IR 05000302/1997301
| ML20197E956 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 12/17/1997 |
| From: | Aiello R, Peebles T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20197E893 | List: |
| References | |
| 50-302-97-301, NUDOCS 9712300110 | |
| Download: ML20197E956 (109) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGION 11 Docket Nos.:
50-302 License Nos.:
DPR-72 Report Nos.:
50-302/97-301 Licensee:
Florida Power Corporation Facility:
Crystal River Nuclear Plant Unit 3 Location:
Crystal River, Florida
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Dates:
November 18, 1997 Examiners:
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Ronald F. Aiello, Chief License Examiner I$ffY Approved by:
Thornas A. Peebles, Chief Operator Licensing and Human Ferformance Branch Division of Reactor Safety Enclosure 1 9712300110 971217 (DR ADOCK 05000302 PDR
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EXECUTIVE SUMMARY Crystal River Nuclear Plant Unit 3 NRC Examination Report No. 50-302/97 301
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On November 18,1997, The facility conducted an announced operator licensing initial written re-take examination in accordance with the guidance of Examiner Standards, NUREG 1021, Interim Rev', don 8. This examination implemented the operator licensing requirements of-10 CFR $55.41, $55.43, and $55.45.
. Operations One Senior Reactor Operator (SRO) candidate received a written re-take examination.
- The licensee administered the written examination on November 18,1997 The candidate passed the examination, (Section 05.1).
Candidate Pass / Fail -
'RO Total Percent Pass
0
100 %
Fail
0
0%
The examiners concluded that candidate performance on the written examination was
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satisfactory. (Section 05.3).
Enclosure 1
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Report Details Summary of Plant Statuji During the period of the examination, Unit 3 was in Mode 5.
LQaetajlo_D1
Operator Training and Qualifications 05.1 General C_pmments The facility (under the guidance of the NRC) developed and conducted a regular, announced operator licensing initial written re-take examination on November 18 under the requirements of an NRC security agreement, in accordance with the guidelines of the Examiner Standards (ES), NUREG-1021, Interim Revision 8. One SRO upgrade re-take applicant received and passed the written examination.
052 Pre-Examination Activities a.
hop.g The NRC reviewed the licensee's examination submittal using the criteria specified for examination development contained in NUREG 1021 Interim Rev 8.
b.
Qbservations and Findinas The licensee developed the SRO written examination. All mcterials were submitted to the NRC on time. The Chief Examiner reviewed, modified, and approved the examination prior to administration. The NRC conducted in-office preparation prior to examination administration. The facility came to the regior'al office to review and discuss the NRC comments during the week of November 10,1997, to validate examination materials and familiarize themselves with the details required for exam administration. No delays were encountered. Fifteen technical and five administrative changes were required to be made on the examination in order to meet the criteria set forth in NUREG 1021 Interim Rev. 8.
(1) Written Examination Development The written examination was submitted on time and in a greatly improved format when compared to the last submittal (report 50-302/97-300). The organization of examination and associated reference material expedited the exam review frocess. The chief examiner noted many improvements in question content and formatting since the last examination (report 50-302/97-300).
c.
Conclusion The NRC concluded that the quality of the examination materials had improved since the 50-302197-300 examination. Greater emrh sis was placed on ensuring the technical accuracy of the examination questions.
Enclosure 1
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Conclusion
.The NRC concluded that the quality of the examination materials had improved since the 50-302/97-300 examination. Greater emphasis was placed on ensuring the technical accuracy of the examination questions.
. 05.3 Examination Results and Related Findinas. Observations. and Conclusions a.
Scope The chief examiner reviewed the results of the written examination, b.
Qhg_qjyMiQDs and Findinas.
The overall performance of the candidate on the written exam was satisfactory.
c.
Conclusion The chief examiner identified no discrepancies.
. V. Manaaement Meetinas X1. Exit Meeting Summary On December 10,1997, the chief examiner discussed the examination results with the Operations Training Supervisor. Dissenting comments were not received from the licensee. No proprietary information was identified.
PARTIAL LIST OF PERSONS CONTACTED Licensee
- A. Kennedy, Nuclear Operations Instructor
- J. Smith, Supervisor Nuclear Licensed Operator Training NRC
- S. Cahill, Senior Resident inspector ITEMS OPENED, CLOSED, AND DISCUSSED Opened:
None p
Closed None Q1pnu_gt None Enclosure 1
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SRO WRITTEN EXAMINATION AND ANSWER KEY
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ENCLOSURE 2
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U.S. NUCL('AR REGULATORY COMMISSION
!!TE-SPECIFIC
WRITTEN EXAMINATION
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APPLICANT INFORMATION Name:
Region:
II Date:
Facility / Unit: Crystal River 3 License Level:
SR0 Reactor Type:
BW Start Time:
Finish Time:
Instructions Use the answer sheets provided to document your answers.
Staple this cover sheet on top of the answer sheets. The passing grade requires a final grade of at least 80%
Examination papers will be collected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the examination starts.
Applicant Certification
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All work done on this examination is my own.
I have neither given nor received aid.
Applicant's Signature RESULTS Examination Value 100.00 Points Applicant's Score Points Applicant's Grade
%
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NRC RULES GElERA1. GUIDELINES 1.
Cheating on any part of the examination will result in a denial of your application and/or action against your license.
If you have any questions concerning the administration of any part of 2..
the examination, do not hesitate asking them before starting that part
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of the test.
3.
SRO applicants will-be tested at the level of responsibility of the senior licensed shift position.
4.-
You must pass every part of the examination to receive a license or to-continue performing license duties. Applicants for an SRO-upgrade license may require remedial training in order to continue their R0 duties if the examination reveal deficiencies in the required knowledge and abilities.
5.
The NRC examiner is not allowed to reveal the results of any part of the
examination until they have been reviewed and approved by NRC management.
Grades provided by the facility licensee are preliminary until approved by the NRC.
You will be informed of the official examination results about 30 days after all the examinatit,ns are complete.
NRC WRITTEN EXAMINATION GUIDELINES 1.
After you complete the examination, sign the statement on the cover sheet (in black ink) indicating that the work is your own and you have not received or given assistance in completing the examination.
2.
To pass the examination, you must achieve a grade of 80% or greater.
Every question is worth one point.
3.
There is a time limit of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for completing the examination.
4.
Use only dark pencil to ensure legible copies on the answer sheets.
5.
Print your name in the blank provided on the examination cover sheet and the answer sheet. You may be asked to provide the_ examiner with some form of positive identification.
6.
Mark your answers on the answer sheet provided and do not leave any question blank.
7.
If the-intent of a question is unclear, ask questions of the examiner only.
8.
Restroom. trips are permitted, but only ore applicant at a time will be
. allowed to leave. Avoid all contact with anyone outside the examination room _to eliminate even the appearance or possibility of cheating.
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When-yee-e6mplete 'the examination, assemble a' package' including the examination questions,-examination aids,-answer-sheets, and-scrap-paper-and give'it to the examiner or proctor.
Remember to sign the statement ~
on the examination cover sheet indicating that the work is your own and
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that you have neither given nor received assistance} in con.pleting the examination. The scrap paper will be disposed of immediately after the
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-examination.
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After you have turned _in'your examination, leave the examination area as
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defined by the examiner.
If you are found in this area while the examination is still in progress, your license may be denied or revoked.-
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Do you have any questions?
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1. ROT 150 001/ B4 TROT-4-10/ NTS/ 015KS.05//4.4/ 33/ Ni A reactor startup is in progress. The following nuclear instrumentation (NI)
readings were initially recorded:
- NI 1 reads 235 cps.
NI 2 reads 230 cps.
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After several rod pulls the NI readings are:
- NI 1 reads 3760 cps.
- NI 2 reads 3G80 cps.
Which of the following describes the condition of the reactor and the Intermediate Range indications?
A.
The reactor is critical with NI 3/4 reading off scale low.
B.
The reactor is critical with NI 3/4 reading 4 x 1011 amps.
v'C.
The reactor is not critical with NI 3/4 reading eff scale low.
D.
The reactor is not critical with NI 3/4 reading 4 x 10'11 amps.
Reasons:
A. & B.
To achieve criticality initial count rate nust double approximately five times.
D. The intermediate range NIs would not be on scale; the bottom of the scale is 1 0'11 amps.
NEW; ROT-410 B4; 0150101005; ROT-410 page 3; Theory Manual pages 1121 and 1122; OP 210 pages 12 and 13; 2/3
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ROT 150 001/ B4 TROT 4-10/ NTS/015K5.05//4.4/33/ NI
' A reactor startup is in progress. The following nuclear instrumentation (NI) -
readings were initially recorded:
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NI 1 reads 235 cps.
-. - NI-2 reads 230 cps.
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After soveral rod pulls the NI readings are:
NI 1 reads 3760 cps, NI 2 reads 3680 cps.
i Which of the following describes the condition of the reactor and the Intermediate Range indications?
A.
The reactor is critical with NI 3/4 reading off scale low.
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The reactor is critical with NI-3/4 reading 4 x 10'11 amps.
vC.
The reactor is not critical with NI 3/4 reading off scale low.
j D,
The reactor is not critical with NI 3/4 reading 4 x 10'11 amps.
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Reasons:
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A. & B.
To achieve criticality initial count rate must double approximately five times.
D. The intermediate range NIs would not be on scale; the bottom of the scale is
- 10'" amps.
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NEW; ROT 410 B4; 0150101005; ROT-410 page 3; Theory Manual pages 11-21 and 1122; OP-210 pages 12 and 13; 2/3
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2. ROT 2-20 0o3/G3/ ROT-4-9o/NTS/2.1.24//3.1/77/PRINTo The following plant conditions exist:
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. Unit is operating at 100% full power.
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A fault on the Offsite Power transformer de energizes the "A" ES 4160V bus.
- After the "A" Emergency Diesel generator loads on the bus the "A" ES 4160V-Undervoltage Lockout is actuated, dI un..n,g y
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ES 4160 LV LOCKEUT CONTROL SCH:rf Which of the following will extinguish the " Actuated" lamp and illuminate the
" Reset / Normal" lamp?
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1-vA.
The undervoltage condition clears and the "Roset" pushbutton is
' depressed.
B.
Breaker 3209 is opened and the " Reset" pushbutton is depressed.
C.
The HPI actuation clears and the " Reset" pushbutton is depressed.
D.
The HPI actuation clears, breaker 3209 is opened and the " Reset" pushbutton is depressed.
Reasons:
B., C. & D. The undervoltage condition must clear prior to any other action.
NEW; ROT 4 90 pages 20 and 21; ROT 2-20 pages 9 through 20; 2/3 FINALNRC.TST Version: 0 Page: 2
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3. ROT 2 32 001/G7//110030',015/2.3.2//2.9/77/ RAD During a post filter pre-planning meeting the following dose rates are listed:
Filter housing'on contact 25 mR/hr.
Filter housing general area 10 mR/hr.
- Room above the filter room (slots in floor) general area 2 mR/hr.
Which of the following v'ould be the best ALARA decision for changing the filters?
VA.
'IM o people in the upper room for 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> using long handled tools.
B.
. Two people; one in the upper room for 0.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> using long handled tools and the other in the filter housing area for 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
C.
Three people, one in the upper room for 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> using long handled tools, one in the filter housing area for 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and the last one in contact with the filter housing for 0.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />.
D.
Three people, two in the filter housing area for 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> each and the last one in contact with the filter housing for 0.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />.
Reasons:
B. The total dose received is 6.5 mR which is greater than 6 mR.
C. The total dose received is 12.25 mR which is greater than 6 mR.
D. The total dose received is 16.25 mR which is greater than 6 mR.
NEW; ROT-2-32 pages 21 through 23; 2/3 FINALNRC.TST Version: 0 -
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4. ROT 3-03 001/B6//0000501009/ EO9EA1.1//3.5/33/NAT CIRC During a loss of off site power which of the fullowing indications identifies symptoms ofgas ' accumulation in the reactor coolant system?
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A.
Steam generator pressure is 825 psig.
When the turbine bypass valves are throttled open the steam generator pressure decreases to 810 psig and then stabilizes.
B.
Steam generator pressure is 850 psig and decreasing.
Steam generator levelis 35% and increasing.
When steam generator level reaches required setpoint steam generator pressure stabilizes.
C.
Steam generator pressure is 825 psig and decreasing.
cold s 523'F and decreasing.
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Core exit thermocouples are 565'F and decreasing.
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Steam generator pressure is 850 psig and decreasing.
cold c 536*F and decreasing.
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hot s 579*F and increasing.
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Reasons:
A. With steam pressure holding when steam demand is increased natural circulation has not been interrupted by gas accumulation.
B. When the steam generator level reaches setpoint and the pressure stabilizes then the steam generator and reactor coolant system aie still coupled.
C. When steam generator pressure is lowered and Tcold and core exit thermocouples follow then the steam generator and reactor conlant system are etill coupled.
NEW; 0020501001; ROT-3-03 pages 14 and 15; 2/3
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J 5. ROT 3-0? 002/B3//0000501009/0055EK1.02//4.4/33/NAT CIRC
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The following plant conditions exist:
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_ A Loss of Offsite Power (LOOP) has occurred.
. The "B" OTSG had a steam leak in containment and was isolated.
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. Core exit therpoccuple temperatures are stable.
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Which of the following methods of core cooling is orc / erred?
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OTSG cooling with Emergency Feedwater (EFW) and Atmospheric Dump valves (ADVs)'
J B.
OTSG cooling with Emergency Feedwater (EFW) and Turbine Bypass
valves (TBVs).
C.
High Pressure Injection (HPI) PORV cooling.
D.
High Pressure Injection (HPI) high point vent (HPV) cooling, Reasons:
B., C. & D.
During a LOOP natural circulation with OTSG cooling using EFW and ADVs is the preferred method. HPI/PORV and HPI/HPV cooling is available but is not the preferred method. TBVs are not available due to the loss of the Condenser.
i MODIFIED BANK: ROT.?-03 5; ROT-3 03 pages 3 and 4; ROT 5100, page 2; 2/3 o
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6. ROT 318 001/B1//0000501001/2.1.7//4A/33/LOCA The following plant conditions exist:
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. Plantis at 100% power.
Pressurizer levelis at 220 inches and stable.
Letdown is at,a constant 80 gpm.
There is no increase in the reactor or auxiliary building sumps.
The Nuclear Services Closed Cycle Cooling (SW) surge tank (SWT 1) levelis -
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slowly increasing.
The liquid radiation monitor for the SW system, RM L3, is in alarm.
What is the cause of these indications?
A.
A CRD stator cooler is leaking, vB.
The primary sample ccaler is leaking.
C.
