recognize

conditi ons have totally reso lved and therefore no declara tion is to be made .

d. Incorrect. Plausible, since the notification to the NRC time is correc t. Cand idate fails to recogn ize

that these conditions have totally resolved prior to declara tion , and there fore that NO declaration

should be made.

29 of 29

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