Text

NRC Operator Licensing Exam Rept 50-298/99-301 (Including Completed & Graded Tests) for Tests Administered on 990212. Exam Results:Four Reactor Operator License Applicants Passed Exam
	ML20205A593
Person / Time
Site:	Cooper
Issue date:	03/09/1999
From:	; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	;
Shared Package
ML20205A573	List: IR 05000298/1999301; ML20205A567; ML20205A581;
References
	50-298-99-301, NUDOCS 9903310024
	Download: ML20205A593 (108)
	v • d • e

"

-

~

ENCLOSURE U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket Nos.: 50-298

'

License Noo.: DPR-46 Report No.: 50-298/99-301- .

t Licensee: Nebraska Public Power District j

Facility: Cooper Nuclear Station j Location: P. O. Box 98 Brownville, Nebraska 68321 Dates: February 12,1999 Inspector: John L. Pellet, Chief, Operations Branch Approved By: Arthur T. Howell lli, Director l Division of Reactor Safety l

ATTACHMENTS: l Attachment SupplementalInformation Attachment 2: Final Written Examination and Answer Key i I

l I

i l

9903310024 990322 PDR ADOCK 05000298 V PDR E

.

-2- j

,

EXECUTIVE SUMMARY q

!

Cooper Nuclear Station NRC Inspection Report 50-298/99 301 NRC examiners evaluated the competency of four reactor operator license applicants for ' I issuance of operating license at the Cooper Nuclear Station. The licensee developed tha retake written examination using NUREG-1021, Interim Revision 8, January 1997. NR(

examiners reviewed, and approved the examination. The written examination was admiinstered on February 12,1999, by facility proctors in accordance with the guidance in NUREG-1021, interim Revision Operations

The four reactor operator license applicants passed the examination (Section 04.1).

The test material was adequate for administration as submitted, with no post-examination changes identified (Section 05.1).

i l

I I

.

, .

3-Report Details Summary of Plant Status The plant operated at essentially 100 percent power for the duration of this inspectio LO_perations 04 Operator Knowledge and Performance 0 initial Written Examination Insoection Sggpe On February 12,1999, the facility licensee proctored the administration of the written examination approved by the NRC to the four individuals who had reapplied for a reactor operator license. The licensee proposed grades for the written examinations l and evaluated the results for question validity and generic weaknessesc The examiner {

reviewed the licensee's result l Observations and Findinas The minimum passing score was 80 percent. The candidates' scores for the written j examination ranged from 81 to 88 percent. The average score was 84.5. The 1 licensee's post-administration analysis identified that the administered examination was I acceptable. There were no post-examination comments or changes to the written j examination recommended by the licensee or identified by examiner review of the results. Licensee review of the examination results indicated five questions identified potential training deficiencies for further internal review. The specific questions were numbers 27,49,55,62, and 94. The examiner reviewed this analysis and the specific questions ano determined that the issues were narrowly focused within scattered systems, such as logic circuit design, control rod hydraulics' and protection system

,

bases, and did not represent generic weaknesse Conchisions l

!

The four applicants passed the written examination. No broad knowledge or training weaknesses were identified as a result of evaluation of the graded examinatio Operator Training and Qualification 0 Initial Licensina Examination Development The facility licensee developed the initial licensing examination in accordance with guidance provided in NUREG-1021,' Operating Licensing Examination Standards,"

Interim Revision 8, dated January 199 f.

+ .

<

-4 05.1.1- Examination Outline a Inspection Scope The facility licensee submitted the initial examination outlines on December 17,199 The chief examiner reviewed the submittal against the requirements of NUREG-1021, Interim Revision Observations and Findinas The chief examiner determined that the initial examination outlines satisfied NRC requirement . Conclusions The licensee submitted an adequate examination outlin .1.2 Examination Packaae Inspection Scope The draft examination was transmitted by the licensee to the NRC on January 21, 1999. The licensee submitted the completed final examination package on February 5, 1999. The chief examiner reviewed the examination against the requirements of NUREG-1021, Interim Revision Observations and Findinas The draft written examination contained 100 questions. The draft examination was technically valid, discriminated at the proper level, and responsive to the outline submitted by the licensee. The written examination was considered adequate for administration as submitted. The examiner developed clarification or enhancement comments on about one fourth of the questions. The licensee staffincorporated these comments in an excellent manner and all proposed enhancements were accepted by the examiner. In contrast to the prior examination (see NRC Inspection Report 50-298/98-301), the submitted examination demonstrated improved question stem focus and distractor credibility. The quality of the submitted examination is reflected in the ,

lack of postadministration changes, especially as compared to the prior examination, )

for which the license requested changes to over one third of the question Conclusions I The test material was adequate for administration as submitted by the hcensee. No changes to examination materials were required as a result of administratio M:-

.c

.Q-5-I V. Manaaement Meetinas l I

X1 Exit Meeting Summary j The chief examiner telephonically presented the inspection results to members of the licensee staff at the conclusion of the inspection on February 25,1999. The licensee '

acknowledged the findings presente !

The licensee did not identify as proprietary any information or materials examined during i this inspectio .

l

-< ,

!

'

0'

ATTACHMENT 1

1 SUPPLEMENTAL INFORMATION

'

PARTIAL LIST OF PERSONS CONTACTED Licensee H.- McDaniel, Lead initial Training Instructor

-

r NRC-

-

.None

'

INSPECTION PROCEDURES USED NUREG-1021 ,

NUREG-1021, " Operating Licensing Examination Standards," Interim Revision 8, January 1997 i

l l

!

<

,.

ATTACHMENT 2 FINAL WRITTEN EXAMINATION AND ANSWER KEY l l

l i

.

l l

l

!

i i

i

r s

NRC CNS Site-Specific j Written Exandnation (RO)

Master Key ,

i

.

-

I i

!

i

!

'

I i

)

,

RO Written Examination Question No.: 1 -

K/A: 288000, K5.02 Importance: Tier: 2 Group: 3 Cognitive Level: 1 Exam Bank No.: new Reference: COR001-08-02,2.2.47 Objective: COR001-08-02, obj.13b Which one of the following describes how the Reactor Building Ventilation System maintains the required 0.25 inches of negative water pressure in the Reactor Building during normal operation of the system? At least one (1) more exhaust fan than supply fan is operate A d/p controller regulates the operating supply fans vortex damper positio !

i A d/p controller regulates the operating exhaust fans vortex damper positio The capacity of the exhaust fans is greater than the capacity of the supply fan )

Answer: c. capacity is not used to maintain d/p d/p controller on suction dampers maintain flow through the filters not in accordance with system design l

l

.

RO Written Examination Question No.: 2 K/A: 261000 A4.02 Importance: Tier: 2 Group: 1 Cognitive Level: 1 Exam Bank No.: new Reference: 2.4.8.4.1,2.2.73, Sect. 3.3, & Objective: COR002-28-02,5.a, )

While operating at full power, a problem with the Reactor Building HVAC system has resulted in loss of Reactor Building Ventilatio WMeh one of the following is required to aid in improving the Reactor Building differential pressure using the "A" Standby Gas Treatment (SGT)?

~

Start the "A" SGT fan, ... verify SGT-AO-249, SGT A INLET, and SGT-AO-251, SGT A DISCHARGE automatically open onl manually open SGT-AO-249, SGT A INLET, and SGT-AO-251, SGT A DISCHARGE and then open AD-R-1B, PRIMARY CONTAINMENT ISOLATIO manually open SGT-AO-249, SGT A INLET and SGT-AO-251, SGT A DISCHARGE, and then open HPCI-AO-275, HPCI GLAND EXHAUST DISCHARGE TO SG verify SGT-AO-249, SGT A INI,ET, and SGT-AO-251, SGT A DISCHARGE automatically opens and then open AD-R-lC, STANDBY GAS TREATMENT ROOM SUPPL Answer l SGT-AO-249 & 251 should automatically open, AD-R-1B, PRIMARY CONTAINMENT l

ISOI.ATION would take a suction on a dead leg of pipe, l SGT-AO-249 should automatically HPCI-AO-275, HPCI GLAND EXHAUST DISCHARGE TO SGT would take a suction on a dead leg of pip AD-R-lC is failed open

.

i

)

.

l RO Written Examination Question No.: 3 K/A: Generic 2. Importance: )

Tier: 3 Group: N/A Cognitive Level: 1 Exam Bank No.: new Reference: SKL008-01-02, Watchstanding Objective: SKL008-01-02, obj.12,13 Principles (RO) SKL010-01-02, .1, Section 3.5,3.6 While placing the "A" SGT train in service to support HPCI surveillance testing, the Control Room Operator recognizes that the procedure step identifies the control switch to open SGT FLOW /RX BLDG DP CONT, as SGT-DPCV-546B, rather than SGT-DPCV-546A.

Which one of the following is required for this identified condition? Complete the start of the "A" SGT train, then make a pen and ink correction to the !

procedure. Notify the procedure owner after placing SGT in servic !

I Stop action, make a pen and ink correction to the procedure, then proceed with starting the "A" SGT train. Notify the procedure owner after placing SGT in servic Discontinue actions to start the "A" SGT train and place SGT back in standby. The procedure shall be revised using the procedure change process prior to placing SGT in servic Discontinue actions to start the "A" SGT train and leave all SGT components in their ,

current position / condition. The procedure shall be revised using the procedure !

change pncess prior to placing SGT in service.

Answer; b. Work s' .aontinue for obvious typographical errors, spelling errors, title changes, procedu.e number changes. Corrective act!on for the procedure is required prior to proceeding. Procedure change prior to implementation is only required if the procedure is adversely affected. SGT would be aligned to standby if procedure was terminated. Procedure change prior to implementation is only required if the procedure is adversely affecte .-

_

. . . .

,

RO Written Examination Question No.: 4 K/A: 295034 A1.03 Importance: Tier: I Group: 2 Cognitive Level: 1 Exam Bank No.: new Reference: 2.4.8.4.1, Objective: COR001-08-02,11.b,1 .1.22, Section In accordance with General Operating Procedure 2.1.22, " Recovering From A Group Isolation,"

which one of the following methods is used to ensure that the Secondary Containment has ISOLATED 7 Verify NO flow is indicated on the reactor building exhaust flow recorde Verify ALL supply fan discharge valves and ALL exhaust fan discharge valves are closed using the valve position indicator Verify ALL reactor building supply and ALL reactor building exhaust fans trip and SGT starts by observing fan status light indicator Verify ALL supply fan discharge dampers are closed locally and verify ALL exhaust fans discharge dampers are closed by observing control room indication Answer: b. Supply fan valves and exhaust fan discharge rales are the components used to isolate secondary containment. Dampers do not provide adequate isolatio Will not verify an isolation C tecking fan status does not ver:fy proper valve isolation has occurre nuse are the fan isolations, not the building isolations

_ _ _ _ _ _ _ _

RO Written Examination Question No.: 5 K/A:290001 A3.01 Importance: Tier: 2 Group: 2 Cognitive Level: 2 Exam Bank No.: new Reference: COR001-08-02 Objective:COR001-08-02, obj. I1b 2.1.22, Seedon .3.2.24 The unit is at 100% power. Irradiated fuel is being arranged in the fuel pool to support rec 'ft :(

new fuel when annunciator 9-4-1/E-4, RX BLDG VENT HI-HI RAD is received.

RONAN CRTs display the following:

-

(1763) RX BLDG VENT MONITOR A HI-HI RAD

-

(1764) RX BLDG VENT MONITOR B HI-HI RAD

-

(1780) RX BLDG VENT MONITOR D HI-HI RAD Which one of the following describes the effect on the Secondary Containment and why?

The Reactor Building HVAC supply and exhaust fans ... trip and the system isolates because a Group 6 isolation is actuate trip and the system isolates because a Group 7 isolation is actuate continue to operate and the system does NOT isolate because the DIV I ogic has NOT trippe continue to operate and the system does NOT isolate because the DIV II logic has NOT trippe j Answer: !

I Group 2 isolation is high drywell pressure and low reactor water level c, d Both DIV I and DIV II are above the setpoint for the group 6 isolatio A group 6 isolation is actuated and the fans trip and the system isolate RO Written Examination Question No.: 6 K/A:295012 A1.02 Importance: ,

Tier: 1 Group: 2 Cognitive Level: 2 Exam Bank No.: new Reference: EOP-3A,5.8.10 Objective: COR002-03-02,13.d,13.e,15.a.1, 1 l A small break LOCA has occurred with the following conditions:

-

Reactor water level +45" indicated on the narrow range instruments

- Reactor pressure 560 psig

-- Drywell pressure 3.1 psig

-

Drywell temperature 195*F

- Primary containment water level 14 feet l

-

FCU control switch positions RUN Which one of the following actions (if any) will operate ALL available drywell FCUs and maintain them running? NO actions are required, all FCUs are mnnin Start all the FCUs by placing their control switches in OVERRID Start all the FCUs by placing their control switches in OVERRIDE and then RU Installjumpers to bypass the high drywell pressure signal and place all the FCUs control switches in RUN or OVERRID Answer: a. FCUs have tripped c. If the control switches are placed in RUN the FCUs will trip

.d. Nojumpers are required

-

__

.1 RO Written Examination Question No.: 7 K/A:223002 A3.02 Importance: Tier: 2 Group: 1 Cognitive Level: 2 Exam Bank No.: new Reference: COR002-03-02 Objeuive: COR002-03-02, Sa, 6a, 6b, 6d, 6f, 2.1.22, Section j 2.2.33, Section .9, Section Given the following parameters:

-

Drywell pressure is 12.7 psig

-

RPV water level lowered to -50 inches on the wide range RPV level instrument

-

Main Condenser vacuum has degraded to 14"Hg

-

Reactor pressure is 85 psig Which one of the following describes equipment that isolates under these conditions? MSIVs and Main steam line drains, HPCI steam supply line, RWCU isolation valve RWCU isolation valves, Drywell floor and equipment drains, HPCI steam supply I lin , RCIC steam supply line, Recirculation loop sample valves, MSIVs and Main steam line drain Recirculation loop sample lines, RCIC steam supply line, Drywell Floor and i Equipment Drain valves Answer: Group 1 isolation is not met, Group 1 and 5 isolations are not me Group 5 isolation is not met.

!

7

,

RO Written Examination Question No.: 8 K/A: 215001 Kl.05 Importance: Tier. 2 Group: 3 Cognitive Level: 1 Exam Bank No.: new Reference: COR.002-31-02,2.1.22, Objective: COR002-31-02, obj. 9e Which one of the following describes the design response of a TIP detector that is in the reactor core when a Group 2 and a Group 6 isolation signal is received? Group 2 Isolation will cause the TIP to withdraw Group 6 closes the ball valv !

