IR 05000458/1999301
| ML20205C541 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 03/29/1999 |
| From: | Hurley L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 50-458-99-301, NUDOCS 9904010255 | |
| Download: ML20205C541 (200) | |
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Jo Ree% UNITED STATES
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[r t NUCLEAR REGULATORY COMM;SSION t S 3 *
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REGION IV
$11 RYAN PLAZA DRIVE, SUITE 400 g ,, ARLINCTON, TExAb 75011-8064 March 29,1999 m
NOTE TO: NRC Document Control Desk
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Mail Stop O-5-D-24 FR W: Laura Hurley, Licensing Assistant
_ Operations Branch, Region IV SUBJECT: OPERATOR LICENSING EXAMINATIONS ADMINISTERED ON
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FEBRUARY 22 THROUGH MARCH 3,1999, AT RIVER BEND STATION, UNrl 1.
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DOCKET #50-458 On February 22 through March 3,1999, Operator Licensing Examinations were .
g. administerd at the referenced facility. Attached you will find the following information L
for processing through NUDOCS and distribution to the NRC staff, including the NRC PDR:
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ltem #1 - a) Facility submitted outline and the initial exam submittal for distribution under RIDS Code A070.
b) As given operating examination, designated for distribution under RIDS
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Code A070.
Item #2 - Examination Report with the as given written examination attached,
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designated for distribution under RIDS Code IE42.
i- if you have any questions, please contact Laura Hurley, Licensing Assistant, Operations
Branch, Region IV at (817) 860-8253.
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r 9904010255 99032Pi PDR ADOCK 05000458 1 l
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UNITED SYATEs
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? REGloNIV
/ 611 RYAN F " 'A DRIVE, GulTE 400 ('* .,,,, ,h ARLi T 11-8064 Randall K. Edington, Vice President - Operations River Bend Station Entergy Operations, Inc.
P.O. Box 220 St. Francisville, Louisiana 70775 SUBJECT: NRC INSPEC flON REPORT NO. 50-458/99-301
Dear Mr. Cdington:
I From February 22 through March 3 1999, an operator licensing inspection was conducted at your River Bend Station reactor facility. The enclosed report presents the results of this l
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inspection.
The inspection included an evaluation of three applicants for a senior operator upgrade, three applicants for a senior operator instant, and eight applicants for a reactor operator license. We determined that all applicants satisfied the requirements of 10 CFR Part 55, and the appropriate licenses have been issued.
In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter and its I
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enclosure will be placed in the NRC Public Document Room (PDR).
Should you have any questions concerning this inspection, we will be pleased to discuss them with you.
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Sinco ely, i l
s Arthur T. Ho ell 11, Director l Division of Reactor Safety Docket No.: 50-458 f/
License No.: NPF-47 .
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Enclosure:
NRC Inspection Report No.
50-458/99-301 (
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Entergy Operations, Inc. -2-l
REGION IV==
D'ocket No.: 50-458 5 License No.: NPF-47 Report No.: 50-458/99-301 Licensee: Entergy Operations, Inc. ,
Facility: River Bend Station ,
Location: 5485 U.S. Highway 61 !
St. Francisville, Louisiana .
t Dates: February 22 through March 3,1999 Inspectors: Michael E. Murphy, Chief Examiner Howard F. Bundy, Senior Reactor Engineer, Operations Branch Thomas R. Meadows, Senior Reactor Engineer, Operations Branch ;
Steve L. McCrory, Senior Reactor Engineer, Operations Branch '
Tom O. McKernon, Senior Reactor Engineer, Operations Branch John L. Pellet, Ch.ef, Operations Branch
- Approved By: Arthur T. Howell Ill, Director Division of Reactor Safety -
ATTACHMENTS:
. Attachment 1: Supplemental Information
. Attachment 2: Final Written Examinations and Answer Keys
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l EXECUTIVE SUMMARY
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River Bend Station NRC Inspection Report No. 50-458/99-301 NRC examiners evaluated the competency of 6 senior operator and 8 reactor operator license applicants for issuance of operating licenses at the River Bend Station. The licensee developed the initial examinations using NUREG-1021, interim Revision 8, January 1997. NRC .!'
examiners reviewed and approved the examinations.1 The initial written examinations were '
administered to all 14 applicants on February 19,1999, by facility proctors in accordance with the guidance in NUREG-1021, interim Revision 8. The NRC examiners administered the ;
operating tests on February 22 through March 3,1999. l
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. All 14 applicants passed the examinations. No broad knowledge or training weaknesses were identified as a result of evaluation of the graded written examinations. The applicants exhibited good oversight, peer checking and communications. (Sections 04.1 )
and 04.2)
- . The examination submitted was a'dequate for administration and required only limited enhancement and editorial corrections. The licensee staff incorporated enhancement .
suggestions developed during the NRC review process. (Section 05.1.2)
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I 3-Report Details i
Summary of Plant Status
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The plant was at approximately 90% power at the start of the inspection and in power coast L down for the upcoming refueling outage during the inspection, l. Operations I
04 Operator Knowledge and Performance l
04.1 initial Written Examination a. Insoection Scope On February 19,1999, the facility licensee proctored the administration of the written l examinations approved by the NRC to eight individuals who had applied for initial
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reactor operator licenses, three individuals who had applied for initial instant senior i operator licenses, and three individuals who had applied for initial upgrade senior i operator licenses. The licensee proposed grades for the written examinations and evaluated the results for question validity and generic weaknesses. The examiners reviewed the licensee's results.
b. Observations and Findinas The minimum passing score was 80 percent. The candidates' scores for the written examination ranged from 84 to 94 percent. The overall average score was 88.9 percent. The licensee's post-administration analysis identified that nine questions were missed by more than 50 percent of the applicants. The questions missed were three common, numbers 36,'49,55, and 59; three reactor operator, numbers 76,79,90 and 97; and, three senior reactor operator, numbers 79,90,92, and 97. The chief examiner's review of this analysis determined that the erroneous answers were generally dispersed and no broad training or knowledge weaknesses were identified.
There were no post-examination comments or changes to the written examination.
c. Conclusions All 14 applicants passed the written examinations. U broad knowledge or training weaknesses were identified as a result of evaluatie .he graded examinations.
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1 04.2 Initial Ope _ratina Test i
a. Inspection Scoce i
The examination team administered the various portions of the operating examination to the 14 applicants on February 22 through March 3,1999. Each applicant participated in l the appropriate number of dynamic simulator scenarios. Each recctor operator and instant senior operator applicant received a walk-through test which consisted of ten I
system and four administrative areas. The upgrade senior operator applicants were tested in five system and four administrative areas.
b. Observations and Findinos All applicants passed all portions of the operating test. Overall, the applicants performed well in the dynamic simulator scenarios with good oversight, peer checks, and communications noted by the examiners. Good crew briefs and status updates were consistently practiced in a form meeting licensee expectations. Communications clearly identifie expected actions with consistent acknowledgment by the operator The applicants researched and applied technical specifications appropriately and correctly applied abnormal and emergency procedure entry conditions.
Applicants correcty located and simulated operating local plant components during the examination. The applicants also displayed alertness and ownership as evidenced when one applicant noted a local fire alarm panel with a trouble alarm and immediately notified the control room. The control room acknowledged the report and advised the applicant that an auxiliary operator was being immediately dispatched to investigate.
The applicants performed well during the walk-through examination, which indicated a depth of associated system knowledge.
c. Conclusions All 14 applicants passed the operating tests. The applicants exhibited good oversight, peer checking and communications.
05 Operator Training and Qualification 05.1 Initial Licensina Examination Development The facility licensee developed the ini" d licensing examination in accordance with guidance provided in NUREG-1001," Operating Licensing Examination Standards,"
Interim Revision 8, dated Janur.ty 1997.
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-5-05.1.1 Examination Outline c. Inspection Scope The facility licensee submitted the initial examination outlines on October 23,1998. The chief examiner reviewed the submittal against the requirements of NUREG-1021, mterim Revision 8.
b. Observations and Findinos The written examination outlines for both the reactor operator and senior reactor operator met the knowledge and abilities distribution prescribed in :4UREG 1021. The administrative section outline was acceptable as submitted. The job performance measure outline was acceptable except that the safety function distribution did not comply with NUREG-1021 guidance. The chief examiner commented that the licensee should review and comply with NUREG-1021 guidance prior to completing the draft examination. The Scenario outlines were acceptable as submitted. The chief examiner advised the licensee to be careful in differentiating between component failures and instrument failures. The chief examiner determined that the initial examination outlines satisfied NRC requirements.
c. Conclusions The licensee submitted adequate examination outlines.
05.1.2 Examination Packaoe-a. Insocction Scope The draft examinations were transmitted by the ,censee to the NRC on December 19,1998. The licensee submitted i, 3 completed final examination package on February 8,1998. The chief examiner reviewed the examinations against the requirements of NUREG-1021, Interim Revision 8.
b. Observations and Findinas The draft written examination contained 125 questions,75 of which were common to both the reactor operator and senior reactor operator examinations. Of the 125 questions,116 were new, eight were modified and one was from the licensee's bank.
The draft examination was considered technically valid, to discriminate at the proper level, and responsive to the outline submitted by the licensee on October 23,1998.
Following two independent NRC examiner reviews, the chief examiner provided editorial and enhancement suggestions for 28 questions. The comments generally related to grammar, spelling. clarity of the question stem and distractor plausibility. After discussion of the suggested enhancements, the licensee modified the examinations as agreed. The chief examiner concurred with the resolution of the comments and the final product.
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-6-The licensee submitted five scenarios, two of which were designated as backups. The five scenarios were reviewed and validated during the week of January 25,1999, with some enhancement and editorial comments to facilitate administration.
To support the system walk-through section of the operating test, the facility licensee provided job performance measures developed to evaluate selected operator tasks that contained written task elements, performance standards, and comprehensive evaluator cues. Thirteen job performance measures were submitted with three designated as backups. Personnel assignments and scheduling precluded any day-to-day repetition of operating tests. The NRC review identified several enhancement and editorial comments related to improved cues, information clarification for the benefit of the examiner and to facilitate administration. The licensee incorporated all comments.
The licensee submitted six administrative job performance mei.sures and two administrative topic questions. This provided one set of five administrative job performance measures for the senior reactor operator applicants and one set of four administrative job performance measures with two administrative topic questions for the reactor operator applicants. The NRC review identified only minor enhancement and editorial comments related to eliminating overlap with operator actions in one of the scenarios, and improving evaluation capability in one of the tasks. The licensee incorporated all comments.
c. Conclug. igg
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The examination submitted was adequate for administration and required only limited enhancement and editorial corrections. The licensee staff incorporated enhancement suggestions developed during the NRC review process.
05.2 Simulation Facility Performance a. Insoection Scope ,
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The examiners observed simulator performance with regard to fidelity during the examination validation and administration.
b. Observations and Findinas The simulation facility supported the validation and administration of the examination well. There was one instance of two anomalies occurring during a scenario, but they fit.
the flow of events and caused neither a disruptim nor a compromise in the examination evaluations. The problem was evaluated at the conclusion of the scenario by the technical support personnel and a potentially bad power supply was identified and replaced. The scenarios were resumed with no further problems and with minimal delay.
l Since the f acihty was scheduled to accomplish a major modification to the simulator at L the conclusion of this examination, no simulation facility report will be included in this inspection report.
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The simulator and simulator staff supported the examination well. No fidelity issues were identified.
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V. Manaaement Meetinas
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X1 Exit Meeting Summary The examiners presented the inspection results to members of the licensee management at the conclusion of the inspection on March 3,1999. The licensee acknowledged the findings presented.
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The licensee did not identify as proprietary any information or materials examined during this inspection.
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L-ATTACHMENT 1 ,
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SUPPLEMENTAL INFORMATION l
PARTIAL LIST OF PERSONS CONTACTED Licensee i
C. Bush Jr., Operations Superintendent
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M. Cantrell, Supervisor, Operations Training B. Heikes, Supervisor, Simulator Support R. King, Director, Nuclear Safety and Regulatory Affairs-D. Looney, Exam Developer D. Mims, General Manager, Operations W. O'Malley, Manager, Operations J. O'Neil, Licensing Specialist M. Rasch, Exam Developer M. Wagner, Supervisor, Operations Training ;
J. Waid, Director, Training ,
L. VVoods, Supervisor Training Standards )
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N. Garrett. Resident inspector i
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ATTACHMENT 2 i
FINAL WR;TTEN EXAMINATION AND ANSWER KEYS l
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U.S. Nuclear Regulatory Commission Site-Specific
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_ Written Examination Applicant Informat:.vu Region: I/ II / III /
Docket #:
Facility / Unit:
j Date: 19 February 1999 RIVER BEND STATION I License Level:(RO] V
/ SRO Reactor Type: W / CE / BW /(GE)
Start Time: Finish Time: _
Instructions
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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 80.00 percent. Examination papers will be collected four hours after the examination starts.
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l Applicant Certification All work done on this examination is my own. I have neither given nor received aid.
Applicant's Signature Results Examination Value Points Applicant's Score Points Applicant's Grade Percent l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 1 The plant has undergone a transient which resulted in a Recirculation Flow Control Valve Runback.
Which one of the following describes the allowable operation of the Recirculation Flow Control Valves, prior to resetting the Flow Control Valve runback?
The Recire Flow Control Valves can:
A. be closed using loop manual operation, however, they can only be opened to the l 12 % valve position.
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B. be closed using loop manual operation, however they can only be opened to the
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point which they ran back. '
C. not be closed any further because they are at the full closed stop and cannot be reopened due to a hydraulic block on the valves.
D. not be closed any further because they are at the full closed stop, however they can be opened to the 22 % valve position.
QUESTION RO 1 NRC RECORD # WRI 1 ANSWER:. B. SYSTEM # 053 K/A 295001 A1.05: 3.3/3.3 LP# RBS-1-LEC GPST-A053 l OBJ. 2b; 12 SROTIER 1 GROUP 2 / RO TIER I GROUP 2 REFERENCE: ARP-P680-4A-A3&A9 NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.10/41.5/
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l U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 l REACTOR OPERATOR l
QUESTION 2 The plant is operating at 100 % power when a short circuit occurs on the DC bus supplying power for ATWS ARI/RPT. This causes all of the power supply breakers to BYS-PNLO2A2 to trip, resulting in a loss of power to ATWS ARI/RPT.
Which one of the following describes the response of the ARI system and the Reactor Recirculation Pumps?
A. ARI will not function, however the Reactor Recirculation pumps will trip to OFF immediately.
B. ARI will actuate causing a depressurization of the scram air header and the Reactor Recirculation pumps will trip to OFF immediately.
C. ARI will not function and the Reactor Recirculation pumps will not trip on an ATWS condition.
D. ARI will actuate causing a depressurization of the scram air header on an ATWS condition, however the Reactor Recirculation pumps will not trip.
QUESTION RO2 NRC RECORD # VMI 2 ANSWER: C. SYSTEM # 052; K/A 295004 AK2.03: 3.3/3.3 053 LP# RBS-1-STM-GPST-A0053 OBJ. 2d LP# RBS-1-STM-GPST-A0052 OILI. 3,4e SROTIER1 GROUP 1/ ROTIER 1 GROUP 1 REFERENCE: PRINTS NEW CLASS MODIFIED BANK
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DIFF 3 DATE USED: RU SRO BOTH CFR 41.7 l
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U.S. NUCLEAR REGULA'IORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 l REACTOR OPERATOR l QUESTION 3 The plant is operating at 100 % power. A rupture in the high pressure leg of the "A" Feedwater line Flow element causes Feed Flow Indications and inputs to the Reactor Recirculation system to change. The At-The-Controls operator swapped to single element control of Reactor Level with little change in level.
Which one of the following describes the response of the Reactor Recirculation System?
A. Both Recirculation Pumps will remain at present speed, however the Recirc Flow Control Valves will runback to minimum position.
B. Both Recirculation Pumps will downshift to slow speed operation with the Recire Flow Control Valves rernaining at present position.
C. Both Recirculation Pumps will trip to OFF due to cavitation interlocks not being met and Recire Flow Control Valves not being at 22%.
D. Both Recirculation Pumps will remain at present speed and the Recirc Flow
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Control Valves will remain at the present position.
QUESTION RO3 NRC RECORD # WRI 3 ANSWER: D. SYSTEM # 053 K/A 202002 K6.04: 3.5/3.5 LP# RBS-1-STM-GPST-A0053 OnI. 2a,b,j; SROTIER 2 GROUP 1/ RO TIER 2 GROUP 1
REFERENCE: AOP-0024 NEW C1. ASS ARP-P680-4A-A3; A9; MODIFIED BANK DIFF 4 C1; C7 NRC 3 DATE USED: RO SRO BOTH CFR 41.6
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 4 The plant was operating at 100 % power, when a Main Turbine trip caused a reactor scram and lift of one (1) Safety Relief Valve. Reactor level increased such that all three (3) Reactor Feed Pumps tripped. The At-The-Controls operator restarted two of the Reactor Feed Pumps and stabilized Reactor Level at +5 inches.
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Which one of the Silowing describes the final status of the Reactor Recirculation System?
A. Both Recirculation Pumps will downshift to slow speed operation with the Recire Flow Control Valves for both loops running back to minimum position.
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B. Both Recirculation Pumps will downshift to slow speed operation with the Recire i Flow Control Valves remaining at the pre-transient positions.
C. Both Recirculation Pumps will trip to OFF with the Recire Flow Control Valves for both loops running back to minimum position.
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D.- Both Recirculation Pumps will trip to OFF with the Recire Flow Control Valves remhining at the pre-transient positions.
J QUESTION RO4 NRC RECORD # WRI 4 ANSWER: A. SYSTEM # 053 K/A 202001 K4.16: 3.3/3.6 LP# RBS-1-STM-GPST-A0053 OBJ. 2b,c,d, SROTIER2 GROUP 2 / ROTIER 2 GROUP 2 j;12 REFERENCE: AOP-0024 NE, , CLASS ARP-P601-19A-H8; H11 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.5/41.6/43.6
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 5 l
The plant is being started up from cold shutdown. Which one of the following describes the secuence of C11-SOV operation on a control rod withdrawal ?
122 withdraw drive 123 insert drive 121 insert exhaust 120 withdraw exhaust A. Solenoid "alves 120 and 122 open to withdraw the control rod, at the end of the movement 122 closes followed by 120 closing last to allow the control rod to settle.
B. Solenoid valve 123 opens to insert the control rod slightly, then 123 closes, and l 122 opens to withdraw the control rod, at the end of the movement 122 closes and 120 opens to allow the control rod to settle.
C. Solenoid valves 121 and 123 open to insert the control rod slightly, then 121 and 123 close, and 120 and 122 open to withdraw the control rod, at the end of the movement 122 closes followed by 120 closing last to allow the control rod to settle.
D. Solenoid valves 121 and 123 open to insert the control rod slightly, then 121 and 123 close, and 122 opens to withdraw the control rod, at the end of the movement 122 closes followed by 120 opening to exhaust water then closing last to allow the control rod to settle.