The reactor coolant pump motor cooler is leaking.
D.
The Reactor Coolant Drain Tank cooler is leaking.
Reasons:
A. The CRD stator cooler leaking would not elevate RM L3 or increase SWT 1 level.
C. The motor cooler leaking would not elevate RM L3 or increase SWT-1 level.
D. The RCDT is at a lower pressure than the SW flow through the cooler. This would decrease SWT l's levei not increase.
NEW; 0000501003; 0000501005; ROT-3-18 pages 1 and 2; ROT 4 56 pages 14 and 15;2/3 FINALNRC.TST Version: O Page: 6
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7. --_ ROT 3-18 002/B1//0000501001/0006AK1.01//3.7/33/ ACC ANAL l
The following plant conditions exist:
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Reactor coo an average teinperature is 579'F.
- + Reactor coolant pressure is 2100 psig and decreasing at 100 psig/ min.
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Pressurizer te,mperature is 643*F and decreasing at 6*F/ min.
. Makeup flow 810 gpm higher than normal.
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. The Reactor Coolant Drain Tank levelis stable.
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PZR levelis stable.
.. Reactor power is 40%.
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hhich of the following is the most probable cause for the above indications?
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-A.
Small break LOCA on the "B" hot leg.
B.
Stuck open PORV.
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vC.
Pressurizer steam space leak.
D.
Letdown line leak.
Re'asons:
A. & D.
Loss of RCS pressure without a los3 : levelis an indication of steam
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space leak.
B. A stuck open PORV would cause the RCDT level to increase.
MODIFIED BANK; ROT 3181; ROT 318 pages 1 through 4; 2/3
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. 8. ROT 3 20 001/A3//3440403001/001 A2.07//4.4/44/SOPP l j
- During a feedwater transient the reactor trips. Eight minutes later the following i?
plant con'ditiop(exist: ;
- Steam generator outlet pressure is 935 psig and stable.
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RCS pressure is 2155 psig and stable.-
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Which of the following will be the approximate value of Teold?
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528'F.-
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538'F.
-C..
548'F.
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D.
558'F.
Reasons:
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A., C.' & D.
With the steam generator at 935 psig T should be approximately c
538*.
NEW; 3440403005; 3440403011; ROT 3 20 pages 911; 2/3 i
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9. ROT 3-22 001/B4/ ROT 5-94/0000501028/E13EK1.2//3.6/33/EOP/AP The following plant conditions exist:
(assume all Immediate Actions are complete for EOP 2, Vital System Status Verification)
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A reactor trip,has occurred.
The "A" steam generator levelis 87% and increasing, i
- The "B" steam generator levelis 43% and increasing.
- All Reactor Coolant Pumps (RCP) are operating.
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Both Emergency Feedwater Pumps (EFW) are operating.
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Reactor coolant temperature is 533'F.
Reactor coolant pressure is 1850 psig.
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Based on the above conditions which of the following describes the appropriate Emergency Operating Procedure and associated rule for this situation?
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A.
EOP 2, Vital System Status Verification; EOP 13 Rule 4, Pressurized Thermal Shock.
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B.
EOP 2, Vital System Status Verification; EOP 13 Rule 3, EFW Cent ol.
C.
EOP 5, Excessive Heat Transfer; EOP 13 Rule 4, Pressurized Thermal Shock, vD.
EOP 5, Excessive Heat Transfer; EOP 13 Rule 3. EFW Control.
Reasons:
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A. & B. Due to RCS temperature the conditions are met for entering EOP 5.
C. RCS temperature is not low enough for entry into Rule 4 (< 380 * F.)
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NEW:: ROT-5 96 B4; EOP 13 pages 7 and 9; EOP 5 page 1; ROT-3 22 page 1; ROT 514 page 25 and 38; 2/3 clNALNRC.TST Version: 0 -
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10. Roi.3-23 001/B3//0000501027/E04EK2.1//4.0/33/EOP 04 -
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The following events have occurred: -
EOP 4, Inadequate Heat Transfer, has been entered due to a loss of Main
. Feedwater and Emergency Feedwater.
- While attempting to establish full High Pressure Injection (HPI) the Reactor.
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' Operator (RO) informs you that the "C" Make up Pump (MUP 1C) will not start.-
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What are your directions to the RO?
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F A.
Open the PORV then establish HPI flow with MUP 1A.
i vB.
Establish HPI flow with MUP 1A then open the PORV.
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C.
Establish HPI flow and open the PORV when the reactor coolant
pressure reaches 2400 psig.
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D.
- Open all High Point Vents then establish HPI flow with MUP 1A.
Reasons:
A. -IIPI flow must be established prior to opening the PORV, C. With only one MUP the PORV is immediately opened.
D. HPVs are only used if the PORV will not open. HPI Dow should be established
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before any vent opening occurs.
NEW; 0020501002; 0590401007; ROT 3 23 pages 10 through 14; ROT-5-102 pages 9 and 10; 2/3 FINALNRC.TST Version: O Page: 10
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11. ROT 4-00 001!B2// 0640101C34/ 064A4.01//4.3/ 33/ EDG Th9 "B" Emergency Diesel Generator (EDG 1B) is running supplying its associated 4160V ES bus, in parallel with the grid, for surveillance testing. The control room operator takes tha EXU VOLT ADJ DIESEL GENERATOR B rheostat to the raise position.
Which of the following would he the expected response of the diesel generator indications?
,
A.
Output KWs would increase.
B.
Output KWs would decrease.
vC.
Output MVARs would increase.
D.
Output MVARs woCd decrease.
Reasons:
A. & B. The speed switch increases / decreases realload.
D. Raising excitation increases generator terminal voltage which would increase MVARs out, not decrease.
BANK; ROT-4 06 23; ROT-4 00 pages 15 and 16; ROT 4 33 page 25; SP-354A pages 28 and 29; 1 FINALNRC.T~T Version: 0 Page: 11
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12. - R07-4-06 002/ B6// 0160101010/ 016K2.01//2.5/33/ NN!
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The following plant conditions exist:
-,
The plant is in Mode 1.
- All power is lost from VBDP 5.
. The Automatic Bus Transfer (ABT) device associated'with NNI X fails as is.
.
Which statement below describes the NNI X response to these conditions?
.
A.
The NNI X S1 and S2 breakers will open due to the loss of AC power io" NNI X.
vB.
NNI X field loads willlose power, but DC loads will still receive power from the alternate DC power supply.
C.
The S1 breaker will open due to the loss of the AC power from VBDP 5, however the S2 breaker will not open since power is still supplied from -
VBDP1.-
.
D.
NNI X DC loads will lose power, but field loads will still.eceive power from the alternate AC power supply.
Reasons:
A. Only the field loads and shunt trip power are supplied by the output of the
_ ABT S1 and S2 will remain closed and the DC loads will be supplied from the
alternate DC power source.
C. S1 cannot open due to the loss of the ABT. S1 and S2 will remain closed and the DC loads will be supplied from the alternate DC power source.
D.- Only the AC loads will be lost.
.
MODIFIED BANK; ROT-4 09 24; ROTS J - T10B; 0160401002: ROT-4 09 pages 25 through 28; 2/3 FINALNRC.TST Version: 0 Page: 12-
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The following plant conditions existi
'
- Plant is operating at 100% full power -
- The RC temperature select switch is selected to the "B" position.
The Auto / Manual T,y, select switch is selected to the "A LOOP" position.
An I & C technician working on the selected "A" loop Thot circuit in the NNI
'
chbinet inadvertently places a test lead in the wrong test jack causing the signal to begin a slow, stesdy decrease.
Which statement below is correct concerning T,y,?
-
vA.
Indicated T on the recorder would NOT be affected. T indicated ave ave on the digitalindicator and supplied to the ICS would decrease.
B.
Indicated. T on the recorder would decrease. T indicated on the ave ave digitalindicator and supplied to the ICS would NOT be affected.
C.
Indicated T n the recorder and digital indicator would NOT be ave affected. T supplied to tbo ICS would decrease.
ave
,
D.
Indicated T on the recorder and digitalindicator would decrease.
ave T
supplied to the ICS would not be affecte'd.
ave Reasons:
B. & D.' The output of the RC temperature select switch will only affect the recorder. With this switch selected to "B" this output would not change.
C. The output froin the Auto / Manual T select switch will affect the digital ave indicator. With this switch selected to "A" Tyy, will decrease.
.
BANK; ROT-4 09 6 ; ROTS J - T7; ROTS K - T2; ROT-4 09 pages 13 and 14; 1 FINALNRC.TST Version: 0 Page: 13
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s14. ROT 410 001/B2//0150101005/0032AK2.01//3.1/33/NI-A plant shutdown is in progress. The intermediate range power supply in the C" Reactor Protection System (RPS) cabinet has failed. Power range detectors indicato as follows:
NI 5 is at 4%.
-
'
NI 6 is at S4.'
.' NI 7 is at 6E
'
- NI 8 is at 6%.
Based on the above conditions which of the following describes the status of the source range detectors?
A.
Only NI 1 is energized.
B.-
Only NI 2 is energized.
.
'
vC.
Both NI 1 and NI 2 are energized. -
D.=
Neither NI I or NI 2 are energized.
Reasons:
A., B. & D.
To energize the SR high volts while shutting down requires ans IR NI to be below 5 x 1010 amps and the following combination of
-
power range NIs: NI 5 and NI G OR NI 7 and NI 8 below 5%.
'
NEW; ROT 410 B3; 0150401004; ROT-410 pages 3,8,16,20, and 21; 2/3
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15. ROT 4-11001/Bs/0000501006/017A4.02//4.1/33/INCORE The Safety Parameter. Display System (SPDS) has a blinking.nagenta "I" and a magenta reading'of 600. The reactor coolant pressure is 500 psia. Which of the following statements describes the status of core cooling as indicated by the SPDS?
~
-
A.
The average of the 12 selected incore thermocouples to SPDS indicate that subcooling margin has been lost but the core is being adequately cooled.
B.
The average of the 5 highest incore thermocouples to SPDS indicate that subcooling margin has been lost but the core is being adequately cooled.
C.
The average of the 12 selected incore thermocouples to SPDS indicate that an inadequate core cooling event is in progress, vD.
The average of the 5 highest incore thermocouples to SPDS indicate that an inadequate core cooling event is in progress.
Reasons:
A. The incore display to SPDS is an average of the 5 highest incores of twelve.
For this pressure and temperature the core is not adequately being cooled.
B. For this pressure and temperature the core is not adequately being cooled.
C. The incore display to SPDS is an average of sne 5 highest incores of twelve.
FEW; 0000501021; 0170101006; 0170101007; ROT-411 pages 11 and 18; E0P-07 page 1; 2/3 FINALNRC.TST Version: 0 Page: 15
15. ROT-4-11001s B9//0000501006/017A4.02//4.1/33/INCORE The Safety Parameter. Display System (SPDS) has a blinking magenta "1" and a magenta reading of 600. The reactor coolant pressure is 500 psia. Which of the
following statements describes the status of core cooling as indicated by the SPDS?
.
A.
The average of the 12 selected incore thermocouples to SPDS indicate that subcooling margin has been lost but the core is being adequately cooled.
,
B.
The average of the 5 highest incore thermocouples to SPDS indicate that subcooling mergin has been lost but the core is being adequately cooled.
C.
The average of the 12 selected incore thermocouples to SPDS indicate that an inadequate core cooling event is in progress, sD.
The average of tho 5 highest incore thermocouples to SPDS indicate that an inadequate core cooling event is in progress.
Reasons:
A. The incore display to SPDS is an average of the 5 highest incores of twelve.
For this pressure and temperature the core is not adequately being cooled.
B. For this~ pressure and temperature the core is not adequately being cooled.
C. The incore display to SPDS is an average of the 5 highest incores of twelve.
NEW; 0000501021; 0170101006; 0170101007; ROT-411 pages 11 and 18; EOP 07 page 1; 2/3 FINALNRC.TST Version: 0 Page: 15
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116.:: ROT-4 12 001/ B3// 0120101006/ 012K2.01/ / 3.7/ 33/ RPS
,
Tall power is lost toVBDP 4l Which of the following describes the response of the -
.'
.
. Reactor Protectkin System (RPS)'and associated Control Rod Drive (CRD):
"
'
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9A. -
RPS channel."B" will tripi and the "B" CRDM breaker will open.
.
- B.
RPS channel "A" will trip, and the "A" CRDM breaker will open.-
!
- ;
C.-
- RPS channel "B" will not trip, but the "B" CRDM breaker will open.
-
D.
! RPS channel "A" will not trip, but the "A" CFDM b'reaker will open.
..
Reasons:
l B. -' VBDP-4 supplias the "B" RPS channel and would have no affect on the "A"
.
'
RPS channel.
C. The loss of power to VBDP 4 would cause the "B" RPS channel to trip and
-
'-
along with it the "B" CRDM breaker.
- D.-fVBDP 4 supplies the 'B" RPS channel and would have no affect on the "A"
'
RPS channel.
- MODIFIED BANK ROT-4-12 48; 0120101010; 0120201003; 0120401003 ROTS J -
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T8; ROTS K T2; ROT-412 pages 14 and 49; I c
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- - 17. -; ROT 412 002/82// 0120101'010/01EA4.03//3.6/33/RPS
!
The following plant conditions exist:
'
,.
. The plant is in Mode 1.
. The "C" Reactor Protection System (RPS) channelis in Bypass with breaker
testing'in progresr. i The "0" Control Rod Drive (CRD) breakers are open.
- A feedwater transient causes the "D" RPS channel to trip.
- The other RPS channels ("A", "B", and "C") do not trip.
Which of the following describes the expected RPS/CRD response?.
.
i.;.
vA.
' A reactor trip should not oc.atr. The RPS system is in a 2 out of 3 logic, but only one channelis tripped.
B.
L A reactor trip should not occur. The "D" breakers will open but the "C" and "D" breakers do not satisfy the reactor trip logic.
C.-
A reactor trip should occur. You have satisfied the 2 out'of 4 matrix and all CRD breakers should open.
D.
A reactor trip should occur. The "C" breaker was open and the "D" channel will trip the "D" breaker.
,
Reasons:
,
,
B. The "D" breakers will not open as the two out of three logic is not met.
C. No reactor trip will occur. Only one channelis tripped.
D. No reactor trip will occur. The "D" breaker will not be open as the two out of
~ threa logic is not met.
.
'
BANK; ROT-4012 53; 0120101011; 0120101013; 0120401003; ( ? ?0401005; 0120401005: ROT-412 pages' 9 and 49; 1 i
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18. noT-412 004/ 03//0120101oowoo29EA1.1&l3 9/33/ AMSAC
,
With rcactor power at.75% the "A" hiain Feedwater Pump trips. During the plant runback the "E", Main Feedwater Pump governor fails to minimum speed. Which of the following describes the plant response to this situation?