1 Group 6 Isolation will cause the TIP to withdraw. Group 6 closes the ball valv l Group 2 Isolation will cause the TIP to withdraw. Group 2 closes the ball valv ! Group 2 DE Group 6 Isolation will cause the TIP to withdraw and close the ball i valve.

Answer: c.

a,b,d The group 2 isolation will cause a group 6 isolation, however, a group 6 isolation has no efTect on TIP _

h o

,

RO Written Examination Question No.: 9 K/A: 211000 K4.08 Importance: 4.2*

Tier: 2 Group: 1 Cognitive Level: 1 Exam Bank No.: new Reference: COR002-29-02,2.2.74 step 11. Objective: COR002-29-02,5.g, The keylock switch for Standby Liquid Control (SLC) Pump "A" is turned to the START positio Aside from starting the "A" SLC pump, what else will this switch movement initiate? Only the "A" squib valve fires, only RWCU-MO-15 isolate Both "A" and "B" squib valve fires, only RWCU-MO-18 isolate Only the "A" squib valve fires, both RWCU-MO-15 and RWCU-MO-18 isolate Both "A" and "B" squib valve fire both RWCU-MO-15 and RWCU-MO-18 isolat Answer: Only the "A" squib fires, RWCU-MO-15 closes, RWCU-MO-18 does not cir s Only the RWCU-MO-15 isolate Only the "A" squib fires, Only the RWCU-MO-15 isolate RO Written Examination Question No.: 10 K/A:268000 A4.01 Importance: Tier: 2 Group: 3 Cognitive Level: 2 Exam Bank No.: new R.eference: COR001-11-02,2.3.2.25 Objective: COR001-11-02, obj. 2d,5 With the unit operating at 80% reactor power, the following annunciators are received:

- . 9-4-2/D-1, DRYWELL EQUIP. SUMP G HIGH LEVEL

-

9-4-2/C-1, DRYWELL EQUIP. SUMP G HI-HI LEVEL

-

9-4-2/B-1, DRYWELL EQUIP. SUMP G HIGH FILL-UP RATE

-

9-4-2/A-1, DRYWELL EQUIP. SUMP G HIGH TEMP

,

The last drywell equipment sump pump to operate was pump 1-G-1.

l

'

Which one of the following describes the response of the Drywell Equipment Drain system?

i l

l Only the 1-G-1 pump is operating and the Sump "G" totalizer value is changing.

Water is NOT recirculated through the heat exchanger.

l Only the 1-G-2 pump is operating and water is recirculated through the heat exchanger. The Sump "G" totalizer value is NOT changin Both the 1-G-1 and the 1-G-2 pumps are operating and water is recirculated through the heat exchanger. The Sump "G" totalizer value is NOT changin Both the 1-G-1 and the 1-G-2 pumps are operating and the Sump "G" totalizer value is changing. Water is NOT recirculated through the heat exchanger.

l Answer: c.

l Both pumps start when the hi-hi level is received. The recirculation valve opens and the i discharge to Radwaste closes on high temperatur ,

' Both pumps start when the hi-hi level is reache The recirculation valve opens and the discharge to Radwaste closes on high temperatur i

i i

I i

I

t

>

.

RO Written Examination Question No.: 11 K/A: 295036 K2.01 Importance: Tier: 1 Group: 3 l Cognitive Level: 2 Exam Bank No.: new Reference: 2.2.27,2.3.2.20, S-1/A-1, B&R- Objective: COR002-03-02,4,1 COR001-11-02,5 l

A 300 gpm leak on the "A" RHR heat exchanger has resulted in flow into the NW quadrant sump

"l A." As the level rises the following annunciator is received:

- S-1/A-1, REACTOR BLDG A SUMP HI HI LEVEL Which one of the following describes how this will effect water level in secondary containment? The sump level will rise and the sump will overflow until the heat exchanger is isolate The discharge valve from the sump will isolate, level in the NW quad will rise at a faster rat The isolation valve on the inlet into the sump will close, directing water from the RHR heat exchanger into the torus are The second sump pump will start. Provided flow to the sump remains constant, sump level will be maintained at the current valu Answer: l l

i i Sump level should lower with the inlets isolate ! The discharge from the sump is NOT isolated, sump level should lowe l The lines into the sump are isolated so level should eventually lower as the water is pumped from the sump and water is directed into the torus are !

- __

RO Written Examination Question No.: 12 K/A: 209001 Kl.13 Importance: Tier: 2 Group: 1 Cognitive Level: 1 Exam Bank No.: new Reference: 2.3.?.23, COR002-06-02 Obj ective: COR002-06-02, 3.d, 5.d, 6.b, During operation at full power the following annunciator is received:

- 7-3-3/A-5, CORE SPRAY B BREAK DETECTION NO other annunciators alarm. A station operator is sent to the d/p indicating switch and reports that the d/p is oscillating at around 4.0 p ;i Which one of the following caused this indication? j i

i Water is leaking by the core spray injection check valv ] A break has occurred in the core spray sparger or core spray piping to the sparger inside the core shrou The core spray piping is broken between its' Outboard Injection valve and its'

Inboard Injection valve, There is a break in the core spray line within the reactor vessel between the core shroud and RPV penetratio Answer: This would also have to be leaking by the injection valves, this would cause an annunciator for high pressure valve leak,9-3-3/C- This would not cause a change in reading or the alar . A break in this location is isolated from the break detection instrumentation by the closed Inboard Injection valve.

<

_ _ _ - - _ _ _ _ _ _ _ - _ _ _ - _ -

-

>

.

RO Written Examination Question No.: 13 K/A: 400000 Kl.02 Importance: Tier: 2 Group: 2 i

Cognitive Level: 2 Exam Bank No.: new

'

Reference: COR002-19 02 Objective: COR002-19-02, obj. 6a, 6i, 6j 2.3.2.16, M-1/A-? Section 3 Which one of the following will cause a high water level in the Reactor Equipment Cooling (REC) Surge Tank? RHR pump seal cooler lea Tube leak in the REC heat exchange Tube leak in the "A" Service Air Compressor heat exchange Tube leak in the Reactor Water Cleanup non-regenerative heat exchange j Answer: Possible leakage sources are the RWCU NRHX, RWCU and Reactor Recirculation seal water coolers and Fuel Pool Cooling heat exchangers

,

k

-

RO Written Examination Question No.: 14 K/A: 205000 K6.08 Importance: Tier: 2 Group: 2 Cognitive Level: 1 Exam Bank No.: new Reference: COR002-27-02 Objective: COR002-27-02, obj. 6,4e 2.3.2.21, 9-3-1/A-3 SKL010-01-02, A.4, B1 2.4.2.4.1, Attachment 3, Section 1.1 The plant is shutdown with the following conditions:

-

"C" RHR Pump operating in Shutdown Cooling Mode

-

"A" RHR SWBP is operating

-

RPV Level is in the prescribed RPV level band for shutdown cooling operation The following annunciator is received:

- 9-3-1/A-3, RHR SWBP A TRIP Which one of the following actions is required to be taken? Manually start the "C" RHR SWB Throttle closed RHR-MO-66A, HX Bypass valv Verify thcane "C" RHR SWBP automatically start i Throttle open SW-MO-89A, HX-A SW Outlet valve.

Answer: a. Valve should be opened, No automatic start. During MODE 4 or 5 operations, a normal SW flow of 4000 gpm can be supplied to HX A or B at service water pressure with SWBPs A and C windmilling. This requires lifted lead {;

1 RO Written Examination Question No.: 15 K/A:295009 Kl.05 Importance: Tier: 1 Group: 1 Cognitive Level: 1 Exam Bank No.: new Reference: 2.4.2.4.1, Attachment 4 Objective: COR002-22-02, obj. Sh COR002-23-02, obj. 9d i

The reactor has been shutdown for 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months and is currently in Cold Shutdown (MODE 4).

A cooldown is in progress with reactor coolant temperature at 162 F. RHR Loop "A" is in Shutdown Cooling with both reactor recirculation pumps trippe A Group 2 isolation signal trips the RHR system and RHR CANNOT be restarte Which one of the following describes where RPV water level is required to be maintained for the current conditions and why7 At least +48 inches on the narrow range RPV water level instruments to promote I natural circulatio l At 0.0 inches on the wide range RPV water level instruments to support alternate heat removal using RWC c. ' Flooded (solid) on the shutdown range RPV water level instruments to support alternate heat removal using the SRV Between +27.5 inches and +42.5 inches on the narrow range RPV water level instruments to minimize thermal stratification in the reactor pressure vesse Answer: J Water level is not high enough to support this method of heat removal Not an approved method of heat removal

- Circulation is needed to minimize thermal stratification i

l

-

RO Written Examination Question No.: 16 K/A:295021 Generic 2. , Importance: Tier: 1 Group: 3 Cognitive Level: 2 Exam Bank No.: new .

Reference: AP 2.4.2.4.1, Attachment 5 Objective: COR002-23-02,7.a The reactor is in Cold Shutdown with RPV metal temperatures at 150 F. The reactor has been shut down for 33 hours3.819444e-4 days 0.00917 hours 5.456349e-5 weeks 1.25565e-5 months with the recirculation pumps secured. RPV water level is just below full i scale on the Narrow Range indicators. The RPV head vents are OPE l A loss ofshutdown cooling occurs.

Assume NO operator actions are taken.

Which one of the following is the approximate time (in hours) the reactor will remain in Cold Shutdown if shutdown cooling CANNOT be restored? I hour .4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months Answer; l hours after shutdown with reactor water level at RPV flange curve days after shutdown with reactor water level at high level trip curve days after shutdown with reactor water level at RPV flange curve Attachments: 2.4.2.4.1, Attachment 5 (all of the attachments)

!

i

16

f I

,

l j

l l

RO Written Examination Question No.: 17 K/A: 204000 Kl.01 Importance: Tier: 2 Group: 2 Cognitive Level: 2 Exam Bank No.: new Reference: COR001-20-02,2.2.6 Objective: COR001-20-02, obj. 4k, 7f, 7h The unit is in MODE 2 with a startup in progress. Reactor pressure is being maintained at 300 psig using the main turbine bypass valves. The "A" reacter recirculation pump trips and than a Group 3 isolation signal is receive Assume NO operator action is take !

Which one of the following describes the consequence on the plant? I Reactor water level will rise outside the allowed band, Reactor water level will lower and a reactor scram will be receive A prerequisite for the "A" reactor recirculation pump start CANNOT be determine RWCU non-regenerative heat exchanger outlet temperature will rise damaging the demineralizer resi Answer: Reactor water level will rise but will be within the required band (a shutdown is not required based on reactor water level) Reactor water level will ris Temperature willlowe l L

!

RO Written Examination Question No.: 18 K/A: 202001 K4.16 Importance: Tier: 2 Group: 2 Cognitive Level: 2 Exam Bank No.: new Reference: 2.2.68, COR002-22-02 Objective: COR002-22-02, 5.d, 6.I,10.1,1 During a plant startup, the "A" recirculation pump trips causing the following conditions:

-

Reactor power is 39%

-

"B" recirculation pump is operating

-

Both recirculation MG sets M/A transfer stations are in MANUAL set at 57 % deman What is the expected scoop tube speed position on the "A" recirculation pump 4 minutes after the recirculation pump trip? (Assume operator actions for the tripped recirculation pump have been completed.) % % % %

Answer: a. c. d. Pump speed is limited by the dual limiter to 22% speed because the discharge valve is closed on the tripped ("A") pum i I

_

J RO Written Examination Question No.: 19 K/A: 202002 A3.03 Importance: Tier: 2 Group: 1 Cognitive Level: 2 Exam Bank No.: new Reference: COR002-22-02, 2.2.68, Attach 1 Objective: COR002-22-02,6.I,6j,10c,10j, step 2.1.4,2.2.68 step 5.8.2, 10n 2.3.2.26 9-4-3/D-1 The "A" Reactor Recirculation Pump is being started. The JOG BYPASS switch is the JOG-IN

, position. After the MG SET switch has been placed to START, the following events occur in sequence:

-

RRMG "A" drive motor breaker close seconds later, RRMG "A" drive motor breaker trips and locks out

-

The field breaker does NOT close during the start attemp Given the above panel indications why did the Reactor Recirculation pump fail to start? RRMG set room ventilation not in operatio Recirculation pump discharge valve failed to ope Recirculation pump suction valve is NOT fully ope Scoop tube positioner failed to ramp to the startup positio Answer: d. RRMG set trips and lockout on incomplete sequence This is not a recirculation pump trip, it is a breaker close permissive, b.- Discharge valve shall be partially open within one (1) minute and full open within two (2)

minutes from time drive motor breaker is closed to prevent pump from trippin This would prevent the drive motor breaker from closing 1 l

l t

,

RO Written Examination Question No.: 20 K/A: 295001 K2.01' Importance: _

Tier: 1 Group: 2 l

Cognitive Level: 2 Exam Bank No.: new Reference: 2.4.2.2.1,2.1.2 Objective: COR002-22-02, 4.h, 5.j, 6.c, 6.j, 7.k,1 A reactor startup in MODE 1 is in progress with reactor power at 15%. A spurious group 6 isolation occurs and all equipment operates as designed. When the isolation is reset, it is observed that the Reactor Recirculation Motor Generator (MG) set ventilation system CANNOT be re-started. Reactor Recirculation MG set internal air temperatures are as follows:

-

Reactor Recirculation MG set "A" motor air temperature is 276 Reactor Recirculation MG set "B" generator internal air temperature is 271 Which one of the following IMMEDIATE operator actions is required for the above plant conditions? Manually scram the reacto Stop power changes in progres Press the Recirculation MG set "A" Scoop Tube Lockout push butto Reduce RRMG speed as necessary to lower RRMG set temperatures below 210 l Answer: I Immediate operator action for 2.4.1.7, " Unexplained Decrease In Reactor Power." Action from 2.4.2.2.2, " Reactor Recirculation Flow Control System Failure."

d. ' Action from 2.3.2.26 alarm card 9-4-3/C-4. Not applicable due to RRMG sets being trippe _-

RO Written Examination Question No.: 21 i

K/A:295014 A2.03 Importance: )

Tier: 1 Group: 1 Cognitive Level: 2 Exam Bank No.: new Reference: 2.4.1.7, Section Objective: COR002-22-02,6b,6d,6h While operating steady state the following indications are observed:

- Reactor power lowers ,

-

Narrow Range reactor water level rises

-

Indicated core plate d/p lowers

-

Indicated core flow rises

-

- "A" and "B" recirculation loop flows rise Which one of the following failures caused the above conditions? One (1) of the Jet pumps has faile A shroud support access hole cover has faile One (1) recirculation pump's speed has raised to maximum.