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QUESTION RO5 NRC RECORD # WRI 5 ANSWER: C. SYSTEM # 052 K/A 201001 A1.03: 2.9/2.8 '
LP# RBS-1-STM-GPST-A0052 OBJ. Ib;2g, SROTIER2 GROUP 2 / ROTIER 2 GROUP 1 h
REFERENCE: prints NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.6 I
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o U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 6 <
The Control Rod Drive mechanism in certain conditions is capable ofinserting a control rod with ONLY the use of Reactor Pressure.
Which one of the following describes the reactor pressure and physical means by which
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this can be accomplished?
A. Reactor Pressure is 150 psig Drive Mechanism under piston area vented B. Reactor Pressure is 625 psig Drive Mechanism under piston area vented.
C. Reactor Pressure is 150 psig Drive Mechanism over piston area vented.
D. Reactor Pressure is 625 psig Drive Mechanism over piston area vented.
QUESTION RO6 NRC RECORD # WRI 6 ANSWER: D. SYSTEM # 052 K/A 201003 Kl.02: 2.9/3.0 >
LP# RBS-1-STM-GPST-A0052 OBJ, 8k; SROTIER2 GROUP 3 / RO TIER 2 GROUP 2 11b j REFERENCE: Tech Spec Bases B3.1.5 NEW CLASS MODIFIED BANK DIFF 2 l DATE USED: RO SRO BOTH CFR 41.2 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 l REACTOR OPERATOR ,
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QUESTION 7 The plant was operating at 100% power at the beginning of the transient.
The At-The-Controls Operator observes the following indications. ;
" Control Rod Drift" annunciator P680-7A-B02 in alarm
" Rod Drin" pushbutton on P680 back-lit !
"Accumu'ator Trouble" annunciator P680-7A-C03 in alarm
"Accum Fault" pushbutton on Po80 back-lit ;
"Ackn Accum Fault" pushbutton on P680 back-lit
" Scram Valves" pushbutton on P680 back-lit APRM power 97 %
l Which one of the following plant conditions was the probable cause?
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A. Single control rod drifling inward.
B. Single control rod drining outward.
C. Control Rod Drop Accident D. Single control rod scram.
QUESTION RO7 NRC RECORD # WRI 7 ANSWER: D. SYSTEM # 500 K/A 201005 A3.01: 3.5/3.5 l
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OBJ. 7,91 SROTIER 2 GROUP 1/ ROTIER 2 GROUP 1 l- REFERENCE: ARP-P680-07-7A-B02; NEW CLASS 7A-C03 MODIFIED BANK 7 DIFF 3
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 8 RBS is operating at 10% rated power with the mode switch in the STARTUP position, and total core flow at 53%. APRM E and H are bypassed due to failed power supplies.
The following is the present status of the APRMs versus LPRM inputs and indicated power:
APRM A B C D E F G II LPRM LVL D 3 4 2 2 2 3 3 3 LPRM LVL C 4 3 3 4 4 4 4 4 LPRM LVL B 2 4 4 3 2 3 3 2 LPRM LVL A 4 2 2 4 4 4 2 4 INDICATED 10 % 13 % 12 % 14 % 0% 11 % 13 % 0%
POWER byp byp LPRM 22-39D has failed downscale and must be bypassed to allow troubleshooting.
With present conditions would this action be allowed?
Attached is the LPRM vs. APRM assignments Attachment of SOP-0074.
A. Yes, conditions are satisfactory.
B. Yes, however an LCO would have to be written on the associated APRM for Administrative inputs.
C. No, this action would result in a half scram and administrative LCO requirements not to be met.
D. No, this action would result in a full reactor scram.
QUESTION RO8 NRC RECORD # WRI 8 ANSWER: C. SYSTEM # 505 K/A 215005 A1.04: 4.1/4.1 LP# RBS-1-LEC-GPST-A0503 A1.02: 3.9/4.0 A1.03: 3.6/3.6 OBJ. 22a,b;23; SROTIER 2 GROUP 1/ RO TIER 2 GROUP 1 26;29a,b;
REFERENCE: SOF-0074 NEW CLASS REP-0037 MODIFIED BANK DIFF 3 Tech Spec Bases 3.3.1.1.2 DATE USED: RO SRO BOTH CFR 41.6
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 9 The plant is in the process of mitigating the impact of a LOCA, which has uncovered fuel, and hydrogen is present in the Containment. The Control Room Supervisor has requested the Hydrogen Recombiners be started.
Using the attached procedure, determine the Required Recombiner Power Setting, and also determine the time required to reach the Required Recombiner Power Setting.
Pre-LOCA Containment Temperature 90 F Post-LOCA Containment Pressure 4 psig Post-LOCA Containment Temperature 120 F A. 50.31 KW at 20 min.
B. 50.31 KW at 25 min.
C. 52.03 VW at 20 min.
D. 52.03 KW at 25 min.
QUESTION RO9 NRC RECORD # WRI 9 I ANSWER: D. SYSTEM # 254 K/A 500000 EA1.03: 3.4/3.2 LP# 2.1.25: 2.8/3.1 OBJ, SROTIER3 GROUP / RO TIER 1 GROUP I REFERENCE: SOP-0040 NEW CLASS l MODIFIED BANK I DIFF 3 DATE USED: RO SRO BOTH CFR 41.10/43.5 j
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 10 The plant is in Mode 4 with RHR "A" in Shutdown Cooling. A misalignment of RPV drain valves has resulted in reactor vessel level lowering. The following are nie present plant parameters:
Reactor Pressure 0 psig Reactor Water Level +34 inches and lowering Reactor Water Temperature 160 F Drywell Pressure O psig Which one of the following describes the operation of the RHR "A" Shutdown Cooling System if Reactor Water Level continues to lower?
A.- At + 9.7 inches RPV water level, E12-F053A (RHR A SDC Injection Valve) will isolate, which will cause a low flow on the RHR A pump automatically opening E12-F064A (RHR Pump A Min Flow to Syp P1).
B. At + 9.7 inches RPV water level, E12-F053A (RHR A SDC Injection Valve),
E12-F008 and F009 (RHR Shutdown Cooling Isol Valves) will isolate causing the
.RHR A Pump to trip.
C. At +31 inches RPV water level, E12-F006A (RHR Pump A SDC Suction Valve)
will isolate, which will cause E12-F004A (RHR Pump A Sup Pl Suction Valve) to open and the low flow on the RHR A pump to open the E12-F064A (RHR Pump I A Min Flow to Sup Pl). I l
D. At +31 inches RPV water level, E12-F008 and F009 (RHR Shutdown Cooling l Isol Valves) will isolate, which will cause the RHR A pump to trip; the RHR A pump trip will cause E12-F053A (RHR A SDC Injection Valve) to close.
l QUESTION RO 10 NRC RECORD # WRI 10 ANSWER: B. SYSTEM # 204 K/A 205000 A2.05: 3.5/3.7 LP# A2.06: 3.4/3.5 I OBJ. SROTIER2 GROUP 2 / ROTIER 2 GROUP 2 REFERENCE: SOP-0031 NEW CLASS AOP-0003 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.7 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR
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QUESTION 11 The plant is in a ATWS with RHR "B" in Suppression Pool Cooling to control the temperature of the Suppression Pool. Reactor water level is being controlled at - 150 inches. Electrical power is lost to ENS-SWG1B. The Diesel Generator did not automatically start. After 30 minutes power is ready to be restored to ENS-SWGlB from the Division 11 Diesel Generator.
Which one of the following describes actions that MUST be taken per procedure prior to the restoration of power to the bus?
A. RHR B system piping would have drained down into the Suppression Pool, such that the RHR B pump circuit breaker must be racked out to prevent a pump start when the bus is re-energized.
B. RHR B system valves will have to be manually realigned for Standby to ensure when the bus is re-energized and the RHR B pump starts the pump is on
+
minimum flow to begin operation.
C. RHR B system will require venting of the system piping prior to the re-energizing of the bus to prevent water hammer of the system piping. I D. RHR B system valves must be manually re-aligned and the RHR B pump circuit breaker must be racked out prior to re-energizing the bus to prevent an uncontrolled restart upon power restoration.
QUESTION RO 11 NRC RECORD # WRI 11 ANSWER: A. SYSTEM # 204 K/A 219000 K6.01: 3.2/3.3 LP#
OBJ. SROTIER 2 GROUP 2 / RO TIER 2 GROUP 2 REFERENCE: SOP-0031 sect.2.1.2 NEW CLASS AOP-0004 sect. 5.2.4.1 MODIFIED BANK DIFF 4 NRC 3 DATE USED: RO SRO ROTH CFR 41.9
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 l REACTOR OPERATOR QUESTION 12 l
The plant is in a Refueling outage with RHR A in Shutdown Cooling. LPCS, RHR B and C are tagged out for maintenance. HPCS is aligned for manual ECCS operations.
Maintenance personnel moving a Recire Pump motor in the Drywell drop the motor on RWCU piping coming from the Bottom Head drain. Reactor level is rapidly lowering.
RPV Level is + 3 inches. All isolations occurred as expected.
Which one of the following describes the operation of RHR "A"? ;
A. The operator can manually initiate LPCI A, which will open E12-F004A (RHR A !
Sup Pl Suction Valve) and re-start RHR A pump and open E12-F042A (RHR A !
LPCI Injection Valve).
i B. After the automatic isolations are complete E12-F006A (RHR A SDC Isolation Valve) will automatically close. Once E12-F006A is started closed, E12-F004A (RHR A Sup Pi Suction Valve) will open, the RHR "A" Pump will start and E12-F042A (RHR A LPCI Injection Valve) will then open. j l
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C. After the automatic isolations are complete E12-F006A (RHR A SDC Isolation Valve) will autornetically close. Once E12-F006A is started closed, E12-F004A (RHR A Sup Pt Guttion Valve) will open, operator will be required to manually start the RHR "A" Pump, and E12-F042A (RHR A LPCI Injection Valve) will then open.
D. The operator will have to close E12-F006A (RHR A SDC Isolation Valve) then open E12-F004A (RHR A Sup P1 Suction Isolation Valve). After the suction is re-aligned, LPCI A can be manually initiated.
QUESTION RO 12 NRC RECORD # WRI 12 ANSWER: D. SYSTEM # 204 K/A 203000 K1.14: 3.6/3.7 LP#
OBJ. SROTIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: SOP-0031 sect. 2.2.3; NEW CLASS sect. 4.3/5.1 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.7
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 13 The plant is operating at 100 % power. The Auxiliary Building SNEO reports there is a major leak on the service water side of the CCP Heat Exchangers and the only way to isolate the leak is to isolate all service water to the CCP Heat Exchangers.
Water temperature on CCP is 110 *F and rising.
Reactor Recirculation Pump Motor temperatures are rising.
Which one of the following describes the actions required to be taken by AOP-0011 LOSS OF REACTOR PLANT COMPONENT COOLING WATER with regard to loss of cooling water to CCP7 A. Shutdown CCP pumps. and isolate the CCP Heat Exchangers on the Service Water side. Repair the leak, un-isolate the Service Water side of the CCP Heat Exchangers and re-start the CCP Pumps.
B. Reduce CCP heat loads by tripping to OFF the operating CRD Pump, and start the standby CCP pump to increase cooling water flow while mechanics effect repairs on the broken piping.
C. Manually scram the reactor and trip and isolate both Reactor Recirculation Pumps, and isolate service water to the CCP Heat Exchangers D. Reduce CCP heat loads by down shifting the Reactor Recirculation Pumps to slow speed, establish a feed and bleed on CCP to remove heat, and isolate the leak.
QUESTION RO 13 NRC RECORD # WRI 13 ANSWER: C. SYSTEM # 115; K/A 295018 AA2.03: 3.2/3.5 118 LP#
OBJ. SROTIER 1 GROUP 2 / RO TIER I GROUP 2
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REFERENCE: AOP-0011 sect 4.0 NEW CLASS AOP-0009 sect. 5.5 MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.4
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U.S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION FEBRUARY 1999 ,
REACTOR OPERATOR QUESTION 14 Certain Safety Relief Valves are designated as " Low-Low Set".
Which one of the following describes the bases and operation of the " Low-Low Set SRVs"?
A. When the first SRV opens on the relief function, two (2) SRV relief setpoints are lowered. This is done to minimize the cyclic stress on the Containment due to SRV lifting and ensures the Containment design basis is met.
B. As reactor pressure increases above the scram setpoint, the relief setpoints on five (5) SRVs are lowered to start them opening well below the design pressure of the reactor vessel to prevent exceeding reactor design pressure.
C. When the first SRV opens on tl.e relief function, five (5) SRVs are opened automatically and their reset pressures are lowered. This minimizes the number of SRV lifts by extending the length of time they are open.
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D. As reactor pressure increases above the scram setpoint, the reset setpoints for five (5) FRVs sre lowered. This minimizes the number of SRV lifts by extending the leng1 J nme they are open.
QUESTION RO 14 NRC RECORD # WRI 14 ANSWER: A. SYSTEM # 050 K/A 295025 EK3.09: 3.7/3.7 LP#-
OBJ. SROTIER I GROUP I / ROTIER 1 GROUP 1 REFERENCE: Tech Spec Bases B3.3.6.4 NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.3/41.5 ,
41.7/43.2 l. . . .
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 15 The following stable conditions exist in the plant:
Reactor Power 0 % (All Rods In)
Reactor Pressure 150 psig Reactor Water Level + 4 inches Drywell Pressure 0.8 psig
, Main Steam Tunnel Temperature 150 F Reactor Mode Switch in SHUTDOWN Given the above plant conditions, determine which one of the following describes the systems which should have received isolation signals.
A. CCP; MSIVs; RCIC; RWCU B. MSIVs; RCIC; RHR to Radwaste; RWCU C. CCP; RCIC; Reactor Sample lines; RWCU D. MSIVs; Reactor Sample lines; RHR to Radwaste QUESTION RO 15 NRC RECORD # WRJ 15 ANSWER: B. SYSTEM # 058 K/A 223002 A1.02: 3.7/3.7 LP#
OBJ. SRO TIER 2 GROUP 1/ ROTIER 2 GROUP 1 REFERENCE: AOP-0003 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.7/41.9 L
U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATIOP' 'BRUAhY 1999 REACTOR OPEta ,OR I
QUESTION 16 i
The plant is operating at 100 % power. The Feedwater Level Control (FWLC) Sys un is j in three element control with the "A" Reactor Water Level Chanael selected. A rupture !
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t occurs on the "A" reference leg causing a level change.
Assuming no other instruments are affected by the rupture, which one of the following describes the required operator action?
The Operator should:
A. transfer the FWLC System to single element control.
B. select the "B" Reactor Water Level Channel.
C. allow the level dominant signal to take control and return level to normal.
D. manually control water level with RCIC and / or HPCS.
QUESTION RO 16 NRC RECORD # WRI 16 ANSWER: B. SYSTEM # 107; K/A 295009 AA1.02: 4.0/4.0 ,
501 LP#
OBJ. SRO TIER I GROUP 1/ RO TIER 1 GROUP 1 REFERENCE: A OP-0006 NEW CLASS ARP-P680-3A-C08 MODIFIED BANK DIFF 4 NRC 3 ~
DATE USED: RO SRO BOTH CFR 41.7
U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 17 The plant is operating at 100 % power. A failure of the in-service Steam Jet Air Ejector has resulted in Condenser vacuum being lost.
Which one of the following describes the sequence of events in response to a loss of Condenser vacuum? ASSUME NO OPERATOR ACTIONS. *
A. Main Turbine Trip Reactor scram on Turbine trip l
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MSIV closure i Main Steam Bypass Valves close Condensate Pumps continue to operate B. MSIV closure Reactor c*m on MSIV closure Main Turbine hip Main Steam Bypass Valves close Condensate Pumps continue to operate l
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C. Reactor scram Main Turbine Trip MSI" closure Main Steam Bypass Valves close Condensate Pumps trip on low suction pressure D. Main Turbine Trip MSIV closure Reactor scram on MSIV closure Main Steam Bypass Valves close Condensate Pumps trip on low suction pressure QUESTION RO 17 NRC RECORD # WRI 17 ANSWER: A. SYSTEM # 104 K/A 295002 2.4.4: 4.0/4.3 LP#
OBJ. SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 I
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REFERENCE: AOP-0005 NEW CLASS ARP-P680-2A-A01; MODIFIED BANK DIFF 3 2A-B01 ,
DATE USED: RO SRO BOTH CFR 41.4 i
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 18 The plant is operating at 100 % power. A rupture of the Instrument Air line entering Containment has caused a loss of air inside the Containment. The Unit Operator has isolated the leak by closing the Containment Instrument Air Isolation Valve (IAS-MOV106)
Which one of the following describes the response of the Control Rod Drive System to a loss ofInstrument Air? ASSUME NO OTHER OPERATOR ACTION.
A. The Scram Discharge Volume ' vent and Drain Valves will open, the Scram valves i will immediately open on a low air pressure signal, and the CRD Flow Control !
Valve will fail closed.
i B. The Scram Discharge Volume Vent and Drain Valves will close, the Scram valves I will immediately open on a low air pressure signal, and the CRD Flow Control l
Valve will fail open. j C. The Scram Discharge Volume Vent and Drain Valves will open, the Scram valves will individually drift open as air pressure drops, and the CRD Flow Control Valve will fail open.
D. The Scram Discharge Volume Vent and Drain Valves will close, the Scram valves will individually drift open as air pressure drops, and the CRD Flow Control Valve will fail closed.
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QUESTION RO 18 NRC RECORD # WRI 18 l ANSWER: D. SYSTEM # 122 K/A 295019 AK2.01: 3.8/3.9 LP# RBS-1-LEC-LP-H052 OBJ. Sg; 6b SROTIER 1 GROUP 2 / RO TIER I GROUP 2 REFERENCE: AOP-0008 NEW CLASS '
MODIFIED BANK DIFF 4 i NRC 3 DATE USED: RO SRO BOTH CFR 41.4/41.7
U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 19 The plant was operating at 100 % power. A rupture of the Instrument Air line on the main header downstream of the Air Compressors has resulted in a complete loss of instrument air to the plant. The At-The-Controls operator has manually scrammed the reactor.
Which one of the following describes the response of the Condensate and Reactor Feedwater System to a loss ofInstrument Air and the ability of the plant to feed the ReactorVessel? ASSUME NO OTHER OPERATOR ACTION.
A. The Condensate demin inlet and outlet valves will fail open and the feed pump min flow valves will fail closed. The Feed Reg Valves fail as is allowing continued feeding of the reactor. )
i B. The Condensate and Reactor Feedwater system min flow valves fail open, l
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however, the lines are sized to allow continued operation of the systems, Feed Reg valves fail open to allow continued injection to the Reactor. ,
C. The Reactor Feedwater pumps will trip on low suction due to min flow valves failing open. Makeup to the reactor will ccme from HPCS and RCIC when they a.ito start.