.
i vA.
AhtSAC will trip the Main Turbine and actuate EFIC if feedwater flow decreases to < 17%.
B.
AMSAC will trip the Main Turbine and actuate EFIC if feedwater flow decreases to < 45%.
-
C.
DSS will trip the Regulating Rods and actuate EFIC if feedwater flow decreases i > < 17%.
D.
DSS will trip the Regulating Rods and actuate EFIC if feedwater flow Jactaases to < 45%.
Roauons:
B., C. & D4 AMSAC requires FW flow <17% concurrent with Reactor power
>4ti%. These conditions will trip the turbine and actuate EFIC.
DSS requires RCS preseure >2450# and will trip the regulating control rods and any other group on the Auxiliary power supply.
NEW: 0120101010; G120201003; 0120401003; ROT 412 pages 40 and 47; 1
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FINALNRC.TST Version:0 Page: 18 I
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19. rot 412 005/ Be//012010 tot 4/o0007A106/4.4/4.6/33/ OSS The following plant conditions exist:
'
Group 2 control rods are being moved per SP.333, Control Rod Exercises.
. The Reactor Protection System (RPS) has failed to actuate on a high RCS pressure trip signal.
. RCS pressure'is 2400 psig and increasing.
Which of the following lists the control rod groups that are expected to fully insert into the core when the Diverso Scram System (DSS) setpoint is reached?
A.
Groups 1,2,3 and 4.
B.
Groups 5,6 and 7.
vC.
Paoups 2, 5,0 and 7.
.
D.
Groups 1, 3, 4,5,6 and 7.
Reasons:
A. & D.
DSS will only trip the regulating rods (groups 5,6 and 7) and any group on the auxiliary power supply.
B. Group 2 will also trip because it is powered from the auxiliary power supply.
BANK; ROT 412 94; ROT 412 pages 43 & 44; 1 FINALNRC.TST Version: 0 Page: 19
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20. rot.4-13 ooilse//0130101001/oooooox4094.3/4.4/33/ES During a plant shutdown the following conditions exist:
P.cactor Coolant Eystem (RCS) pressure is 1200 psig.
RCS temperature is 400' F.
,
High Pressure Injection (HPI) has been bypassed per procedure.
An RCS pressure transient occurs which increases RCS pressure to 1750 psig.
Which of the following describes the status of the HPI system?
.
A.
The actuation bistable will automatically reset and the bypass bistable will automatically reset. There will be no actuation of HPI.
B.
The actuation bistable will automatically reset but the bypass bistable will not automatically reset. There will bc no actuation of HPI.
vC.
The actuation bistable will not automatically reset but the bypass bistable will automatically reset. There will be a full actuation of HPI.
D.
The actuation bistable will not automatically reset and the bypass histable will not automatically reset There will be a full actuation of HPI.
Reasons:
A. & B.
The actuation bistable will not automatically reset.
D. The bypass bistable will automatically reset.
Bank Question ROT 413, #5 FINALNRC.TST Version: 0 Page: 20
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21. ROT.413 002/96//0130401001/013A1.06//3.9/33f ES The following plant conditions exist:
'
- A large break Loss of Coolant Accident (LOCA) has occurred.
Reactor Building (RB) pressure is 17 psig.
i
'
i Reactor Coolant (RCS) pressure is 550 psig.
. Borated Water Storage Tank (BWST)levelis 18 feet.
!
.
Based on the above conditions which of the following describes the suction sourco for the High Pressure Injection (HPI) pumps as prescribed by EOP.3, Inadequate i
Subcooling Margin?
i 4 -
y i
vA..
MUV 58 and MUV 73, make up pump supply vawes from the BWST.
'
B.
DHV.42 and DHV 43, decay heat pump supply valves from the RB
!
sump.
!
!
C.
MUV.04, make.up pump supply valve from the make up tank.
D.
DHV 11 and DHV.12, make up pump supply valves from the decay heat l
system.
!
Reasons:
I
.
B., C. & D. _ With the BWST level above 15 feet HPI pump suction will be from the BWST, not the make.up tank, Piggy back operation /RB sump
'
suction does not occur until the BWST is below 15 feet.
.
NEW; 0130501002; ROT.413 pages 14 and 10; ROT.4 54 pages 3 and 0; 2/3
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FINALNRC.TST Version:0 Page: 21
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22. ROT 413 003/BS//0130101001/2.4 46//3.6/33/ES l
The following plant conditions exist:
'
Reactor cooYan't pressure is 1400 psig.
l
. Reactor building pressure is 1.5 psig.
j
.
!
The following aimunciators are in alarm on the ESA panel:
LOAD SEQUENCE ACTUATION A, BLOCK 2,3,4,5, AND G
'
. DIVERSE CONTAINMENTISOLATION A
,
.
C Based on the above information which of the following alanns on the ESA panel
should also'be in alarm?
!
l i
r A.
ES A ACTUATION NOT RESET.
!
B.
<C..
ES A ACTUL' ION NOT BYPASSED.
,
I i
D; RB ISOLATION ES A ACTUATION.
'
Reasons:
!
A. The conditions for having this alarm is RCS pressure > 1640 psig and HPI not
.
reset, j
!
B. The conditions for this alarm would require RCS pressure to be < 500 psig.
- D. The conditions for this alarm would require RB pressure to be greater than 4
,.
.psig.
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NEW; AR 301 pages 2 through 8,16 through 18, and 44; 2/3
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FINALNRC.TST Version: O Page: 22
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23. nor 414 ootta1// 041010100/ossM.10//3.a/3rics
!
The following plant conditions exist:
!
-
. The unit is at 45% power.
.
. Both "A" and "B" Main FW pumps are running in automatic.
. Both "A" and *8" Main FW block valves, FWV.29 and FWV.30, are closed.
. The FW cross.*over valve, FWV.28, is open.
. A" loop AP is 83 pai.
"
"B" hop AP_is 75 psi.
I
.
!
Which of the following signals is controlling the "A" Main FW pump?
-
-
r l-A.
"A" loop AP.
vB,
"B" loop AP.
C.
"A" loop OTSG level error.
!'
D.
"A" loop feedwater flow error.
Reasons:
!
A., C. & D.
For the configuration in the stem the "A" Main FW pump is controlled by the lowest loop AP.
<
NEW; ROT 414 pages 34 30; 2/3
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FINALNRC.TST Version: 0 Page: 23
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24. ROT 414 002/ B 1/ / 0410101001/ 03$K4.011/ 3.8/ 33/ ICS l
A plant shutdown is bi progress with three Reactor Coolant Pumps (RCPs)
operating.
,
The OTSO with the single RCP has reached its low level limit (LLL).
,
As the shutdown continues what is the expected plant response?
,
A.
A ATeold will develop; both main feedwater pumps will revert to flow error control.
B.
A ATeoki will devel p; RCS flow input to the A AT control circuit will be c
dienbled.
C.
A ATcold will develop; one third of the feedwater flow error will be
-
'
applied to the OTSO not on low levellimits.
vD.
A ATeold will develop; all feedwater flow error will be applied to both OTSGs.
Reasons:
A. The Main Block valve on the levellimited OTSO will be closed. This will switch the associated Main Feedwater pump to AP control.
B. The RCS flow input is not disabled.
C. All flow error is applied across both OTSGs.
BANK; ROT 41418; NRC 1193; ROT.414 page 28; 1 FINALNRC.TST Version: 0 Page: 24
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2 5. ROT 414 003/B1//0410101001/041K6.03//2.9/33/lCS l
During a unit startup the fobowing plant conditions exist:
'
Turbine header pressure is 950 psig and decreasing.
Turbine controlis in operator auto with the generator output breakers closed l
and a megawa,tt output of 50 megawatts.
Reactor power is 13%.
!
All bypass valves are open.
If the turbine header pressure continues to decrease, the bypass valves should be fully closed when header pressure reaches:
'
VA.
885 psig B.
035 psig C.
1010 psig D.
1025 psig Reasons:
'
A.,B.& C.
With the turbine and reactor not tripped and none of the 50 psi bias conditions apply (ULD demand > 15% or all turbine bypass valves closed with header pressure error less than 10 psi) then the setpoint is 885.
BANK; ROT 414 27; ROT 414 pages 18 and 10; ROTS J T10B; 1
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26. - nor.414 oo4/ 81// o410101ootlooo3AK3 02// 3.7/33/ICS The following plant conditions exist:
'
The reactor is producing 1788 MWthermal.
Three reactor coolt,at pumps are operating.
Control rod group 7 is 60% withdrawn.
I Which of the following describes the ICS response if control rod 7 3 dropped fully
-
into the core?
'
A.
ICS will automatically runback to 45% ULD demand.
E vB.
ICS will automatically runback to 60% ULD demand.
C.
ICS will not runback due to NI power being below the runback setpoint.
,
D.
ICS will not runback due to one OTSG being on Low Level Limits.
Reasons:
A. The ICS runback for this condition is 60% ULD demand.
>
C. NI power is above the runback setpoint.
D. Ono OTSG is probably not on LLL at this power level. Even ifit is the runback will still occur with the " Total Flow Control" circuit ratioing FW flow to the OTSGs.
NEW; ROT-414 pago 12; 2/3
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27. ROT 414 005/ B1// 0410101001/ 059K3.04/ / 3.8/ 33f ICS While operating at 70% reactor power extraction steam is lost to feedwater heat exchanger 6A Which statement below is correct concerning the ICS feedwater control subsystem for this condition?
'
.
L A.
The total fe sdwater flow control circuit will decrease total feedwater demand to maintain heat removal from the reactor compatible with the current reactor power.
vB.
The feedwater temperature compensation circuit will decrease total feedwater demand to maintain heat removal from the reactor compatible with the current reactor power.
C.
The total feedwater flow control circuit willincrease total feedwater demand to maintain heat removal from the reactor compatible with the current reactor power.
D.
The feedwater temperature compensation circuit willincrease total feedwater demand to maintain heat removal from the reactor compatible with the current reactor power.
Reasons:
A. & C.
The total feedwater flow control circuit is used for 3 RCP operation when at least one of the steam generators is on low levellimits.
D. FW demand will be decreased for this condition.
13ANK; ROT.414100; ROT 414 pages 20 through 30,32 and 33 ;1 FINALNRC.TST Version: 0 Page: 27
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28. ROT-415 001/09//0190101001////31'EFIC The following plant conditions exist:
.
Both Main Feedwater Pumps have tripped and cannot be restarted.
The "A" Once Through Steam Generator (OTSG) level is 47 inches and rising.
- The "B" Once Through Steam Generator (OTSG) levelis 30 inches and stable.
EFV 56, flow control valve, has lost its DC power.
Which of the following actions would be required to establish "A" OTSG level control?
,
.
vA.
Select " Manual Permissive" for both EFW trains and close EFV 11, block valve for EFV 50, at the control board.
B.
Reset all EFW actuations and close EFV 11, block valve for EFV 50, at the control board.
C.
Select closed EFV 11, block valve for EFV 50, at the control board.
D.
Direct the Auxiliary Building Operator to close EFV 56 locally.
Reasons:
B. An EFW actuation will not reset. Without selecting " Manual Permissive" for both trains the block valve cannot be closed.
C. EFV 11 has an automatic open signal applied to it until both EFW actuation signals have been bypassed.
D. EFV 5G is a target rock valve. There is no local control station and this valve cannot be closed manually.
.
NEW: ROT 415 pages 8 and 9; 2/3
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29. ROT 4-15 002/ 38// 0190101001/ 061 A1.01//4.2/ 33/ EFIC Given the following sequence of events:
i
i
. A loss of offsite power has occurred.
Subcooling margin was less than 30'F.
. Emergency Feedwater (EFW)is running and feeding both Once Through Steam Generators (OTSGs)
!
. Offsite power has been restored.
Subcooling margin has been restored to > 50*F.
. One Reactor Coolant Pump (RCP) per loop has been restarted.
Which of the following describes OTSG level response?
A.
Decrease from 95% to the 65% level setpoint, vB.
Decrease from 95% to the Low Level Limit setpoint.
.
C.
Decrease from 65% to the Low Level Limit setpoint.
D.
Remain at the Low Lovel Limit setpoint.
Reasons:
A. With at least 1 RCP running the Low Level Limit setpoint is automatically selected.
C. & D., Initially, when sebcooling margin was lost, all RCPs were secured and the setpoint of 95% is selected.
BANK; ROT 415 23; NRC 1193; ROT-415 pages 17 and 18; 1 FINALNRC.TST Version: 0 Page: 29
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30. ROT 415 003/ B7// 0190101001/ E05EK1.1// 3.8/ 33/ EFIC
!
The following plant conditions exist:
-
A large steam leak has occurred.
OTSG "A" indicates 590 psig.
OTSO "B" indicates 460 psig.
-
No operator actions have been performed.
-
Which of the following describes the expected Emergency Feedwater Initiation and Control (EFIC) response?
A.
EFIC vector logic will send open commands to both OTSG feedwater valves and will feed both OTSGs.
B.
EFIC vector logic will send close command to "A" OTSG feedwater valves, an open command to "B" OTSG feedwater valves, and will feed only "B" OTSO, vC.
EFIC vector logic will send an open command to "A" OTSG feedwater valves, a close command to "B" OTSO feedwater valves, and will feed only "A" OTSO.
D.
EFIC vector logic is bypassed under these conditicaa therefore neither OTSG will isolate.
l Reasons: A., B. & D.
Vector logic is as follows:
ERESSUBE STATUS SG A SG B VALVE COMMANDS A & B > 600 PSIG OPEN OPEN A > 600 PSIG & SG B < 600 PSIG OPEN CLOSE A < 600 PSIG & SG B > 600 PSIG CLOSE OPEN A & B < 600 PSIG AND
,
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A & B WITHIN 125 PSIG OPEN OPEN A 125 PSIG > SG B OPEN CLOSE B 125 PSIG > SG A CLOSE OPEN l
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30. ROT-415 003/B7//0190101001/E05EK1.1//3.8/33/EFIC BANK; ROT 415 25; NRC 1193; ROTS J - T10B; ROT 415 pages 23 and 25: 1
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31. ROT 416 001/D1//1010$01001/0068AK2.01//4.0/33/RSP A plant transient has occurred requiring the use of the Remote Shutdown Panel (RSP). llow eenyou confirm that the four Remoto Shutdown transfer switches are transferred to the RSP7
.
A.
All four transfer switches have status lights on the RSP.
B.
All four transfer switches are located on the RSP, VC.