, Flow through a control cell (four fuel bundles) has been blocke Answer: l Loop flows will only rise in one loop and reactor water level change would not be discernible, l Would not provide these indications This would lower core flow i

,

'

i RO Written Examination Question No.: 22 K/A: 295025 K3.02 Importance: Tier: 1 Group: 1 Cognitive Level: 1 Exam Bank No.: new l

Reference: 2.3.2.28,9-5-2/C-8 Objective: COR002-33-02,8c While the plant is operating at full power a Turbine trip causes RPV pressure to peak at 1095 l psi Which one of the following describes the effect on the Recirculation Pumps, including why? Both field breakers trip to insert negative reactivit Both field breakers trip to reduce the differential pressure across the steam drye Both drive motor breakers trip to prevent an overcurrent lockout of the Startup transformer, Both drive motor breakers trip to prevent over-pressurizing the Recirculation pump discharge pipin Answer: The field breakers do trip, however, the MSL flow restrictors limit the differential pressure across the steam dryer and dryer dp is not a basis for the field breaker tri The drive motor breakers do not trip. The drive motors may draw extra current at the elevated pressure, but this is not the basis for the field breaker tri The drive motor breakers do not trip. The discharge pressure will rise during the event, but the basis is not to ;>mtect piping l

>

t

!

i

!

,

RO Written Examination Question No.: 23 K/A: 215005 K4.07 Importance: Tier: 2 Group: 1 Cognitive Level: 2 Exam Bank No.: new Reference: COR002-01-02 Objective: COR002-01-02, obj. 8a, 8e,9c 4.1.3, Section .3.2.27, 9-5-1/F-4 The reactor is operating at 100% rated thermal power. The following instrument readings are observed:

-

APRM A 102 % -

APRM B 99 %

-

APRM C 100 % -

APRM D 99 %

-

APRM E 100 % -

APRM F 99 %

-

Flow Unit A 98% -

Flow Unit B 87 %

For the conditions above, which one of the following describes the effect on the plant if all trips and alarms occur at their design setpoint(s)? An APRM Upscale alarm and a control rod withdrawal bloc An APRM INOP alarm and a RPS Trip System B trip (% scram). A Flow Reference Off Normal alarm and a control rod withdrawal bloc , A Flow Reference Off Normal alarm and a RPS Trip System A trip (% scram).

Answer: i Reactor power is below the APRM rod block setpoin ' The requirements for an APRM Inop trip are not satisfie _ A flow reference off normal does not cause a scram. Reactor power is below the APRM scram setpoint.

If Flow Unit A if greater than flow unit B by more than 10%, comparator A trip unit trips causing a rod block, a FLOW REF OFF NORMAL annunciator, a COMPAR white indicating light on panel 9-5, a COMPAR amber light on the respective flow unit and on panel 9-14.

Scram setpoint is .58w + 61 = 111.46%

Rod Block setpoint is .58w + 49.5 = 99.96%

,

RO Written Examination Question No.: 24 K/A: 219000 A4.02 Importance: 3.7*

Tier: 2 Group: 2 Cognitive Level: 2 Exam Bank No.: new Reference: COR002-23-02 Objective: COR002-23-02, Obj. 3p, 5b 2.2.69.3, Section 4.0 The following conditions are present following a LOCA: I

-

Drywell pressure 12 psig and slowly rising

-

RPV level (Fuel Zone instrument) + 5 inches and slowly rising

-

Reactor pressure 20 psig

-

RHR Loop A and B Secured

)

-

CS Loops A and B Injecting following automatic initiation

-

Three (3) minutes have elapsed since annunciator RX LOW PRESS 291-436 PSIG alarme l The Control Room Operator is directed to place RHR Loop B into Suppression Pool Cooling.

For the current plant conditions, which of the following operator action (s) must be performed to manually close the RHR-MO-27B, Outboard Injection valve? Wait two (2) additional minute Depress the Loop B Initiation Logic Reset pushbutto Place the containment cooling valve control permissive switch is in MANUAL, Place the containment cooling permissive switch in MANUAL and the containment l cooling 2/3 core permissive switch in MANUAL OVERRID i Answer: a. 5 minute timer must time out, Cannot be reset with the current conditions. No input into MO-27B logi l No input into MO-27B logi l l

,

RO Written Examination Question No.: 25 1 K/A: 295016 K3.03 Importance: Tier: 1 Qoup: 2

._

!

Cognitive Level: 1 Exam Bank No.: inw Reference: 5.4. Objective: COR002-34-02, Emergency Procedure 5.4.3.2," Post-Fire Shutdown to Mode 4 Outside Control Room," requires j the ASD panel isolation switches be placed in the ISOLATE Positio j

!

Which one of the following describes the reason for this action? I To disconnect control room control circuit l To ensure automatic operation of ECCS remains availabl ' To isolate wire runs to meet divisional physical separation criteri i To prevent overloading the associated DG during a design basis LOC I l

Answer: l HPCI automatic features (except overspeed) are disabled when operated from the ASD panel.

l The ISOLATE position does not change the physical routing or location of equipmen These valves draw the same power from the ASD panel as from the control roo <

f I

I l

i b

.S RO Written Examination Question No.: 26 K/A:206000 K2.01 Importance: 3.2*

Tier: 2 Group: 1

_

Cogniti/e Level: 2 Exam Bank No.: new Reference: 2.2.33. Sect 2. Objective: COR002-11-02, 5.g, 5.h, 6.a, 6.c, 10.b With NO AUTOMATIC HPCI initiation signal present, which one of the following describes the effect that a loss of ALL AC power has on HPCI system operation and why?

HPCI can be started for RPV...

1 injection only. The Pressure Control Mode is unavailable because HPCI-MO-21, Test Bypass to ECST valve, CANNOT be opened due to loss of powe injection only. The Pressure Control Mode is unavailable because HPCI-MO-24,

'

ECST Test Line Shutoff valve, CANNOT be opened due to loss of powe level control and/or Pressure Control. The HPCI-MO-15, Steam Supply Inboard Isolation valve, will NOT re-position due to loss of powe level control or pressure control. The interlock between the HPCI-MO-19, Injection valve, and the HPCI-MO-24, Outboard ECST Test Line Shutoff valve, will NOT function due to loss of power.

Answer: c.

a, b, d The only AC powered valve in the HPCI system is HPCI-MO-1 RO Written Examination Question No.: 27 K/A: 206000 A4.12 Importance: Tier: 2 Group: 1 Cognitive Level: 2 Exam Bant:No.: new l

Reference: 2.2.33, Sect. 4. Objective: COR002-11-02, 5.b, 8.a, 8.c,1 )

HPCI initiated due to low RPV water level due to a loss of Reactor Feedwater Pumps. HPCI ;

subsequently tripped on RPV high water level. The following conditions are observed:

-

RPV water level is +20 inches and slowly lowering

-

Drywell pressure is 1.0 psig and slowly rising Which one of the followingi 4c ibes how HPCI will respond as water level continues to lower?

.If the operator depresses the ...

INITIATION SIGNAL RESET pushbutton on panel 9-3, HPCI will start an'! inject

' into the RPV regardless of RPV water level and Drywell pressur HI REACTOR WATER LEVEL TRIP RESET pushbutton, HPCI will start and I inject into the RPV regardless of RPV water level and Drywell pressur ! INITIATION SIGNAL RESET pushbutton, HPCI will start and inject into the RPV ]

if Drywell pressure rises to 1.84 psig regardless of RPV water leve ' HI REACTOR WATER LEVEL TRIP RESET pushbutton, HPCI will restart only

'

after water level lowers to -42 inches nr Drywell pressure rises to above 1.84 psi Answer: Does not reset the high level trip, removes the "open" signal to HPCI-MO-1 Will not reset the high level tri These conditions do not require low water level or high drywell pressure to start HPC ____ _ _ _ _ .

-

RO Written Examination Question No.: 28 K/A: 295020 Al.01 Importance: Tier: 1 Group: 2 Cognitive Level: 1 Exam Bank No.: new Reference: 2.1.22 Objective: COR002-11-02,8.b A false high drywell pressure signal has caused an automatic initiation of HPCI. An operator then depresses the Manual Isolate pushbutton instead of the Turbine Trip pushbutton on the 9-3 panel when attempting to secure HPCI.

Which of the following will occur? HPCI Inboard Steam Isolation valve, HPCI-MO-15 closes, the ECST Suction valve HPCI-MO-17 receives a close signal and the HPCI turbine trips, HPCI Outboard Steam Isolation valve, HPCI-MO-16 closes, the Suppression Pool suction valve HPCI-MO-58, receives a close signal and the HPCI turbine trips, Both HPCI Inboard and Outboard Steam Isolation valves, HPCI-MO-15 and HPCI-MO-16, close and both HPCI Suction valves, HPCI-MO-17 and HPCI-MO-58 receive a close signal and the HPCI turbine trip Both HPCI Inboard and Outboard Steam Isolation valves, HPCI-MO-15 and HPCI-MO-16, close and both HPCI Suction valves, HPCI-MO-17 and HPCI-MO-58 receives a close signal, HPCI turbine coasts down but does NOT trip.

Answer: b. This is logic A which is not tripped by the manual pushbutton. The manual pushbutton only trips logic B The manual pushbutton only trips logic B and the turbine trips on an isolation signa .

_- _

.,

RO Written Examination Question No.: 29 K/A: 218000 A4.02 Importance: 4.2*

Tier: 2 Group: 1 Cognitive Level: 2 Exam Bank No.: new Reference: COR002-16-02, 2.2.1 Section 4.1,. Objective: COR002-16-02, obj. 5b,6a 2.4.4.1, Section 4. The following conditions have been present for 2 minutes:

-

RPV water level indicates -148 inches on the wide range RPV level instrument

-

Reactor pressure is 300 psig

-

Drywell pressure is 22 psig Assume ALL equipment operates as designe Which one of the following describes the current status of the ADS valves, and the actions necessary to close or maintain them closed?

The ADS valves are...

! open.

'

The ADS A INHIBIT and the ADS B INHIBIT switches must be placed in INHIBIT.

, closed.

l The ADS A INHIBIT and the ADS B INHIBIT switches must be placed in INHIBI close The ADS LOGIC A TIMER and the ADS LOGIC B TIMER pushbuttons must be depressed at least every 90 second ope The ADS A INHIBIT and the ADS B INHIBIT switches must be placed in INHIBIT and then the ADS LOGIC A TIMER and ADS LOGIC B TIMER pushbuttons must be depresse Answer: ADS valve logic is satisfied and the valves are open ADS valve logic is satisfied and the valves are open Depressing the reset push buttons is not required i I

t i

i l

29 i

_

RO Written Examination Question No.: 30 K/A: 295007 A1.04 Importance: Tier: 1 Group: 1 Cognitive Level: 2 Exam Bank No.: new Reference: 2.4.2.3.3, Sect 2.1, COR002-16-02 Objective: COR002-16-02, The plant is operating at full power when all MSIVs close. All control rods fully insert into the reactor. Reactor pressure rises to 1093 psi Assume that the Safety Relief Valves (SRVs) and the Safety Valves (SVs) function at their design set point (t 0.0 psig).

Which one of the following describes the response of the nuclear pressure relief system?

~

~ SRVs Open th'en CloseJ ' SVs Open then Close - SRVs Cycling :- or 3 or 2

)

Answer: d. 5 will open based on SRV setpoints, then pressure will be controlled by one or two of the low-low set SRVs SRVs will open No SV open SRVs will open. No SV open !

l l

l

i

, ,

RO Written Examination Question No.: 31 K/A: 217000 K6.04 Importance: Tier: 2 Group: 1 Cognitive Level: Exam Brik No.: new Reference: COR002-18-02 Objective: COR002-18-02,10c,11b 2.2.67, Sections 4.2.1.5,4.2.1.8,4.2.1.10 &

4.2.1.12'

A reactor scram due to MSIV closure while at 100% power occurred. RPV water level reached a minimum of-20 inches indicated on the wide range RPV water level instrument. RCIC has been placed in a test lineup, recirculating water to the ECSTs. The following conditions exist:

-

RPV water level +20 inches indicated on wide range RPV level instrument

]

(stable)

-

Suppression Pool water level 0.0 inches indicated on the narrow range RPV level instrument

-

ECST A LEVEL 1.8 feet above the bottom of the tank

-

ECST B LEVEL 1.7 feet above the bottom of the tank Which'one of the following describes the effect on the RCIC valve alignment? ECST suction valve MO-18 remains open and the suppression pool suction valve MO-41 remains closed. Test Bypass to ECST valve MO-30 and ECST Test Line Shutoff valve MO-33 remain ope Suppression pool suction valve MO-41 fully opens, and then ECST suction valve MO-18 closes. Test Bypass to ECST valve MO-30 and ECST Test Line Shutoff valve MO-33 clos ECST suction valve MO-18 remains open and the suppression pool suction valve MO-41 remains closed. Test Bypass to ECST valve MO-30 and ECST Test Line 1 Shutoffvalve MO-33 clos Suppression pool suction valve MO-41 fully opens, and then ECST suction valve MO-18 closes. Test Bypass to ECST valve MO-30 and ECST Test Line Shutoff valve MO-33 remain open Answer: A low level in either ECST causes the ECST suction valve MO-18 to automatically close when the open limit switch of the suppression pool suction valve MO-41 is actuated. The MO-30 and MO-33 auto close when the MO-41 open !

RO Written Examination Question No.: 32 __

K/A: Generic 2.1.32 Importance: Tier: 3 Group: N/A Cognitive Level: 1 Exam Bank No.: new Reference: Objective: COR002-18-02, 0 .2.67, Section 5.19 & 5.20 SKL012-42-18,05 SIL 548 In accordance with 2.2.67, " Reactor Core Isolation Cooling System," which one of the following states the reason for securing the RCIC Gland Seal Vacuum Pump whenever RCIC is NOT available for injection? To reduce contamination levels in the North East Qua To eliminate addition of oxygen to the Suppression Chambe To reduce post-LOCA DC loads within the assumptions of the load calculations, To eliminate unnecessary wear on the Gland Seal Vacuum Pump mechanical seal Answer: Would not cause a eduction in contamination levels; potentially could raise contamination levels is secured shortly after RCIC operatio This is not the reason for placing the pump in PT The precaution has nothing to do with pump seals. The pump seals are not mechanical, they are packing typ .

_ - __

7- RO Written Examination Question No.: 33 l

l K/A: 295024 Generic 2. Importance: *.3

Tier: 1 Group: 1 Cognitive Level: 1 Exam Bank No.: new Reference: EOP-1 A, EOP-3A, EOP-5A Objective: INT 008-06-13,1 INT 008-06-02

.