D. The Condensate and Reactor Feedwater flow to the reactor will stop due to the !
Feed Reg Valves failing closed. Makeup to the reactor will come from HPCS and i RCIC when they auto start.
QUESTION RO 19 NRC RECORD # 'WRI 19 ANSWER: C. SYSTEM # 122; K/A 300000 K3.02: 3.3/3.4 104;107 LP# HLO-030 OBJ. 8 SROTIER 2 GROUP 2 / RO TIER 2 GROUP 2 REFERENCE: AOP-0008 NEW CLASS i MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.4 l
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U.S. NUCLEAR REGULATORY COMMISSION l l
WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 20 i
The plant is in a refueling outage with the reactor vessel disassembled. The Reactor cavity is filled to 23 feet above the flange. Fuel movement is in progress. The refueling i cavity bellows ruptures. )
In accordance with AOP-0027 FUEL HANDLING MISHAPS which one of the following is NOT an allowed safe position for an irradiated fuel bundle?
A. The Upper Containment Fuel Pool Fuel Rack.
B. The Cattle Chute hanging on the fuel grapple.
C. The Fuel Transfer Mechanism carriage rack.
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D. The Reactor Vessel in an area of the core which has no fuel. !
l QUESTION RO 20 NRC RECORD # WRI 20 ANSWER: B. SYSTEM # 055 K/A 295023 AK3.01: 3.6/4.3 LP#
OBJ. SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 3 I REFERENCE: AOP-0027 NEW CLASS FHP-0003 MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.2/41.10/
41.12/43.4/43.5/
43.6/43.7
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUEt IION 21 The plant is in an ATWS with reactor level being maintained low.
The following parameters are indicated in the Main Control Room:
Reactor Pressure 450 psig Reactor Level Wide Range - 135 inches Reactor Level Upset Range + 6 inches Reactor Level Shutdown Range + 10 inches Reactor Level Narrow Range 0 inches Reactor Level Fuel Zone - 180 inches Drywell Temperature 145 ft 310 F Containment Temperature 119 ft 165 F 2 SRVs open Which one of the following describes the Reactor Level instruments allowed to be used?
A. Fuel Zone and Upset Range only.
B. Upset Range and Wide Range only.
C, Fuel Zone and Wide Range only.
D. All Reactor Level instruments are invalid.
QUESTION RO 21 NRC RECORD # WRI 21 ANSWER: C. SYSTEM # 051 K/A 295027 EK1.02: 3.0/3.2 LP#
OBJ. SRO TIER 1 GROUP 1/ ROTIER 1 GROUP 2 REFERENCE: EOP-0001 NEW CLASS Caution 1 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.5/41.9/
41.10/43.5
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l U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 22 The plant is in a LOCA with attempts being made to restore reactor water level. Level is offscale low on all le. el instruments except Fuel Zone.
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The following parameters are indicated in the Main Control Room: *
Reactor Pressure 50 psig Reactor Level Fuel Zone - 2.20 inches Drywell Temperature 145 ft 310 F Containment Temperature 119 ft 165 'F Suppression Pool Level 15.0 ft 7 SRVs open Which one of the following describes the method to be used for determining Suppression Pool Temperature?
A. Suppression Pool Temperature indicators on H13*P808.
B. Safety Parameter Display System (SPDS) computer. ,
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C. Remote Shutdown Panel Suppression Pool Temperature indicators. l D. RHR temperature recorder with RHR operating. ,
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QUESTION RO 22 NRC RECORD # WRI 22 ANSWER: D. SYSTEM # 057 K/A 295026 EA2.01: 4.1/4.2 LP#
OBJ. SRO TIER I GROUP 1/ ROTIER 1 GROUP 2 REFERENCE: EOP-0001 NEW CLASS Caution 9 MODIFIED BANK 1
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DIFF 3 DATE USED: RO SRO BOTH CFR 41.7/41.9/ I l
41.10/43.5 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 23 The plant is operating at 4 % power in a reactor startup. The B CRD pump is tagged out with the oil sump drained for maintenance.
The A CRD pump trips. The CRS dispatches a SNEO to investigate the pump circuit breaker. The SNEO reports that the breaker has over current trip flags and the lockout device is tripped.
Electrical Maintenance is called to investigate.
The following parameters are indicated in the Main Control Room:
Reactor Pressure 450 psig Reactor Water Level + 34 inches Main Steam Bypass valves are fully closed.
With present plant conditions, which one of the following describes the actions to be taken?
A. Increase reactor pressure to > 600 psig and wait for electrical maintenance to ,
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repair the CRD Pump.
B. If two or more control rod accumulator faults exist on withdrawn control rods, fully insert the control rods within 20 minutes or place the reactor mode switch in !
SHUTDOWN.
C. If one or more control rod accumulator faults exist on withdrawn control rods, which cannot be inserted, immediately place the reactor mode switch in SHUTDOWN.
D. Increase reactor pressure to > 600 psig, and restore charging water header pressure to >l520 psig within 20 minutes or place the reactor mode switch in l SHUTDOWN.
l QUESTION RO 23 NRC RECORD # WRI 23 l ANSWER: C. SYSTEM # 052 K/A 295022 AK3.01: 3.7/3.9 l l
LP#
OBJ. SROTIER 1 GROUP 2 / RO TIER 1 GRCUP 2
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REFERENCE: ARP-P60122A-A01 NEW CLASS
! Tech Specs 3.1.5 MODIFIED BANK l. DIFF 3 ,
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DATE USED: RO SRO BOTH CFR 41.5/41.6/
43.2
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U.S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR l QUESTION 24
The plant is operating at 90 % power.
Which one of the following descriptions of plant conditions will result in a Main Turbine Trip and describes the basis for the trip?
A. The Main Turbine will trip when the selected Reactor Narrow Range Level Instruments has level at + 51 inches. This is to prevent the erosion of the Main Steam piping and Main Control Valves' seats, from moisture carryover.
B. The Main Turbine will trip when two of the Reactor . Narrow Range Level Instruments have level at + 51 inches. This is to prevent the erosion of the Main Steam piping and Main Control Valves' seats, from moisture carryover.
C. The Main Turbine will trip when two of the Reactor Narrow Range Level Instruments have level at + 51 inches. This is to prevent the erosion of the Main Turbine blades, from moisture carryover.
D. The Main Turbine will trip when the selected Reactor Narrow Range Level Instruments has level at + 51 inches. This is to prevent the erosion of the Main Turbine blades, from moisture carryover.
i QUESTION RO 24 NRC RECORD # WRI 24 ANSWER: C. SYSTEM # 110 K/A 295008 AK1.01: 3.0/3.2 LP# 295005 AA2.07: 3.5/3.6 I OBJ. SRO TIER 1 GROUP 2 / ROTIER 1 GROUP 2 l REFERENCE: AOP-0002 NEW CLASS TRM 3.3.7.3 MODIFIED BANK l l DIFF 2 l
DATE USED: RO SRO BOTH CFR 41.5 l
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l U.S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 25 A plant transient has caused Suppression Poollevel to reach 20 ft. 8 in. The HPCS l suction valves have transferred to the Suppression Pool suction alignment.
l The plant is operating at 90 % power.
With present plant conditions, wSch one of the following describes the actions to be i taken? !
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A. Leave HPCS suction aligned to the Suppression Pool until the Suppression Pool level can be lowered and the transfer logic reset.
B. If desired, transfer suction back to the CST since a low level in the CST will transfer suctions back to the Suppression Pool if required.
C. Transfer the HPCS suction back to the CST since the CST is the required suction source for HPCS to remain operable. I
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l D. Leave HPCS suction aligned to the Suppression Pool since HPCS will transfer to j the CST on a Low Suppression Pool Level to ensure an adequate source.
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QUESTION RO 25 NRC RECORD # WRI 25 ANSWER: A. SYSTEM # 203 K/A 295029 EK2.03: 3.5/3.3 LP# ;
, OBJ. SRO TIER 1 GROUP 2 / RO TIER I GROUP 2 REFERENCE: ARP-P60116A-C04; C05 NEW CLASS 16A-F03 MODIFIED BANK DIFF 3 Tech Spec 3.5.1 & bases l DATE USED: RO SRO BOTH CFR 41.7 i
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U.S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION FEBRUARY 1999
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REACTOR OPERATOR QUESTION 26 l
The plant is in mode 2. The following parameters are indicated in the Main Control Room:
l APRMs (
A B C D E F G H 11 % 12 % 13 % 11 % 14 % 15 % 13 % 13 %
IRMs (range / reading)
A B C D E F G H R9/ 36 R9/ 39 R9/ 37 R9/ 37 R9/ 36 R9/ 36 R9/ 37 R9/ 37 Reactor Water Level + 36 inches Reactor Pressure 950 psig Main Turbine speed 1800 rpm With present plant conditions, which one of the following is correct with regard to the status of the Reactor?
A. No RPS actuation.
B. Half scram on Division I.
l C. Half scram on Division II.
D. Full scram.
QUESTION RO 26 NRC RECORD # WRI 26 ANSWER: C. SYSTEM # 508 K/A 212000 K5.02: 3.3/3.4 LP# HLO-061 OBJ. 5 SROTIER 2 GROUP 1/ ROTIER 2 GROUP 1 l REFERENCE: A0P-0001 NEW CLASS
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Tech Specs 3.3.1.1 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.6 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR i
QUESTION 27 The plant is in mode 2 afler a normal refueling outage. The following parameters are !
indicated in the Main Control Room:
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IRMs (range / reading)
l A B C D E F G 11 R2/100 R3/ 39 R2/ 75 i R3/ 30 R2/ 80 R3/15 R3/18 R3/ 36 SRMs (cps)
A B C D 3 2 5 2.0 x 10 3.0 x 10 2.5 x 10 Bypassed With present plant conditions, which one of the following is correct with regard to the status of the Reactor?
A. No RPS actuation and no Control Rod Blocks B. Control Rod Block only.
C. Half scram and Contro! Rod Block.
D. Full scram and Control Rod Block.
QUESTION RO 27 NRC RECORD # WRI 27 ANSWER: C. SYSTEM # 500; K/A 215004 A3.04: 3.6/3.6 508; 503; 504 LP# RBS-1-LEC-GPST-A0503 OBJ. 4; 13 LP# HLO-061 OBJ. 3; 5 SROTIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: A0P-0001 NEW CLASS t Tech Specs 3.3.1.1; MODIFIED BANK DIFF 2 3.3.2.1 DATE USED: RO SRO BOTH CFR 41.6
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 28 The plant is in mode 3. The Control Room personnel are in the process of cooling down the reactor from rated conditions.
Which one of the following is used to reduce the possibility of erroneous Reactor Vessel Levelindications and trips?
A. Temperature / Pressure compensation inputs from the ERIS computer.
B. Continuous reference leg fill to each of the reference legs from CRD.
C. Reference leg purge to a common line from Reactor Recirculation loops.
D. Manual Temperature compensation by I&C via resetting calibration conditions.
QUESTION RO 28 NRC RECORD # WRI 28 ANSWER: B. SYSTEM # 051; K/A 216000 K5.06: 3.4/3.6 052 LP# RBS-I-LEC-LP-H052
)
j OBJ. lg- SROTIER 2 GROUP 1/ ROTIER 2 GROUP 1 REFERENCE: S01'-0001 NEW CLASS SOP-0002 MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.2
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 29 Which one of the following describes the hazards associated with the injection of RCIC into the Reactor head region?
A. RCIC injection will cause pressure fluctuations on all of the Reactor Vessel Level reference legs.
B. RCIC injection will cause a downshift of the Reactor Recirculation Pumps due to a change ofinlet subcooling.
C. RCIC injection will cause a rapid decrease in Reactor Pressure due to the collapse of the steam bubble thus causing a reactivity excursion.
D. RCIC injection will result in excessive carryover of moisture into the steam lines that in turn will cause impingement on the main turbine blading.
QUESTION RO 29 NRC RECORD # WRI 29 ANSWER: D. SYSTEM # 209 K/A 217000 2.1.32: 3.4/3.8 LP#
OBJ. SROTIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: S0P-0035 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.7/41.10 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999
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, REACTOR OPERATOR QUESTION 30 Which one of the following is NOT a basis fu the normal Suppression Pool Level?
A. Prevent the introduction of steam into the suction of the ECCS pumps by providing subcooling for ECCS suctions.
B. Provide a mechanism for the limiting of fission prcduct release in a LOCA i condition.
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C. Prevent excessive clearing loads (Level in SRV Tailpipe) on the SRVs that could
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result in tail pipe damage.
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D. Provide sufficient water for the absorption of steam energy before heating up excessively.
QUESTION RO 30 NRC RECORD # WRI 30 ANSWER: A. SYSTEM # 057 K/A 223001 K4.02: 3.6/3.7 K4.01: 3.7/3.8 LP# 2.2.25: 2.5/3.7 OBJ. SROTIER 2 GROUP 1/ ROTIER 2 GROUP 1 REFERENCE: Tech Spec 3.6.2.2 bases NEW CLASS USAR 6.2.1 MODIFIED BANK !
DIFF 3 DATE USED: RO SRO BOTH CFR 41.7/43.2 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 l REACTOR OPERATOR QUESTION 31 l During an ATWS, with stable conditions, the Control Room Supervisor has ordered Standby Liquid Controlinitiated.
l l Which one of the following would be an indication that Standby Liquid Control is
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injecting into the core? (CONSIDER. EACH ANSWER SEPARATELY.)
l A. Reactor pressure is rising.
B. Main Steam Bypass valves are throttling closed.
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C. Reactor levelis lowering.
D. Feedwater flow is rising.
QUESTION RO 31 NRC RECORD # WRI 31 ANSWER: B. SYSTEM # 201 K/A 211000 A1.09: 4.0/4.1 A1.01: 3.6/3.7
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A1.02: 3.8/3.9 l A1.03: 3.6/3.6 i
LP# HLO-016 A1.04: 3.6/3.7 OBJ. 6 SROTIER 2 GROUP 1/ ROTIER 2 GROUP 1 REFERENCE: SOP-028 NEW CLASS MODIFIED BANK DIFF 4 NRC 3 DATE USED: RO SRO ROTH CFR 41.6 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999
- REACTOR OPERATOR QUESTION 32 i The plant was operating at 80 % power. A transient in the Fancy Point Switchyard resulted in a Main Turbine trip. All control rods did not fully insert. The following are the plant parameters at present
Reactor Pressure 900 psig Reactor Level - 50 inches wide range Reactor Power 22.5 %
Suppression Pool Temperature 112 *F Feedwater flow is stable 2.79 Mlbm/hr.
IRMs and SRMs detectors have been inserted.
All Main Steam Bypass valves are fully open.
Two (2) SRVs are open.
To avoid exceeding the Heat Capacity Temperature Limit (HCTL) curve, the Control Room Supervisor has ordered Reactor Pressure lowered to 700 psig using SRVs.
l Which one of the following describes the reaction ofindicated Reactor Power immediately following the opening of the SRVs, and why?
- A. Reactor power will rise due to the lowering of the reactor coolant temperature l l adding positive reactivity. !
l B. Reactor power will rise due to the water level inside the core rising causing more moderation of neutrons.
C. Reactor power will drop due to the voiding of the water in the core as it flashes to
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steam.
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l D. Reactor power will drop due to the moderator temperature rising with the low flow through the core.
QUESTION RO 32 NRC RECORD # WRI 32 ANSWER: C. SYSTEM # K/A 295037 EKl.01: 4.1/4.3 I ATWS I LP#
OBJ. SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 I REFERENCE: Applied Reactor Theory NEW CLASS EOP Bases MODIFIED BANK I DIFF 3 DATE USED: RO SRO BOTH CFR 41.1/41.2/41.6 43.6 I
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 33 Due to an ATWS, Standby Liquid Control has been initiated from an initial tank level of 1985 gallons.
Which one of the following describes when Standby Liquid Control Injection can be terminated?
A. Reactor Temperature has been lowered to 68 'F with no control rod motion.
B. All Control Rods have been inserted fully with the exception of two control rods on opposite sides of the core.
C. Reactor Pressure has been lowered to 0 psig and RHR Shutdown Cooling interlocks have been met.
D. Standby Liquid Control Boron tank level has lowered from its initial level to 441 gallons.
QUESTION RO 33 NRC RECORD # WRI 33 ANSWER: D. SYSTEM # 201; K/A 295037 EA2.03: 4.3/4.4 ATWS EK1.04: 3.4/3.6 LP# EKl.05: 3.4/3.6 l OBJ. SROTIER I GROUP 1/ ROTIER 1 GROUP 1 .
REFERENCE: EOP-0005 Encl.15 NEW CLASS EOP-1 A step RQA-22 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.6/41.10 43.5/43.6 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 34 The plant is performing a reactor startup from cold shutdown. The reactor isjust at the point of adding heat. The CRS instructed the operators to stop the startup to perform a surveillance. During this time, the reactor went suberitical and power had dropped to range 3 of the IRMs. The At-The-Controls operator noting power selected the next control rod and withdrew the control rod from 00 to 48 with continuous motion, resulting in a sustained 20 second period. The following are the plant parameters at present:
Reactor Pressure 80 psig Reactor Level + 40 inches l
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Which one of the following describes the action the At-The-Controls operator should take? {
A. Monitor IRMs and range them according to the power increase to keep them on scale.
B. Perform the coupling checks for the Control Rod, and inform the Reactor Engineer of an increase in power rise.
C. Insert the Control Rod to a position which causes Reactor Period to be > 30 seconds.
D. Withdraw the next in sequence Control Rod to maintain the power increase to achieve the point of adding heat.
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l QUESTION RO 34 NRC RECORD # WRI 34 ANSWER: C. SYSTEM # 500; K/A 295014 A1.04: 3.2/3.3 503; 504 LP#
OBJ. SROTIER 1 GROUP 1/ RO TIER 1 GROUP 1 REFERENCE: GOP-0001 Att 1 step 3.3 NEW CLASS
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Susquehanna reactivity MODIFIED BANK DIFF 3 Event 7/98 l DATE USED: RO SRO BOTH CFR 41.1/41.2 l
41.6/43.6
U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 35 The plant was operating at 100 % power at the end of cycle when a failure of the A Reactor Feedwater line in the Turbine Building has caused reactor water level to drop.
The reactor scrammed however all control rods did not insert. The CRS is directing actions per EOP-0001 A - RPV Control - ATWS.
The following are the plant parameters at present:
Reactor Power 25 %
Reactor Pressure 940 psig Reactor Level - 180 inches Fuel Zone Range HPCS injection is overridden The MSIVs are closed.
RCIC, CRD, and SLC are injecting into the Reactor.
Four(4) SRVs are open.
Which one of the following describes the method of Adequate Core Cooling being employed?
A. Core submergence.
B. Steam Cooling with injection.
C. Steam Cooling without injection.
D. Adequate Core Cooling is NOT being assured such that Emergency Depressurization is required.
QUESTION RO 35 NRC RECORD # WRI 35 ANSWER: B. SYSTEM # RPV K/A 295031 EA2.04: 4.6/4.8 Control y LP# !