Two of the transfer switches, "A" and "B", have status lights on the RSP and two of the switches, "AB" and "NS", are located on the RSP.
D.
Two of the transfer switches, "AB" and "NS", have status lights on the RSP and two of the switches, "A" and "B", are located on the RSP.
Roni,ons:
A., B. & D. The "A" and "B" switches have status lights on the RSP while the
"AB" and "NS" switches are located on the RSP.
NEW; ROT.410 page 1: 1
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32. rot 4 22 001/02/i0480104006/045K4.37//16/77/EHC An oilleak has developed in the turbine lube oil system. Auto stop oil pressure is
,
40 psig. Which.of the following describes the condition of the main turbine and the non return valves?
.
A.
The main turbino is not tripped; the non return valves are closed.
,
B.
The main turbino is not tripped; the non return valves are open.
'
vC.
The main turbino is tripped; the non return valves are closed.
'
D.
The main turbine is tripped; the non return valves are open.
Reasons-A., B. & D.
With autostop oil pressure < 45 psig the interface valve will open causing the turbine to trip. The Air /EH pilot valve will vent causing the non return valves to close.
NEW: ROT 4 22 page 15; I
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33. rot 4 25 ooitsol/07204oiootlooooAk2.02//3.1/33/RM During a waste gas release, with a normal"B" side ventilation lineup, RM.A2, j
Auxiliary Bunding Purge Exhaust Radiation Monitor, goes into high alarm.
Which of the following groups of fans trip?
I
.
A.
AHF.10, Fuel Handling Area Fan.
AHF.11B, Auxiliary Building Supply Fan.
AHF.9B, Penetration Cooling Fan.
-
AHF.14A and 140, Auxiliary Building Exhaust Fans.
.
AHF.30, Chem Lab Supply Fan.
'
B.
AHF 10, Fuel Handling Area Fan.
AHF.14B and AHF 14D, Auxiliary Building Exhaust Fans.
AHF 9B, Penetration Cooling Fan.
l
!
AHF 34A, Hot Machine Shop Welding Hood Exhaust Fan.
AHF.30, Chem Lab Supply Fan.
i
.
C.
AHF 10, Fuel Handling Area Fan.
AHF.11B, Auxiliary Building Supply Fan.
t AHF.9B, Penetration Cooling Fan.
AHF.44A, Chemistry Hood Exhaust e an.
!
AHF.30, Chem Lab Supply Fan.
vD.
AHF.10, Fuel Handling Area Fan.
AHF 11B, Auxiliary Building Supply Fan.
AHF.9B, Penetration Cooling Fan.
AHF.34A, Hot Machine Shop Welding Hood Exhaust Fan.
AHF.30, Chem Lab Supply Fan.
Reasons:
A., B. & C. A high alarm on RM A2 will trip tho following fans: AHF.10, AHF.11B, AHF DB, AHF 34A, and AHF 30.
NEW; ROT-4 25 pages 27 and 29, t
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t 34. ROT 425 002/ F2/ / 0720106001/ 0061 AA1.01/ / 3.6/ 88/ RM The Makeup Pump Area Radiation Monitor, RM.G 0, goes into alarm.
Which of the fodowing describes the actions associated with this alarm?
.
.
A.
Local alarm sounds and ventilation dampers for the makeup pump area close.
B.
Local alarm sounds and operating auxiliary building supply fans trip.
vC.
The local alarm sounds and the associated control room annunciator goes into alarm.
D.
The local alarm sounds and the associated waste disposal panel
'
annunciator goes into alarm.
Hensons:
A.,B.& D.
The only automatic actions associated with RM 010 are the alarm functions at the local monitor (born and lights) and in the control room (annunciator).
NEW; ROT 4 25 pages 5,6 and 20; 1 FINALNRC.TST Version: 0 Page: 35
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3 5. ROT 426 0011 B1// 0340101003/ 034K6.02// 3.3/ 44/ FH l
During refueling operations the following radiation monitors come into alarm:
-
RM.017, Reactor Building Personnel Hatch.
. RM.018, Reactor Building Incore Instrument Area.
Which of the foliowing actions are required?
.
A.
Secure rafueling operations, vB.
Secure the RB purge ifin operation.
C.
Perform the actions required in the ODCM, D.
Perform the actions required in Techr.ical Specincations.
Reasons:
A. Refueling operations are not required to be secured due to these RMGs being in alarm.
C. & D. There are no ODCM or Technical Specification requirements for these RMGs.
NEW; 0340101022; 0340101023; 0340204001; 0340204002; FP.203 pago 7; I FINALNRC.TST Version: O Page: 36
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36. ROT 426 002/ 01// 0340101003/ 2.2.31// 3.8/ 33/ FH During refueling operations the following events occur:
'
The main bIidge operator is placing a fuel assembly into the core.
'
'
The " Underload Limit" light is flashing in and out.
. The spotter is,giving mstructions to the bridge operator on the use at.d direction of the inching motor.
Which of the following is the Refueling Area Supervisor's responsibility in this situation?
sA.
Approve all further actions prior to them taking place.
B.
Assume the role of the spotter and give directions.
,
C.
Obtain permission from the Refueling Engineer in the control room to continue.
D.
Initiate a "Stop Work" order for a significant fuel handling event.
Reasons:
B. The Refueling Area Supervisor shall not assume the duties of the spotter.
C. Only the Itefueling Aren Supervisor has this authority.
D. This does not meet the criteria for a significant fuel handling event.
NEW: 0340101022; 0340101023; 0340201001; 0340201002; FP 203 pages 15 and 10;I FINALNRC.TST Version: 0 Page: 37
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37. rot 4 28 ootistitoo10101oo1/ootk2.o2/isa/33/eno t
With the plant operating at 100% power the "B" Control Rod Drive (CRD) breaker fails open.
--
Which of the following conditions,in conjunction with the above failure, will cause a reactor trip?,
,
VA.
The breaker from Reactor Auxiliary bus SA to the CRD system is opened.
.
B.
The breaker from Plant Auxiliary bus to the CRD system is opened.
C.
Vital bus distribution panel (VBDP) 4 is de energized.
D.
Vital bus distribution panel (VBDP) 0 is de energized.
,
Reasons:
B. This would de energize CRD breakers "B", "D" and "F" which will not make up the trip logic.
C. This will de energize the "B" RPS cabinet which will not make up the trip logic.
D. This will de energize the "D" RPS cabinet which will not make up the trip logic.
NEW; ROT 4 28 pages 33 and 60; 2/3
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38. rot 42s 002/A1/1341oic.,037/ooosAK3 04//4.1/sstcRD While performing SP 333, Control Rod Exercises, at 98% power, control rod 6 3 does not move. Jhe operators attempt to move the rod individually with no success and the rod is determined to be untrippable. Which of the following would be the required actions for this condition?
A.
Reduce power to 60% until the problem is resolved then return to full power.
B.
Verify shutdown margin using the double stuck rod curve; if shutdown margin is adequate then remain at 98% power.
C.
Align all control rods in the group to within 6.5% of the group averago height while maintaining rod insertion limits, group sequence and overlap limits.
.
v'D.
Ensure adequate shutdown margin and be in Mode 3 within six hours from the time of discovery.
Reasons:
A. The rod is not only inoperable but untrippable which requires being in Mode 3 within six hours.
B. & C. The specifications requiro a shutdown for an untrippable rod.
NEW; Technical Specification pages 3.16 through 3.18; 2/3 i
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39. ROT 452 001/ B10/10040101003/ 004 K4.08// 3.2/ iil MU The Make up Tank (MUT) low pressure alarm has.iust actuated with the level at 80 inches. Which of the followingis the amount that the pressure should be increased from its current value to operate in the preferred region without receiving further MUT alarms?
A.
12 psig, vB.
14 psig.
C.
10 psig.
D.
18 psig.
Reasons:
A. The low pressure alarm is at 3 psig. Raising the pressure by 12 psig (15 psig)
is in the acceptable region.
C. At 19 psig the computet high pressure alarm will be actuated.
D. At 21 psig both the computer and annunciator high pressure alarms will be actuated.
NEW; 0040101015; 0040101024;OP 103B pages 30 33; ROT 4 52 pages 14 10; 2/3 FINALNRC.TST Version: 0 Page: 40
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40 noT-4 53 ootl83//0060501001/oo6A1.16//4.2/33/CF I
Which of the following describes the response of the Reactor Coolant System JS)if core 41ood tank (CIT) injection was delayed as a result oflow CFT pressure during a large break Loss of Coolant Accident (LOCA) at EOL?
.
'
A.
Incore temperature will reach saturation for the existing pressure and continue to decrease due to steam cooling.
B.
Incore temperature willincrease due to decay heat generation. CFTs will not inject.
C.
Incore temperature will reach saturation for the existing pressure and will reach equilibrium due to steam cooling.
VD.
Incore temperature will reach saturation for the existing pressure then increasc until CFT injection.
Reasons:
A.. B. & C. The RCS pressure for a LBLOCA will decrease rapidly to saturation for existing temperature. If CFTinjection is late Incore temperature will begin to heatup.
NEW; ROT 4 53 pages 10 and 11; ROT 3 21 page 32; 1 FINALNRC.TST Version: 0 Page: 41
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41. noT454 oo1/etyloos0101oo3/oo25AK2.02//3.2/3NEOP 11 The following plant conditions exist:
The reactor is shut down.
g Core cooling is provided by Decay Heat Pump 1A (DHP 1A).
i,,;
DHP 1B is in stan' by.
?
d DHP 1A discharge flow is oscillating between 500 gym and 2500 gpm.
g Based on the above conditions which of the following action (s),if any, should be taken to maintain core cooling?
A.
Throttle closed on DHV 3 (reactor coolant outlet) to limit flow to DHP1A.
3.
Start DHP 1B to increase core cooling flow.
C.
Start DHP 1B and trip DHP 1A.
vD.
Trip DHP 1A and ttart DHP 1B after resolution of the problem.
Reasons:
A.
DlIP 1A is cavitating and must be tripped.
B. DHP 1A is cavitating and must be tripped. DHP 1B should not be started until the reason for cavitation of DHP 1A is resolved.
C. DHP 1B should not be started until the reason for cavitation of DHP 1A is resolved.
NEW; ROT 4 54 B8; 0050101008; OP 301, page 5; OP 404, page 7; 2/3 FINALNRC.TST Version: O Page: 42
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42. ROT 4-$5 001/B1//0050101022/00$K6.11//2.7/33/DC The following plant conditions exist:
-,
The plant is in Mode 5.
A" decay heat train is in service.
"
The operator l>as been instructed to initiate a plant heatup.
Which of the following describes the response of the Decav Heat Heat Exhanger (DHHE) cooling water inlet, outlet and bypass valves as the operator reduces cooler demand?
.
A.
The inlet valve is throttled in the closed direction, the bypass valve is throttled in the open direction and the outlet valve is open.
B.
The inlet valve is throttled in the closed direction, the bypass valve is closed and the outlet valve is throttled in the closed direction.
vC.
The inlet valve is open, the bypass valve is throttled in the open direction and the outlet valve is throttled in the closed direction.
D.
The inlet valve is open, the bypass valve is throttled in the closed direction and the outlet valve is throttled in the closed direction.
Reasons:
A. & B.
The inlet valve is always full open and does not throttle in the closed direction.
D. The bypass valve should be throttled in the open direction.
MODIFIED BANK; ROT 4 55 5; 0050101033; 0050101024; 0050101025; 0050201008; ROT 4 55 pages 10 and 11: 1 FINALNRC.TST Version: 0 Page: 43
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43. Rot 4 se oot/stilooooiotoost0640ooksot/s.s/4.1/sstsw i
During a Loss Of Off site Power (LOOP), concurrent with'an ES actuation, the low i
pressure auto-e6ert interlocks fot SWP 1A and SWP.1B are defeated. What is the
purpose of defeating these interlocks during a LOOP?
l
.
>
.
A.
Prevent a simultaneous start of SWP.1A and SWP.1B when tho 4160V
>
ES buses are re energized by the off site power transformer.
B.
Prevent a simultaneous start of SWP.1A and SWP.1B to prevent i
overpressurizing the SW header.
C.
Prevent a simultaneous pump start from overloading the 480V ES buses.
- D.
Prevent the pump starts from presenting an untimaly and abnormal loading sequence to the emergency diesel generators.
i
t t
t Reasons:
!
l
'
A. Normally only one ES bus is supplied by the off site power transformer however a simultaneous start of both pumps is not a transformer concern.
B. Both pumps are commonly in service for a period of time when rotating
,
operating equipment.
C. SWP.1A & IB are fed from the ES 4160V buses.
P t
Bank Question ROT 4 56 #2
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FINALNRC.TST Version: 0 Page: 44
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44. - noT4ss 001/ s 1// 0000101002/ 0024AA1.20t / 3.3/ 3s/ cA
!
Following a reactor trip control rods 21 and 6 7 have not fully inserted. Which of
the following actions, if any, should be taken to assure adequate shutdown margin
!
exists?
!
l
i
l
A.
No operator action is required with two rods partially inserted.
Adequato shutdown margin exists.
l VB.
Open CAV.60, the emergency boration valve and start CAP.1A or CAP 1B, boric acid injection pumps.
!
,
-C.
Open CAV.58, the normal boration valve and start CAP 1A or CAP.1B,
- boric acid injection pumps.
D.
Open MUV 58 and MUV.73, makeup pump suction valves from the borated water storage tank (BWST), and start the second makeup pump.
Reasons:
,
A. E0P 02 requires emergency boration if > 1 control rod is not fully inserted.
,
C. CAV 58 is not in the boration flow path using CAP.1A or CAP.1B.
D. There is no requirement for starting a second makeup pump.
.
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' 4 5. ROT 4 59 001/ B2/ / 0680101009/ 068A3.02/ / 3.6/ 33/ UQ WST -
"/hile pumping the Reactor Coolant Drain Tank (RCDT) to the Miscellaneous Waste Storage Tank (MWST) a small break LOCA occurs. Reactor coolant system pressure has dropped to 1470 psig and the reactor building pressure has increased to three (3) psig. Based on these conditions which of the following describes the status of the RC.DT 6utflow?
4 A.
OutDow continues to the MWST.
B.
OutGow is diverted to a Reactor Coolant Bleed Tank.
C.
Outflow is diverted to the Reactor Building Sump.
vD.
Outflow is terminated.
Reasons:
,
A., B. & C. An HPI actuation will close all diverse containment isolation valves.
In the flow path for pumping the RCDT to anywhere one of these isolation valves exists.