Which one of the following describes the EOP(s) required to be entered if drywell pressure rises

. to 2.0 psig? - EOP-1 A,"RPV Control" onl EOP-3A, " Primary Containment Control" only, EOP-1 A,"RPV Control," and EOP-3A," Primary Containment Control" onl EOP-1 A, "RPV Control," EOP-3 A, " Primary Containment Control," and EOP-5 A,

" Secondary Containment Control". j Answer: Entry into EOP-1 A and EOP-3A is required. EOP-5A has no entry condition me j Note: Any EOPs supplied for exam need to have entry conditions blanked ou RO Written Examination Question No.: 34 K/A: 295026 K3.01 Importance: Tier: 1 Group: 2 Cognitive Level: 1 Exam Bank No.: new Reference: EOP-3A, INT 008-06-13 Objective: INT 008-06-13,4 INT 008-06-18 Which one of the following describes why an emergency depressurization is required per EOP-3 A, " Primary Containment Control," when suppression pool temperature cannot be maintained below the Heat Capacity Temperature Limit (HCTL)?

Ensures initiati9n of RPV depressurization will NOT result in exceeding ... Net Positive Suction Head (NPSH) limits for low pressure ECCS when they are required for adequate core coolin Boron Injection Initiation Temperature (BIIT) when Hot Shutdown Boron Weight has NOT been injected into the reacto l I Suppr,:ssion Chamber Spray Initiation Pressure (SCSIP) while sufficient energy remains in the RPV to exceed containment limits, Primary Containment Pressure Limit (PCPL) while the rate of energy transfer to the suppression pool exceeds containment vent capacity.

Answer: d. Not related to NPSH for pumps. HSBW is not the reason for emergency depressurization SCSIP is used to preclude cyclic condensation of steam (chugging) at the downcomer openings of the drywell vent .

O'

RO Written Examination Question No.: 35 K/A: 295029 K3.03 Importance: Tier: 1 Group: 2 Cognitive Level: 2 Exam Bank No.: new Reference: EOP-3A, INT 008-06-13, II. Objective: INT 008-06-13,4 l

Which one of the following describes why a reactor scram is required per EOP 3A, " Primary 1 Containment Control," when suppression pool water level CANNOT be maintained below sixteen (16) feet?- l A subsequent LOCA blowdown above this suppression pool level will exceed the torus design peak pressur To ensure the reactor is shutdown by control rod insertion prior to performing an emergency depressurization, To ensure the reactor is shutdown by control rod insertion prior to covering the Torus to Drywell vacuum breaker A subsequent LOCA blowdown above this suppression pool level will exceed the suppression pool downcomer design differential pressure l Answer: b. EOP set-point basi Not the reason for the scram. EOP terms that sound technica .5' is the number for the Torus to Drywell vacuum breaker Not the reason for the scram. EOP terms that sound tecimica RO Written Examination Question No.: 36 K/A: 226001 K3.01 Importance: Tier: 2 Group: 2 Cognitive Letel: 1 Exam Bank No.: new Reference: INT 008-06-13 Objective: INT 008-06-13, obj. 3 -

A LOCA is in progress and drywell sprays have been initiated.

Which one of the following will result if drywell sprays are NOT terminated and drywell pressure lowers below 0.0 psig? Chugging at the outlet of the downcome Partial de-inerting of the Primary Containmen Mechanical failure (collapse) of the Torus downcomer ring heade Mechanical failure of the Reactor Building to Torus vacuum breakers.

Answer: F ! Phenomenon associated with initiation of DW sprays. Phenomen associated with evaporative cooling due to spraying while in the unsafe region of the DW spray initiatiori limit curve, Event is within the design of the vacuum breaker l l

l l

-

RO Written Examination Question No.: 37 K/A: 500000 A2.03 Importance: Tier: 1 Group: 1 Cognitive Level: 2 Exam Bank No.: new Reference: EOP-3A Objective: INT 008-06-13,4 COR002-03-02,14e A LOCA has occurred and the following conditions exist:

-

Drywell H2 concentration is 7%

-

Torus H2 co tcentration is 4%

-

Drywell 02 concentration is 4%

-

Torus 02 concentration is 6%

-

In accordance with the EOPs, which one of the following describes the Primary Containment H2/02 combustible limit status and required actions?

The Primary Containment H2/02 concentration is ...

1 below the combustible limit. A Reactor scram and emergency depressurization is l require below the combustible limit. A Reactor scram and emergency depressurization is NOT require above the combustible limit. A Reactor scram and emergency depressurization is require I 1 above the combustible limit. A Reactor scram and emergency depressurization is i NOT require l Answer: !

The limits,6%, H2 and 5%,02 in either torus or drywell are the limits for the primary containment. Combustible limit exceeded requires a reactor scram and emergency depressurizatio Attachments: EOP-1A & EOP-3A with Entry Conditions blanked out and all Cautions except Caution 1 blanked ou I l

37 )

i l

l l

RO Written Examination Question No.: 38 l K/A: 295032 K3.03 Importance: Tier: 1 Group: 3'

Cognitive Level: 1 Exam Bank No.: new Reference: EOP-5A, INT 008-06-17 Objective: INT 008-06-17,4 Which one of the following describes the FOP-5A, " Secondary Containment Control," basis for isolating a system discharging into the secondary containment? To minimize RPV inventory losse j l To backup PCIS automatic function To maintain the Recirc MG set room accessible to personne , To terminate rising temperatures, radiation levels, and water level Answer: This is covered by other EOPs j

' PCIS automatic actions may not have been required . Secondary Containment Control does not maintain habitability for all areas. The Max Safe I

values are based on equipment operability and persocnel access necessary for EOP actions the Recirc MG set room is not one of the areas requiring acces l

l

!

i

.-

.

-

RO Written Examination Question No.: 39 K/A: 295013 Generic 2. Importance: Tier: 1 l Group: 2 Cognitive Level: 2 Exam Bank No.: new Reference: EOP-3 A, EOP-5A Objective: INT 008-06-13,1; INT 008-06-17,1 The plant is operating at 100% power with HPCI testing in progress. A loss of cooling for the HPCI Room has occurred. The following conditions exist:

-

Annunciator 9-3-1/E-10, AREA HIGH TEMP alarms

-

Ronan Annunciator (1522), HPCI ROOM (E 878') AREA TEMP HIGH, is displayed 4

-

Temperature Module HPCI-TS-105D indicates 180 F on the Plant Area Temperature Monitor Panel

-

Average Suppression Pool temperature is 98 F Which one of the following describes the EOP(s) required to be entered? EOP-5A, " Secondary Containment Control" only, EOP-1 A, "RPV Control," and EOP-3A, " Primary Containment Control" onl EOP-3 A, " Primary Containment Control," and EOP-5 A, " Secondary Containment Control" onl EOP-1 A, "RPV Control," EOP-3 A, " Primary Containment Control," and EOP-5A,

" Secondary Containment Control".

Answer: l EOP-I A has no entry condition me !

RO Written Examination Question No.: 40 K/A:295033 A2.01 Importance: Tier: 1 Group: 2 Cognitive Level: 2 Exam Bank No.: new Reference: EOP-5A Objective: INT 008-06-17,3 Which one of the following describes the conditions in secondary containment that require a reactor shutdown if a primary system is NOT discharging into the Secondary Containment? RMA-RA-7," Neutron Monitor Sys Drive Mech Area," and RMA-RA-10,"HPCI Pump Room," alarm and both indicate upscale hig RMA-RA-27, " Torus HPV Area (Southwest)," and RMA-RA-4, "RWCU Precoat j Area," alarm and both indicate upscale hig I RMA-RA-3,"New Fuel Area," and RMA-RA-5,"RWCU Sludge and Decant Pump I Area," alarm and both indicate upscale hig l RMA-RA-8,"CRD Hydraulic Equip Area (South)," and RMA-RA-9,"CRD Hydraulic Equip Area (North)," alarm and both indicate upscale high.

Answer: b.

Two areas must exceed max safe, b. is the only one that meets this. RMA-RA-3 and RMA-RA-7 are not on the table.

Attachments: EOP 5A with entry conditions blanked ont. EOP Tables 9,10 & 11 l

-.

-

RO Written Examination Question No.: 41 K/A: 295028 Generic 2.4.20 Importance: Tier: 1 Group: 2 Cognitive Level: 1 Exam Bank No.: new Reference: EOP-7A . Objective: INT 008-06-10,3 While performing EOP-7A,"RPV Level / Failure to Scram," with power below 3%, which one of the following CAUTIONS applies as reactor water level is controlled,

)

\

Lowering RPV water level to ... j i inches will result in an ADS initiation if ADS is NOT inhibite l 1 will result in low pressure ECCS injection unless it is stopped and prevente )

l l-110 inches will result in an MSIV isolation and loss of the main condenser as a heat sin inches will result in injection from low pressure ECCS systems NOT required l for RPV level contro Answer: I a,b,d These cautions do not exist in EOP-7A Attachments: EGP-7A with all CAUTIONS blanked out (Verify CAUTION 6 blanked out as it is in the flowchart steps).

l l

I i

k l

I RO Written Examination Question No.: 42 1 K/A: 295037 K1.07 Importance: l Tier: 1 Oroup: 1 Cognitive Level: 2 .-- Exam Bank No.: new Reference: 2.4.1.1.1, Tech Spec Objective: COR002-04-02,2 Immediately following a reactor scram, it is determined that seven (7) control rods located randomly throughout the core are stuck between positions 06 and 34. Conditions are as follows:

-

Reactor pressure 920 psig

-

Reactor water level +25 inches (stable on the narrow range)

-

Drywell pressure 1.0 psig

-

Drywell temperature 130 F

-

Torus temperature 85 F

-

The seven control rods will NOT respond to RMCS

'

In accordance with EOP-6A and EOP-7A," Failure to Scram," which one of the following describes the condition allowing exit from EOP-6A and EOP-7A7 i Cold shutdown boron weight has been injected into the rer.ctor cor ALL control rods except 26-27 are fully inserted into the reactor cor Reactor power will remain below 3% under ALL conditions without boro The reactor will remain shutdown with RHR in time Shutdown Cooling mod Answer: The only condition allowing exit of EOP 6A & 7A is the reactor will remain shutdown under all conditions without boron I

) l

-

_

RO Written Examination Question No.: 43 K/A: Generic 2. Importance: Tier: 3 Group: N/A Cognitive Level: 2 Exam Bank No.: new Reference: OI-7 Objective: SKLO10-10-01, A3 I

i During an ATWS, the Reactor Operator is directed to perform alternate control rod insertio J The Reactor Operator will be performing the actions to insert control rods by resetting RPS and inserting a manual reactor scra Assume the CRS has NOT suspended any peer check requirement !

Which one of the following describes the peer checking requirements to perform this task? I l

Peer checking is required for ... , all steps of the tas all steps except for panel 9-5 actions onl all steps exc mt forjumper installation onl the jumper installation, and is waived for all other step Answer: c. Jumper installation is waived in accordance with 01-7 as it is a back panel actio Peer check will be performed by operators in the Control Room for front panel manipulations prior to manipulating controls. This verification will be consistently performed during steady state manipulations and whenever reasonably possible during abnormal and transient condition Immediate operator actions shall not be delayed to wait for peer check. Peer check can be suspended for specific tasks during transients by the CRS as he deems reasonable and necessary.

I

f RO Written Examination Question No.: 44 K/A: 295030 K2.04 Importance: Tier: 1 Group: 2 Cognitive Level: 3 Exam Bank No.: new Reference: EOP-1 A, EOP/ SAG Graphs Objective: INT 008-06-18, l., A LOCA has occurred with the following conditions:

-

All control rods are fully insened

-

RHR pumps "A," "B" and "D" are unavailable

- HPCI, and RCIC are unavailable

-

RHR "C" operating in LPCI mode is being used to maintain RPV water level

-

Actual RPV water level is twenty (20) inches above the top of active fuel (TAF) and rising ten (10) inches per minute

-

RHR flow rate 8500 gpm (maximum available)

-

Torus pressure 4 psig

-

Suppression Pool temperature 185 F

-

Suppression Poollevel 7 feet Which one of the following describes the use of RHR as an injection system? The RHR pump must be secure Reduce RHR flow to 5500 gp Reduce RHR flow to 7000 gp Maintain RHR flow at 8500 gpm.

'

Answer: Answer if 0# curve is use AnswerifCore Spray curve is use Answer if10# curve is use With 4 psig pressure and 3 feet of water above the suctions there is 5.29 psig overpressure requiring that flow be reduced to no more than 7000 gpm for NPS11 concerns. Flow is also in the UNSAFE region of the Vortex limit curve, but this curve is less limiting than the NPSH curve in this case. 10 inches per minute RPV rise means that an excess flow of 1500 to 2000 gpm is available and flow could be reduced to 6500 and still maintain RPV water level above TA Attachments: EOP and FULL SIZE SAG graphs must be attached to the exa Calculators and rule .

I J

l RO Written Examinatica Question No.: 45 K/A: 295010 K2.01 Importance: Tier: 1 Group: 1 Cognitive Level: 2 Exam Bank No.: new Reference: EOP-3 A, INT 008-06-18, Graph 10 Objective: INT 008-06-18,1. A reactor scram due to a LOCA has occurred. The following conditions exist:

-

Reactor pressure 400 psig

-

Reactor waterlevel +35 inches (stable on the narrow range)

-

Drywell pressure 27 psig

-

Drywell temperature 250 F

-

Torus pressure 28 psig

-

Torus temperature 200 F

-

Torus water level 13 feet l

Which one of the following is assured by emergency depressurizing the reactor under current l plant conditions? !

To prevent exceeding the ... i Drywell Spray Initiation Limit (DWSIL). Heat Capacity Temperature Limit (HCTL). Primary Containment Pressure Limit (PCPL). Safety Relief Valve Tailpipe Level Limit (SRVTPLL).

Answer: PCPL is not related to DWSIL. We are in the safe region of DWSI PCPL is not related to HCTL. We are in the safe region of HCT FCPL is not related to SRVTPLL, the SRVTPLL is 16', not 16' 6" (right side of PSPL).

Torus level is 13', and does not challenge SRVTPL Attachments: All the EOP Graphs, FULL SIZE (EOP 5.8 Attachment 2) and rule .

.