OBJ. SRO TIER 1 GROUP 1/ RO TIER I GROUP 1 REFERENCE: EOP-0001A RLA-12 NEW CLASS EOP Bases MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.10/43.5 I
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 36 Which one of the following describes the basis for the maximum design internal pressure of the Drywell?
A. Maximum Drywell Pressure is + 20 psid based on a double-ended shear of a .
Recirculation Pump discharge pipe.
B. Maximum Drywell Pressure is + 20 psid based on a double-ended shear of a Main Steam Line upstream of the MSIVs.
C. Maximum Drywell Pressure is + 25 psid based on a double-ended shear of a ;
Recirculation Pump discharge pipe.
D. Maximum Drywell Pressure is + 25 psid based on a double-ended shear of a Main i Steam Line upstream of the MSIVs. !
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QUESTION RO 36 NRC RECORD # WRI 36 i ANSWER: D. SYSTEM # 057 K/A 295024 EKl.01: 4.6/4.2 l LP# HLO-013 (057) )
OBJ. 3; 4 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 REFERENCE: Tech Spec Bases 3.6.5.4 NEW CLASS .
USAR 6.2.1.1.1 MODIFIED BANK l DIFF 3 )
l NRC 2 DATE USED: RO SRO BOTH CFR 41.9 i
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 37 The plant is operating at 100 % power. A leak on the Service Water Header in the Drywell requires the isolation of the Service Water piping inside the Drywell.
Which one of the following describes the reaction of the plant to this isolation?
A. Drywell temperature will rise along with Drywell pressure such that eventually the scram and isolation setpoint for Drywell pressure will be reached.
B. Drywell temperature will remain stable due to the evaporation of water inside the Drywell sumps absorbing heat energy.
C. Drywell temperature will rise and stabilize at the point where evaporation of the water in the Drywell will absorb the heat and Drywell pressure will stabilize I
< l.68 psig.
D. Drywell temperature will remain stable due to the continued circulation of the
' Drywell atmosphere through the Drywell Coolers and the transfer of heat to any i residual water remaining in the Service Water piping. l QUESTION RO 37 NRC RECORD # WRI 37 ANSWER: - A. SYSTEM # 404; K/A 295010 AK2.05: 3.7/3.8 l 118; 057 I LP# RBS-1-LEC-GPST-A0118 OBJ. 7; 8 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 REFERENCE: SOP-0060 step 2.2 NEW CLASS AOP-0009 MODIFIED BANK DIFF 3 ,
DATE USED: RO SRO ROTH CFR 41.4/41.9 l I
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j U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR I
QUESTION 38 The plant is at S % power. Chemistry samples taken indicate that fuel damage is present in the core. Radiation levels in Offgas and the Main Steam lines have risen drastically.
Which one of the following describes the reaction of the plant if the Main Steam Line l
Radiation Levels reach 3 times the normal background readings?
A. The Reactor will scram, the Main Steam Lines and the Reactor Sample Lines will l isolate, and the Condenser Air Removal Pumps will trip.
l B. Initiation of Standby Gas Treatment and Annulus Mixing, and an isolation of the l Main Steam Lines and Reactor Sample Lines.
C. The Reactor Sample Lines will isolate and the Condenser Air Removal Pumps will trip and isolate.
D. The Reactor will scram, Standby Gas Treatment and Annulus Mixing will initiate, and the Condenser Air Removal Pumps will isolate.
QUESTION RO 38 NRC RECORD # WRI 38 ANSWER: C. SYSTEM # 058; K/A 295033 EK3.03: 3.8/3.9 511 LP# '
OBJ. SROTIER 1 GROUP 2 / RO TIER 1 GROUP 2
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REFERENCE: AOP-0003 Att 1 E NEW CLASS MODIFIED BANK L DIFF 2 DATE USED: RO SRO BOTH CFR 41.11/41.12 43.4
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 39 The plant is at 5 % power. Chemistry samples taken indicate that fuel damage is present in the core.
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Which one of the following will NOT automatically initiate measures to control an Offsite Radiation release?
A. Fuel Building Ventilation Radiation Monitors.
B. Control Room Ventilation Radiation Monitors.
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C. Offgas Post-Treatment Radiation Monitor.
D. Reactor Building Annulus Ventilation Exhaust Radiation Monitor.
QUESTION RO 39 NRC RECORD # WRI 39 ANSWER: B. SYSTEM # 058; K/A 295034 EK1.02: 4.1/4.4 511; 606; 402; 403 LP#
OBJ. SRO TIER 1 GROUP 2 / RO TIER I GROUP 2 REFERENCE: AOP-0003 Att 1 E; Z; AA NEW CLASS BB; CC; FF MODIFIED BANK DIFF 3 AOP-0039 DATE USED: RO SRO ROTH CFR 41.7/41.11 41.13/43.4
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t l U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 49
The plant is in a reactor startup Mode 2. Due to an I&C error, HPCS initiated. The CRS directed the Control Room Operator to override close the HPCS injection Valve E22*F004. The Control Room Operator closed E22*F004. The maximum Reactor Water Level reached was + 56 inches. I&C can NOT reset the trip unit that caused the l
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Initiation Signal.
l Reactor Water Level is now + 36 inches and lowering.
Which one of the following describes the operation of E22*F004 HPCS INJECT ISOL VALVE with water level now in the normal band?
l A. The valve will automatically open on receipt of a Reactor Water Level - Low ,
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Level 2 signal.
B. The valve can only be opened using the valve hand switch in the open position.
C. The valve can only be opened if the HPCS High Reactor Water Level signal is manually reset, and then the valve hand switch is taken to the OPEN position. l D. The valve will automatically reopen if the HPCS Manual Initiation Pushbutton is depressed.
QUESTION RO 40 NRC RECORD # WRI 40 ANSWER: C. SYSTEM # 203 K/A 209002 A4.03: 3.8/3.8 LP#
OBJ. SROTIER 2 GROUP 1/ ROTIER 2 GROUP 1 REFERENCE: SOP-0030 NEW CLASS ARP-P60116A-B04 MODIFIED BANK DIFF 3 16A-F02 DATE USED: RO SRO BOTH CFR 41.7/41.8 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999
- REACTOR OPERATOR l
' QUESTION 41 The plant is in a LOCA condition. The leak is inside the Drywell from the "A" Main Steam Line.
The following plant parameters exist:
Reactor Power 0 % (All Rods In)
Reactor Level +15 inches !
Reactor Pressure 600 pig l Drywell Pressure 10 psig '
Drywell Temperature 230 F l Containment Temperature 102 F l Containment Pressure 3.2 psig Suppression Pool Level 21.0 feet Suppression Pool Temperature 102 F Which one of the following describes the method that should be employed to control Containment Pressure? ,
A. Align the Containment Ventilation Systems for Normal Containment Venting.
B. Operate all available Containment Cooling fans and coolers.
C. Use Emergency Containment Venting, if radioactive release rates are acceptable.
D. Perform an Emergency Containment Venting irrespective of radioactive release.
QUESTION RO 41 NRC RECORD # WRI 41 ANSWER: B. SYSTEM # 057; K/A 226001 2.1.27: 2.8/2.9 403 LP#
Olu. SROTIER 2 GROUP 1/ ROTIER 2 GROUP 1 REFERENCE: EOP-0002 CP-2,3,4; NEW CLASS CT-2,3, 4 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.9 I
U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINAT50N FEBRUARY 1999 REACTOR OPERATOR QUESTION 42 The plant is operating at 100 % power. The CCP line inside Containment going to the RWCU Non-Regenerative Heat Exchangers has ruptured. An operator in the area has manually isolated CCP to the Non-Regenerative Heat Exchangers.
Which one of the following describes the plant response with No further operator actions?
A. The RWCU Filter Demins will isolate and go into hold due to Low CCP Flow through the Non-Regenerative Heat Exchangers.
B. The RWCU Filter Demins bypass valve will open and the Filter Demins will go into Hold due to High Filter Demin Inlet temperature.
C. The RWCU pumps will immediately trip on High Filter Demin Inlet Temperature and G33*MOVF004, RWCU PUMPS OUTBD SUCTION VALVE will isolate to protect the Filter Demins.
D. G33*MOVF004, RWCU PUMPS OUTBD SUCTION VALVE will isolate on High Filter Demin Inlet temperature causing the RWCU pumps to trip on low flow.
QUESTION RO 42 NRC RECORD # WRI 42 ANSWER: D. SYSTEM # 601; K/A 204000 K4.04: 3.5/3.6 115 LP#
OBJ. SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2 REFERENCE: ARP-P680 01 A-B01 - NEW CLASS 01A-A01 MODIFIED BANK DIFF 3 SOP-0090 DATE USED: AOP-0011 RO SRO BOTH CFR 41.4
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 43 The plant is operating at 30 % power.
Met Tower Data indicates 40 * F.
The Circulating Water Pumps IB and IC are tagged out for repairs.
Suddenly the 1 A Circulating Water Pump trips due to a phase to phase short.
Which one of the following describes the expected response of the Main Condensers?
(Assume NO operator actions.)
A. Main Condenser vacuum will initially decrease then return to original value due to the one remaining Circulating Water Pump.
B. Main Condenser vacuum will decn.ase and stabilize above the turbine trip setpoint, since power is within the capabilities of one Circulating Water Pump.
C. Main Condenser vacuum will increase due to the increased flow rate of the remaining Circulating Water Pump.
D. Main Condenser vacuum will remain stable at its present value, as the Steam Jet Air Ejectors will control Main Condenser Vacuum.
QUESTION RO43 NRC RECORD # WRI 43
. ANSWER: B. SYSTEM # 104; K/A 256000 K6.02: 3.1/3.1 103 LP#
OBJ. SROTIER 2 GROUP 3 / ROTIER 2 GROUP 2 REFERENCE: AOP-0005 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.4
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 44 The plant is in a Refueling outage with the Refueling Platform located over the Cattle Chute area.
Which one of the following WILL ALLOW movement of the Refueling platform over the Reactor Vessel core?
A. The Main Holst unloaded, Control Rod 24-37 is selected at position 24 on H13*P680, and the Reactor Mode Switch is in REFUEL.
B. The Main Hoist loaded, Control Rod 24-37 is at position 24 on H13*P680, and the Reactor Mode Switch is in REFUEL.
C. The Main Hoist is unloaded, no Control Rods are selected on H13*P680, and the Reactor Mode Switch is in STARTUP.
D. The Main Hoist is loaded, no Control Rods are withdrawn as indicated on H13*P680, and the Reactor Mode Switch is in STARTUP.
QUESTION RO 44 NRC RECORD # WRI 44 i ANSWER: A. SYSTEM # 055 K/A 234000 K5.02: 3.1/3.7 ;
LP# HLO-022 OBJ. 2 SRO TIER 2 GROUP 2 / RO TIER 2 GROl'P 3 REFERENCE: FHP-0003 NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.4/43.7
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR I
- QUESTION 45 The plant has just returned to 100 % power following completion of Refueling Outage 7.
The Refueling Outage began 45 days ago.
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The Fuel Pool Cooling Pump SFC-PI A seal blew out and started dropping level in the Lower Fuel Pool, the pump has been secured and isolated.
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The Fuel Pool Cooling Pump SFC-P1B is tagged out for seal replacement.
Using the Decay Heat curves, determine the present conditions of the Spent Fuel Pool.
Time to Boil Decay Heat Heat-up rate A. 33 hrs 6.4 Mbtu/hr 2.0'F/hr B. 48 hrs 8.' Mbtu/hr 2.0'F/hr C. 33 hrs 8.5 Mbtu/hr 1.8 F/hr D. 48 hrs 6.4 Mbtu/hr 1.8'F/hr
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QUESTION RO 45 NRC RECORD # WRI 45 ANSWER: D. SYSTEM # 602 K/A 233000 A4.05: 2.7/3.1 A2.07: 3.0/3.2 LP# 2.1.25: 2.8/3.1 OBJ. SRO TIER 2 GROUP 3 / RO TIER 2 GROUP 3 REFERENCE: OSP-0037 Att.9 NEW CLASS AOP-0051 Att.1-5 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.10/43.5
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 46 The plant is shutting down for a refueling outage.
The followhg are the Temperature and Pressure parameters as the plant was cooled
?Osn.
Time Rx Press Rx Temp.
1000 103 psig 340 F 1030 74 psig 320 *F i100 45 psig 292 F 1130 18 psig 256 *F 1200 2 psig 220*F 1230 0 psig 182 *F 1300 0 psig 145 F 1330 0 psig 107 F 1400 0 psig 69 F Which of the following statements is correct concerning the Reactor Coolant System?
A. All RPV pressure and temperature limits are within specifications B. RPV pressure vs. temperature limits are satisfied, but the cooldown rate has been eveeeded.
C. RPV pressure vs. temperature limits have been violated, but the cooldown rate is satisfactory.
D. RPV pressure vs. temperature limits and the cooldown rate have been exceeded.
QUESTION RO 46 NRC RECORD # WRI 46 ANSWER: C. SYSTEM # 050 K/A 280002 K5.05: 3.1/3.3 LP# 2.1.25: 2.8/3.1 OBJ. SROTIER 2 GROUP 3 / ROTIER 2 GROUP 3 REFERENCE: Tech Specs 3.4.11 NEW CLASS Steam Tables MODIFIED BANK DIFF 3 STP-050-0700 DATE USED: GOP-0002 RO SRO BOTH CFR 41.3/43.2
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l U.S. NUCLEAR REGULATOPY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR l l
QUESTION 47 The plant was scrammed at 28 % power as part of a normal shutdown.
Which one of the following statements is true regarding the ability to reset RPS?
A. The scram signal is unable to be reset, until RPV water level is restored to below Level 8.
B. The scram signal can be reset by taking the RPS reset switches to RESET.
C. The scram signal can be reset by resetting the ARI logic channels then taking the RPS reset switches to RESET.
l D. The scram signal can be reset by taking the SDV Bypass switches to BYPASS j and then placing the RPS reset switches to RESET. '
QUESTION RO 47 NRC RECORD # WRI 47 ANSWER: D. SYSTEM # 508 K/A 212000 A4.14: 3.8/3.8 i LP#
OBJ. SROTIER 2 GROUP 1/ ROTIER 2 GROUP 1 i REFERENCE: Tech Specs 3.3.1.1 NEW CLASS AOP-0001 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.6/41.7 I
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 48 The plant has experienced an ATWS condition.
The following parameters exist at the present time:
Reactor Mode Switch is in Shutdown Reactor Pressure is 600 psig Reactor Water Level is - 120 inches Reacto- Power is <1 %
Suppression Pool Tempcrature is 130 F MSIVs are closed.
IRMs and SRMs are inserted. I Under which one of the following conditions would it be appropriate to exit EOP-1 A -
RPV Control- ATWS7 !
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A. Standby Liquid Control has injected Cold Shutdown Boron Weight into the reactor with an additional 25 % injected to allow for imperfect mixing and leakage.
B. Standby Liquid Control has injected sufficient quantities to allow the SLC pumps to be secured as directed by EOP-1 A.
C. The Reactor Engineer has performed shutdown margin determinations and has determined that adequate shutdown margin exists for all conditions.
D. Chemistry and the STA have determined that the combination of Boron and Control Rods has brought the reactor suberitical for all conditions.
QUESTION RO 48 NRC RECORD # WRI 48 ANSWER: C. SYSTEM # K/A 295015 AA2.02: 4.1/4.2 EOP-0001A l LP#
OBJ. SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 REFERENCE: Tech Specs 3.1.1 bases NEW CLASS EOP-0001A step RCA-1 MODIFIED BANK DIFF 3 EPSTG*0002 step RCA-1 DATE USED: Tech Spec 3.1.7 bases RO SRO BOTH CFR 41.1/41.2/41.6 ;
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR
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QUESTION 49 The Design Basis Accident for Maximum Drywell Atmosphere Temperature is ...
A. SRV tailpipe failure within the Drywell.
B. A small steam leak break in the Drywell.
C. A double ended shear of the Recirculation System discharge piping D. A double ended shear of a Main Steam pipe upstream of the Inboard MSIV.
QUESTION RO 49 NRC RECORD # WRI 49 ANSWER: B. SYSTEM # 057 K/A 295012 2.4.18: 2.7/3.6 LP#
OBJ. SROTIER 1 GROUP 2 / ROTIER 1 GROUP 1 REFERENCE: Tech Specs 3.6.5.5 bases NEW CLASS USAR 6.2.1.1 MODIFlED BANK DIFF 3 6.2.1.1.',.1.7 NRC 2 DATE USED: 6.2.1.1.3.1.7.4 RO SRO BOTH CFR 41.9 j
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MISSION Y COM UARY 1999 REGULATORFEBR TION TOR OPERA U.S.NUCLEARWRITTEN REA CTOR EXAMINA t)areincreasing.
ondary Containmen t and QUESTION Auxiliary Building 3S(Sec econdary Containmen visorhas enteredEOP 000 mined aleakexists.
The temperaturesinthe tion ofthe trol,andhas deterEmergency Depressuriza TheControlRoom SuperRadioactive Release (
o F,Emergencyafrombeing Whichone Why? f thefollow Reactor and excecc's e 200he quipmentinthe are ,
heRCICRoom Sump If the temperatureintrequired to protectt F andtheRCICRoom A. Depressurization extreme is conditions.
exceeds 200CIC Room, the RCICEmergency exposed to R energy releaseinto h
oRCICRoomver the floor ieof theextreme If thetemperatureint e the
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ired tolimit ove B.
Levelis five(5)inchesDepressurizationis F and e the to rem MainSteamL requ exceeds 144urizationisan ptionconditions.
extreme Room. RCICRoom d n Emergency Depresswhich could cause the C.
F If thetemperatureintheTunnel i eTunnelbothexceew exceeds 13 the drivinghea d of the and MainSteamL nduce the flo h RCICRoomizationis required to re D.
If thetemperatureint eF, Emergency # W RI 50 Depressur 200 by reducingthe therma NRC RECORDEK3.01: 3.5/3.8 K/A 295032 EOP-3 I GROIi 3 RO 50 TIER CLASS SYSTEM # 2 / RO BANK QUESTION D. 1 GROUP NEW ANSWER: SRO TIER 19; 41.4/41.9 LP#
MODIFIED CFR OBJ. EOP 0003 STEPS SC-20 RO SRO SC-BOTH 41.10/43.5 REFERENCE: EPSTG*0002 SC 19; SC 20
DIFF USED:
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 50 The temperatures in the Auxiliary Building (Secondary Containment) are increasing.
The Control Room Supervisor has entered EOP-0003 Secondary Containment and Radioactive Release Control, and has determined a leak exists.
Which one of the following would require the Emergency Depressurization of the Reactor and Why?
A. If the temperature in the RCIC Room exceeds 200 'F, Emergency Depressurization is required to protect the equipment in the area from being exposed to extreme conditions.
B. If the temperature in the RCIC Room exceeds 200 F and the RCIC Room Sump Level is five (5) inches over the floor of the RCIC Room, Emergency Depressurization is required to limit the extreme energy releese into the RCIC Room. .