NEW; 0680101011; ROT 4 59 pages 2 and 7; ROT-413 pages 71 and 72; 2/3 FINALNRC.TST Version: 0 Page: 46
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46. ' ROT-4-60 002/ L j / 0020101018/ 0015AA2.10/ / 3.7/ 33/ RC.
- The following plant conditions exist:
- ,
MUV 16, seal injection Dow control valve, failed closed while in automatic.
MUV 16 manual control has been selected and sealinjection flow is being i.
restored.
-
.
e Sealinjection flows to each Reactor Coolant Pomp (RCP) are:
RCP-1A. 8 gpm.
RCP 10 0 gpm.
RCP 1D 8 gpm.
,
,
What action (s) should be taken with respect to the RCPs if SW Dow to the RCPs is lost?.
A.
- RCP 10 should be tripped within Sve minutes. All other RCPs can continue to operate.
VB.
RCP 10 should be tripped immediately. All other RCPs should be
,
tripped within five minutes.
C.
All RCPs must be tripped immediately.-
,
D.
All RCPs should be tripped within five minutes ofloss of sealinjection.
Reasons:
A. & D. RCP 10 must be tripped immediately due to l'oss of both SI and SW.
C. RCP 1A, IB & ID may operate a maximum of five minutes with a loss of SW.
NEW: OP 302 pages 6 and 6; 2/3
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47. ROT-4-60 003/09//0020101018/003K6.02//3.1/33/RCS
'
The following conditions are observed for the "A" Reactor Coolant Pump, RCP 1A.
~
First seal cavity pressure is 2150 psig.
Second seal cavity pressure is 1100 psig.
- Third seal cavity pYessure is 1055 psig.
Controlled bleed off flow has increased.
Seal leakage flow has not changed.
Which of the following would cause the above indications?
.
-
A.
Seal number 1 has failed.
vB.
Seal number 2 has failed.
C.
Seal number 3 has failed.
D.
Restriction bushing has failed.
Reasons:
A. First seal cavity pressure should be reactor coolant pressure (2155 psig) if working correctly.
C. When the second seal has failed then the cavity pressures of the second and third seal should be approximately one half of reactor coolant pressure.
D. The restriction bushing limits reactor coolant Gow when a major seal failure has occurred. The data does not support bushing failure.
B.ANK; ROT-4-60119; NRC 5 93; ROTS J T10B & FPC Final 96; ROT-4-60 pages 14 and 15; 1 FINALNRC.TST Version; O Page: 48
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48. ROT-4-6o 004/B4//0020101009/011A1.03//3.2/11/MU Reactor power is 5% with a plant startup in progress. Tave is 547'F and pressurizer levelis 150 inches.' Which of the following describes the status of pressurizer level and the actions, if any, that should be taken.
~
.
VA.
Pressurizer level is high; increase letdown flow.
B.
Pressurizer level is high; no actions are required unless pressurizer level exceeds 290 inches.
C.
Pressurizer level is low; increase make up flow.
D.
Pressurizer level is low; no actions are required unless pressurizer level decreases to < 40 inches.
Reasons:
B. Pressurizer levelis high and must be lowered.
C. and D. The curve in OP 103A indicates that pressurizer levelis no high but low.
NEW; ROT-4-60 page 6; OP 203 page 15; OP 103A page 8; 2/3 FINA8 NRC.TST ' Version: 0 Page: 49
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49. - rot 441 ooit rel/ 071050s001/ 071 A2.02// 3.s/ sat wo '
_i L During a release of the'"B" Waste Gas Decay Tank, (WDT 1B), the release header.
>
- outlet isolation valve (WDV 437) closes.
-
t Which of the following could cause WDV 437 to close?
.
I
.A.
RM A8, auxiliary building exhaust radiation monitor, in high alarm.
,
B..
One of the auxiliary building supply fans has tripped.
,
- C.
WDV-478, release isolation valve, has closed.
r vD..
High flow as indicated at WD 19 FE, waste gas release flow recorder.
-Reasons:
'
A. RM A8 does not have any interlocks.
B. Both fans have to be tripped to close WDV 437.
,
C. WDV 478 is a manual valve with no interlocks.
.
'
NEW; 0710106002; 0710106006; 0710106007; 0710106009; 0710106011;
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50. ROT 4-62 001/B8//1190301015/026K3.01//4.1/11/ BS
,
The following equipment is inoperable at 85% power:
. BSV 3, the "A" train Reactor Building Spray control valve.
_
. Containment cooling units AHF.1B and AHF.10..
'
,
-What would be the required actions for these conditions?
.
J A.
Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> restore BSV-3 and AHF 1B to operable.
-B.
Within 7 days restore BSV-3 and AHF.1B to operable,
,
VC. -
Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> restore BSV-3 to operable and within 7 days restore
,
AHF 1B to operable.
D.
Within 7 days restore BSV 3 to operable and within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> restore AHF 1B to operable.
Reasons:
.
,
A. With one required containment cooling train inoperable restore to operable in
' 7 days.
B. With one reactor building spray train inoperable restore in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
D. -With one required containment cooling train inoperable restore to operable in
7 days. With one reactor building spray train inoperable restore in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
. NEW; Technical Specifications pages 3.617 and 3.618; 2/3
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51. ~ ROT 443 001/ B6//1030101002/022A3.01//4.3/33/ RB VENT A steam leak has developed in the reactor btiilding (RB). The following conditions -
exist:
--
. RB pressure is 4.2 psig.
RB average temperature is 180'F.
- RCS pressure is 2155 psig.
,
Which of the following describes the status of the RB main fans (AHF-1A,1B, and
IC)?-_
.-.
.
,
_
[A.
- The ES selected RB main fans are in fast speed and being cooled by
,
Nuclear Services Closed Cycle Cooling.
B.-
The ES selected RB main fans are in fast speed and being cooled by
.
-Industrial Cooling.
vC.
The ES selected RB main fans are in slow speed and being cooled by Nuclear Services Closed Cycle Cooling.
D.
' The ES selected RB main fans are in slow speed and being cooled by
'ndustrial Cooling.
,
Reasons:
A.. RBIC has actuated which cascades a signal for an HPI actuation. HPI causes the ES selected RB main fans to start in slow speed.
B. RBIC has actueted which cascades a signal for an HPI actuation. HPI causes the ES selected ItB main fans to start in slow speed. - RBIC will also cause the fans cooling water to transfer from CI to SW.
D. RBIC has actuated and will also cause the RB main fans cooling water to transfer from CI to SW.
NEW; ROT-4 63 pages 15 and 23; 2/3 FINALNRC.TST Version: O Page: 52
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' 52. - rot 443 002/ sill 1030101oo2/1osA1.01//4.1/ 33r AH
-
The following pla.nt conditions exist:
I A Loss of Cool nt Accident (LOCA) has occurred.
'
- Reactor coolant pressure is 730 psig.
!
' L Reactor building p'ressure is 7 psig.
l
.
. Reactor building temperature is 230'F.
.
-
Under these conditions what will prevent Reactor Building pressure and temperature from exceeding design limits?
'
.
l A.
Two reactor building operating floor fans running.
- l VB.
Two reactor building air handling units running.
C.
' Two reactor building cavity cooling fans running.
-
+
D..
Two reactor building steam generator compartment cooling fans l
running.
Reasons:
A. The reactor building operating floor fans will not prevent the RB from
!
exceeding design limits.
~
C. The reactor building cavity cooling fans will not prevent the RB from
- exceeding design limits.
L D. The steam generator compartment cooling fans will not prevent the RB from
'
exceeding design limits.
NEW; ROT 4-63 pages 21 through 23; 2/3
g H FINALNRC.TST Version: O Page: 53
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53. Rowes 003/ B1//1030101002/ 2.3.9/ / 3 A/ 33/ PURGE A reactor building (RB) purge is in' progress when the following occurs:
,
-..
AHF 7A, the "A" purge exhaust fan trips (AHF 7B is still operating).
RM Al gas channel has reached its warning setpoint.
What is the condition of the purge supply fans (AHF GA/0B) and AHV 10 (supply inside RB valve)?
A.
Both fans a-o operating; AHV 10 is open.
B.
Neither fan is operating; AHV 1C is open.
vC.
Neither fan is operating; AHV-10 is closed.
D.
One fan is operating; AHV-10 is closed.
Reasons:
A. Both purge exhaust fans must be running for a supply fan to remain running.
AHV-10 automatically closes with the loss of both supply fans.
B. AHV-10 automatically closes with the loss of both supply fans.
D. Both purge exhaust fans must be running for a supply fan to remain running.
NEW; 1030101003; 1030101004; 1030101005; 1030101006; 0220101003; 0220101004; 0220101008; ROT-4 63 pages 31 through 33; OP-417 page 4; 2/3 FINALNRC.TST Version: O Page: 54
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- 54. ' ROT-4-64 001/F2//0630106001/063K4.04//2.9/88/DC
'
The following readings were taken on the "A" battery charger following a
- BATTERY A Ol6 CHARGE HIGH alarm in the control room
>
- - 120 volta
- 60 amps :
.
Shortly after these readings were taken the amp meter increases to > 350 amps.
.
Which of the following action (s) will occur following this increase?
A.
The "A" inverter will trip.
sB.
The "A"_ battery charger will trip.
.
C.
The "A" inverter will not trip but will swap to the AC input.
D.
The "A" battery charger will trip and the "C" battery charger will automatically be placed in service.
Reasons:
A. There may be problems with the inverter but it sli'ould not trip.
C. There may be problems with the inverter but it should not trip. The inverter is normally supplied from its AC source.
D. The "A" charger will trip but the "C" charger does not have any automatic closing feature.
NEW: ROT-4 64 pages 4 and 5; AR-701 pages 62 and 84: OP-705 page 3; 2/3
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' FINALNRC.TST Version: 0 Page: 55
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' 5 5. ROT 4-66 001/G5//0390104004/039K1.02//3.3/77/MS
- The following plant conditions exist:
- - The plant was initially operating at 100% power near the end of core life.
- A sudden pressure fault on the Startup transformer has caused a reactor trip.
-- Main steam (MS) pressure is stable at 1025 psig.
"
Which of the following conditions is causing the elevated MS pressure?.
.
A.
The ICS header pressure setpoint is set too high.
B..
Feedwater flow is too low causing elevated reactor coolant temperatures.
.C.
Turbine bypass valve demand is at maximum due to high decay heat.
,
YD.
' The atmospheric dump valves are controlling steam pressure.
Reasons:
J A., B. & C.
With a loss of the start up transformer, no Circulating Water Pumps (CWPs) will be operating. The Turbine Bypass Valves
_
(TBVs) are interlocked with the operation of the CWPs. Pressu e
~ increased and is being controlled by the Atmospheric Dump Vs.ves (ADVs).
F
- MODIFIED BANK; ROT-4 66 22; ROTS J Final 96; ROT-4 66 pages 4 and 5; 2/3
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FINALNRC.TST Version: 0 Page: 56
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56. ROT-4-69 001/ B2/ / 0560101007/ 056K5.03/ / 2.6/ 44/ CD At full power operation annunciator window N 04 02, CNDSTE HTR 1A/B DUMP lVALVES NOT PULL CLOSED, comes into alarm.. Several minutes later the SPO -
calls to report that' CDHE 1A is below its normal water level (NWL). -What are.
the implications whe,n CDHE 1A is operating at a lower than normallevel?
,
_
A.
. The heater subcooling section overcools the heater drains increasing the
'
extraction steam flow to the heater.
o
'
B..
The heater subcooling section does not extract enough heat'from the heater drains which will decrease the condensate temperature.
_
C.
- The heater subcooling section overcools the heater drains decreasing the
.
possibility of water hammer.
vD.
.The heater subcooling section does not extract enough heat from the heater drains increasing the possibility of water hammer.
Reasons:
'
A. & C. A low levelin the heater subcooling section does not overcool the' drains.
Not enough heat is extracted.
B. If the heater subcooling section did not extract enough heat then the condensate temperature vrould increase.
-NEW; 0560101012; 0560101013; 0560101014; 0560101019; 0560401002; 0560401003; ROT-4 69 pages 47 and 48; AR-602 page 29; I f
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5 7. ~ ROT-4-69 002/ B8// 0550401001/ 055K3.01// 2.7/ 33/ AR
.The following plant conditions exist:
_.
.
- The "A" Air Removal Pump (ARP 1A) has tripped.
-- ARP 1B auto-starts.
-- Condenser vacuuni is 26 inches Hg and decreasing.
Unit load is at 85% full power.
What will be the mode of operation of ARP-1B and the status of the plant it condenser vacuum decreases to 25 inches Hg? -
A.
ARP 1B will be in the holding mode and the plant will be tripped.
B.-
ARP 1B will be in the hogging mode and the plant will be tripped.
vC.
ARP-1B will be in the holding mode and the plant will be at 85%.
D.
ARP-1B will be in the hogging mode and the plant will be at 85%.
Reasons:
A. ' Condenser vacuum did not get low enough for the plant to trip.
B. At 25 inches Hg ARP 1B should be in the holding mode. Condenser vacuum did not get low enough for the plant to trip.
D. At 25 inches Hg ARP 1B should be in the holding mode.
- NEW; ROT;4-69 pages 52, and 55; OP 607 pages 3,10 and 11; AR-602 page 2; 2/3
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58. ROT 4 77 001/G4//07401040041075K1.08//3.2/77/SC The following plant conditions exist:
.-
-- A plant heatup has commenced.
Secondary Services Closed Cycle Cooling (SC) is unavailable.
'
What components should be aligned to the Nuclear Services Closed Cycle Cooling System (SW) to support the start of a Circulating Water Pump (CWP)?
A.
The "C" station air compressor and a waterbox priming pump. -
v8.
The "A" instrument air compressor and a waterbox priming pump.
C.
Tha "C" station air compressor and a condenser air removal pump.
D.-
The "A" instrument air compressor and a condenser air removal pump.
.
Reasons:
,
A. SAP 10 is air cooled.
'
C. SAP 10 is air cooled. ARP 1A cannot be cooled by SW.
^
D. ARP 1A cannot be cooled by SW.
NEW; 0740104005; 0740104007; 0740104008; 0740104009; 0740104010; OP 408 page 20; AP-770 page 57; 2/3
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5 9. ROT-4-91 001/ A1// 3410103036/ 062A3.01/ / 3.1/ 55/ VB -
The "A" inverter,.VBIT-1A, is in service with the following meter indications:
,
- Battery Input-220 amps-
.
DCInput 0 amps
-
'
-DC Volts-120 volts
.
Inverter Output 200 amps Inverter Output
- 120 volts
.
Inverter Output --
_60 hertz
- Which of the following descrwes the c.ondition of the inverter and the actions, if any, that should be taken? -
,
i-A.