RO Written Examination Question No.: 46 K/A: 295031 Generic 2.4.48 Importance: Tier: 1 Group: 1 Cognitive Level: 2 Exam Bank No.: new Reference: EOP-1 A, RPV Control Objective: INT 008-06-18,2 During conduct of the EOPs, the following parameters exist:

-

Reactor pressure 20 psig

'- Drywell pressure 8 ps'g l

l

-

Drywell temperature 300 F l

-

Torus temperature 105'F l

-

Rx Building temperature 150 F )

l If actual reactor water level is at the top of active fuel (TAF) and NO instrument run boiling is observed, which one of the following describes the RPV level instrumentation that can be used to l confirm reactor water level?

l RPV level CANNOT be determine ' Fuel Zone level instruments can be use Wide Range level instruments can be use l Narrow Range level instruments can be use ;

Answer: NEW Caution 1. Although in the unsafe region of Graph 1, instrument can be used .

no boiling is observed, Although in the unsafe region of Graph 1, instrumem can be used as long as no boiling is observed. below minimum indicated level below minimum indicated level Attachments: All the EOP Graphs. Caution I blanked out on all EOP chart _

-

RO Written Examination Question No.: 47 K/A: 295031 K1.01 Importance: Tier: 1 Group: 1 Cognitive Level: 3 Exam Bank No.: new Reference: INT 008-06-02, EOPs-1 A,6A Objective: INT 008-06-02,8 Which one of the following conditions assures adequate core cooling?

Note: All RPV levels are as INDICATED on the Fuel Zone instruments.

% All control rods are fully inserted, Reactor Pressure 128 psig, RPV level -40 inches, NO SRVs open, the only available injection is ECCS pressure maintenanc l All control rods are fully inserted, Reactor Pressure 200 psig, RPV level -50 inches, )

NO SRVs open, the only available injection is one (1) Core Spray pum I ATWS with reactor power at 5%, Reactor Pressure 60 psig, RPV level -20 inches, !

Three (3) SRVs open, the only available injection is one (1) RHR pum ATWS with reactor power at 14%, Reactor Pressure 385 psig, RPV level-50 inches, One (1) SRV open, the only available injection is (1) Condensate pum Answer: c. Level is above -30 inches for adequate steam cooling and 3 SRVs are open with l Minimum Alternate RPV Flooding Pressure met inches is too low for adequate steam cooling RC/L-16 : inches corrected is below minimum steam cooling level inches corrected is below minimum steam cooling level but above old minimum steam cooling level Attachments: All the EOP Graphs, EOP 1A, 6A & 7A with entry conditions," Exit" override, and all Cautions blanked ou :

l I

, j i

l RO Written Examination Question No.: 48 J

K/A: Generic 2.3.10 Importance: l

Tier: 3 Group: N/A Cognitive Level: 1 Exam Bank No.: new Reference: Objective:SKL010-01-02, .8.3, Section INT 008-06-02, 7 Given:

I

-

The Emergency Director declared a Site Area Emergency five (S) minutes ago.

-

The TSC is NOT operational.

-

EOP actions outside the control room are necessary to perform manual draining of the SD ALL Area Radiation Monitors (ARMS) on the Reactor Building 903' elevation are j alarming and indicate off-scale hig Drywell radiation monitor RMA-RM-40A and RMA-RM-40B indicate 2 x 10 rem / hour.

In addition to a TLD and PD-1," Digital Alarming Dosimeter," which one of the following describes the requirements to perform the directed actions per 5.8.3, " Alternate Rod Insertion Methods?" l

I The operator shall be accompanied by a Radiological Protection Technicia ] The operator may NOT enter Secondary Containment until the TSC is operationa The operator shall ca ry a survey instrument capable of monitoring radiation dose rates, The operator may perform the actions independently with NO additional radiological protection.

Answer: a. This is true if the Drywell radiation monitor RMA-RM-40A or RMA-RM-40B indicate

>10 4rem / hou I This is not an option when the ARM is off-scale hig ) This is true ifNO ARMS are in alar .

RO Written Examination Question No.: 49

,

K/A:295022 Kl.01 Importance: Tier: 1 Group: 2 Cognitive Level: 2 Exam Bank No.: new Reference: 2.4.1.1.4, Section Objective: COR002-04-02, 8.a,10.b,13.g During a plant startup RPV pressure is 850 psig when a loss of suction causes the "A" CRD pump to trip. The "B" CRD pump is out-of-service for maintenance.

After one (1) minute, which one of the following statements is correct?

Rod motion control with RMCS is ... availabl Scram times will exceed technical specification limit availabl Scram times will be within technical specification limit I NOT availabl ' Scram times will exceed technical specification limit NOT availabi Scram times will be within technical specification limits.

Answer: d. Normal rod motion is lost. Scram times are OK as long as accumulators are charge ! Normal rod motion is lost ;

' Scram times are OK as long as accumulators are charge .

_

,

RO Written Examination Question No.: 50 K/A:201002 A2.02 Importance: Tier: 2 Group: 1 Cognitive Level: 1 Exam Bank No.: new Refea:nce: 2.4.1.1.3, Section Objective: COR002-20-02,4.c, .3.2.27, 9-5-1/C-4 Reactor Power is 50%. With NO control rod selected, a padially withdrawn control rod is slowly moving into the core.

Which one of the following is required in accordance with procedure 2.4.1.1.3, " Failure of Drive to Latch?"

Select the specific control rod and ... fully insert the specific control rod with the EMERGENCY IN switc insert a scram on the specific control rod for a minimum of five (5) second raise CRD cooling water pressure and monitor the specific control rod movemen )

i attempt to stop the specific rod movement by placing the ROD MOVEMENT l CONTROL switch in WITHDRAW, Answer: a. This action would be taken if the control was drifting out after fully inserted. Would cause the control rod to drift in faster if d/p was the problem and not covered by CNS procedures. Unacceptable yet, similar in concept to the control rod drifting out action J RO Written Examination Question No.: 51 K/A:201002 A1.02 Importance: Tier: 2 Group: 1 Cognitive Level: 2 Exam Bank No.: new Reference: COR002-20-02 Objective: COR002-20-02,4.a, IOP 4.3, Sect - 12.3.2 With the plant at 65% power, a control rod is single notch withduwn from notch 24 to notch 26.

While the control rod is being withdrawn, a malfunction of the R vlCS timer causes a continuous withdrawal signal to be sent to the selected control rod.

Assume NO additional operator actions.

Which one of the following describes the final position of the cor trol rod? Notch 0 Notch 28 Notch 3 Notch 48.

Answer: Timer malfunction deselects the rod after 2 seconds (which is % second longer q than the normal timer), causing the control roc: to be de-selected From the given power level a Rx scram will not occu ; An extra % second will not result in a two (2) notch chang j Rod will be de-selected after 2 seconds of motio J

!

I

!

i I

i l

RO Written Examination Question No.: 52 K/A: 214000 A3.03 Importance: Tier: 2 Group: 2 Cognitive Level: 2 Exam Bank No.: new Reference: COR002-20-02, Objective: COR002-20-02, obj. I1,13b 2.3.2.27 9-5-1/B-4,4.3 A non-selected control rod at position 36 is uncoupled. The CRDM will be fully withdrawn.

While fully withdrawing the CRDM to position 48, which one of the following describes when the uncoupled control rod can be identified using RPIS? As soon as the rod is selecte When the CRDM is moved from position 3 When the CRDM coupling check is performe When the RMCS timer times out at position 48.

Answer: c.

An uncoupled control rod cannot be detected by RPIS until it is withdraw i to the overtravel position. This is done during the coupling chec I

,

RO Written Examination Question No.: 53 K/A: 215004 Generic 2. Importance: Tier: 2 Group: 1 Cognitive Level: 1 Exam Bank No.: new Reference: 4.1.1,2.4.1.1.2, COR002-30-02 Obj ective: COR002-30-02, 3.b, 5.a, 5.b During a reactor startup with the reactor close to criticality, control rod 18-19 is withdrawn from position 08 to 12. During movement of the Control Rod Drive Mechanism (CRDM), ALL of the SRM count rate meters remain at 4 x 104 cps.

Which one of the following is the cause of this indication? The SRM detectors have been withdrawn too far out of the core, Source neutrons contribution is NOT measurable at this power level, This control rod is uncoupled from it's control rod drive and is stuck, This control rod is located too far from any SRM for this movement to be detected.

Answer: c. This would not prevent an indicated flux change from occurring. 4 x 10* cps is well within the required value for detection of changes, Source neutrons are the major contributor at this power level.. This control rod is right next to the SRM. Any rod movement near criticality would be detected by at least one SRM detecto l

,

RO Written Examination Question No.: 54 K/A: 201001 K3.03 Importance: Tier: 2 C:oup:1 Cognitive Level: 2 Exam Bank No.: new Reference: 2.4.1.1.1, COR002-04-02, Fig. 5 Objective: COR002-04-02,12.c The plant is operating at 50% power with the following CRD system indication Drive water differential prer.sure 265 psid

-

Drive flow 0.0 gpm

-

Charging Header pressure 1450 psig

-

CRD system flow 50 gpm While attempting to insert control rod 18-19, drive water flow is observed to be 0.0 gpm. When attempting to withdraw control rod 18-19, drive water flow is observed to be 2.0 gpm. The !

control rod does NOT move.

Which one of the following describes the cause of the above indications?

Directional Control Valve ... is stuck ope is stuck ope l is stuck close )

I is stuck close l

.

Answer: d. stuck open would provide continuous withdrawal flow stuck open would provide continuous insert flow stuck closed would prohibit withdrawal flow but allow insert flow.

Attachment: Provide Fig 5 of reference

J RO Written Examination Question No.: 55 K/A: 201003 A1.01 Importance: .

Tier: 2 Group: 2 Cognitive Level: 1 Exam Bank No.: new Reference: COR002-05-02, COR002-04-02 Objective: COk002-05-02, obj. Ilb 2.4.1.1.3 Section COR002-04-02, obj.12c The unit operating at 100% power near the end of cycle with all control rods fully withdraw The scram inlet valve (CRD-AOV-126) for control rod 30-31 open '

Which one of the following describes the response of the plant over the next five (5) minutes, including why?

Reactor power will ... be downscale on APRMs. The reactor will scram due to high Scram Discharge Volume leve remain at 100% reactor power. NO control rod motion will occur. NO leakage into the Scram Discharge Volume will occu J l, lower, but the plant will continue to operate at power. The associated control rod will insert. NO leakage into the Scram Discharge Volume will occu j lower, but the plant will continue to operate at power. The associated control rod

-

will insert. The scram valve will leak into the Scram Discharge Volume, but NO scram will occur as the Scram Discharge Volume drain capacity exceeds the leakage from the scram valv Answer: A reactor scram will not occur. The SDV level will not chang The control rod will insert into the core. A single control scram will reduce reactor powe No leakage will occur into the SD l l

'lan slu .aa-._

.

RO Written Examination Question No.: 56 K/A: 215002 Generic 2.1.28 Importance: I Tier: 2 Group: 2 )

(

Cognitive Level: 2 Exam Bank No.: new

{

Reference: COR002-24-02 Objective: COR002-24-02, obj.1,4a )

4.1.5, Section A plant startup is in progress with reactor power at 32%. When withdrawing a contro' rod 22-23 from position 08 to position 16, the Reactor Operator mistakenly continues to withdraw control rod 22-23 using Rod Out Notch Override control when a control rod block is receive )

Which one c' the follo cing is the reason for this control rod block?

Prevent exceeding... the MCPR safety limit, a peak fuel enthalpy of 280 cal /g % plastic strain on the fuel claddin analyzed amount of energy that must be absorbed following a LOC Answer: a. The rod block is a result of the RBM. The design of the RBM is to prevent exceeding the MCPR safety limi RWM basis LHGR basis APHLGR basis

_

!

RO Written Examination Question No.: 57 K/A: 201006 A2.03 Importance: Tier: 2 Group: 2 Cognitive Level: 1 Exam Bank No.: new Reference: COR002-26-02,2.4.1.1.3, Objective: COR002-26-02, obj. 8a,9 The plant is operating at 9% reactor power. All control rods in the current rod group are at their insert limit of 36. One of the control rods in the current group drifts in from position 36 to position 00.

Which one of the following describes the effect on the Rod Worth Minimizer (RWM) if the drifting control rod is selected?

RWM will identify the control rod as ... an Insert Error only. A control rod block will NOT be enforce a Withdrawal Error only. A control rod block will NOT be enforce l an Insert Error and a Select Error. A control rod block will bc enforce an Insert Error and a Withdrawal Error. A control rod block will be enforced.

Answer: a.

A Select Error occurs when a non-error rod is selected. Ths drifting rod is an error rod. The rod will not be a Withdrawal Error at position 00. A rod past its' insert limit is an insert error. No rod block occurs for a single insert erro :

i

!

j

!

i

!

)

.

57 i

!

I RO Written Examination Question No.: 58 K/A: 234000 K5.02 Importance: Tier: 2 Group: 3 Cognitive Level: 2 Exam Bank No.: new Reference: COR001-21-02 Objective: COR001-21-02, obj. Sb 2.2.31, Attachment 3 Conditions during a core OFFLOAD are as follows:

-

The Reactor Mode Switch in the REFUEL positio l

-

ALL rods are fully inserted into the reactor core.

-

The refueling bridge is over the spent fuel pool with the Main Hoist in the Normal-Up position.

- A fuel assembly will be removed from the reactor core and transferred to the spent fuel pool.

Which one of the following describes when a control rod withdrawal block is FIRST received during the fuel movement sequence? i When the refueling bridge is moved over the reactor core ... with the Main Hoist in the Normal-Up positio and the Main Hoist is lowered from the Normal-Up position, The fuel assembly is grapple The fuel assembly is grappled, and the Main Hoist is loaded.

Answer; b. Main hoist must be loaded with fuel at NOT Normal-Up with the refueling bridge over the l reactor cor ' The Rod Block is received when the Main Hoist is lowered. The Rod Block is received when the Main Hoist is lowere _

RO Written Examination Question No.: 59 l l

K/A:295023 A2.04 Importance: i Tier: 1 Group: 3 l Cognitive Level: 1 Exam Bank No.: new

{

Reference: 10.25 section 4.1.2.5, 2.4.2. Objective: SKLO10-01-02, A4 l Section 4.2,2.4.8.6, T.S. 3.3. J Refueling activities are in progress with a new fuel bundle being lowered into reactor core location 21-4 Which one of the following requires the Control Room Monitor to direct fuel loading be immediately terminated per 10.25, " Refueling - Co.e Unload, Reload, and Shur 1e?" Failure of two (2) or more APRMs within the same trip syste SRM "A" and SRM "B" count rates rise by a factor of ten (10) to 300 cp Shutdown Cooling is lost with less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months estimated for " time to boil." Fuel Pool Cooling is lost with less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months estimated for " time to boil."

Answer: b. Note below step 4.1.2.4 states "SRM count rates normally do not exceed 100 cps."

a, c, d - None of these conditions require fuel loading be terminated per 10.2 APRM are not referenced in 10.25 and are not required to be operable for fuel handlin Subsequent action of 2.4.2. Similar to "c," but not required. 2.4.8.6 does not specifically call for terminating fuel handling, but does have refuel floor evacuation required as a subsequent action if fuel pool cooling cannot be establishe i

, '

)

_

,

.*

RO Written Examination Question No.: 60 K/A: 212000 K2.01 Importance: Tier: 2 Group: 1 Cognitive Level: 1 Exam Bank No.: new Reference: 2.4.6.6, 2.4.6.12, COR002-21-02 Objective: COR002-21-02,9.a I The reactor is operating at 100% reactor power when the feeder breaker to MCC-L trip Which one of the following describes the response of the Reactor Protection System (RPS) to this feeder breaker trip?