C. If the temperature in the RCIC Room exceeds 144 F ar.d the Main Steam Line Tunnel exceeds 135 *F, an Emergency Depressurization is an option to remove the driving head of the only system which could cause the extreme conditions.
D. If the temperature in the RCIC Room and Main Steam I ine Tunnel both exceed 200 F, Emergency Depressurization is required to reduce f.he flow from the break by reducing the thermal driving head.
QUESTION RO 50 NRC RECORD # WRI 50 ANSWER: D. SYSTEM # EOP-3 K/A 295032 EK3.01: 3.5/3.8 LP#
OBJ. SRO TIER I GROUP 2 / ROTIER 1 GROUP 3 REFERENCE: EOP-0003 STEPS SC-19; NEW CLASS SC-20 MODIFIED BANK DIFF 3 EPSTG*0002 SC-19; DATE USED: SC-20 RO SRO BOTH CFR 41.4/41.9 41.10/43.5
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l U.S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 51
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All Power to the Division I DC bus has been lost.
Concerning the Low Pressure Core Spray System operation, which one of the following .
statements is true?
A. In the event of an actual LOCA condition, LPCS will NOT operate automatically, j
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however, the system can be manually initiated from the Main Control Room and inject into the Reactor.
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B. In the event of an actual LOCA condition, LPCS will automatically start, l however, the injecti;n valve must be manually opened due to the loss of the I
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automatic opening feature of the pressure permissive.
C. Low Pressure Core Spray is unable to be initiated manually or automatically, however, the LPCS pump can be manually started from the Main Control Room and placed on minimum flow or can be aligned for injection.
D. Low Pressure Core Sprav is unele to be initiated manually or automatically, and the LPCS pump will not operate from the Main Cor51 Room, if the pu y i s start locally "' vill operate on minimum flow.
QUESTION RO51 NRC REJORD # WRI 51 ANSWER: D. SYSTEM # 205 K/A 209001 K2.03: 2.9/3.1 1 LP# )
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OBJ. SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: ARP-P60121 A-H08 NEW CLASS ,
Electrical Dwgs MODIFIED BANK i DIFF 4 828E535AA sh'3,4,6,10 I 3 NRC 3 !
DATE USED: ESK5CS.LOI RO SRO BOTH CFR 41.7/41.8 l ESK6CSL01
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOp nPERATOR QUESTION 52 During maintenance in the H13*P628, an I&C Technician shorted ajumper causing a loss of DC power to the Division I SRV solenoids.
Concerning the operation of the Automatic Depressurization System, which one of the following statements is true?
A. The ADS SRVs have opened on a logic initiation signal and must have their !
handswitches taken to OFF to close the ADS SRVs.
i B. The ADS SRVs are unable to be opened using manual or automatic initiation i signals, however, they will still function by placing either division handswitch in 1 the OPEN po:ition.
C. The ADS SRVs will operate in automatic for all modes of operation using the Division II logic system or the Division II handswitch in the OPEN position.
- D. The ADS SRVs are unable to be opened using manual and automatic initiation logic, however, the valves may be opened using the Division 11 handwitches.
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l QUESTION RO 52 NRC RECORD # WRI 52 ANSWER: C. SYSTEM # 202; K/A 218000 K3.02: 4.5/4.6 305 K4.03: 3.8/4.0 l K5.01: 3.8/3.8 P- A2.05: 3.4/3.6 OBJ. SRO T.lER. 2 GROUP 1/ RO TIER 2 GROUP 1 dEFE"ENCE: ARP-P601 19A-A07; B08 NEW CLASS 19A-Bil; E08; H08 MODIFIED BANK DIFF 4 Elect DWGs 851E225AA NRC 3 )
DATE USED: RO SRO ROTH CFR 41,7/41.8
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 53-Which one of the following is used to operate the Safety Relief Valves? l The SRVs are operated by:
A. at least one solenoid valve must energize admitting air pressure OR Reactor pressure overcomes spring pressure.
B. two solenoids must energize for the air valve to admit air pressure OR Reactor pressure overcomes spring pressure.
C, at least one solenoid must de-energize to close the air admitting valve which allows Reactor pressure to open the SRV.
I D. two solenoids must de-energize to open the valve admitting air pressure to open the SRV OR Reactor pressure overcomes spring pressure. ;
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QUESTION RO 53 NRC RECORD # WRI 53 I ANSWER: A. SYSTEM # 202; K/A 239002 K2.01: 2.8/3.2 305; 109 LP#
l OBJ. SROTIER 2 GROUP 1/ ROTIER 2 GROUP 1
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REFERENCE: Elect Dwgs 851E225AA NEW CLASS l P&ID 3-1C Syst.109 MODIFIED BANK ,
DIFF 3 ADS Logic !
DATE USED: RO SRO BOTH CFR 41.3 i
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 54 Standby Gas Treatment has started on a high Drywell pressure. The Unit Operator has placed the "B" Standby Gas Train in standby.
Which one of the following describes the response of the Standby Gas Treatment System to a High-High Annulus Exhaust Radiation signal on both divisions?
A. The "B" Standby Gas Treatment Train will automatically restart from standby.
B. The "A" Standby Gas Treatment Train will shutdown, then both Standby Gas Treatment Trains will re-initiate..
C. The "A" Standby Gas Treatment Train v,ill remain operating and the "B" Standby Gas Treatment Train will remain in st'.ndby.
D. Both Standby Gas Treatment Traias shutdown and isolate awaiting further l operator action.
l QUESTION RO 54 NRC RECORD # WRI 54 ANSWER: C. SYSTEM # 257; K/A 261000 Kl.08: 2.8/3.1 403 K4.01: 3.7/3.8 LP#
OBJ. SRO TIER 2 GROUP 1/ P.O TIER 2 GROUP 1 REFERENCE: Elect Dwgs NEW CLASS ESK-06GTS01 sh 1,2 MODIFIED BANK DIFF 3 ESK-06GTS02 sh 1,2 l NRC 2 DATE USED: SOP-0043 & 0059 RO SRO BOTH CFR 41.13 ARP-P863 71 A-C07; G07 73A-C04; D05; E05; F04 :
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 55 ACB28 for E12-C002C RHR Pump "C" Motor on bus ENS-SWG1B (4160 volt) has to be racked out to allow an electrical surveillance to be completed.
Which one of the following describes the safety materials required to be utilized when racking out the circuit breaker?
A. Face shield; approved eye protection; switchingjacket; and high voltage rubber gloves only.
B. Face shield; approved eye protection; switching jacket; and leather gloves only.
C. Face shield; switchingjacket and leather gloves only.
D. Face shield; switching jacket; and high voltage rubber gloves only.
QUESTION RO 55 NRC RECORD # WRI 55 ANSWER: B. SYSTEM # 303 K/A 262001 2.1.30: 3.9/3.4 LP# 2.1.26: 2.2/2.6 OBJ. SROTIER 2 GROUP 1/ RO TIER 2 GROUP 2 REFERENCE: SOP-0046 Att.5 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.4
U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 56 The plant electrical distribution system is in a normal lineup (on the preferred transformers).
The plant has scrammed.
Reactor Water Level dropped to - 82 inches and is risine The Entergy Grid has experieniced transients due to ses *her in the area.
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Plant 4160 Volt bus voltages DROPPED to 3500 volts for 10 seconds. ;
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Which one of the following statements is the condition of the Division I, II, III buses after this voltage transient? !
l A. Division I bus is being supplied from iRTX-XSR1C Division II bus is being supplied from IRTX-XSRID l Division III bus is being supplied from 1RTX-XSR1D !
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B. Division I bus is being supplied from Div I D/G Division II bus is being supplied from Div II D/G l Division III bus is being supplied from Div III D/G
C. Division I bus is being supplied from 1RTX-XSRIC Division II bus is being supplied from 1RTX XSRID Division III bus is being supplied from Div III D/G
D. Division I bus is being rupplied from Div I D/G l Division II bus is bcmg supplied from Div II D/G Division III bur is being supplied from IRTX-XSRID l
! QUESTION RO Sb NRC RECORD # WRI 56 l
ANSWER: C. SYSTEM # 309 K/A 264000 A3.01: 3.0/3.1 LP# 11LO-037
. OILI. 3 SRO T1ER 2 GROUP 1/ ROTIER 2 GROUP 1
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REFERENCE: Tech Specs 3.3.8.1 NEW CLASS Tech Specs TR3.3.8.1 MODIFIED BANK DIFF 4 Tech Specs TR3.3.5.1 DATE USED: RO SRO BOTH CFR 41.8
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U.S. N UCLEAR REGULATORY COMMISSION :
WRITTEN EXAMINATION FEBRUARY 1999 i REACTOR OPERATOR QUESTION 57 i
The plant is operating at 70 % power. The Master Level Control System is selected for three (3) element control and Rx Level A.
The D004A Condensing pot has ruptured. !
Which one of the following statements defines tiu response of the Feedwater System and Reactor Level?
A. The Feedwater Level Control System will close the Feed Reg Valves to a lower position and ac+'tal Reactor Level will lower and stabilize at a point just above the !
scram setpoint. I B.' The Feedwater 1.evel Control System will open the Feed Reg Valves to full open and actual Reactor Level will rise to the point to pick up the Main Turbine trip on !
high level.
C. The Feedwater Level Control System will close the Feed Reg Valves to a lower position and actual Reactor Level will lower to below the Reactor scram setpoint resulting in a Reactor scram. l D. The Feedwater Level Control System will shift the Master Level Controller to I
manual and lock up the Feed Reg Valves at their present position and actual Reactor level will stabilize at a slightly higher level.
QUESTION RO 57 NRC RECORD # WRI 57 ANSWER: C. SYSTEM # 107 K/A 259002 K5.01: 3.1/3.1 LP# K5.03: 3.1/3.2 OBJ. SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: ARP-H13-P680 03A-C08 NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.5
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 58 The plant is operating at 100 % power. Condensate Pump CNM -P1B has tripped on over-current.
The pressure at the suction of the Reactor Feed Pumps has dropped to 200 psig for 12 seconds.
Which one of the following would be the response of the Reactor Feed Water System?
l A. The Feed Reg Valves will throttle back to increase suction pressure resulting in a low Reactor Level.
B. The "A" Reactor Feed Purap will trip and cause Reactor Feed Pump suction pressure rise.
C. The "A and B" Reactor Feed Pumps will trip and cause a Reactor scram on low Reactor Level.
D. All three Reactor Feed Pumps will trip causing Reactor Level to lower and result in a Reactor scram.
QUESTION RO 58 NRC RECORD /.' WRI 58 ANSWER: B. SYSTEM # 107; K/A 259001 A3.10: 3.4/3.4 104 LP#
OBJ. SRO TIER 2 GRt*UP 2 / ROTIER 2 GROUP 1 l REFERENCE: ARP-H13-P680 03A-A01 NEW CLASS MODIFIED BANK DIFF 3 )
. DATE USED: RO SRO BOTH CFR 41.4 ;
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l U.S., NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 l REACTOR OPERATOR QUESTION 59
! The plant is Shutdown in Mode 4.
The Main Cond:nsers are drained and open for maintenance.
RHR "A" is in Shutdown Cooling with Service Water at 82 'F.
HPCS, RHR "B", and RHR "C" pump motors are tagged out for repairs.
l Reactor Recirculation Pump "A" is operating in Slow Speed. Reactor Recirculation l Pump "B" is tagged out for seal replacement.
Reactor Engineeting has calculated decay heat as 17 x10 Btu /hr at 130 F.
RHR "A" Pump has tripped due to unknown reasons.
Based on present plant conditions, which one of the following would be the minimum Alternate Shutdown Cooling Methods required to maintain present plant conditions?
(OSP-0041 is available.)
A. CRD and RWCU ONLY B. MSL Flooding ONLY.
C. CRD and SFC ONLY.
I D. ADHR ONLY.
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QUESTION RO 59 NRC RECORD # WRI 59 ANSWER: D. SYSTEM # 204 K/A 295021 AA1.04: 3.7/3.7 LP#
OBJ. SRO TIER 1 GROUP 2 / ROTIER 1 GROUP 3 REFERENCE: AOP-0051 NEW CLASS OSP-0041 MODIFIED BANK DIFF 3 l DATE USED: RO SRO BOTH CFR 41.5/43.5
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR l
QUESTION 60 Plant conditions are as follows:
Reactor Power 42 %
Actual Turbine Generator Load 420 MWe Turbine Generator Load Set 1040 MWe The At-The-Controls Operator withdraws a control rod which causes Reactor thermal power to change by + 30 M'Vth.
Which one of the following would occur based on the changes made?
A. The Turbine Control Valves will open as required to maintain Reactor pressure as Reactor power increases.
B. The Turbine Control Valves will close as required to maintain Reactor pressure as Reactor power increases.
C. The Turbine Control Valves will open as required to turn Reactor power by lowering Reactor pressure. l D. The Turbine Control Valves will close as required to turn Reactor power by raising Reactor pressure.
Q'UESTION RO 60 NRC RECORD # WRI 60 ANSWER: A. SYSTEM # 110 K/A 241000 K5.04: 3.3/3.3 LP# KS.03: 3.5/3.6 OBJ. SROTIER 2 GROUP 1/ ROTIER 2 GROUP 1 REFERENCE: EHC Functional Diagram NEW CLASS MODIFIED BANK 1)IFF 3 DATE USED: RO SRO BOTH CFR 41.5 l
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g U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 61 The plant is operating at 100 % power.
A leak in the RWCU pump room caused an isolation of the RWCU System.
G33*MOVF040, RWCU RETURN TO FW failed to close.
Which one of the following actions is REQUIRED to be taken?
(Tech Specs attached, if needed.)
A. The penetration is allowed to remain unisolated if the remainder ofisolation valves in the rest of the RWCU system have isolated.
B. Verify another valve in the associated penetra: ion is closed and is also de-activated.
C. The penetration is allowed to be unisolated during present conditions as long as the RWCU pumps have tripped.
D. The plant must shutdown to cold shutdown and shutdown the RWCU system.
QUESTION RO 61 NRC RECORD # WRI 61 ANSWER: B. SYSTEM # 110 K/A 290001 A2.06: 3.7/4.0 LP#
OBJ. SROTIER 2 GROUP 1/ RO TIER 2 GROUP 2 REFERENCE: Tech Specs 3.6.1.3 NEW CLASS TR3.6.1.3-1 MODIFIED BANK DIFF 3 i DATE USED: RO SRO BOTH CFR 41.9 i
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U.S. NUCLEAR REGULATORY COMMISSION
, WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR L
QUESTION 62 The plant is operating at 100 % power when B21-AOVF028B, an outboard MSIV fails closed due to a rupture of the va!ve actuator air supply.
Which one of the following describes the response of the reactor?
ASSUME NO OPERATOR ACTION.
A. RPV pressure will increase and stabilize at a higher pressure.
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Reactor power will increase and stabilize at a higher power.
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RPV water level will decrease and then return to normal level. i B. RPV pressure will increase and then decrease folowing the scram.
Reactor power will increase and cause a reactor scram on power.
RPV water level will decrease and then stabilize at a lower level. I C. RPV pressure will decrease and stabilize at a lower pressure.
Reactor power will decrease and stabilize at a lower power.
RPV water level will increase and then return to normal level. 1 l
D. RPV pressure will decrease and stabilize at a lower pressure.
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Reactor power will increase and return to the original power.
RPV water level will increase and then return to normal level.
QUESTION RO 62 NRC RECORD # WRI 62 ANSWER: B. SYSTEM # 109; I K/A 295007 AK1.03: 3.8/3.9 LP# 107 OBJ. SROTIER I GROUP 1/ ROTIER 1 GROUP 1 REFERENCE: USAR 15.2.4.1.2.2. NEW CLASS MODIFIED BANK DIFF 3 ID 308 DATE USED: RO SRO BOTH CFR 41.5/41.14 l
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U.S. NUCLEA3. REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR -
QUESTION 63
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The plant was operating at 100 % power when a tornado caused a complete loss of Offsite Power.
Division I, II, and III diesel generators failed to start resulting in a Station Blackout.
Which one of the following describes the response of the RCIC system?
ASSUME NO OPERATOR ACTION.
I A. RCIC steam supply willisolate. j RCIC will be unable to be unisolated until AC power is restored.
B. All AC powered components of RCIC will fail as-is.
RCIC will respond to all initiation signals and inject to the reactor. j C. All RCIC isolations are still available.
RCIC will respond to all initiation signals and inject to the reactor.
D. All AC powered components of RCIC will fail as-is.
RCIC will ONLY start on manual initiation and alignment signals.
QUESTION RO 63 NRC RECORD # WRI 63 j ANSWER: B. SYSTEM # 303 K/A 295003 AA1.03: 4.4/4.4 l LP# 209 OBJ. SROTIER 1 GROUP 1/ ROTIER 1 GROUP 2 REFERENCE: SOP-0035 Attachment 3 NEW CLASS MODIFIED BANK DIFF 3 ID 287 j DATE USED: RO SRO BOTH CFR 41.7
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 64 The plant is operating at 100 % power when a leak causes reactor water level to lower to
- 50 inches.
Which one of the following will be the affects on the Containment and Containment Cooling System?
ASSUME NO OPERATOR ACTION.
A. Standby Gas Treatment System and Annulus Mixing will initiate.
Containment Cooling System will operate without cooling water causing Containment temperatures to rise.
B. Standby Gas Treatment System and Annulus Mixing will initiate.
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Containment Cooling System will shutdown all fans causing Containment temperatures to rise. 1 C. Containment Cooling System will operate without cooling water causing ,
Containment temperatures to rise. l Containment Ventilation will align to purge the Containment. i I
D. ~ Containment Cooling System will shutdown all Containment Cooling fans i causing Containment temperatures to rise.
Containment Ventilation will align to purge the Containment.
QUESTION RO 64 NRC RECORD # WRI 64 ANSWER: A. SYSTEM # 403 K/A 295020 AA1.03: 2.9/3.1 LP#
OBJ, SROTIER I GROUP 2 / ROTIER 1 GROUP 2 REFERENCE: AOP-0003 NEW CLASS ;
MODIFIED BANK DIFF 3 ID 518 i DATE USED: RO SRO BOTH CFR 41.9
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U.S, NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 65 The Control Room Supervisor has entered EOP-0003 Radioactive Release Control.
Turbine Building Ventilation has shutdown.
Why is the Control Room Supervisor directing an operator to restart Turbine Building Ventilation? l l
A. Provide for the filtration of the Turbine Building atmosphere to prevent the release of any radioactive material from the Turbine Building.
B. Provide for the filtration of the Turbine Building Exhaust to ensure radioactive releases do not exceed General Emergency Levels.
C. Provide for the monitoring of the air released from the Turbine Building instead I of unmonitored release to the environment.
D. Provide for the condensing of steam in the Turbine Building which may contain radioactive particulate thus preventing any release.