VBIT-1A is operating properly with the normal source supplying power.
,
- No action is required.
.-_ B.
VBIT 1A is operating properly with the battery source supplying power.
No action is required, vC.
VBIT 1A is not operating properly. Bypassing should be considered.
D..
VBIT 1A is not operating properly. The inverter must be bypassed.
Reasons:
'
-
.
A., B & D.
Battery input amps is abnormally high, The battery source is supplying power to the inverter, The AR suggests that consideration should be given to bypassing the inverter.
NEW;: ROT-4 91, pages 2,3 and 6; AR 701 page 61; 2/3
,
1 FINALNRC.TST Version: 0 Page: 60'
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60. ROT 4-91002/F2// 0620100002/0068AA2.01//4.1/88/VB:
.
_
The "A" Vital Bus Inverter (VBIT 1A) is powered up and operable when DC power
~
is lost.- Whictrof the following is the status of VBIT-1A's indicating li ; hts?.
t
.
,
.
.-
vA.
~ Normal Source Available ON: Battery Source Available OFF; Normal
{
- Source Input ON; Battery Source Input OFF; In Sync ON.-
B.-
Normal Source Available ON; Battery Source Available OFF; Normal
[
,
Source Input ON; Battery Source input OFF; In Sync OFF.
'
C.
Normal Source Available ON Battery Source Available OFF; Normal Source Input OFF; Battery Source Input ON; In Sync ON.
D; Normal Source Available OFF; Battery Source Available OFF; Normal-Source Input OFF; Battery Source Input OFF; In Sync OFF.
Reasons:
B., C. & D.
If the battery input is lost while the VBIT is powered up then the
'
lights will have the following con 5guration: Normal Source Available ON: Battery Source Available OFF; Normal Source Input
'-
ON; Battely Source Input OFF; In Sync ON.
,
NEW; ROT-4 91 pages 3 and 19; I
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61~. ROT 5-01001/A9//1190201001/2.2.21//3.5/55/TS
- Following maintenance on the First Ifvel Undervoltage Relays (FLUR) and-Second Level-Undervoltage Relays (SLLR) for the."A" Emergency Diesel Generator, EDG 1A, the electricians perform SP 907A, Monthly Functional Test of 4160V ES Bus "A" Undervoltage and Degraded Grid Relaying, with the following results:
~
SLUR relay A 3947 volts in 4,7 seconds.
- SLUR relay B 3932 volts in 5.3 seconds.
SLUR relay C 3953 volts in 5.1 seconds.
FLUR relay A
- loss of voltage in 7.9 seconds.
,
. FLUR relay B loss of voltage in 7.6 seconds.
FLUR relay C loss of voltage in 8.0 seconds.
What are the required actions, if any, for these resulta?
A.
All results are within specification - no action ned be taken, vB.
Trip one channel of SLUR.
C.
Trip one channel of FLUR.
D.
Declare EDG 1A inoperable.
Reasons:
A. One SLUR relay does not meet its surveillance requirements.
J C All FLUR relays are within their surveillance requirements.
D. Only one SLUR relay does not meet its surveillance requirements therefore,-
EDG 1A does not have to be declared inoperable.
NEW; Technical Specifications pages 3.3 20 and 3.3-21; 2/3 FINALNRC.TST Version: 0 Page: 62 L
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- 62. ROT 5-01002/Aill1150101003/2.2.24//3.8/55/TS The following events.have occurred:-
-
. One bank ofIkalon bottles has discharged..
No fire is indicated and the reason for the discharge is unknown.
The back up b3nk bf halon bottles has not discharged.
~ What action (s) must the Shift Supervisor On Duty, SSOD, take?
A.
- Select the back up bank for automatic fire suppression. Restore operability as soon as possible.
B.
Select the b'ack up bank for automatic fire suppression. Provide the
.
- unprotected area with backup fire suppression equipment, vC.
Establish a continuous fire watch with backup fire suppression
-
equipment and restore operability.
D.
Notify American Nuclear Insurers (ANI), restore operability and submit
'
a special report to the NRC.
Reasons:
"
A. & B.
The back up bank should not be selected until. the cause of the initial bank discharging is known.
D. In addition to the reports it is required to establish some backup form of fire suppression.
1 ^
'_
NEW; 1150401010; 1190101015; 1190101017; 1190201001; 1190201002; Fire Protection Plan pages 135 and 136; 2/3 i
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63. - ROT 5-01003/ A1//1150101003/2.2.22//4.1/55/TS
. The plant is operating.at 10% FP when the following events occur:-
. The operating RM.A0 (containment monitor) pump trips.
.
- RM-A6 Dow alarm actuates.
. The back up RM A6 pump fails to start.
Which of the following. actions are required if neither pump can be started?
VA.
Analyze grab samples each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and have one of the pumps operable within 30 days.
,
B.
Perform SP 317, Reactor Coolant System Water Inventory Balance, and analyze grab samples every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
!
C.
Analyze grab samples each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or have one of the pumps operable within 30 days.
D.
Perform SP 317, Reactor Coolant System Water Inventory Balance, every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or have one of the pumps operable within 30 days.
Reasons:
B. To complete the required actions of TS 3.4.14 either of these actions are taken and one of the pumps must be made operable within 30 days.
C. & D.-
To be correct these statements would have to be "and" not "or".
,
MODIFIED BANK; ROT-5 0182; 1150401010; 1190101015; 1190101017; 1190201001; 1190201002; Technical Specifications pages S.4 27 and 3.4-28; 2/3
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64. ROT 5-01004/ A1//1150101003/0076AK3.05//3.6/55/TS The plant is at 100% full power when the letdown radiation monitor, RM L1, aiarms. Chemistry is notified and, after sampling, returns with the following data:
Dose equivalent I 131 is 0.02 pCi/gm.
Reactor coolant gross specific activity is 150/E bar pCi/gm.
What technical specification action, if any, should be taken?
VA.
Be in Mode 3 with Tave < 500'F in six hours.
B.
Verify dose equivalent I 131 within acceptable region and restore within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
C.
Verify gross specific activity within acceptable region and restore within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
D.
No technical specification applies fo: these conditions.
Reasons:
B. Dose equivalent I 131 is within its technical specificatior, limit.
C. Gross specific activity does not have an acceptable region, it is either within limit or not.
D. Gross specific activity is outside its technical specification limit, need to go to surveillance to determine what the limit is.
NEW; Technical Specifications pages 3.4 30 through 3.4 33; 2/3 FINALNRC.TST Vercion: 0 Page: 65
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- 65. - R0T.5-01005/ 99//1190302001/ 0033AK3.01// 3.6/11/TS.
- --While taking critical data following a reactor startup, the operator discovers that'
-
,
NI 3 has faile7 to 10"11 amps.J Which of the following describes the limifations, if
'
.
'
any, on reactor power until NI 3 intermediate range can be fixed? -
.
-
A.
Within one hour action rust be taken to place the unit in Hot Standby.
B.
-Within one hour power must be reduced to < 1010 amps.
'
,
vC.
Power may not be increased above 5% full power.
-D.
Power may not be increased above the POAH.
Reasons:
A., B. & D.
If one of the two intermediate range NIs are inoperable, Technical
Specifications require restoring the inoperable NI to operable prior to entry into Mode 1.
,
'l BANK; ROT 5-0185; Technical Specifications pages 3.3 24; 1
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66. ROT 5-14 001/88//NTS/2.4.0//4.0/33/EOP/AP During the performance of EOP 04, Inadequate Heat Transfer, the operator notes that subcoolinganargin is lost. Which of the following actions should be taken?
~
.
vA.
Immediately go to EOP 03, Inadequate Subcooling Margin, and return to EOP 04 only when directed.
B.
Complete the actions of EOP 04 then go to EOP 03 ifinadequate subcooling margin still exists.
,
C.
Continue in EOP 4 until directed to go to EOP-3.
D.
Go to EOP 03 until adequate subcooling margin is regained and then return to EOP 04.
Reasons:
B. & C.
EOP-03 is a higher symptom than EOP 04. The operator should exit EOP 04 once the inadequate subcooling margin is noted and enter EOP-03.
D. Returning to EOP 04 should only occur if the procedure directs it.
BANK; ROT-51413; ROT-514 pages 25 through 27; 1
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T 67.. ROT 514 002/B10f /NTS/2.4.2//4.1/33/EOP/AP-
,
Which of the following.would require the operator to manually trip the reactor? -
-,
f
'
-A.-
The turbine trips at 60% full power. -
-B.
The nuclear services closed cycle cooling surge tank, SWT 1, has a level.
'
of 4 feet.
VC.
Pressurizer level is 95 inches at 100% full power.
D.
One MSIV closes at 80% full power.
Reasonr A. The reactor protection system would trip the reactor for this condition.
-
B. The manual trip for low SWT 1 levelis 2 feet.
.
D.. Power operation may continue if reduced to 60% full power.
i-NEW; AI 505 pages 15 and 10; 2/3 FINALNRC.TST: Version: 0:
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68. - ROT 514 003/ BS// NTS/2.4.11// 3.6/ 33/ EOP/AP The following plant conditions existi j
.,
.
a
- A reactor trip has occurred..
' RM.A3, Auxiliary Building Exhaust (Waste Gas Area), is in high alarm. -- _
-
-
- The NNI X status light on the Redundant Instrument Panelis extinguished.
"
. ' All the lights on the NNI X power supply monitor are lit.
'
,
,
_
Which of the following describes the Procedure Director's response to these indications? =
j
'
,
v-I
'
A.
Enter EOP 02, Vital System Status Verification, and complete prior to performing AP 250, Radiation Monitor Actuation.
,
B.
Concurrently perform EOP 02 and AP 581, Loss of NNI X, by completing first the immediate actions in both procedures and then performing the
follow ups. -
i vC.
Enter EOP 02 and once the immediate actions are complete concurrently perform AP 250 and AP 581.-
D.
Enter EOP 02 and once the immediate actions are complete concurrently
- perform AP 250.
_
Reasons:
A. EOP 02 does not need to be completed prior to entry into AP 250.
B; EOP 02 immediate actions should be completed prior to entering another
.
procedure.
D. A loss of NNI X has also occurred. AP 581 should be concurrently performed.
,
n
..NEW; ROT-514 page 25; 2/3 o
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69. ROT 5-29 001/ 01// 0450401001/ A04AA2.1// 3.7/33/ AP-660
,
The following plant conditions exist:
.
,
. The plant is at 30% power.
Turbine vibration on bearing #7 is 10 mils.
- Stator bar dischari;e gas temperature is 78'C.
Average cold gas temperature is 60*C.
Based on the above conditions which of the following actior.(s) should be taken?
A.
Enter AP 510, Rapid Power Reduction, and perform a plant shutdown.
B.
Enter OP 204, Power Operations, and perform a plant shutdown.
C.
Enter EOP 02, Vital System Status Verification, and trip the reactor.
vD.
Enter AP 660, Turbine Trip, and trip the turbine.
Reasons:
A., B. & C. Average cold gas temperature is above the limit of 55'C per AP 660.
The first immediate action of this AP is to trip the turbine.
NEW; 0510401001; 0550401001; 1150401004; ROT 5 29 pages 1 and 2; 1 FINALNRC.TST Version: 0 Page: 70
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..70. -' ROT 5-30 001/ A1/ ROT 457/ 3440403001/ 076K2.01// 2.7/ 88/ RW '-
Following a loss of off site power which of the raw water pumps (RWPs), if any,-
'
would be operating??
'
.
~
.
A,
' Heat RWPs, RWP 3A and RWP 3B.
,
B.
"B" Emergency Duty Nuclear Servi.;es RWP, RWP 2B; "B". Decay' Heat RWP, RWP 3B.
!
C.
Normal Duty RWP, RWP-1.
,
,
vD.
No RWPs would be operating.
Reasons:
A. and B. There was no ES actuation so RWP 2A,2B, SA, and 3B would not receive a start command.
C. RWP 1 is powered from the "A" Unit 4160V bus which is not powered from the -
.
diesels.
NEW; ROT 4 57 F2; ROT 5 30 page 25; ROT 4 57 pages 9,10, and 14; 2/3
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71. ROT 5-30 002/B3i/0640401003/A05AK1.1//3.7/30/AP 770 During an ES 4160V bus undervoltage, bus feeder breakers are stripped prior to the emergency diesel generator re-energizing the bus. Which of the following is the reason for this action?
-
.
A.
To ensure that there is no load on the bus when it is re energized.
B.
To prevent paralleling both Emergency Diesel generators.
vC.
To prevent any fault on the system from feeding to the ener'gizing source and damaging it.
D.
To ensure that block loading will occur correctly.
Reasons:
A. Some loads ride the bus down and do not cause a problem for breaker closure.
B. The cross tie blocking circuitry prevents this possibility.
D. Block loading occurs after the diesel breaker closes. The diesel breaker will not close unless all feeders are open.
BANK; ROT 5-301; ROTS J T10A; ROT 5 30 pages 6 and 7; I FINALNRC.TST Version: 0 Page: 72
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.72. ROT.5 31001/ Ai/ RCT-4-16/ 34404030011004A3.02/ / 3.6/ 33/ MU '
l
' The following plant conditions exist: -
-1An immediate control room evacuation is required..
I The plant was initially at 75% full power.
- _.The only actions tsken by the operators prior to evacuating the control room is -
to announce that the control room is_being _ evacuated, trip the reactor and
,
actuate EFIC.
Which of the followi' a describes how letdown should be isolated after the control room has been evacuated?
-
VA.
Close MUV 49 and 50 Oetdown isolation valves) locally.
'B.
Close MUV 49 and 50 (letdown isolation valves) at tha Remote Shutdown Panel.
-
C.
Close MUV 40, MUV 41 and MUV-505 Oetdown cooler outlet isolation valves) at the Remote Shutdown Panel.
D.:
Close MUV 40 and MUV-41 at their breakers and MUV 505 locally.
Reasons:
B.: MUV-50 is not controlled from the RSP.
C. Besides the control room the only other place to operate MUV 10, MUV 41 and MUV-505 is from their respective breakers.
D. MUV 40 and MUV-41 can be closed from their breakers but MUV-505 must be closed from its breaker also; MUV 505 has no local control.
.
BANK; ROT-4-52 63; 0040106012; ROT-4-52 page 9; ROT-416 page 14; 3440403007:3440403008:1
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l 73. noT+3a oot/A5// NTS/2.1.1//3 8/44/OI-04 The Shift Supervisor on Duty (SSOD) has star.ed a shift crew briefing when the 311 phone rings, Which of the following personnel,if anyone, should answer this phone?
A.
The SSOD should interrupt the briefing and must answer the call personally.