Power is lost to the ... "A" RPS MG set, and a % scram is receive "B" RPS MG set, and a % scram is receive "A" RPS MG set. The "A" RPS bus automatically shifts to the alternate power supply. NO % scram is receive "B" RPS MG set. The "B" RPS bus automatically shifts to the attemate power supply. NO % scram is receive Answer: MCC-L powers the Div 1 "A" RPS bu The RPS system does not automatically transfer power supplies, The RPS system does not automatically transfer power supplies.

I i

-

<

_ _ _ _ - _ _ _ _ _ _

__

RO Written Examination Question No.: 61 K/A: 239001 K2.01 Importance: 3.2*

Tier: 2 Group: 2 Cognitive Level: 2 Exam Bank No.: new Reference: COR002-14-02 Objective: COR002-14-02, obj. 2a, 3h, 7j 2.1.22, Section 8.1,2.4.6.12, The plant is operating at 100% power with the RPS "A" Power Transfer Switch and the RPS "B" i Power Transfer Switch in the MG-SET position. The 480V Station Service Transformer IG is de-energize Which one of the following describes the position of the MSIVs following the power loss? Inboard and outboard MSIVs CLOS Inboard and outboard MSIVs remain OPE Inboard MSIVs CLOSE. Outboard MSIVs remain OPE Outboard MSIVs CLOSE. Inboard MSIVs remain OPL Answer: a,c.d The AC and DC solenoid power that remains maintains the MSIVs ope ,

t

-

.

RO Written Examination Question No.: 62 l K/A: 295005 K3.01 Importance: l Tier: 1 Group: 1 Cognitive Level: 1 Exam Bank No.: new Reference: Tech. Spec Bases, Sect. B3.3. Objective: COR001-14-02,2.c, i Which one of the following is the bases for a reactor scram on a main turbine trip when reactor j power is at 35%7 l

a Protects the reactor from the effects of a loss of heat sin Anticipates a reactor power rise due to the colder feedwate j Ensure the bottom of the RPV steam dryer separator skirt is NOT uncovere I 1 Provide a backup scram to the RPV pressure and APRM high reactor s: ram I Answer: ~ No, this is a concern below 30% power

_ Prevents exceeding the MCPR safety limit These scrams backup the turbine trip scram ,

l l

!

,

!

i

.u

e.--:

.v RO Written Examination Question No.: 63 K/A: 212000 K5.02 Importance: Tier: 2 Group: 1 Cognitive Level: 1 Exam Bank No.: new i

Reference: COR002-21-02, Sect. II. D. 4 & . Objective: COR002-21-02,4.m, 5.a,1 Figure 7 I

The reactor is operating at 25% reactor power. Both the "A" Inboard MSIV and the "D"

)

Outboard MSIV close. All others MSIVs remain fully ope j

]

Which one of the following describes the design effect on the RPS logic?

a A full scram is receive b A % scram on RPS Trip System "A" is receive A % scram on RPS Trip System "B"is receive Neither a % scram, nor a full scram will be receive Answer: a,b,c. A combination of MSIV valve closure in the "A" and "D" or "B" and "C" steam lines will not result in RPS % or full scram.

,

j RO Written Examination Question No.: 64 i l

K/A: 201001 K4.04 Importance: l

Tier: Oroup: 1 l Cognitive Level: 1 Exam Bank No.: new Reference: COR002-04-02, Objective: COR002-04-02,4.d,

)

Which one of the following describes the operation of the Backup Scram Valves? Both valves must be energized to depressurize the scram air heade Both valves must be de-energized to depressurize the scram air heade Only one (1) valve must be energized to depressurize the scram air header, Only one (1) valve must be de-energized to depressurize the scram air header.

Answer: c. Only one valve is required by design. Only one valve is required by design.- These are DC valves, energized to actuat _

RO Written Examination Question No.: 65 K/A:295015 K2.04 Importance: Tier: 1 Group: 1 Cognitive Level: 2 Exam Bank No.: new Reference: 2.3.2.28, COR002-21-02, Figure 3 Objective: COR002-21-02, While operating at full power a power excursion to 125% occurs and the following annunciators are received:

- 9-5-2/A-3, REACTOR SCRAM CHANNEL B

- 9-5-2/B-1, NEUTRON MONITORING TRIP NO control rods moved. At the 9-5 vertical panel, you observe the following:

-

White CRD acram Solenoid Group lights for RPS Trip System "A" are li White CRD Scram Solenoid 3roup lights for RPS Trip System "B" are of NG operator actions have been taken in response to the conditions stated abov If the 5A-K15A and the SA-K15C relays will NOT change state, which one ofihe following operator actions will cause ALL control rods to fully insert? Depressing the "A" manual scram pushbutto Placing the Reactor Mode Switch to SHUTDOW Resetting RPS and then inserting a manual reactor scra Placing "A" and "C" RPS trip channel test switches to TRI Answer: K15A and K15C must both actuate to ins.-rt all control rods KI5A and K15C must both actuate to insert all control rods KI5A and KISC must both actuate to insert all control rods

%

Attachments: RPS Trip System A figure (COR002-21-02, Figure 3)

65 ..-

,

RO Written Examination Question No.: 66 K/A: 295006 A1.05 Importance: 4.2*

T Tier: 1 Group: 1 Cognitive Level: 1 Exam Bank No.: new Reference: 2.1.5, Section Objective: COR002-12-02, Following a scram from full power when all control rods fully insert, which one of the following meets the requirements of General Operating Procedure 2.1.5," Emergency Shutdown from Power?" Fully insert SRM detectors and verify lowering SRM reading Fully insert IRM detectors and verify IRM Range 6 to Range 7 overla Fully insert SRM detectors and verify lowering SRM reading Fully insert IRM detectors, range IRMs on scale and verify lowering IRM reading l 3 5 Partially insert SRM detectors to maintain 10 to 10 cp Fully insert IRM detectors and verify IRM Range 6 to Range 7 overla Partially insert SRM detectors to maintain 10 to 10 cp !

Fully insert IRM detectors, range IRMs on scale and verify lowering IRM reading I Answer: Insert both SRMs and IRMs, then check power lowering. Range 6 to 7 overlap is done on a 5 5 startup and is not required per 2.1.5. SRM detectors are withdrawn maintain 10 to 10 cps during a startu !

1

66

,

RO Written Examination Question No.: 67

.

K/A:215003 A2.02 Importance: Tier: 2 Group: 1 Cognitive Level: 1 Exam Bank No.: new Reference: 4.1.2, Sect Objective: COR002-12-02, 3.b, 3.d, 5.a, 5.b, 6.b, 6.c, 7.e A normal plant startup is in progress with the reactor mode switch in STARTUP. The following Intennediate Range Monitor system conditions exist:

-

IRM Channel"A"is failed downscale

-

IRM Channel"A" is bypassed

-

IRM Channel "A" Mode Switch is in STANDBY

-

All the IRM range switches, including IRM channel "A", are on Range 2.

Which one of the following describes the automatic action (s) that occur when IRM "A" is teken out of bypass? % scram onl Control rod block only, IRM downscale alarm onl Control rod block and % scram.

Answer: d. It's Mode switch is out of OPERATE Yes, but it also generates a rod block Normally a downscale is a rod block, but in this case it's been disabled with the Mode Switch. INOP generates a % scram A rod block and % scram are receive l j

J RO Written Examination Question No.: 68 K/A: 295018 Kl.01 Importance: Tier: 1 Group: 2 l I

Cognitive Level: 2 Exam Eank No.: new Reference: 5. Objective: COR002-19-02,2.b,4.a,4.c 5.a, 5.c, 6.b, 6.d, The unit is operating at 85% power with REC pumps "A", "B" and "C" operating. REC pump control switches are positioned as follows:

-

"A" REC pump STANDBY

-

"B" REC pump NORMAL

-

"C" REC pump STANDBY

- "D" REC pump NORMAL An operator mistakenly de-energizes MCC-K and ten (10) seconds later re-ener;,izes MCC-K.

Twenty (20) seconds after MCC-K is re-energized, which one of the following will restore REC cooling with three (3) REC pumps in operation?

Manually start ... two (2) REC pumps only ("A," "B" and/or "D").

!

' "A" or "B" REC ptunps and verify REC pump "D" automatically start two (2) REC pumps only ("A,""B" and/or "D") and then open the non-critical l

'

header supply, drywell supply isolation, HX outlet, and augment, d radwaste suppl l "A" or "B" REC pumps, verify REC pump "D" automatically starts, and then open I the non-critical header supply, drywell supply isolation, HX outlet, and augmented radwaste suppl ,

l Answer; a.

No pumps auto start. If two of "A," "B" and or "D" pumps are started within 40 seconds of the pump trips, no header isolation valves clos )

.

RO Written Examination Question No.: 69 K/A: 262001 K3.01 Importance: Tier: 2 Group: 2 Cognitive Level: 2 Exam Bank No.: new Reference: COR002-27-02 Objective: COR002-27-02, obj. 8c,3g,4c 2.2.71, Section .2.5, Section l The unit is operating at 100% reactor power. SW pump alignment is as follows:

l l

-

SW pumps "A," "B" and "C" are operating 1

-

Mode Selector switches for the "A" and "B" SW pumps are in STANDBY

-

Mode Selector switches for the "C" and "D" SW pumps are in AUTO A loss of offsite power occurs. Both DGs start and energize busses IF and IG.

Assume NO operator actions are taken.

Which one of the following describes the Service Water pumps that will be operating by design two (2) minutes after offsite power was lost? i A and B A and C B and D C and D Answer: a.

Only the SW pumps selected to standby start 13 seconds aner buses IF and IG are energized from an emergency power sourc , _ -

RO Written Examination Question No.: 70 K/A:203000 K2.01 Importance: 3.5* ,

Tier: 2 Group: 1 Cognitive Level: 2 Exam Bank No.: new Reference: 2.2.69,2.2.69.1, COR002-23-02 Objective: COR002-23-02, 2.a, 3.f, 8.a, The following sequence of events has occurred:

-

A LOCA has occurred resulting in LPCI injection

-

20 seconds later, all offsite power is lost

-

DG1 will NOT start

-

DG2 starts and loads as designed Under these conditions, which one of the following describes the status of the RHR pumps thirty (30) seconds after the loss of power and why? "A" and "B" pumps are available but NOT operating because the pump stop signal has sealed i "A" and "B" pumps are operating because the breaker anti-pump circuitry was sealed l in when the associated 4160 volt bus was re-energize "C" and "D" pumps are available but NOT operating because the breaker anti-pump )

circuitry has sealed i i "C" and "D" pumps are operating because the breaker anti-pump circuitry was reset when the associated 4160 volt bus was de-energize ' Answer: Not available as they're powered by DG 1. Stop signal has NOT been energized (switch

'

not taken to OFF) Not operating as they're powered by DG 1 Anti pump will NOT prevent the pumps from starting

RO Written Examination Question No.: 71 K/A:264000 A2.10 Irnportance: Tier: 2 Group: 1 Cognitive Level: 2 Exam Bank No.: new Reference: COR002-08-02 Objective: COR002-08-02,13c,9b I

DG2 has been started and loaded to 3850 KW for the monthly surveillance when a reactor scram l due to high drywell pressure occurs. Two (2) minutes following the LOCA, ALL offsite sources I are los Which one of the following describes the effect the above conditions will have on DG2 and 4160 Bus 1G7 DG2 engine and output breaker will NOT trip. DG2 will remain connected l to Bus 1 ;

i DG2 output breaker will trip when offsite power is log. DG2 is NOT available until the Diesel Generator over current lockout is manually rese DG2 engine and output breaker will trip when the LOCA signal is received. DG2 will automatically start and re-connect to Bus 1G when offsite power is los DG2 output breaker will trip when the LOCA signal is received. DG2 output breaker will close when offsite power is lost.

Answer: d.

The DG output breaker receives a trip signal opening the breaker when the LOCA signal occurs.

The DG would then run unloaded. The DG will pick up 4160 Bus 10 when it is de-energized (LOOP).

i i

l

!

RO Written Examination Question No.: 72 K/A: 264000 Generic 2.1.32 Importance: Tier: 2 Group: 1 Cognitive Level: 1 Exam Bank No.: new I

Reference: COR002-08-02 Objective: COR002-08-02,9g '

2.2.20, Section COR002-34-02,2b The following Diesel Generator Isolation Switches are positioned to ISOLATE: I

-

IS/DG-1 A and IS/DG-1B l

-

IS/EGI and IS/EGI-CT I

$

Which one of the following describes the effect on DGl?

DG1 must be started ... locally. DGl output breaker must be manually closed from panel"C". locally. DG1 output breaker must be manually closed from the local pane ) from the ASD Panel. DG1 output breaker must be manually closed from panel"C". from the ASD Panel. DG1 output breaker must be manually closed from the local panel.

Answer: b.

a,c,d All remote and automatic start features of DG1 are disabled. Panel "C" indications and controls are disable l l

i

RO Written Examination Question No.: 73 K/A: 262002 K6.02 Importance: i Tier: 2 Group: 2

'

Cognitive Level: 1 Exam Bank No.: new Reference: COR002-07-02 . Objective: COR002-07-02, obj. 8q 2.4.6.7, Section l l

The 250 VDC supply to the No Break Power Panel (NBPP) Irverter is los I

'

Which one of the following describes how the NBPP is powered after this event? The NBPP will automatically transfer to MCC- The inverter will automatically transfer to the alternate 250 VDC Bu The NBPP will NOT automatically transfer but can be manually transferred to MCC- The inverter will NOT automatically transfer but can be manually transferred to MCC-L Answer: ;

'

NRC exam #94 evaluated location of a transfer switc Does not have an alternate supply from DIV II D Automatically transfer and cannot be manually powered from MCC- Automatically tranrfer and cannot be manually powered from MCC-L l

,

I RO Written Examination Question No.: 74 K/A: 295003 A1.01 Importance: Tier: 1 Group: 2

-

Cognitive Level: 2 Exam Bank No.: new Reference: 2.2.13, COR001-01-02 Objective: COR001-01-02, 6.b, 7.a,1 The reactor is operating at 100% power when the Auto-Transformer becomes de-energize !

Which one of the following will occur?

J Power will be lost to ...