QUESTION RO 65 NRC RECORD # WRI 65 ,
ANSWER: C. SYSTEM # 408 K/A 295038 EK2.04: 3.9/4.2 i LP#
OBJ. SRO TIER 1 GROUP I / RO TIER I GROUP 2 REFERENCE: EPSTG*0002 NEW CLASS EOP-0003 RR-1 MODIFIED BANK DIFF 2 ID 311 DATE USED: RO SRO BOTH CFR 41.13/43.4
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f l U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 66 l
The plant is operating at 100% power.
I&C is performing a surveillance on the APRMs and causes an inadvertent Reactor Scram.
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! Which me of the following describes the response of Reactor Water Level on the Reactor Scrant ' (The Scram was NOT caused by Reactor Water Level.)
A. Level will rise tojust above the High level alarm then drop back to the normal level setpoint.
B. Level will rise to above level 8 tripping all Reactor Feed Pumps causing level to drop.
C. Level will drop to level 3 causing Setpoint Setdown to take effect and return level to 18 inches.
D. Level will drop to just below the Low level alarm and then rise back to the normal level setpoint.
l QUESTION RO 66 NRC RECORD # WRI 66 ANSWER: C. SYSTEM # 501 K/A 2950% AK3.01: 3.8/3.9 LP# HLO-060 OBJ. 4 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 REFERENCE: A OP-0001 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.5/41.14 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 67 The plant has scrammed due to an isolation of the MSIVs.
Suppression Pool Tempereture is 95 F.
Which one of the following describes the approved method to prevent localized heating of the Suppression Pool?
A. Cycle all SRVs going Main Steam Line to Main Steam Line in order. l B. Cycle the Low-Low Set SRVs in a specified order.
l C. Cycle non-Low-Low Set SRVs in any order and start RHR C in Suppression Pool Cooling.
D. Use SRV B21 *F051D to control reactor pressure and start Suppression Pool Cleanup to circulate water.
QUESTION RO 67 NRC RECORD # WRI 67 ANSWER: B. SYSTEM # 057; K/A 295013 AK1.03: 3.0/3.3 LP# 109 OBJ. SRO TIER 1 GROUP 1/ RO TIER I GROUP 2 REFERENCE: Operator Aid H13-P601 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.9/41.10
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WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 68 The Suppression Pool is leaking into the crescent area of the Auxiliary Building.
Suppression Pool Level has dropped to 12 ft 6 inches.
Concerning Emergency Depressurization, which one of the following describes the significance of this Suppression Pool Level?
l A. A vortex will be formed when an SRV is opened drawing air into the SRV tailpipe causing water hammer.
B. If an SRV is opened while any ECCS pump is drawing a suction from the Suppression Pool the pump will draw in steam.
C. The SRV discharge quencher may not be covered, such that SRV operatian may directly pressurize Containment with steam.
D. The horizontal vents are not covered such that SRV operation will admit steam into the Drywell.
QUESTION RO 68 NRC RECORD # WRI 68 ANSWER: C. SYSTEM # 057; K/A 295030 EK2.08: 3.5/3.8 LP# 109 OBJ. SROTIER 1 GROUP I / RO TIER I GROUP 2 REFERENCE: EPSTG*0002 NEW CLASS EOP-0004 ED-3; AED-2 MODIFIED BANK DIFF 3 ID23 DATE USED: RO SRO BOTH CFR 41.9
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 69 The Unit Operator notices the AUX BLDG FL DRAIN SUMP LEVEL EXTREME HIGH/ LOW ANNUNCIATOR (P870-51 A-G3) is in alarm.
The Auxiliary Building SNEO has reported the sump level in the HPCS Pump Room is overflowing onto the floor.
Which one of the following describes the expected equipment operation and procedural requirements?
A. Both sump pumps should be operating and the Control Room Supervisor should be entering EOP-0003.
B. Only one of the sump pumps should be operating and the Auxiliary Building Operator should be locating the source of the leakage.
C. Both sump pumps should be operating and the Control Room Supervisor should evacuate the Auxiliary Building. ,
D. Only one of the sump pumps should be operating and the Control Room
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Supervisor should be entering EOP-0003.
QUESTION RO 69 NRC RECORD # WRI 69 ANSWER: A. SYSTEM # 604 K/A 295036 EK3.04: 3.4/3.8 LP#
OBJ. SROTIER 1 GROUP 2 / RO TIER 1 GROUP 3 REFERENCE: ARP-P870 51 A-G03 NEW CLASS EOP-0003 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.4
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REAC10R OPERATOR QUESTION 70 The plant is operating at 100 % power.
Both Offgas Post Treatment Radiation monitors have alarmed on a High-High-High Radiation signal.
Which one of the following describes the effects on the Offgas System and the Main Condenser?
A. Offgas will shift into a bypass mode of operation causing Main Cond< nser vacuum to be lost.
B. Offgas will isolate only the charcoal adsorbers inlet and outlet valves causing Main Condenser vacuum to be lost.
C. Offgas will continue operation allowing Main Condenser vacuum to remain constant.
D. The Offgas System will isolate causing Main Condenser vacuum to be lost.
QUESTION RO 70 NRC RECORD # WRI 70 ANSWER: D. SYSTEM # 606 K/A 271000 K3.01: 3.5/3.5 LP#
OBJ. SRO TIER 2 GROUP 2 / ROTIER 2 GROUP 2 REFERENCE: ARP-P60122A-A03 NEW CLASS AOP-0005 MODIFIED BANK DIFF 3 DATE USED: RO SRO B07N CFR 41.13
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I U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999
! REACTOR OPERATOR QUESTION 71 l
The Main Control Room has been evacuated due to a fire.
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All Immediate Operator Actions have been completed perAOP-0031.
l l Control of the plant has been established at the Remote Shutdown Panel. l l Which one of the following systems could have the possibility of overfilling AND
over-pressurizing the Reactor?
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C. HPCS D. LPCS QUESTION RO 71 NRC RECORD # WRI 71 ANSWER: C. SYSTEM # 200 K/A 295016 AA1.06: 4.0/4.1 LP#
OBJ. SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 2 REFERENCE: AOP-0031 Caution NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.7
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 72 The plant is operating at 100 % power, j l
Radwaste is discharging a Recovery Sample Tank.
Circulating Water blowdown is stopped due to a failure of the Circulating Water Blowdown Isolation Valve CWS-MOV104.
l The Aux Control Room Operator notifies the Main Control Room of alarm LWS-PNL187 4-A4 BLOWDOWN WATER FLOW LOW has been received.
Which one of the following actions should occur?
A. LWS-AOV257, RCVY SAMPLE DISCH V TO CRCLT WATER BLWDN will auto close, and LWS-AOV258, RCVY SAMPLE DISCHARGE DIVERTING will auto open.
B. The Aux Control Room Operator should manually secure the discharge by closing LWS-AOV257, RCVY SAMPLE DISCH V TO CRCLT WATER BLWDN, and securing the discharge lineup.
C. Continue the discharge, and monitor the Recovery Sample Process Radiation Monitor, and only secure the discharge if the radiation levels reach the High alarm setpoints.
D. Continue the discharge, and inform Chemistry to take grab samples, and secure the discharge only if the Radiation levels are above the limits of the Discharge Permit.
QUESTION RO 72 NRC RECORD # WRI 72 ANSWER: B. SYSTEM # 603; K/A 272000 A3.03: 3.1/3.5 103 LP#
OBJ. SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2 REFERENCE: ARP-LWS-PNL187 4-A4 NEW CLASS 4-C4 MODIFIED BANK DIFF 3 SOP-0113 DATE USED: RO SRO BOTH CFR 41.11/41.13/
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 73 The plant is operatMg at 100 % power.
A fire erupts in the Division I Diesel Generator room causing the sprinkler system to i initiate.
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Fire Water header pressure has dropped to 115 psig. ;
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Which one of the following actions would be expected to occur?.
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A. The Motor Driven Fire Pump will auto start and the Diesel Driven Fire Pump "A" and "B" will start immediately if the Motor Driven Fire Pump fails to start. J B. The Motor Driven Fire Pump will auto start, if header pressure is still at 115 psig aflerl5 seconds AND the Motor Driven Fire Pump failed to start, then the Diesel l Driven Fire Pump "A" will start. !
C. The Diesel Driven Fire Pump "A" will auto start,11 fire water header pressure remains at 115 psig for 10 seconds, whether the Motor Driven Fire Pump starts or j NOT. I D. The Diesel Driven Fire Pump "A" will auto start, if fire water header pressure remains below 140 psig for 10 seconds and the Motor Driven Fire Pump is ,
running.
QUESTION RO 73 NRC RECORD # WRI 73 ANSWER: D. SYSTEM # 251 K/A 286000 '*" l.03: 3.3/3.4
LP#
ORJ. SROTIER 2 GROUP 2 / RO TIER 2 GROUP 2 REFERENCE: SOP-0037 NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.4 l
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U.S. NUCLEAR REGl'LATORY COMMISSION 1 WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 74 The plant is operating at 100 % power.
The following annunciators have been received in the Main Control Room:
H13-P8'70 55A-B01, TURB CMPNT CLG WTR SYS SURGE TK LOW LEVEL H13-P870 55A-B02, TURB CMPNT CLG WATER SYSTEM LOW PRESSURE H13-P870 55A-E01, TURB CMPNT CLG WATER PUMP BRKR AUTO TRIP H13-P870 55A-E02, TURB CMPNT CLG WATER PUMP LOW DISCH PRESS H13-P870 55 A-G02, TURB CMPNT CLG WATER PUMP 1B OVERLOAD i
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The Turbine Building SNEO reports that there is a large ld m t'.e con. mon discharge piping of the TPCCW pumps which is spraying the IB and IC TPCCW Pu.. ns. The SNEO reports the leak is unisolable. l l
When checked, TPCCW Surge Tank level is offscale low.
l Which one of the following actions is required to take place under these conditions?
A. Verify TPCCW Pump 1C has started and MWS-AOV132, TPCCW SURGE TK MAKE-UP valve is full open u add water to the TPCCW Surge Tank.
' B. Lower power to 45 Mlbm/hr Core Flow, verify TPCCW Pump 1C has started, and MWS-AOV132, TPCCW SURGE TK MAKE-UP valve is full open to add water to the TPCCW Surge Tank.
C. Manually scram the reactor, shutdown the equipment which is being supplied by TPCCW, and secure the remaining TPCCW pumps and isolate the leak.
D. Verify TPCCW Pump 1C has started, and MWS-AOV132, TPCCW SURGE TK MAKE-UP valve is full open, and shutdown equipment as necessery on TPCCW.
QUESTION RO 74 NRC RECORD # WRI 74 ANSWER: C. SYSTEM # 116 K/A 400000 K6.04: 3.0/3.1; K4.01: 3.3/3.9 LP# A2.01: 3.3/3.4; A2.02: 2.8/3.0 OBJ. SRO TIER 2 GROUP 2 / ROTIER 2 GROUP 2 REFERENCE: AOP-0012 NEW CLASS ARP P870 55A-b01; CO2 MODIFIED BANK DIFF 2 55A-E01; E02; G02 DATE USED: RO SRO BOTH CFR 41.4
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 75 The plant is operating at 40 % power.
The following annunciator has been received in the Main Control Room:
H13-P680 07A-B01, ROD CONTROL AND INFO SYS INOPERATIVE 1&C has determined the power supply to the Division 2 RACS cabinet has failed.
Which one of the following describes the At-The-Controls Operators ability to move control rods with present plant conditions?
A. Control rods may be inserted or withdrawn normally.
B Control rods may be inserted normally however, are unable to be withdrawn.
C. Control rods may be inserted or withdrawn by bypassin: the control rod in the Division 2 RACS Cabinet.
D. Control rods are unable to be inserted or withdrawn normally, the only control rod movement is by scram.
NRC RECORD # WRI 75
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QUESTION RO 75 ANSWER: D. SYSTEM # 500 K/A 262002 K3.07: 2.6/2.8 LP#
OBJ. SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2 REFERENCE: SOP-0071 NEW CLASS ARP P680 07A-B01 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.4 )
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 76 Which one of the following statements describes the reason for scramming the reactor on a Main Turbine Trip?
A. In the event of a Main Turbine trip, the Main Steam Bypass valves are too small l to handle the capacity of reactor steam flow. ,
B. Scramming the reactor prevents excessive steam pressues from causing a Main )
Turbine overspeed condition due to leakage of the Main Turbine Stop and Control l Valves.
C. The scram anticipates the decrease in Reactor Recirculation Flow that at high !
power would result in thermal hydraulic instability l
D. The scram inserts a large amount of negative reactivity to prevent exceeding j thermal limits at the end of core life because of the distance control rods must i insert to be effective. I l
QUESTION RO 76 NRC RECORD # WRI 76 )
ANSWER: D. SYSTEM # 508; K/A 295005 AA2.03: 3.1 '
LP# 053 212000 K1.10: 3.2 OBJ. SRO TIER GROUP / ROTIER 1 GROUP 1 REFERENCE: Tech Specs 3.3.1.1 bases NEW- CLASS 3.3.4.1 bases MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.4/41.5
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999
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REACTOR OPERATOR -
QUESTION 77 Which one of the following plant conditions would result in the initiation of a RCIS control rod withdrawal block?
A. IRM H detector is Full IN and the Reactor Mode Switch is in RUN.
B. IRM H detector is Full IN and the Reactor Mode Switch is in STARTUP. ,
C. IRM H detector is Full OUT and the Reactor Mode Switch is in STARTUP.
D. IRM H detector is Full OUT and the Reactor Mode Switch is in RUN.
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QUESTION RO 77 NRC RECORD # WRI 77 ANSWER: C. SYSTEM # 504 K/A 215003 K6.05: 3.1 A1,03: 3.6 LP# A1.04: 3.4 OBJ. SRO TIER GROUP / ROTIER 2 GROUF 1 REFERENCE: Tech Specs TR3.3.2.1-1 NEW CLASS MODIFIED BANK l DIFF 2 ID87 DATE USED: Rp SRO BOTH CFR 41.6 I i
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 78 The plant is operating at 100 % power in the extended operating area (MELLL) of the Power to Flow Map at the 106 % rod line.
Electricians performing preventive maintenance cause the Reactor Recirculation Pumps to downshift to slow speed. The Reactor scram setpoints have NOT been exceeded.
Reactor Power is 71.5 %.
Reactor Thermal power is 2070 MWth.
Total Core Flow is 41 Mlbm/hr.
Which one of the following describes the actions to be taken?
A. Continue to monitor the core thermal limits and FCBB (Fractional Core Boiling Boundary), with no further actions required. '
B. Initiate an immediate Reactor Scram.
C. Insert control rods using the Shutdown Control Rod Sequence Package and shifl Reactor Recirc Pumps to Fast Speed.
D. Insert control rods using the Shutdown Control Rod Sequence Package and raise core flow by opening Recire Flow Control Valves.
QUESTION RO 78 NRC RECORD # WRI 78 4 ANSWER: D. SYSTEM # 053 K/A 202002 K3.05: 3.2 !
LP# 216000 K4.10: 2.9 i OBJ. SRO TIER GROUP / RO TIER 2 GROUP 1 l REFERENCE: AOP-0024 NEW CLASS MODIFIED BANK DIFF 3 DATE USED: CFR 41.2/41.6 RQ SRO BOTH L. j
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 79 Operators are shutting down the Reactor.
Reactor Power is at 50 %.
Control Rod 20-25 is being inserted. The Position Indication is totally lost on Division II.
Control Rod 20-25 still indicates position on Division I.
All other control rods are operating normally.
I Which one of the following will allow the normal Reactor Shutdown to continue?
I A. The control rod position must be bypassed in both RACS cabinets. !
i B. The control rod position can be bypassed in Division II RACS cabinets only.
C. The control rod position must be bypassed in the RGDS cabinet.
D. The control rod may only be inserted by using the local individual scram switches.
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QUESTION RO 79 NRC RECORD # WRI 79 l ANSWER: B. SYSTEM # 500 K/A 201005 K3.03: 3.0
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A2.02: 2.8 l A2.03: 3.2 LP# A2.04: 3.2 l
OBJ. SRO TIER GROUP / RO TIER 2 GROUP 1 REFERENCE: SOP-0071 NEW CLASS Tech Specs 3.1.3 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.6/41.7 l
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U.S. NUCLEAR REGUL ATORY COMMISSION l
WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 80 The plant has scrammed due to a loss of offsite power.
HPCS and RCIC WILL NOT start. I Approximately 5 minutes after RPV water level decreases below - 143 inches, the "DIV i 2 ADS LOGIC TIMER INITIATED" annunciator illuminates.
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The Unit Operator is directed to " INHIBIT ADS" per EOP-0001.
Later the Unit Operator Arms and Depresses the ADS B MANUAL INITIATION l pushbuttons. I What is the response of the ADS System in this situation? l
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ADS will initiate:
A. immediately, if any DIV II low pressure ECCS subsystem pressure permissive is sat sfied.
B. in 105 seconds, if any DIV II low pressure ECCS subsystem pressure permissive l is satisfied.
l C. immediately, regardless oflow pressure ECCS subsystem status.
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D. in 105 seconds, regardless oflow pressure ECCS subsystem status. '
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QUESTION RO 80 NRC RECORD # WRI 80 j ANSWER: A. SYSTEM # 202 K/A 218000 A4.02: 4.2
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LP# HLO-064 OBJ. SRO TIER GROUP / ROTIER 2 GROUP 1 -
' REFERENCE: ARP P60119A-A10 NEW CLASS ,
l MODIFIED BANK DIFF 3 ID3448 l DATE USED: RO SRO BOTH CFR 41.7 l
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l U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 81 The plant is operating at 100 % power.
The air supply fails to both Annulus Pressure Control Systems (APCS) isolation dampers, HVR-AOD23A/B, APCS FAN A/B SUCTION.
How will the plant / system respond?
A. The APCS continues to maintain a negative pressure of 3.0 inches W.G.
B. The Annulus Mixing and the Standby Gas Treatment System start automatically.
to maintain negative pressure of at least 0.5 inches W.G.
C. The Standby Gas Treatment System must be started manually to maintain negative pressure of at least 0.5 inches W.G.
D. The APCS continues to operate and the Standby Gas Treatment System start automatically to maintain a negative pressure of 3.0 inches W.G.
I QUESTION RO 81 NRC RECORD # WRI 81 ANSWER: B. SYSTEM # 403 K/A 288000 A3.01: 3.8 LP# HLO-038 OBJ. 5 SRO TIER GROUP / ROTIER 2 GROUP 3 REFERENCE: ARP P863 72A-C01 NEW CLASS MODIFIED BANK DIFF 3 ID3536 DATE USED: RO SRO BOTH CFR 41.13/43.4
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REAC10R OPERATOR QUESTION 82 The plant has scrammed.
Reactor level has dropped 10 - 50 inches and is recovering.
Reactor pressure is being controlled with Turbine Bypass Valves.
Which one of the following describes the response of the Control Room Ventilation System?
A. Normal outside air intakes close, the air inside the Control Rou.i is just recirculated until the ventilation system is reset.