B.
The call should be forwarded to the clerks's office where the clerk can take a message, vC.
The control board nuclear operator or other avabable operator.
D.
The call should be forwarded to Security until the briefing is complete.
Reasons:
A. The SSOD is not required to answer the call personally.
B. & D. 311 is the emergency phone and should not be transferred to the clerk's office or Security but answered in the control room.
NEW; 0104 pages 4 and 5; 1 FINALNRC.TST Version: 0 Page: 74
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74. ROT.$-43 001/A4//3410103031/2.31//3.0/44/ RAD Which of the following radiation exposures would require immediate notification to the Nuclear Hegualtory Commission (NRC)?
.
A.
,
B.
vC.
D.
Reasons:
A. The immediate notification limit for CEDE is 25 Rem.
B. The immediate notification limit for TEDE is 25 Rem.
D. The immediate notification limit for SDE.WB is 250 Rom.
NEW; ROT 5 43 pages 10 and 11; I FINALNRC.TST. Version: 0 Page: 75
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75. noT+43 002/ e5/i 1100301013/ 2.3.4/ / 3.1/ 33/ RAD A 29 year old radiation worker at CR 3, with a complete NRC Form 5, has received a totaHifetime TEDE of 37.125 rem which includes 0,150 rem he has
received this year. What would be the maximun additional TEDE exposure he could receive during the remainder of this year and still be within CR 3 administrative guidelines?
VA.
0.850 rem B.
2.850 rem C.
3.850 rem D.
4.850 rem Reasons:
B., C, & D.
The annual administrative dose is 1 rem TEDE if the lifetime dose is
> age.
BANK; ROT.5 4310; ROTS J T10B; ROT 5 43 page 3; 1 FINALNRC.TST Version: 0 Page: 76
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76. HOT 5-61001/All/3440403001/0026AA2.01//3.6/44/SW The following plaint conditions exist:
-,
. Reactor power is 100%
. The SW surge tank, SWT 1,is at 6.6 feet and decreasing.
. All SW cooled component temperatures are s, able and within normal parameter %
'
No sump kvels ura increasing.
l
. No SW dib.rential flow meters are indicating leakage.
No SW differential flow switches are in alarm.
'
Based on the above conditions what is the most probable location for the leak? :
A.
The i'sactor Coolant Drain Tank cooler.
v8.
An SW heat exchanger.
,
C.
\\ Post Accident Sample System cooler.
D.
A Reactor Coolant Pump seal cooler.
Reasons:
A.11 SW is leaking in the RCDT, its associated differential flow meter would indicate a leak and the switch would have caused an alarm in the control room.
C. If the PASS cooler was leaking then reactor coolant would be leaking into the SW system.
D. If SW is leaking in the RCP seal cooler, its associated differential flow meter would indicate a leak and the switch would have caused an alarm in the control room.
.
NEW; 3440403008; ROT 4 56 pages 19 through 21; ROT 5 61 pages 17 through 20; 213 FINALNRC.TST Version: O Page: 77
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l 77. R07 5 61002/A2/ ROT.514/3440403001/0062AA2.04//2.9/33/AP-330
.
The following plant conditions exist:
The differential pressure across the in service control rod drive (CRD) filter is 32 pei.
'
Three control rod drive stators have a temperature of 182'F.
'1 One control rod drive stator has a temperature of 180'F.
All the other control rod drive stators have temperatures between 173'F and 178'F.
Based on the above indications which of the following actions is required?
VA.
Trip the reactor.
B.
Place the standby CRD filter is service.
.
C.
Start the standby SW booster pump.
D.
Start the emergency SW pump (SWP.1N1B).
Reasons:
B.
The CRD filter has exceeded its DP and should be changed but tripping the reactor needs to be done immediately since multiple CRD stators are > 180*F.
C. & D. The reactor needs to be tripped immediately since multiple CRD stators are > 180*F.
MODIFIED BANK; ROT.5 612; ROTS J T10A; ROT.514 B2; ROT.5 61 pages 3, 14,15, and 16; 2/3
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78. ROT.5-68 0011 A2// 3440403001/ A01 AK1.3// 3.7/ 44/ AP 545 Which of the following. sets of conditions would require a plant runback to 55%?
,
A.
"A" Main Feedwater pump lube oil pressure of 10 psig and a deacrator level of two feet.
B.
"B" Main Feedwater pump lube oil pressure of 0 psig and a deaerator level of two feet, vC.
"A" Feedwater Booster ptnap lube oil pressure of 0 psig and a deaerator level of six feet.
D.
"B" Feedwater Booster pump lube oil pressare of 10 psig and a deacrator level of six feet.
Reasons:
A. Low Denerator level will trip both booster pumps, which will trip both main feedwater pumps which will cause a reactor trip, not a plant runback.
B. The "B" main feedwater pump will trip on low lube oil pressure but low Denerator level will trip both booster pumps, which will trip both main feedwater pumps which will cause a reactor trip, not a plant runback.
D. Oil pressure and donerator level are above booster pump trip setpoints.
NEW; ROT 5 68 pages 3 and 4; AR 504 page 30; OP 605, page 4: 2/3 FINALNRC.TST Version: O Page: 79
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79 R07 5 72 001/ A2//3440403001/033A2.03//3.5/44/ AP 1080 Twenty minutes prior to completion of defueling activities the Spent Fuel (SF)
,
poollevel war 168 feet. Defueling has now been completed and the following -
plant conditions exist:
'
Spent Fuel (SF) pool level is 156 feet and decreasing.
. The SF pool low level annunciator is in alarm.
The transfer tube valves are open.
.
- The auxiliary building sump levelis increasing.
. All other building sumps are stable.
'
. SFP.1B is operating; SFP 1A is secured.
. Reactor building pressure is 1 psig.
Which of the following could stop the decrease in Spent Fuel poollevel?
A.
Close the transfer tube valves.
VB.
Secure SFP.1B.
C.
Trsnsfer SF heat exchangers.
D.
Secure the reactor building purgo.
'
Reasons:
A. The leak is in the AB, Closing the transfer tube valves would only prevent the transfer canal level from decreasing.
C. If the heat exchanger had a leak then SW would be leaking into the SF system.
D.' The purge may cause the pressure to lower in the reactor building, which would lower SF poollevel slightly, but would not cause the AB sump level to increase.
--
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t 80 ROT 5-41001183//0160401002/A02AK3.3//3.2/33/ AP 581
!
A step in AP 581, Loss of NNI X, states-
!
-..
Control RCS pressure based For loss of NNI X DC the following on plant conditions using:
equipment is available for pressure
!
controlin the MANUAL mode *
i
~
.
-
-
o PZR (pressurizer) heaters o-PZR htr banks D and E j
o PZR spray o __ RCV.14
o PORV/
o RCV.10
'
.-
-
!
Which of the following is the basis for taking manual control of the PORV?
i
,
A.
RCS pressure input to the PORV has failed low.
B.
The PORV annunciator window and accelerometer are disabled.
C.
The PORV has transferred to its low pressure setpoint.
vD.
The automatic control of the PORV is disabled.
Reasons:
A., B. & C. Power for automatic operation of the PORV is unavailable with a loss
of NNI X DC.
,
NEW: ROT 5 81 pages 3 and 4: 1
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81. ROT-5-82 001/A2/ ROT-4-09/3440403001/0057AA2.15//4.1/44/NNI-Y The following plant conditions exist:
.
The white "Y Power On" indicating liel>t on the Redundant Instrument Panel has extinguished.
Letdown flow contiol valve (MUV 51) has failed to 50% open.
Alllights are illuminated on the NNI X and NNI Y power supply monitor.
- Condensato pump control has transferred to manual.
Based on the above indications which of the following failures have occurred?
.
A.
Loss of NNI Y +24 VDC.
B.
Loss of NNI Y 24 VDC.
4.
Loss of NNI Y 120 VAC.
D.
Loss of NNI Y 120 VAC and 24 VDC.
Reasons:
A., B. & D. If either 24 VDC power supply was lost then some combination of lights on the power supply monitor would be extinguished.
NEW; ROT-4 09 BG; ROT 5 82 pages 2,5,6. 7, and 8; ROT 4 09 page 34; 2/3
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i 82. ROT 5-83 001/B1//0750401002/A07AK2.1//3.5/33/AP-1050 Which of the following. situations require entry into AP.1050, Flooding?
.
'
A.
The control room receives a COND PUMP PIT SUMP LEVEL HIGH annunciatur alarm on both the "A" and "B" Condensato Pump (CDP) pit sumps.
B.
A Building Serviceman calls the control room and reports water running out of the in service Secondary Services Closed Cycle Cooling
'
heat exchanger. Shortly after the call CDP 1A trips.
C.
CDP 1A and CDP 1B decouple and a high level alarm is received for the turbine building sump.
vD.
The control room receives a COND PUMP PIT SUMP LEVEL HIGH annunciator alarm on the "A" CDP pit sump and CDP 1A suddenly
decouples.
Reasons:
A.
Entry into AP 1050 requires this alarm and a running CDP decoupling/ trip, or an indication of flooding.
B. & C. Without the alarm the entry conditions are not met for AP 1050.
MODIFIED BANK; ROT 5 83 3: ROT 5 83 pages 1 and 2; 1 FINALNRC.TST Version: 0 Page: 83
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83. ROT.5-84 001/D3//0780401001/0065AA1.C3//3.1/33/AP-470 Instrument Air (IA) pressure has decreased to 80 psig. Which of the following describes the-Sorrect operation and function ofIAV 30?
.
A.
IAV 30 opens to supply the IA system from Station Air (SA).
8.
IAV 30 opens to provide JA flow from the berm compressors to the IA header.
vC.
IAV.30 closes to prevent flow from the IA system to the SA system.
D.
IAV.30 closes to isolate non essential IA loads.
Reasons:
A., B. & D.
The purpose ofIAV.30 is to close on low Instrument Air pressure to isolate IA from SA. If the cauto ef the low air pressure is in the SA system then it needs to be isolated from IA.
NEW: ROT 5 84 pages 15 and 10; I FINALNRC.TST Version: 0 Page: 84
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I 84. ROT 5 84 002/83/ ROT 4-81/0780401001/079K4.01//3.2/33/lA A step in AP 470, Loss ofInstrument Air, states:
and first loop isolation valves (IAV.2I,22,26 & 27),
'
THEN isolate the leak,
'
.
'
Which of the following describes the purpose of this step?
,
A.
Isolates the in house compressors to provide Instrument Air (IA) from the berm air compressors.
B.
Isolates the berm air compressors to provide IA from the in house compressors, s
'
C.
Isolates the 95' and 119' TB loops from SA to allow repressurization from IA.
- VD.
Isolates the 95' and 110' TB loops from IA to allow repressurization from SA.
Reasons:
A., B. & C. The purpose of this step is to isolate the leak (between IAPs and first loop isolations) from the TB loop and then repressurize with SA.
NEW; ROT 5 84, pages 9 & 10;- ROT-4 81, page 25; I
FINALNRC.TST: Version: O Page: 85
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85. ' ROT 5-45 001/ A1// 3440403001/ E03EK1.3//4.0/44/ EOP-03 -
The following plant conditions exist:
A Loss of Coolant Accident (LOCA) has occurred:
i
'
Reactor coolant pressure is 1330 psig.
'
- Indicated subcooling margin has been 25' for the past three minutes.
Two reactor coolant pumps (RCPs) are operating.
. Reactor Building (RB) pressure is 6 psig.-
.
Based on the above conditions which of the following actions should be taken?
.
A.
Bypass the high pressure injection actuation and restore seal return to
. the operating RCPs.
,
VB.
Bypass the RB isolation actuation and restore seal return to the operating RCPs.
.
C.
Bypass the RB isolation actuation and restore Nuclear Services Closed
'
Cycle Cooling (SW) to the operating RCPs.
D.-
Trip the operating RCPs.
Reasons:
A An RBIC actuation isolates RCP seal return. To restore seal return RBIC must be bypassed.
C. There is no need to restore SW to the RCPs it is only isulated if SWT 1 level is low.
D. The RCPs have been running with inadequate subcooling margin for > 2 minutes.- One RCP per loop should be left running.
.
1 NEWi ROT 5 85 pages 15 and 16; AR 301 page 44; 2/3
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86. ROT $-$$ 002/A1//3440403001/E03EK2.2//4.3/44/EOP-03 The following plant conditions exist:
-
-
A large steam line brer.k on the "B" OTSO has occurred, j
The "B" steam generator has been isolated.
. Reactor coolant pressure is 1300 psig,
)
Reactor coolant temperature is 532'F.
. Full high pressure injection (HPI)is established.
!
Based on the above conditions why was full HPI established and when can it be
'
throttled?
I
,
r
.A.
HPI is required for core cooling and can be throttled when low pressure injection (LPI) had been established for 20 minutes.
'
B.
HPI is required for core cooling and can be throttled when heat removal
through the "A" OTSO is established.
.
C.
HPI is required to compensate for RGS contraction and can be throttled when the pressurizer level exceeds 40 inches.
vD.
HPI is required to compensate for RCS contraction and can be throttled when adequate subcooling margin is recovered.
Reasons:
A. and B High pressure injection is required to compensate for RCS contraction due to the overcooling.
C. HPI cannot be throttled unless adequate subcooling margin is recovered.
'
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87. nor.s 9e 005/D4//////33/EOP.13 Under which of the following conditions MUST HPI be throttled?
-
vA.
During a plant startup with Tave at 532*F and 4 RCPs in operation a steam leak results in a cooldown to 372'F in 20 minutes. HPI is being used to control pressurizer level at 80 inches and subcooling margin is 112*F.
B.
A small break LOCA has resulted in a loss of adequate SCM. All RCPs have been stopped and HPI is supplying flow to the RCS through all 4 injection lines. Subcooling margin is currently 43*F with RCS pressure at 398 psig.
C.
Following a LOCA full HPI has been initiated in accordance with EOP Rule #2. Adequate SCM does not exist. Per the guidance of EOP Rule
- 1 the flow through each of the 4 HPI injection lines is balanced at 260
gym.
D.
In response to a MODE 3 overfeed event the operations staff has started two HPI pumps with suction from the BWST. Flow has been initiated through all four injection lines at 200 gpm per line. Normal makeup flow has been isolated and both MUP recirc valves are open.
Reasons:
B., C. & D. If adequate SCM exists then HPI must be throttled to prevent exceeding NDT limits and PTS guidelines Bank Question ROT 5 90 #14 FINALNRC.TS7 brsion: 0 Page: 88
.
__.____ _._ ___.________ _.-__ _ _ ____ _..____ _
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.