I

\

! one (1) of the Reactor Recirculation pumps, requiring single leop operation.

the intake structure equipment, requiring a shutdown in accordance with GOP 2. the 12.5 KV system, requiring the system to be restored from the Cornfield I substatio one (1) Condensate and one (1) Condensate Booster pump, resulting in a low RPV water level reactor scram.

! Answer: c.

l The startup transformer will be supplied by the 161KV Auburn line l The intake stmeture is not effecte The normal transformer is NOT effected l i l

\

i 74 l

~

!

RO Written Examination Question No.: 75 l

K/A: Generic 2. Importance: Tier: 3 Group: N/A Cognitive Level: 1 Exam Bank No.: new Reference: 5.1.3, Objective: SKL010-01-02, A.4, USAR Volume II,Section II, Subsection 4.2.2 A reactor scram has occurred. Which one of the following describes conditions needed to enter i Emergency Procedure 5.1.3," Flood?" The Missouri river level is 880' MSL and risin Notification is received that local levees have faile The Missouri river level is forecast to reach 890' MS Notification is received that the Gavins Point dam has fhiled.

Answer: d. Actual river level must be at least 890 to require entry into 5. l This does not require entry into 5.1.3, but does require actions in 5.1.3 if the procedure is in progress. The forecast must be at least 902' to reg .e entry into 5. l

!

,

i

!

l

!

i

,

.

i RO Written Examination Question No.: 76 K/A: 286000 KS.05 Importance: Tier: 2 Group: 2 Cognitive Level: 2 Exam Bank No.: new

'

Reference: COR002-08-02, COR001-05-02 Objective: COR002-08-02, obj. 6d 2.2.2, Section COR001-05-02, obj. 5f Both Emergency Diesel Generators are running following a start on a LOCA signa Which one of the following signal (s) will actuate one (1) of the Emergency Diesel Generator CO2Fire Suppression System, including the effect on the associated DG room ventilation? Actuation of the manual release station at the exit to the Turbine Building will initiate CO2 immediately and trip the DG room ventilatio . One (1) thermal detector in the DG room sensing high temperature will initiate CO2 immediately and DG room ventilation continues to operate, One (1) thermal detector in the DG fuel oil day tank room sensing high temperature will initiate CO2 after a time delay and trip the DG room ventilatio I Two (2) of the four (4) DG area smoke detectors activated in a DG room will initiate CO2 after a time delay and DG room ventilation continues to operat Answer: DG room ventilation will not trip because a LOCA signal is present l There are no thermal detectors in the DG roo DG room ventilation will not trip because a LOCA signal is present

1 s

RO Written Examination Question No.: 77 K/A: 239002 K6.04 Importance: Tier: 2 Group: 1 Cognitive Level: 1 Exam Bank No.: new Reference: COR002-16-02 Objective: COR002-16-02, obj. 8f,2b 3 2.4.2.3.1, Sect 4.8, 2.3.2.219-3-1/E-1, E-2 During normal operation at 100% power,125 VDC panel "A" is lost. Which one of the following describes the effect on the Low-Low Set SRVs?

All Low-Low Set SRVs ... remain powered from their normal power suppl automatically transfer to their alternate power suppl are de-energized with NO alternate power supply availabl are de-energized and must be manually transferred to their alternate power suppl Answer: Both LLS logic channels are normally powered from 125 VDC panel AA2, with an alternate supply from 125 VDC panel BB2. On a loss of power (panel AA2), both channels will automatically transfer to the alternate suppl !

l

<

l i

- I 77 i o

RO Written Examination Question No.: 78 K/A:295004 K1.02 Importance: ,

Tier: 1 Group: 2 Cognitive Level: 1 Exam Bark No.: new Reference: 2.2.25A, pp. 9,10,12 Objective:

2.2.24A, p. 2 COR002-07-02, obj: 6j,6h, Sc,9a,15b COR002-07-02, Figures 1 & 2 The plant is operating at 75% power when a fault causes a complete loss of 125 VDC bus "B." I Which one of the following describes equipment that has been de-energized and has the ability to be manually transferred to its' alternate electrical power source? Main Turbine Emergency Oil pump and the Air Side Seal Oil Backup Pum The ASD Panel and HPCI-MO-16," Steam Supply Outboard Isolation valve." MS-MO-77, " Outboard Isolation valve" and RWCU-MO-18, " Outboard Isolation l valve."

RCIC-MO-41," Torus Pump Suction valve" and RCIC-MO-131," Steam Supply to Turbine valve."

Answer: c. These are transferrable, but are powered by 250 VDC and are not de-energized. Cannot be ;

transferred during power operations due to SBO calculation l The ASD Panel and HPCI-MO-16 do not have the ability to be transferred to alternate. RCIC is normally powered from Division 1 125 VD I

i RO Written Examination l Question No.: 79 K/A: 263000 Generic 2.4.11 Importance: Tier: 2 Group: 2 Cognitive Level: 2 Exam Bank No.: new Reference: COR002-07-02 Objective: COR002-07-02, obj. 6f 2.4.6.10, Section {

SKL010-01-02, A.4, B.1, The unit is operating at 60% power when a loss of 125 VDC Panel bbl occur In accordance with Abnormal Procedure 2.4.6.10,"125 VDC System Failures," which one of the j following is required?

I Operate the "B" CRD pum !

! Transfer DG1 control power to its alternate sourc Transfer the "B" recirculation pump to the startup transforme l Entry into single loop operations on the "A" recirculation pum Answer: The "A" CRD pump is operate Cannot be performed with current plant desig l The "A" recirculation pump needs to be transferred to the startup transforme l

!

l

L

,

___

RO Written Examination l Question No.: 80 K/A: 223001 A1.09 Importance: Tier: 2 Group: 1 Cognitive Level: 2 Exam Bank No.: new Reference: 2.6.1, sect 6. Objective: SKL012-42-03,02j 2.2.25, sect 2. Given the following conditions:

-

ALL 4160 volt busses are de-energized

-

VBD-li Manual Transfer switch is in ALTERNATE

-

ALL Division I DC power sources are unavailable

Which one of the following describes the indicators available to be used as an information source j to take action regarding Suppression Pool Temperature without reliance on other indications? l l

I PMIS/SPDS onl Alternate Shutdown Panel instruments onl One (1) of the Suppression Chamber Water Temperature recorders onl Both PMIS/SPDS and one (1) of the Suppression Chamber Water Temperature recorders.

,

i Answer: I PMIS cannot be used as a sole sourc NBPP is not availab!: as DIV I DC is de-energized and no AC power is availabl l NBPP is not available as DIV I DC is de-energized and no AC power is available to the l

temperature recorder, PMIS cannot be used as a sole sourc .

l

)

i RO Written Examination Question No.: 81 _

K/A: 295019 Generic 2.4.11 Importance: Tier: 1 Group: 2 Cognitive Level: 1 Exam Bank No.: new Reference: 5. Objective: SKL010-01-02, A.4, B.3 While the plant is operating at power, a failure causes Service Air pressure to lower.

If Service Air pressure continues to lower below 90 psig, which one of the following requires a manual reactor scram per Emergency Procedure 5.2.8, " Loss ofInstnunent Air?" Only one (1) control rod starts to insert or instrument air pressure lowers to 84 psig, More than one (1) control rod starts to insert or instrument air pressure lowers to 70 psi Less than two (2) compressors running and Intake Bldg Control Air Low Pressure alarm is receive The in-service instrument air dryer becomes clogged and Drywell Pneumatic Header Low Pressure alarm is received.

Answer: b.

Procedure requires a scram when more than one rod drifts in or < 77 psig instrument air pressur l j

I i

J RO Written Examination Question No.: 82 K/A: 300000 K3.02 Importance: Tier: 2 Group: 2 Cognitive Level: 1 Exam Bank No.: new Reference: 5. Objective: COR002-02-02, obj. 5d,8a COR002-02-02, p. 45 COR002-32-02, obj. 8a COR002-32-02, p. 31 The unit is operating at 100% reactor power. The instrument air header completely rupture Assume NO operator action is take l Which one of the following describes the effect on the Condensate & Feedwater system valves? MC-AOV-FCV17 " System Minimum Flow" valve fails ope MC-AOV-FCV11 A/B "RFP Minimum Flow" valves fail close MC-AOV-B1 " Condensate Demin System Bypass " valve fails' close RF-AOV-FCV11 AA/BB "RFPA/B "Startup Flow Control" valves fail open.

Answer: a. valves fail open l valve fails open l

' valves fait close !

!

I

l l

l

l l

RO Written Examination Question No.: 83 K/A: 271000 A2.04 Importance: Tier: 2 Group: 2 Cognitive Level: 1 Exam Bank No.: new Reference: COR001-16-02 Objective: COR001-16-02, obj. 8g,10b 2.3.2.24 9-4-1/C-4,9-4-1/C-4 2.4.7.1 section l l

l The plant is operating at 75% power when the following indications are received:

I

- 9-4-1/C-5, OFFGAS HIGH RAD alarm j

- 9-4-1/C-4, OFFGAS TIMERINITIATED alarm

- K-1/A-4, OFFGAS FILTER HIGH D/P alarm j

- Off-gas flow indicates 100 cfm on Recorder AR-FR-47, SJAE AIR FLOW j If the above conditions are sustained for 20 minutes, which one of the following automatic actions will occur? AOG-AO-901 "AOG Supply valve" closes, AOG-AO-902 "AOG Return valve" close OG-AO-254 "Offgas System Isolation valve" open AR-AO-12 "30 Minute Holdup Pipe Drain valve" opens.

Answer: !

I AOG-AO-901 remains open l' OG-AO-254 closes AR-AO-12 closes I

l

..

I RO Written Examination Question No.: 84

'

K/A: 295017 A2.04 Importance: Tier: 1 Group: 2 Cognitive Level: 2' Exam Bank No.: new Reference: 2.3.2.24,2.4.7.1,2.4.1.2 Section Objective: COR001-16-02, 5, 7.b,1 .1,2.4.1.6 Section 6.1

. While operating at full power the following alarms are received:

- 9-4-1/C-4, OFFGAS TIMER INITIATED

- 9-4-1/C-5, OFFGAS HIGH RAD Which one of the following caused these alarms?

Operation with ... MFLPD of 0.85 at 90% powe l MAPRAT of 1.2 at 75% powe flow blockage to at least one (1) fuel assembly, power / flow in the Stability Exclusion Region of the Power to Flow ma Answer: MFLPD limit is < l.0, At 0.85, operation is well within limits, APLHGR limit violations will not cause fuel damage during operation. It is a post LOCA concer A manual reactor scram is no longer required to protect the core. The automatic APRM flow-biased scram provides protection from fuel damag ' \

RO Written Examination Question No.: 85 ,

K/A: 259002 Kl.05 Importance: Tier: 2 Group: 1 Cognitive Level: 1 Exam Bank No.: new Reference: 2.2.28.1, Sect. 8.2.11, Objective: COR002-32-02, .3.2.28, 9-5-2/G-4 2.3.2.1, A-1/F-6 A Startup is in progress with the "A" RFP maintaining RPV water level. The following conditions exist:

-

Feedwater flow Ix106lbm/hr

- "A" Feedpump Controller (RFC-MA-84A) AUTOMATIC

-

Master Level Controller (RFC-LC-83) AUTOMATIC

-

Startup Master Controller (RFC-LC-130) MANUAL

-

RF-MO-29, RFP A Discharge Viv CLOSED

-

RF-MO-30, RFP B Discharge Viv CLOSED Which one of the following would cause the "A" feedwater pump to go into " Track and Hold?" Startup Master Controller output slowly fails upscal Startup Master Controller output slowly fails downscale, Selected RPV water level instrument slowly fails upscal Selected RPV water level instrument slowly fails downscal Answer: d. The selected instrument failure will cause a < 6 ma. output to be sensed by the track ;

and hold circui The startup level controller output is not sensed by the track and hold circuit. Under l provided conditions, the SULCV would be mid-position. The RFP will reduce speed as l RPV water level rises, but the Master Controller cannot lower the MA station output low enough to cause a Track and Hol The startup level controller output is not sensed by the track and hold circuit. This is the reverse of"a" above, The selected level transmitter output must drop < 6 ma. to initiate the track and hold circuit. An upscale failure would result in a 50 ma. outpu !

!

! 85 l

l E__

F RO Written Examination Question No.: 86 K/A: 259002 A1.02 Importance: Tier: 2 Group: 1 Cognitive Level: 2 Exam Bank No.: new Reference: COR002-32-02 Objectiw;: COR002-32-02, obj. Sb, 7a,7b 2.4.5.1, Section 4.4, The plant is operating at power with the following reactor vessel level control alignment:

- RFC-LC-83, MASTER LEVEL CONTROLLER in balance

- RFC-MA-84A, FW CONTROLLER STATION A in balance

- RFC-MA-84B, FW CONTROLLER STATION B in balance

Feedwater flow is approximately 9.6x10 lbm/hr Steam flow is approximately 9.6x106 lbm/h RPV water level is +35 inche The Master Controller OUTPUT slowly fails downscale. RPV water level lowers to +27 inches when the operator places the "A" and "B" RFP controllers to MANUA Assume NO additional action is taken by the operator.

'

Which one of the following describes the response of Feedwater Flow and RPV water level?

Feedwater flow will ...

6 rise to 9.6x10 lbm/hr. Level will rise to +42 inche rise to 9.6x10 lbm/hr. Level will remain at +27 inche rise above 9.6x10 lbm/hr. Level will rise to +42 inche remain below 9.6x10 lbm/hr. Level will continue to lowe Answer: ' Level will not ris Feed flow will not rise above 9.6x10 lbm/hr. Level will not ris Feed flow rises to 9.6x10 lbm/hr. . Level does not lower.

l

,

C6 l

l

RO Written Examination Question No.: 87

'

K/A: Generic 2.4.13 Importance: Tier: 3 Group: N/A Cognitive Level: 1 Exam Bank No.: new Reference: INT 008-06-02,Section II. Objective: INT 008-06-02, Obj. 9 5.8, Attachment 4, section 1.12

While performing Abnormal Procedures 2.4.9.4.1,"RFP Turbine Control Failure," an entry I condition into the Emergency Operating Procedures (EOPs) is me l l

Which one of the following describes the Abnormal Procedure and Emergency Operating !

Procedure (EOP) use for this condition? )

Enter all applicable EOPs and execute ... I all flow paths concurrently for the EOPs entered. The Abnormal Procedures are exited when the EOPs are entere the flow path for the most degraded plant parameter first. The Abnormal Procedures are exited when the EOPs are entere all flow paths concunently for the EOPs entered. Execute the remaining steps of the Abnormal Procedures when the plant is stabl the flow path for the most degraded plant parameter first. Execute the other flow paths and the remaining steps of the Abnormal Procedures when the plant is stabl Answer: When EOPs are entered, all paths are pursued simultaneously. Abnormal procedures are not exited just because EOPs are entere ____ _ -_ _ _ __--___- _ ____ __ - -

-.