B. Outside air intakes remain open with the charcoal filter train drawing from the control room and exhasts to the outside atmosphere.
C. Normal outside air intakes close, remote outside air is supplied to the control room.
D. Outside air is filtered through the charcoal filtration trains before it is supplied to the control room.
QUESTION RO 82 NRC RECORD # WRI 82 ANSWER: D. SYSTEM # 402 K/A 290003 A3.01: 3.3 LP# HLO-049 OBJ.1I SRO TIER GROUP / ROTIER 2 GROUP 2 l REFERENCE: Tech Specs 3.3.7.1 NEW CLASS l MODIFIED BANK DIFF 3 ID2412 DATE USED: RO SRO BOTH CFR 41..i l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 83 The plant is operating at 100 % power.
All the Reactor Wide Range Level instruments have failed upscale due to a common component failure.
All other level instruments are responding nonnally.
Which one of the following describes the response of the plant?
A. HPCS will send an isolation signal to its injection valve.
B. The Main Turbine and Reactor Feed Pumps will trip.
C. Reactor Feed Water will close the Feed Reg Valves.
D. Reactor Recirculation pumps will automatically downshift to slow speed.
QUESTION RO 83 NRC RECORD # WRI 83 ANSWER: A. SYSTEM # 051; K/A 216000 K5.01: 3.1 LP# 203 OBJ. SRO TIER GROUP / ROTIER 2 GROUP 1 REFERENCE: ARP P60116A-B04 NEW CI. ASS MODIFIED BANK DIFF 2 DATE USED: R_q SRO BOTH CFR 41.5
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U.S. NUCLEAR REGULATORY COMMISSION WKITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR
. QUESTION 84 The plant was operating at 100 % power when a grid transient caused the Main Turbine to trip.
Which one of the following describes the response of Reactor Pressure Control?
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A. The Safety Relief Valves will open and relieve pressure, then close WITHOUT any actuation of the Turbine Bypass Valves.
B. The Main Steam Isolation Valves will isolate causing all the Safety Relief Valves to open then control Reactor Pressure on Low-Low Set.
C. The Safety Relief Valves and Turbine Bypass Valves will open with the Safety Relief Valves closing followed by the Turbine Bypass Valvec closing.
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D. The Turbine Bypass Valves will open and control press re WITHOUT any Safety Relief Valve action.
i QUESTION RO 84 NRC RECORD # WRI 84 ANSWER: C. SYSTEM # 109 K/A 239001 A1.01: 3.6 i A1.08: 3.8 LP# A1.10: 3.8 OBJ. SRO TIER GROUP / RO TIER 2 GROUP 2 REFERENCE: ARP P60119A-H11 NEW CLASS GOP-0001 MODIFIED BANK l DIFF 3 l DATE USED: RO SRO BOTH CFR 41.4
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 85
The plant is operating. !
Main Generator is operating ~ at 900 MWe.-
Main Generator is carrying + 50 MVARs, ,
Main Generator Hydrogen Pressure is at 65 psig. l
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The System Dispatcher has requested that River Bend Station pick up an additional + 450 1 MVARs to increase voltage in our area.
Which one of the following describes the allowances to assume this amount of generator I load?
A. The generator is allowed to pick up this amount ofload with NO restrictions.
B. The generator is allowed to pick up this amount ofload as long as generator real load is raised to compensate.
C. The generator is NOT allowed to pick up this amount ofload because the generator stator windings will overheat.
D. The generatoi- is NOT allowed to pick up this amount ofload because the generator is overexcited.
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QUESTION RO 85 NRC RECORDil WRI 85 ANSWER: A. SYSTEM # 110 K/A 245000 K5.02: 2.8 LP#
OBJ. SRO TIER GROUP / RO TIER 2 GROUP 2 REFERENCE: SOP-0080 NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.4 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 86 DC Power is lost to Bus NNS-SWGI A (4160 volt).
Which one of the following describes the operation of circuit breakers supplying loads from NNS-SWGI A7 A. The circuit breaker can be closed from the Main Control Room but opened only at the local cubicle.
B. The circuit breaker can only be closed and opened locally using manual means for one closing and opening cycle.
C. The circuit breaker can only be closed locally however, all circuit breaker trips are available local and remote.
D. The circuit breaker can be closed and opened from the Mais Control Room however, all automatic breaker closures and trips are disabled.
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QUESTION RO 86 NRC RECORD # WRI 86 ANSWER: B. SYSTEM # 302; K/A 263000 K3.02: 3.5 LlY 305 OILI. SRO TIER GROUP / RO TIER 2 GRO!!P :t REFEitENCE: AOP-0014 NEW CLASS MODIFIED BANK DIFF 2 DATE 'USED: CFR 41.4 RQ SRO BOTH
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U.S. NUCLEAR REGULATORY COMMISSION
! WRITTEN EXAMINATION FEBRUARY 1999
[ REACTOR OPERATOR QUESTION 87 l The plant is at 135 "F.
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All ECCS systems are in standby.
The Reactor Mode Switch is in SHUTDOWN.
The Reactor Head has been de-tensioned for removal.
Primary and Secondary Containment are SET (OPERABLE).
With the above conditions, which one of the following is the Plant Operational Mode?
A. Mode 2 - Startup B. Mode 3 - Hot Shutdown C. Mode 4 -Cold Shutdown D. Mode 5 - Refueling
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QUESTION RO 87 NRC RECORD # WRI 87 ANSWER: D. SYSTEM # Tech K/A Generic 2.1.22: 2.8 LP# HLO-0009 Specs OBJ. 4 SRO TIER GROUP / RO TIER 3 GROUP REFERENCE: Tech Specs Definitions NEW CLASS Table 1.1-1 MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.10/43.2/
43.5
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U.S. NUCLEAR REGULATORY COMMISSION i WRITTEN EXAMINATION FEBRUARY 1999 l REACTOR OPERATOR l
l QUESTION 88 l
The High Pressure Core Spray Diesel Generator is operating for a surveillance run.
All systems are in their normal lineup.
Offsite poweris lost, i
All systems responded as expected.
Which one of the following describes the response of the Cooling Water for the High Pressure Core Spray Diesel Generator? -
A. Normal Service Water will continue to supply the High Pressure Core Spray Diesel Generator.
B. Standby Service Water pumps "A & C" will start and supply the High Pressure Core Spray Diesel Generator.
C. Standby Service Water pumps "B & D" will start, but must be manually aligned to supply the High Pressure Core Spray Diesel Generator.
D. The Unit Operator must manually start the division of the Standby Service Water System to be aligned to the High Pressure Core Spray Diesel Generator.
QUESTION RO 88 NRC RECORD # WRI 88 ANSWER: B. SYSTEM # 256; K/A 264000 K1.04: 3.2 ,
LP# 309 OBJ. SRO TIER GROUP / ROTIER 2 GROUP 1 REFERENCE: AOP-0009 NEW CLASS SOP-0018 MODIFIED BANK DIFF 3 DATE USED: RO.SRO BOTH CFR 41.8 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 89 Plant conditions are as follows:
Mode 1 Reactor Power 15 %
Generator was placed on-line ~2 minutes ago.
Which one of the following describes the response of the RC&IS system if the Main Turbine were to trip with no reactor scram?
RC&lS will: :
i A. implement the constraints of the Rod Withdrawal Limiter allowing rod I movements of up to 4 notches. l l
B. implement the constraints of the Rod Pattern Controller and Insert and/or Withdraw blocks are initiated as necessary.
C. be between the Rod Pattern Controller and the Rod Withdrawal Limiter indicating the Low Power Alarm Point with NO constraints on rod motion. ;
D. implement the constraints of the Rod Withdrawal Limiter allowing rod movements of up to 2 notches.
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QUESTION RO 89 NRC RECORD # WRI 89 ANSWER: 3. SYSTEM # 500 K/A Generics 2.2.12: 3.0 LP#
OBJ. SRO TIER GROUP / RO TIER 3 GROUP REFERENCE: Tech Specs 3.1.6 NEW CLASS 3.3.2.1 MODIFIED BANK DIFF 3 DATE USED: Rg SRO BOTH CFR 41.10/43.6 l^
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 90 Which one of the following describes the basis for the Tech. Spec. minimum number of OPERABLE Safety Relief Valves?
The Safety Relief Valves are required to have a minimum number of OPERABLE valves:
A. to be able to dissipate 90 % of the power produced by a full power reactor following a turbine trip.
B. to allow for the removal of decay heat during post LOCA conditions.
C. to prevent over-pressurization of the reactor during ATWS conditions.
D. to prevent over-pressurization of the reactor during the most severe pressure transient.
QUESTION RO90 NRC RECORD # WRI 90 ANSWER: D. SYSTEM # 109 K/A Generics 2.2.25: 2.5 LP#
OBJ. SRO TIER GROUP / ROTIER 3 GROUP REFERENCE: Tech Specs Bases 3.4.4 NEW CLASS MODIFIED BANK 3 DIFF 3 NRC 2 DATE USED: RO SRO BOTH CFR 41.6/43.2 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 91 Given the following plant conditions:
Reactor Power 0% (all rods in)
Reactor Level +33 inches Reactor Pressure 890 psig Containment Temperature 98 *F Containment Pressure 1.3 psig Suppression Pool Temperature 97 *F Drywell Pressure 4.5 psig Drywell Temperature 208 F Which one of the following describes the Emergency Operating Procedures that are to be implemented?
A. EOP-1 only B. EOP-1 and 2 C. EOP-1 and 3 l
D. EOP-1,2, and 3 QUESTION RO 91 NRC RECORD # WRI 91 l ANSWER: B. SYSTEM # EOP K/A Generics 2.4.4: 4.0 LPd
' OBJ, SRO TIER GROUP / RO TIER 3 GROUP ,
REFERENCE: EOP-0001; 0002; 0003 NEW CLASS l MODIFIED BANK DIFF 2 '
DATE USED: RQ SRO BOTH CFR 41.10/43.5
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l U.S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION FEBRUARY 1999 l REACTOR OPERATOR
I QUESTION 92 Given the following plant conditions:
l l Reactor Power 100 %
l Reactor Level +36 inches l Reactor Pressure 1025 psig l Containment Temperature 85 F Containment Pressure 0.03 psig l Suppression Pool Temperature 81 F Drywell Pressure 1.1 psig Drywell Temperature 110 F i Drywell Area Sumps show no unusual changes in level, flow, or temperature, j l
Drywell Atmosphere radiation monitor show no changes.
' The Unit Operator has noted that Drywell Pressure is rising slowly.
Drywell atmosphere radiation levels are steady.
Which one of the following describes a possible cause of the conditions as noted above?
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i A. Small leak on the Main Steam Line Flow Elbows Instrument Line.
B. Small leak on the RWCU suction from the Reactor Bottom Head. I C. Small leak on the Instrument Air header inside the Drywell.
D. Small leak on Normal Service Water to a Drywell Unit Cooler. j
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QUESTION RO 92 NRC RECORD # WRI 92 ANSWER: C. SYSTEM # ~ K/A Generics 2.4.21: 3.7 LP# Diagnostic l OBJ. SRO TIER GROUP / RO TIER 3 GROUP REFERENCE: GGNS IE NEW CLASS L
MODIFIED BANK j DIFF 2 L NRC 3-i DATE USED: CFR 41.5 R_O SRO BOTII i.
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! U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999
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REACTOR OPERATOR l
l QUESTION 93
' The Control Room Supervisor has determined there is a fire in the Offgas Charcoal Adsorbers.
Reactor Power 100 %
Reactor Level +36 inches Reactor Pressure 1025 psig Which one of the following describes the actions to be taken to mitigate the fire?
A. Initiate a reactor shutdown and connect fire hoses to the inlet of the Charcoal Adsorbcrs and flood the adsorber beds.
B. Initiate a reactor shutdown and connect Nitrogen bottles to the Charcoal l Adsorbers and purge the adsorber beds.
C. Maintain stable reactor power to prevent introduction of further ignition sources and connect fire hoses to the inlet of the Charcoal Adsorbers and flood the adsorber beds.
D. Maintain stable reactor power to prevent introduction of further ignition sources and connect Nitrogen bottles to the Charcoal Adsorbers and purge the adsorber beds.
QUESTION RO 93 NRC RECORD # WRI 93 ANSWER: B. SYSTEM # Fire K/A Generics 2.4.25: 2.9 LP#
Olu. SRO TIER GROUP / ROTIER 3 GROUP REFERENCE: AOP-0039 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.10/43.4/
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l U.S. NUCLEAR REGULATORY COMMISSION i WRITTEN EXAMINATION FEBRUARY 1999 l REACTOR OPERATOR l
QUESTION 94 l
An SNEO is being sent out on ajob in a High Radiation Area.
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The Dose rate in the area of thejob is 120 mrem /hr.
The job b expected to take I hour and 15 minutes. l I
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The operator's exposure history to date for the year is 1800 mrem.
Can the operator be utilized for thisjob and WHY?
A. Yes, however the operator will have to have an approved extension on dose limits before thejob.
i B. Yes, the operator will NOT exceed his administrative limits. I C. No, the operator will exceed his federal dose limits.
D. No, the operator will exceed administrative dose limits which are NOT allowed to
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be extended.
I QUESTION RO 94 NRC RECORD # WRI 94 l ANSWER: B. SYSTEM # Rad K/A Generics 2.3.1: 2.6 i LP# Limits OBJ. SRO TIER GROUP / RO TIER 3 GROUP REFERENCE: GET Hand Book - Rad NEW CLASS Worker Training MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.10/43.4 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 95 The plant is operating at 100% power. -
The Control Room Fupervisor has a tagout that requires independent verification.
Under which one of the following conditions can the Operations Shill Superintendent waive independent verification?
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A. The components are located inside the Containment over the Hydraulic Control Units. I B. The valves to be tagged out are located around the Main Turbine Stop Valves. ,
1 C. The components involve a Temporary Alteration on the HPCS Diesel Generator Air Start System.
D. The components to be tagged are required to continue power operation.
QUESTION RO 95 NRC RECORD # WRI 95 ANSWER: B. SYSTEM # K/A Generics 2.3.2: 2.5 LP# ALARA OBJ. ERO TIER GROUP / RO TIER 3 GROUP REFERENCE: ADM-0076 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.10/43.4
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1939 REACTOR OPERATOR QUESTION 96 Given the following conditions:
Reactor Power 20 % l l Reactor Level - 170 inches I l Reactor Pressure 900 psig l Suppression Pool Temperature 125 *F l Suppression Pool Level 17 feet 1 3 SRVs are open l
Which one of the following best describes the correct actions to be taken for the above given conditions?
l A. Commence an emergency depressurization per EOP-0001 A.
l B. Makeup to the Suppression Pool using High Pressure Core Spray. l L
C. Close the open SRVs and allow Reactor Pressure to rise to 1065 psig. I
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D. Lower Suppression Pool Level to maintain within the safe region of HCLL. 1
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l QUESTION RO 96 NRC RECORD # WRI 96 l ANSWER: B. SYSTEM # K/A Generics 2.4.20: 3.3 ,
LP# EOP-0002 OBJ. SRO TIER GROUP / ROTIER 3 GROUP REFERENCE: EOP-0001 NEW CLASS EOP-0002 MODIFIED BANK DIFF 4 HCLL DATE USED: RO SRO BOTH CFR 41.10/43.5
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 97 Under which one < ~ uie following conditions is concurrent verification allowed?
l A. A valve line up for the RHR following operation in Shutdown Cooling.
B. Restoration of an electrical line up following a Division 1 Diesel Generator Surveillance.
C. I&C is lifting leads in preparation for a Surveillance.
D. Verification of alignment of a locked rising stem valve at the Diesel Generator.
QUESTION RO 97 NRC RECORD # WRI 97 ANSWER: C. SYSTEM # K/A Generics 2.1.23: 3.9 LP# Procedure OBJ. SRO TIER GROUP / RO TIER 3 GROUP REFERENCE: ADM-0076 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: CFR 41.10 R_O SRO BOTII l- _
o U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 98 One result of the RPV Level 1 LOCA signal is the shedding ofloads, and the subsequent sequencing back on of required loads. The purpose for this automatic action can be summarized by:
A. Sequencing safety related loads minimizes the time the diesel generator runs in an unloaded condition.
B. Shedding loads allows time for the diesel generator to warm up prior to placing loats on the bus.
C. Sequencing safety related loads ensures that starting current loads do not overload the diesel generator.
D. Snedding loads ensures that non essential loads do not operate in an unmonitored condition for long periods of time.
QUESTION RO 98 NRC RECORD # WRI 98 ANSWER: C. SYSTEM # K/A Generics 2.1.28: 3.2 LP# Procedure OBJ. SRO TIER GROUP / ROTIER 3 GROUP REFERENCE: Technical Specifications NEW CLASS Bases 3.8.1 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.7
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR QUESTION 99 A reactor startup is in progress. The following personnel are in the Main Control Room:
Duty Manager Manager Operations Superintendent Operations Operations Shift Superintendent Control Room Supervisor Shift Technical Advisor (SRO licensed)
Unit Operator At-The-Control Operator The Unit Operator has identified two steps in GOP-0001, PLANT STARTUP, that need to be combined and one step that should be bypassed until later in the startup. It is acceptable to combine and bypass steps in GOP-0001 provided that:
A. the Unit Operator gets concurrence from the At-The-Control Operator and it is documented appropriately.
. B. it is authorized by the Operations Shift Superintendent and documented appropriately.
C. it is authorized by the Manager Operations and documented appropriately.
D. the Duty Manager has been notified and the notification has been documented appropriately.
QUESTION RO-99 NRC RECORD # WRI-99 i ANSWER: B. SYSTEM # GOP- K/A Generics 2.2.1: 3.7 l 0001 j LP#
OBJ. SRO TIER GROUP / RO TIER 3 GROUP REFERENCE: GOP-0001 NEW CLASS MODIFIED BANK L DIFF 3
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NRC 2 DATE USED: -RO SRO BOTH CFR 41.10
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U.S. NUCLEAR REGULATORY COMMISSION ;
WRITTEN EXAMINATION FEBRUARY 1999 REACTOR OPERATOR j QUESTION 100 l The ' Unit Operator has just completed venting the Primary Containment in accordance with the SOP (System Operating Procedure). All equipment operated as expected.
What type of Control Board Line-up is required following this evolution?
A. Documentation on the Unit Operator Rounds (OSP-0028), with the line-up verified by a different operator.
B. Documentation on the Unit Operator Rounds (OSP-0028), however the line-up is l not required to be verified by a different operator.
C. Documentation in the Main Control Room Log Book, with the line-up verified by a different operator. j D. Documentation in the Main Control Room Log, however the line-up is not required to be verified by a different operator.