88. MOT $ 96 001/ A1//344040300110011EA2.01//4.7/66/EOP Os The following plant conditions exist:
.
,
. A large break Loss of Coolant Accident (LOCA) has occurred.
,
. Reactor coolant pressure is 54 psig.
.
-. Reactor coolant teinperature is 205'F.
-
. Reactor building pressure is 2.2 psig.
The steam generators are being fed with Emergency Feedwater (EFW).
. For 35 minutes low Pressure Injection now has been approximately 1800 gpm in each train.
i i
.
.
.
,
. Based on the above conditions which of the following is an appropriate action and-l
'
the reason for it?.
!
i
)
.
!
vA.
Isolate EFW to the OTSGs; the core is being cooled by Low Pressure t
Injection.
I B.
Isolate EFW to the OTSGs; the core is being cooled by High Pressure
.
Injection.
C.
Ensure EFW continues to feed the OTSGs; the core is being cooled by a i
'
combination of EFW and Low Pressure Injection.
'
D.
Ensure EFW continues to feed the OTSGs; the core is being cooled by a
>
combination of EFW and High Pressure Injection.
'
Reasons:
.
B., C. & D.
The generators are isolated because they are not needed for core cooling. RCS pressure is low enough and LPI has been flowing long
'
enough that LPI is cooling the core.
,
'
!
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,
L 89.' rot +95 00vA1//344040300ilE06EA1.3//3.8/$5/EOP-06
- The following plant conditions exist:
'
.
A Loss of Coolant Accident (LOCA) has occurred.
.- High Pressure Injection (HPI) has actuated and been bypassed.
. Low PressureJnjection (LPI) has actuated and been bypassed.
!
. Reactor coolant pressure is 145 p6;g.
MUP.1A is in operation.
.
. LPI flow is 1000 gpm.
i
- What directions should be given concerning the Decay Heat Pumps (DHP.1A and i
DHP.1B)?
.
A.
Both pumps should be shut down.
.
B.
Reduce the running DHPs to one.
!
VC.
Leave both DHPs running.
D.
Reduce the running DHPs to one and establish piggy back operation.
Reasons:
A. & B.
With RCS pressure < 500# and flow > 900 gpm both DHPs should
j remain operating.
D. With > 900 gpm LPI flow piggy back operation is not required.
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90. ROT 5 90 001/ A2//3440403001/002K$.09//4.2/44/ EOP 02 The following plant conditions exist:
A reactor trip has occurred.
. Immediate Actions of EOP 2, Vital System Status Verification, have been
completed.
.
Reactor coolant (RCS) pressure is 1785 psig.
i
.Thot s 588'F and decreasing slowly.
Which of the following action (s) should be taken?
vA.
Transition to EOP.10, Post Trip Stabilization, following the completion of EOP.02.
B.
Transition immediately to EOP 03, Inadequate Subcooling Margin.
C.
Transition to EOP 4, Inadequate Heat Transfer, following the completion of EOP 2.
.,
D.
Transition immediately to EOP 05, Excessive Heat Transfer.
Reasons:
B., C. & D.
The RCS is subcooled and for this pressure there is adequate subcooling margin. The proper course of action is to finish EOP.02 and then transition to EOP.10. The subcooling margin is not excessive so there ia no reason to enter EOP 05.
NEW: ROT 5 96 pages 27,29 and 59; 2/3 FINALNRC.TST Version: 0 Page: 91
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91. ROT.S 96 002/ Ai/ ROT $ 14/ 3440403001/2.1.20//4.2/44/ Al 505 A step in E0P 02, Vital System Status Verification, states:
IE, at any time while performing this procedure, ES systems have or should have actuated, THEN Ensure ES* status lights indicate components aligned to their ES position.
If reactor coolant pressure is 1850 psig, which of the following indications and associated operator responses is in compliance with this stop?
.
A.
The decay heat inlet valve to the reactor coolant system, DHV 5, has a green ES status light and the control board operator rotates the valve's switch to open it.
B.
The "B" Building Spray Pump, BSP 1B, has an amber status light and the control board operator rotates the pump's control handle to start it.
C.
High pressure injection valve, MUV 23, has a green status light and the control board operator rotates the valve's switch to open it.
vD.
The "A" Decay Heat Closed Cycle Cooling Pump, DCP 1A, has an amber status light and the control board operator rotates the pump's control handle to start it.
Reasons:-
A. Pressure is too high for an LPI actuation to open DHV 5 A green light '
indicates it has opened inappropriately.
B. An HPI actuation will only give a permit for BSP 1B to start. The amber light was correct and the pump should not be started.
C. A green status light indicates the valve is already open, There is no need to manipulate the switch.
NEW: ROT 514 01; ROT-413 pages 51,69 and 70; ROT 514 page 20; EOP 02 page 23; 2/3 FINALNRC.TST Version: 0-Page: 92 m
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92. ROT 5 96 003/ A1//3440403001/ E02EK1.2//4.0/ $5/EOP-02
,
The following plant conditions exist:
A reactor trip has occurred.
EOP 02, Vital System Status Verification, immediate actions are complete.
The pressurizer level setpoint is selected to 100".
. Makeup flow to the pressurizer has increased.
. Reactor coolant pressure is 2000 psig and slowly decreasing.
. Reactor coolant temperature is stable at 550*F.
The PORV block valve (RCV 11), pressurizer block valve (RCV 13), and the letdown cooler inlet valves (MUV.38, MUV 39 and MUV 498) are each cycled closed and then opened to check for possible leaks. No change in parameters were noticed.
Based on the above conditions which of the following describes the correct procedure implementation?
.
A.
Continue on in EOP.02. Additionalleakage isolation guidance will be provided.
vB.
Concurrently perforta OP 301, Operation of the Reactor Coolant System.
C.
Exit EOP 02 and on.
OP 8, LOCA Cooldown, for leakage isolation guidance.
D.
Exit EOP 2 and perform OP 301, Operation of the Reactor Coolant System.
Reasons:
A. Further guidance for RCS leakage is found in OP 301.
C. There is no guidanco in EOP-02 to enter EOP 08 for these conditions.
D, EOP 02 should not be exited.
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93. ROT 5-96 004/ A2// 3440403001/ 0009EK3.24//4.6/44/ EOP.13 The following plant conditions exist:
. A reactor t ip has occurred.
A small break LOCA is in progress.
.
l
. The reactor trippe'd on 4# RB pressure.
. Subcooling margin is 95'F and increasing.
RCS pressure is 2050 psig and increasing.
,
HPI flow is 800 gym.
-
Pressurizer level is 180" and increasing.
-
RCPs are operating.
-
Based on the above indications which of the following describes the correct
!
operation of High Pressure Injection (HPI) system and reason for this decision?
A.
HPI may be throttled because PTS guidelines are applicable.
v8.
HPI may be throttled to stabilize PZR level because adequate subcooling margin exists.
C.
HPI must be throttled because NDTlimits are being challenged.
D.
HPI must be throttled to prevent pump runout.
Reasons:
A. PTS guidelines do not apply to this situation.
C. The NDT limit is not being challenged.
D. The runout for two HPI pumps is 1080 gpm.
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94. nor4.e7 ooit A2//344040)ootloo74EK1.o1//4,7/44/EOP-07 The following plant conditions exist:
. The reactor has tripped from 100% power.
. Reactor coolant pressure is 1185 psig.-
.T is 589?F. *
incore Immediate Actions of EOP.2 have been completed.
-
Which of the following procedures is applicable for these conditions?
t i
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A.
- EOP.02, Vital System Status Verification.
B.
EOP.03, Inadequate Subcooling Margin.
vC.
EOP 07. Inadequate Core Cooling.
D.
EOP 04, Inadequate Heat Transfer.
!
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Reasons:
i A., B. & D.
When the reactor tripped, EOP 02 was entered. With these superheated conditions the appropriate procedure is EOP.07.
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95. hot +s7 002/AillM4040201/0074EK2.N//4.1/44/EOP 01 The following plant conditions exist:
A small break LOCA has occurred.
. Reactor coolant pressure is 1445 psig..
. Incores are at.614'F.
Nuclear Services Closed Cycle Cooling (SW)is lost and cannot be restored.
Based on the above conditions which of the following actions should be taken?
!
.
A.
Ensure MUP.1A and 10 are operating.
B.
Shutdown MUP.1A and start MUP.1B.
l
!
VC.
Place MUP 1A on Decay Heat Closed Cycle Cooling (DC) and ensure MUP 1A and MUP.10 are operating.
-
D.
Place MUP.1B on Decay Heat Closed Cycle Cooling (DC) and ensure MUP.1B and MUP.10 are operating.
Reasons:
,
,
A. MUP 1A has no cooling water.
B. MUP.1B has no cooling water.
'
D. MUP.1B cannot use DC for cooling.
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96. RCT 5100 001/ A1//3440403001/0051AK3.01//3,1/44/EOP 12 The following planc conditions exist:
--
. A loss of off site power (LOOP) has occurred.
. The "A" Emergency Diesel Generator (EDG 1A) has failed to start.
. The "B" Emergency Diesel Generator (EDG 1B)is running and supplying the
- B" 4160V ES Bus.
All"B" side emergency safeguards components are operating properly.
How is the reactor core being cooled?
.
A.
The Electric Driven Emergency Feedwater Pump is feeding the OTSGs.
Steamir.g is through the Atmospheric Dump Valves.
B.
The Electric Driven Emergency Feedwater Pump is feeding the OTSGs.
Steaming is through the Turbine Bypass Valves, vC.
The Steam Driven Emergency Feedwater Pump is feeding the OTSGs.-
Steaming is through the Atmospheric Dump Valves.
D.
The Steam Driven Emergency Feedwater Pump is feeding the OTSGs.
Steaming is through the Turbine Bypass Valves.
Reasons:
A. EDG 1A is not operating therefore the Electric Driven Emergency Feedwater Pump is not operating.
B. EDG 1A is not operating therefore the Electric Driven Emergency Feedwater
Pump is not operating. If all o. site power is lost then there is no vacuum and rr the TBVs cannot be used.
D. If all off site power is lost then there is no vacuum and the TBVs cannot be used.
- NEW: ROT 5100 pages 3,4,15 and 10; 2/3 FINALNRC.TST Version: O Page: 97
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- 97. Roi 51oo oo2/ A1/ noT.51oot 344o403ootloo66EA1.o2/ rot 4M4.4/44/ EOP 12
[
A station blackout has just occurred and EOP.12, Station Blackout, has been entered. The"A" Emergency Diesel Generator (EDG 1A) starts and then trips due
'
to low lube oil pressure. The following steps have been completed.
'
. The "NormaVAt. Engine" switch is selected to "At Engine".
!
. The low lube oil pacesure condition has been corrected.
[
-
. The " Reset" pushbutton on the diesel gage board was depressed three minutes
'
ago.
. The " Start Mode Select" switch is selected to " Auto".
Which of the following actions will start EDG 1A7
.
.
I A.
Reset the fuel racks.
[
,
<B.
Select the "NormaVAt Engine" switch to " Normal".
.
C.
Depress the manual start pushbutton or, the main control board.
D.-
Select the " Start Mode Select" twitch to " Manual" and select the
,.
"NormaVAt Engine" switch to "Noa mal".
-
Reasons:
,
A. The fuel racks are not tripped.
O, The start pushbutton on the MCB is defeated when the "NormaVAt Engine" switch is selected to "At Engine".
D. Selecting " Manual" willinput a 500 rpm signal to the EDG and defeat the EDO auto start circuitry.
D.
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98. nonioi cot /A2//34404 cat /2.4.is//3 6/3WEOP42-l The following plant corditions exist:
-
--
.
i An OTSG tube rupture has occurred.
Reactor coolant pressure is 1705 psig.
i
+ Incore temperpture is 570'F.
-
i No reactor coolant pumps are operating, j
!
EOP.6, Steam Generator Tube Rupture, Step 3.23 states " Start RCS. _
depressurization". As procedure director which of the following would describe l
your recommendation for a means of depressurization and what is the lowest
!
pressure you would attempt to achieve?
i e
.-
'
A.-
Pressurizer spray; = 1528 psig.
i
- v'B.
High pressure auxiliary spray; = 1528 psig.
,
a C.
- Open the PORV; = 963 psig.
,
' D.
De energize the pressurizer heaters; = 963 psig.
,
Reasons:
.
. A. T lR spray is unavailable.
l
,
C. & D.1528# is the minimum pressure allowablo without losing submohng margin.
,
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i 99. noT fr101002/ A2//3440403001/003BEA2.01//4.7/$$/EOP-6 The following plant conditions exist:
. A tube rupture has occurred in the "A" OTSG.
. Prior to the tubo leak dose equivalent I.131 activity was 0.5 pCi/cc.
. As the reactoris being shutdown a Startup transformer protective relay actuates and power is lost to the Unit 4160V buses.
. Chemistry samples indicate a transient peak dose equivalent 1131 activity of 97
-
pCi/cc.
Based on the above information which of the following describes the requirements for operation of the OTSGe?
vA.
The "A" OTSG must be isolated as soon as possible. Steam the "B" OTSG to atmosphere.
B.
The "A" OTSG must be isolated as soon as possible. Steam the "B" OTSG to the condenser.
C.
Neither OTSG needs to bo isolated. Continuo the cooldown steaming both OTSGs to the condenser.
D.
Neither OTSG needs to be isolated. Continue the cooldown steaming both OTSGs to atmosphere.
Reasons:
B. The condenser is not availablo.
C. The "A" OTSG's peak dose equivalent I.131 is too high and must be isolated as soon as possible. The condenser is not available.
,
D. The "A" OTSG's peak dose equivalent I.131 is too high and must be isolated as soon as possible.
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100. ROT +102 001/ A1//M4040201/0054AA1,01//4.4/44/ EOP 04 r-The following events have occurred:
,
. The plant has tripped due to a loss of Main Feedwater.
. ASV.50, steam supply valve to the Emergcency Feedwater Pump, EFP.2, elps.
. The Off site Powei Transformer has an electrical fault that de energizes the "A" l
4160V ES bus.
. The "A" Emergeacy Diesel Generator fails to start, j
Based on the sbove conditions which of the follo ving actions will ensure core cooling?
ll i
I A.
_ Verify automatic feed from AFW and stemming through the Atmospheric
,
Dump Valves.
'
i 4.
- Establish manual feed from AFW and steaming through the Turbine
'
Bypass Valves.
'
!
C.
Verify automatic feed from EFW and steaming through the Atmospheric
[
Dump Valves.
D.
Establish manual feed from EFW and steaming through the Turbine Bypass Valves.
- Reasons:
i
!
- A. AFW has no automatic featurets.
-
.
. C. & D. EFW is not available.
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