RO Written Examination Question No.: 88 K/A:295008 A2.02 Importance Tier: 1 Group: 2 ,

-

Cognitive Level: 3 Exam Bank No.: new Reference: 2.4.5.1, Section 2. Objective: COR002-32-02,8d,9c Given the following conditions:

-

Reactor poweris ste.ady at 50%

-

Reactor Vessel Level Control System is in 3-element control

-

Reactor level detector channel "B" is selected The Channel "B" feedwater flow SIGNAL faile to ZERO.

, Which one of the following describes the result (s) and why?

Actual Reactor level will ... lower, then return to the original level due to the level error signal overriding the steam flow / feed flow error signa lower, and stabilize at a lower level due to a mismatch between the level error signal and the total feed flow signal, rise, and stabilize at a higher level due to a mismatch between the total steam flow signal and the total feed flow signa rise until the main turbine and feedpumps trip due to a mismatch between the total steam flow signal and the total feed flow signa Answer: Level rises Level rises Level rises but should not reach the high level setpoint at this steam flo .

Attachments: Provide Figure 2 from COR002-32-0 i

i

-

RO Written Examination Question No.: 89

,

K/A:259001 A3.10 Importance: Tier: 2 Group: 1 Cognitive Level: 1 Exam Bank No.: new

_

Reference: COR002-02-02 Objective: COR002-02-02, obj. 6g 2.4.9.4.3 step 5.1.1 During operation at 70% power a breaker electr: cal fault causes a trip of one (1) Condensate Booster Pump. RFP suction pressure step changes to 250 psig for 13 seconds. Assume all equipment operates at e h design setpoint(s).

Which one of the following describes the response of the feedwater system? Both RFPs tri Both RFPs continue to operat The "A" RFP will continue to operate. The "B" RFP will tri The "B" RFP will continue to operate. The "A" RFP will trip.

Answer: d.

When RFP suction lowers to 260 psig, the "A" RFP trips after c 10-second time delay and the

"B" KFP trips aner a 15-second time dela U RO Written Examination Question No.: 90 K/A: 216000 K1.13 Importance: Tier: 2 Group: 1

' Cognitive Level: 3 Exam Bank No.: new Reference: COR002-15-02 Objective: COR002-15-02, obj. 2f, 6e, 4a, 5a 4.6.1, Section The plant is operating at 100% reactor power with NBI-LT-52C level transmitter (Narrow Range Reactor Water level instrument) failed upscal Prior to removing the NBI-LT-52C level transmitter from service for maintenance, the equalizing valve for NBI-LT-52A is fully opened by I& Assume NO operator actions are take Which one oithe following desci : effect of these failures? The RFPs and the Main Turbine will tri Only a low reactor water level alarm is receive Only a high reactor water level alarm is received, Only a % scram is received on RPS trip system "A".

Answer: A full scram is received on lowleve A high level trip occur A full scram is received on low leve l l

90 i

RO Written Examination Question No.: 91 K/A: Generic 2.4.11 Importance: ;

_

Tier: 3 Group: N/A Cognitive Level: 1 Exam Bank No.: new Reference: COR001-14-02 Objective: SKL010-01-02, .4.9.3.5, Section The unit is operating at 100% reactor power with three (3) Circulating Water Pumps operatin The "A" Circulating Water Pump trips and the TG LOW VACUUM PRE-TRIP annunciator is received. Recirculation flow is mduced to slow .Jte rate ofloss of vacuum but condenser vacuum continues to slowly degrad In accordance with Abnormal Procedure 2.4.9.3.5," Loss of Condenser Vacuum," what IMMEDIATE action is required? Bypass AOG charcoal bed ) Start a third circulating water pum Place a second set of SJAE in servic ! Manually scram the reactor, then trip the Main Turbine Answer: l not an action for the condition provided not an immediate action for loss of condenser vacuum. CNS does not place a second SJAE in servic not an immediate action for loss of condenser vacuum i

l t

I

I j

i l

RO Written Examination Question No.: 92 K/A:295002 K2.08 Importance: Tier: 1 Group: 2 Cognitive Level: 2 Exam Bank No.: new l Reference: 2.4.9.3.5,2.3.2.4 & 2.3. Objective: COR001-02-02, 3.b, 4.c, I l

While operating at 96% power a backwash sequence is initiated on the "l Al" condelar. As the backwash sequence starts the following annunciators are received:

-

A-4/E-1, CONDENSER A/B BACKWASH TROUBLE I

- B-1/B-3, TG LOW VACUUM PRE-TRIP l Main condenser vacuum is slowly degrading. Which one of the following is the cause of degrading main condenser vacuum?

' The backwash sequence initiated and ... the "l Al" condenser water box inlet valve did NOT clos b the "l A2" condenser water box inlet valve did NOT ope the "l Al" condenser water box outlet valve did NOT clos the "l A2" condenser water box outlet valve did NOT ope Answer: c. With the outlet valve open circ water will bypass the l Al condense This normally occurs during a backwas This valve does not reposition from ope This normally occurs during a backwas !

l l

RO Written Examination ls " ion 14o.: 93 K/A: 241000 K3.06 Importance: 4.l * ,

i Tier: 2 Group: 1 Cognitive Level: 2 Exam Bank No.: new Reference: 2.4.5. Objective: COR002-09-02, 6.h, 7.b, 7.e, During a plant startup, the turbine has just been synchronized per SOP 2.2.14 "22 KV Electrical System." At this point, the digital controller fails and prevents generator load from automatically ramping up, this causes a turbine generator trip on reverse powe Which one of the following statements describes how reactor pressure control will respond?

The pressure setpoint ... remains in automatic. The bypass valves remain in their pre-tripped position until opened by the operator, transfers to manual at the existing setpoint. The bypass valves automatically maintain reactor pressure at that setpoin remains iu automatic. The bypass valves transfer to manual and must be manually positioned to control reactor pressur transfers to manual at the existing setpoint. The bypass valves remain in their pre-tripped position until opened by the operato Answer: ! DEH shifts to manual and closes the control valves. BPVs will open to control pressure at ]

the preset pressur ! Pressure control shifts to manual and the bypasses respond to control pressure because the pressure transducer has NOT faile BPVs will open to control pressure at the preset pressur !

. -

i PO Written Examination Question No.: 94 K/A: 245000 K4.06 Importance: l Tier: 2 Group: 2 l

Cognitive Level: 2 t Exam Bank No.: new l Reference: 2.2.14, Section 4.4,4.6, Objective: COR001-13-02, Obj. 6c, 6d, 7a i 2.4.9. l The plant is operating at 100% reactor power when the following conditions are observed:

-

Armunciator C-3/G-1, MAIN GEN VOLTAGE REG TROUBLE, alarms

'

-

Ronan Annunciator (4022), MAIN GEN VOLT REG FORCING ALARM, is displayed

-

Generator reactive load has risen by 300 MVARS out and continues to rise Assume NO operator action is taken.

Which one of the following describes the design response of the Main Generator to the above conditions?

Field excitation current will ... raise to correct the problem. If the problem continues, the voltage regulator will remain in AUTOMATIC. After a time delay, the Main Generator will tri lower to correct the problem. If the problem continues, the voltage :egulator will remain in AUTOMATIC. After a time delay, the Main Generator will tri raise to correct the problem. If the problem continues, after a time delay, the voltage regulator will trip (shift to MANUAL). If the problem continues with the regulator in MANUAL for an additional time delay, the Main Generator will tri lower to correct the problem. If the problem continues, after a time delay, the voltage regulator will trip (shift to MANUAL). If the problem continues with the 4 regulator in MANUAL for an additional time delay, the Main Generator will trip.

Answer: Forcing Alerm Overexcitation (OXP-2) protects the Main Generator filed windings from excessive temperature due to prolonged overexcitation. OXP-2 lowers the field excitation l

current. If the field excitation current is not reduced to a safe value within a specified time, then l the automatic voltage regulator is tripped, and a second timer starts. If the over-excited condition still exists when the second timer times out, then the Main Generator trip li

J RO Written Examination Question No.: 95 K/A: Generic 2.2.22 Importance: Tier: 3 " Group: N/A Cognitive Level: 2 Exam Bank No.: new Reference: Technical Specifications 3. Objective: INT 007-05-06,1 & 3 The unit is operating at 100% power when the following Technical Specification conditions are discovered:

-

February 1,1999 at 1200 the "A" RHR pump is declared inoperabl February 3,1999 at 1200 the "C" RHR pump is declared inoperabl February 6,1999 at 0600 the "A" RHR pump is restored to OPERABLE statu February 6,1999 at 0800 the HPCI system is declared inoperabl Apply any extensions that are permitted by Technical Specifications. Assume NO other equipment will be restored to OPERABLE statu Which one of the following describes the time and date when the unit shall be in MODE 37 February 6,1999 at 210 l February 8.1999 at 240 February 9,1999 at 200 February 9,1999 at 240 Answer: When HPCI is declared inoperable, entry into Condition D is required. After ,

I 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , entry into Condition G is required. The unit shall be in MODE 3 within the next 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months )

' Assumes entry into Condition H and LCO 3.0.3 which requires MODE 3 within 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months of HPCl 'oecoming inoperabl Assumes entry into Condition B following the 7 day allowed outage time for the first inoperable pump. Incorrect because a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months completion time extension is permitte Assumes entry into Condition B following the 7 day allowed outage time for the first inoperable pump plus an extension of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for the second pump. This time is greater than that for Condition D and Condition Attachment: Provide Technical Specification 1.0,3.0,3.5 and 3.6. Do not provide the Base u____- _ _-____

.

RO Written Examination Question No.: 96 K/A: Generic 2.1.20 - Importance: Tier: 3 Group: N/A Cognitive Level: 1 Exam Bank No.: new Reference: 01-7, Attachment E Objective: SKL010-01-02, .3.1, section A Surveillance Procedure is being performed on the ADS system. The expected annunciators have been " flagged" with translucent colored tape per Operations Instruction #7. The CRS has been informed of all expected alarms. The operator has referred to the alarm card for all expected alarm Which one of the following describes the required actions when one of the " flagged" alarms is received as expected at the appropriate time?

The operstor shall acknowledge the alarm and ... is NOT required to report the alarm to the CRS. The operator does NOT have to refer to the associated alarm card, is NOT required to report the alarm to the CRS. The associated alarm card shall be referred to and performe the armunciator shall be reported to the CRS. The operator does NOT have to refer to the associated alarm card.

l the annunciator shall be reported to the CRS. The associated alarm card shall be

'

referred to and performe I Answer: The Alarm Procedure does not have to be referred to under these conditions per Ol#7 and i 2. No report is required to the CRS per Ol# No report is required to the CRS per Ol#7. The Alarm Procedure does not have to be referred to under these conditions per OI#7 and 2. RO Writtt.i Examination Question No.: 97 K/A: Generic 2. Importance: Tier: 3 Group: N/A ;

Cognitive Level: 1 Exam Bank No.: new Reference: 9.RADO Ob,iective: SKL010-01-02, A4 & A5 ITS 5.7.1 & 5. When comparing 9.RADOP.3," Area Posting and Acccu Control," requirements for a High Radiation Area to the requirements for a locked High Radiation Area, which one of the following ONLY applies to the locked High Radiation Area? The Control Room shall be contacte A special work permit, SWP, shall be read and understoo An Administrative Technical Specification LCO shall be entere An alarming dose rate meter that continuously integrates the dose rate shall be in the possession of the operator.

Answer; a. A RWP is required for entry into an unlocked and a locked High Radiation Area. High Radiation Area and Locked High Psdiation Area requirements are described in Technical Specification 5.7, however, entry intc, a Radiation Area or locked High Radiation area does not constitute violation of Technical Specifications unless the specific requirements are not met. This is an optional requirement for entry into both area i

r

'

RO Written Examination Question No.: 98 ,

_ ;

K/A: Generic 2. Importance: Tier: 3 Group: N/A Cognitive Level: 2 Exam Bank No.: new

\

Reference: GEN 001-01-03 Objective: GEN 001-01-03, Limiting Radiation Dose Obj. D, E q 9.ALARA.1, Section 7. SKL010-01-02, l l

A station operator has an accumulated TEDE of 1.6 rem for the year as permitted by a previous extension. Because of dose projections during the assigned outage work, the individual is

)

l expected to receive an accumulated TEDE of 2.4 re In accordance with 9.ALARA.1, " Personnel Dosimetry and Occupational Radiation Exposure Program," which one of the following describes the authorization required for the worker to receive the expected dose? Plant Manager Outage Manager Radiological Manager Site Vice President -Nuclear i

l Answer: I a,b,d Authorizations are required by the Radiological Manager above 2000 mrem. Site

, V.P. is required above 3000 mre ,

,

l

.

RO Written Examination Question No.: 99 K/A: Generic 2.1.29 Importance: Tier: 3 Group: N/A l

Cognitive Level: 1 Exam Bank No.: new i Reference: SKL008-01-02, Watchstanding Objective: SKL008-01-02, obj.10 Principles (RO) SKL010-01-02, .31, Section 8.2 Note i

In accordance with Administrative Procedure 0.31," Equipment Status Control," which one of l the following set of conditions permit the concurrent verification for a procedure step to be waived? The valve requires the use of a ladder so that it is accessibl The valve location makes egress difficult should the valve malfunctio The valve is required to be locked and is locked in position by the performe The verification will result in a radiation exposure of 12 mrem to the verifie Answer: Not a permitted waiver for procedure steps Not a permitted waiver for procedure steps Not a permitted waiver for procedure steps

)

l

i

.

I RO Written Examination Question No.: 100 K/A: Generic 2.2.13 Importance: : Tier: 3 Group: N/A '

Cognitive Level: 1 Exam Bank No.: new Reference:0.9, Section Objective:SKL010-01-02, Which ore of the following describes when a CAUTION TAG shall be posted on a control switch iocated in the Control Room? To identify that non-operations personnel can operate the control switc To prevent the operation of a component so that maintenance can be performed on the componen To provide protection to personnel or equipment when a component is undergoing a design modidcatio To provide instructions regarding the safe operation of a component as a result of an abnormal conditio Answer: a blue test tag would be used for this purpose a danger tag would be used for this purpose caution tags are not used for personnel protectio i 100 L_

v t e Inspection Report - Cooper - 1999301 (Operator Exams)
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \|
1999 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 7(0) 8(0) 9(0) 10(0) 11(0) 12(0) 13(0) 14(0) 15(0) 16(0) Security Operator Exams 301(0) Emergency Planning Other

IR 05000298/1999301

Text

REGION IV

Navigation menu

Search