QUESTION RO-100 NRC RECORD # WRI-100 '
ANSWER:. C SYSTEM # K/A Generics 2.1.29: 3.4 Procedure LP# j OIM. SRO TIER GROUP / RO TIER 3 GROUP REFERENCE: ADM-0022 NEW CLASS ;
MODIFIED BANK i DIFF 3 i NRC 2 DATE USED: RO SRO BOTH CFR 41.10 l l
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i U.S. Nuclear Regulatory Cosmnission Site-Specific Written Examination Applicant Information Region:
I/ II /III/(IV)
Docket #:
Facility / Unit:
Date: 19 February 1999 _ RIVER BEND STATION I License Level: RO / ($RO] Reactor Type: W / CE / BW /(GE)
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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 80.00 percent. Examint4 tion papers will be collected four hours after the examinatior. starts.
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Applicant Certification All work done on this examination is my own. I have neither given nor received aid.
Applicant's Signature Results Examination Value Points Applicant's Score Points Applicant's Grade Percent
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR
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QUESTION 1 The plant has undergone a transient which resulted in a Recirculation Flow Control Valve Runback.
Which one of the following describes the allowable operation of the Recirculation Flow Control Valves, prior to resetting the Flow Control Valve runback? !
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l The Recire Flow Control Valves can:
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l A. be closed using loop manual operation, however, they can only be opened to the l 12 % valve position. l B. be closed using loop manual operation, however they can only be opened to the point which they ran back. ,
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C. not be closed any further because they are a' the full closed stop and cannot be l reopened due to a hydraulic block on the valves.
l D. not be closed any further because they are at the full closed stop, however they can be opened to the 22 % valve position.
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QUESTION SRO 1 NRC RECORD # WRI 1 ANSWER:. B. SYSTEM # 053 K/A 295001 - A1.05: 3.3/3.3 LP# RBS-1-LEC-GPST-A053 l
OBJ. 2b; 12 SROTIER 1 GROUP 2 / ROTIER 1 GROUP 2 l REFERENCE: ARP-P680-4A-A3&A9 NEW CLASS MODIFIED BANK
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DATE USED: RO SRO .BOTH CFR 41.10/41.5/
43.5
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 2 The plant is operating at 100 % power when a short circuit occurs on the DC bus supplying power for ATWS ARI/RPT. This causes all of the power supply breakers to BYS-PNLO2A2 to trip, resulting in a loss of power to ATWS ARI/RPT.
Which one of the following describes the response of the ARI system and the Reactor Recirculation Pumps?
A. ARI will not function, however the Reactor Recirculation pumps will trip to OFF immediately.
B. ARI will actuate causing a depressurization of the scram air header and tbc Reactor Recirculation pumc= will trip to OFF immediately.
C. ARI will not function and the Reactor Recirculation pumps will net trip on an ATWS condition.
- D. ARI will actuate causing a depressurization of the scram air header on an ATWS condition, however the Reactor Recirculation pumps will not trip.
QUESTION SRO 2 NRC RECORD # WRI 2 i ANSWER: C. SYSTEM # 052; K/A 295004 AK2.03: 3.3/3.3 053 LP# RBS-1-STM-GPST-A0053 OBJ. 2d LP# RBS-1-STM-GPST-A0052 OBJ. 3,4e SROTIER I GROUP 1/ RO TIER 1 GROUP 2 REFERENCE: PRINTS NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.7
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 3 The plant is operating at 100 % power. A rupture in the high pressure leg of the "A" Feedwater line Flow element causes Feed Flow Indications and inputs to the Reactor Recirculation sysiem to change. The At-The-Controls operator swapped to single element control of Reactor Level with little change in level.
Which one of the following describes the response of the Reactor Recirculation System?
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A. Both Recirculation Pumps will remain at present speed, however the Recire Flow I Control Valves will runback to minimum position.
B. Both Recirculation Pumps will downshift to slow speed operation with the Recire Flow Control Valves remaining at present position. !
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C. Both Recirculation Pumps will trip to OFF due to cavitation interlock . sot being met and Recirc Flow Control Valves not being at 22%.
D. Both Recirculation Pumps will remain at present speed and the Recirc Flow Control Valves will remain at the present position. l QUESTION SRO 3 NRC RECORD # WRI 3 ANSWER: D. SYSTEM # 053 K/A 202002 K6.01: 3.5/3.5 LP# RBS-1-STM-GPST-A0053 Olu. 2a,b,j; SROTIER2 GROUP 1/ ROTIER 2 GROUP 1
REFERENCE: AOP-0024 NEW CLASS ARP-P680-4A-A3; A9; MODIFIED BANK '
DIFF 4 C1; C7 NRC 3 DATE USED: RO SRO BOTH CFR 41.6 l
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l U.S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION FEBRUARY 1999 l
SENIOR REACTOR OPERATOR QUESTION 4 !
The plant was operating at 100 % power, when a Main Turbine trip caused a reactor scram and lift of one (1) Safety Relief Valve. Reactor level increased such that all three (3) Reactor Feed Pumps tripped. The At-The-Controls operator restarted two of the Reactor Feed Pumps and stabilized Reactor Level at +5 inches.
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Which one of the following describes the final status of the Reactor Recirculation I System?
A. Both Recirculation Pumps will downshift to slow speed operation with the Recire Flow Control Valves for both loops running back to minimum position.
B. Both Recirculation Pumps will downshift to slow speed operation with the Recire l Flow Control Valves remaining at the pre-transient positions.
C. Both Recirculation Pumps will trip to OFF with the Recire Flow Control Valves I for both loops running back to minimum position.
D. Both Recirculation Pumps will trip to OFF with the Recirc Flow Control Valves remaining at the pre-transient positions.
QUESTION SRO 4 NRC RECORD # WR14 ANSWER: A. SYSTEM # 053 K/A 202001 K4.16: 3.3/3.6 LP# RBS-1-STM-GPST-A0053 OBJ. 2b,c,d, SRO TIER 2 GROUP 2 / RO TIER 2 G; JJP 2 j;12 REFERENCE: AOP-0024 NEW CLASS ARP-P601-19A-H8; H11 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.5/41.6/43.6 l
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- UcS. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 5 The plant is being started up from cold shutdown. Which one of the following describes the sequence of Cl 1-SOV operation for a control rod withdrawal?
122 withdraw drive 123 insert dnve 121 insert exhaust 120 withdraw exhaust i
A. Solenoid valves 120 and 122 open to withdraw the control rod, at the end of the movement 122 closes followed by 120 closing last to allow the control rod to l settle. l I
B. Solenoid valve 123 opens to insert the control rod slightly, then 123 closes, and 122 opens to withdraw the control rod, at the end of the movement 122 closes and 120 opens to allow the control rod to settle. )
C. Solenoid valves 121 and 123 open to insert the control rod slightly, then 121 and 123 close, and 120 and 122 open to withdraw the control rod, at the end of tLe movement 122 closes followed by 120 closing last to allow the control rod to j
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settle.
D. Solenoid valves 121 and 123 open to insert the control rod slightly, then 121 and .
123 close, and 122 opens to withdraw the control rod, at the end of the movement !
122 closes followed by 120 opening to exhaust water then closing last to allow the control rod to settle.
QUESTION SRO 5 NRC RECORD # WRI 5 ANSWER: C. SYSTEM # 052 K/A 201001 A1.03: 2.9/2.8 LP# RBS-1-STM-GPST-A0052 Olu. Ib;2g, SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 1 h
REFERENCE: prints NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.6
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h l U.S., NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR l
QUESTION 6 j l The Control Rod Drive mechanism in certain conditions is capable ofinserting a control
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rod with ONLY the use of Reactor Pressure.
Which one of the following describes the reactor pressure and physical means by which this can be accomplished? 1 l
l A. Reactor Pressure is 150 psig Drive Mechanism under piston area vented B. Reactor Pressure is 625 psig ;
Drive Mechanism under piston area vented. ;
I C. Reactor Pressure is 150 psig .
Drive Mechanism over piston area vented.
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D. Reactor Pressure is 625 psig Drive Mechanism over piston area vented.
QUESTION SRO 6 NRC RECORD # WRI 6 l ANSWER: D. SYSTEM # 052 K/A 201003 K1.02: 2.9/3.0 LP# RBS-1-STM-GPST-A0052 OBJ. Sk; SROTIER2 GROUP 3 / RO TIER 2 GROUP 2 lib REFERENCE: Tech Spec Bases B3.1.5 NEW CLASS ,
MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.2 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 7 The plant was operating at 100% power at the beginning of the transiert.
The At-The-Controls Operator observes the following indications.
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" Control Rod Drift" annunciator P680-7A-B02 in alarm
" Rod Drift" pushbutton on P680 back-lit
" Accumulator Trouble" annunciator P680-7A-C03 in alarm !
"Accum Fault" pushbutton on P680 back-lit I
"Ackn Accum Fault" pushbutton on P680 back-lit
" Scram Valves" pushbutton on P680 back-lit APRM power 97 % 1 Which one of the following plant conditions was the probable cause?
I A. Single control rod drifting inward.
B. Single control rod drifting outward.
C. Control Rod Drop Accident D. Single control rod scram.
QUESTION SRO 7 NRC RECORD # WRI 7 AhSWER: D. SYSTEM # 500 K/A 201005 A3.01: 3.5/3.5 LP# HLO-057-6 A3.02: 3.5/3.5 A3.04: 3.3/3.3
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OBJ. 7,9-1 SROTIER2 GROUP 1/ ROTIER 2 GROUP 1 REFERENCE: ARP-P680-07-7A-B02; NEW CLASS 7A-C03 MODIFIED BANK 7 DIFF 3 DATE USED: RO SRO BOTH CFR 41.6
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 8 RBS is operating at 10% rated power with the mode switch in the STARTUP position, and total core flow at 53%. APRM E and 11 are bypassed due to failed power supplies.
The following is the present status of the APRMs versus LPRM inputs and indicated power:
APRM A B C D E F G II LPRM LVL D 3 4 2 2 2 3 3 3 LPRM LVL C 4 3 3 4 4 4 4 4 LPRM LVL B 2 4 4 3 2 3 3 2 LPRM LVL A 4 2 2 4 4 4 2 4 INDICATED 10 % 13 % 12 % 14 % 0% 11 % 13 % 0%
POWER byp byp l
LPRM 22-39D has failed downscale and must be bypassed to allow troubleshooting. ]
With present conditions would this action be allowed?
Attached is the LPRM vs. APRM assignments Attachment of SOP-0074.
A. Yes, conditions are satisfactory.
B. Yes, however an LCO would have to be written on the associated APRM for Administrative inputs.
C. No, this action would result in a half scram and administrative LCO requirements not to be m et.
D. No, this action would result in a full reactor scram. l l
QUESTION SRO 8 NRC RECORD # WRI 8 ANSWER: C. SYSTEM # 505 K/A 215005 A1.04: 4.1/4.1 LP# RBS-I-LEC-GPST-A0503 A1.02: 3.9/4.0 A1.03: 3.6/3.6 OBJ. 22a,b;23; SROTIER 2 GROUP I / ROTIER 2 GROUP 1 l 26;29a,b;
REFERENCE: SOP-0074 NEW CLASS REP-0037 MODIFIED BANK DIFF 3 Tech Spec Bases 3.3.1.1.2 DATE USED: RO SRO BOTH CFR 41.6
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l l U.S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION FEBRUARY 1999 '
l SENIOR REACTOR OPERATOR
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The plant is in the process of mitigating the impact of a LOCA, which has uncovered
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fuel, and hydrogen is present in the Containment. The Control Room Supervisor has requested the Hydrogen Recombiners be started.
Using the attached procedure, determine the Required Recombiner Power Setting, and also determine the time required to reach the Required Recombiner Power Setting.
Pre-LOCA Containment Temperature 90 F Post-LOCA Containment Pressure 4 psig Post-LOCA Containment Temperature 120 *F
' A. 50.31 KW at 20 min.
i B. 50.31 KW at 25 min. j C. 52.03 KW at 20 min.
D. 52.03 KW at 25 min.
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QUESTION SRO 9 NRC RECORD # WRI 9 ANSWER: D. SYSTEM # 254 K/A 500000
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EA1.03: 3.4/3.2 LP# 2.1.25: 2.8/3.1 OBJ. SRO TIER 3 GROUP / ROTIER 1 GROUP 1 REFERENCE: SOP-0040 NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.10/43.5
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 10 The plant is in Mode 4 with RHR "A"in Shutdown Cooling. A misalignment of RPV drain valves has resulted in reactor vessel level lowering. The following are the present plant parameters:
Reactor Pressure 0 psig Reactor Water Level +34 inches and lowering Reactor Water Temperature 160 F Drywell Pressure 0 psig Which one of the following describes the operation of the RHR "A" Shutdown Cooling System if Reactor Water Level continues to lower?
A. At + 9.7 inches RPV water level, E12-F053A (RHR A SDC Injection Valve) will isolate, which will cause a low flow on the RHR A pump automatically opening l
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E12-F064A (RHR Pump A Min Flow to Sup PI).
B. At + 9.7 inches RPV water level, E12-F053A (RHR A SDC Injection Valve), ,
E12-F008 and F009 (RHR Shutdown Cooling Isol Valves) will isolate causing the '
RHR A Pump to trip.
i C. At +31 inches RPV water level, E12-F006A (RHR Pump A SDC Suction Valve) l will isolate, which will cause E12-F004A (RHR Pump A Sup Pl Suction Valve) to open and the low flow on the RHR A pump to open the E12-F064A (RHR Pump A Min Flow to Sup P1).
D. At +31 inches RPV water level, E12-F008 and F009 (RHR Shutdown Cooling ,
Isol Valves) will isolate, svhich will cause the RHR A pump to trip; the RHR A l pump trip will cause E12-F053A (RHR A SDC Injection Valve) to close. )
QUESTION SRO 10 NRC RECORD # WRI 10 ANSWER: B. SYSTEM # 204 K/A 205000- A2.05: 3.5/3.7 LP# A2.06: 3.4/3.5 OBJ. SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2 REFERENCE: SOP-0031 NEW CLASS AOP-0003 MODIFIED BANK DIFF 3 l DATE USED: RO SRO BOTH CFR 41.7 1.
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR. REACTOR OPERATOR QUESTION 11 The plant is in a ATWS with RHR "B" in Suppression Pool Cooling to control the temperature of the Suppression Pool. Reactor water level is being controlled at - 150 inches. Electrical power is lost to ENS-SWGlB. The Diesel Generator did not automatically start. After 30 minutes power is ready to be restored to ENS-SWGlB from the Division II Diesel Generator.
Which one of the following describes actions that MUST be taken per procedure prior to the restoration of power to the bus?
A. RHR B system piping would have drained down into the Suppression Pool, such that the RHR B pump circuit breaker must be racked out to prevent a pump start when the bus is re-energized.
B. RHR B system valves will have to be manually realigned for Standby to ensure when the bus is re-energized and the RHR B pump starts the pump is on minimum flow to begin operation.
C. RHR B system will require venting of the system piping prior to the re-energizing i of the bus to prevent water hammer of the system piping. l D. RHR B system valves must be manually re-aligned and the RHR B pump circuit breaker must be racked out prior to re-energizing the bus to prevent an uncontrolled restart upon power restoration.
l QUESTION SRO 11 NRC RECORD # WRI 11 I ANSWER: A. SYSTEM # 204 K/A 219000 K6.01: 3.2/33 l LP#
OBJ. SRO TIER 2 GROUP 2 / ROTIER 2 GROUP 2 l g'i- REFERENCE: SOP-0031 sect.2.1.2 NEW CLASS
' - AOP-0004 sect. 5.2.4.1 MODIFIED BANK l DIFF 4 l NRC 3 j j DATE USED: RO SRO BOTH CFR 41.9 j
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i-U.S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION FEBRUARY 1999 L SENIOR REACTOR OPERATOR QUESTION 12 l
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The plant is in a Refueling outage with RHR A in Shutdown Cooling. LPCS, RHR B and l C are tagged out for maintenance. HPCS is aligned for manual ECCS operations.
Maintenance personnel moving a Recirc Pump motor in the Drywell drop the motor on RWCU piring coming from the Bottom Head drain. Reactor level is rapidly lowering.
RPV Level is + 3 inches. All isolations occurred as expected.
'Which one of the following describes the operation of RHR "A"?
A. The operator can manually initiate LPCI A, which will open E12-F004A (RHR A Sup Pl Suction Valve) and re-start RHR A pump and open E12-F042A (RHR A LPCI Injection Valve).
B. After the automatic isolations are complete E12-F006A (RHR A SDC Isolation Valve) will automatically close. Once E12-F006A is started closed, E12-F004A (RHR A Sup Pl Suction Valve) will open, the RHR "A" Pump will start and E12-F042A (RHR A LPCI Injection Valve) will then open.
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C. After the automatic isolations are complete E12-F006A (RHR A SDC Isolation Valve) will automatically close. Once E12-F006A is started closed, E12-F004A (RHR A Sup P1 Suction Valve) will open, operator will be required to manually start the RHR "A" Pump, and E12-F042A (RHR A LPCI Injection Valve) will then open.
D. The operator will have to close E12-F006A (RHR A SDC Isolation Valve) then open E12-F004A (RHR A Sup Pl Suction Isolation Valve). After the suction is re-aligned, LPCI A can be manually initiated.
QUESTION SRO 12 NRC RECORD # WRI 12 ANSWER: D. SYSTEM # 204 K/A 203000 K1.14: 3.6/3.7 LP#
OBJ. SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 1 REFERENCE: SOP-0031 sect. 2.2.3; NEW CLASS sect. 4.3/5.1 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.7
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U.S. NUCLEAR REGULATORY COMMISSION ,
WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 13
- The plant is operating at 100 % power. The Auxiliary Building SNEO reports there is a l major leak on the service water side of the CCP Heat Exchangers and the only way to isolate the leak is to isolate all service water to the CCP Heat Exchangers.
Water temperature on CCP is 110 F and rising.
Reactor Recirculation Pump Motor temperatures are rising.
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Which one of the following describes the actions required to be taken by AOP-0011 1 LOSS OF REACTOR PLANT COMPONENT COOLING WATER with regard to loss I of cooling water to CCP?
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A. Shutdown CCP pumps, and isolate the CCP Heat Exchangers on the Service Water side. Repair the leak, un-isolate the Service Water side of the CCP Heat Exchangers and re-start the CCP Pumps.
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B. Reduce CCP heat loads by tripping to OFF the operating CRD Pump, and start the l standby CCP pump to increase cooling water flow while mechanics effect repairs on the broken piping.
C. Manually scram the reactor and trip and isolate both Reactor Recirculation ;
Pumps, and isolate service water to the CCP Heat Exchangers '
D. Reduce CCP heat loads by down shifting the Reactor Recirculation Pumps to slow speed, establish a feed and bleed on CCP to remove heat, and isolate the leak.
QUESTION- SRO 13 NRC RECORD # WRI 13 ANSWER: C. SYSTEM # 115; K/A 295018 AA2.03: 3.2/3.5 118 LP#
OBJ. SROTIER 1 GROUP 2 / ROTIER 1 GROUP 2 REFERENCE: AOP-0011 sect 4.0 NEW CLASS AOP-0009 sect. 5.5 MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.4 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 14 Certain Safety Relief Valves are designated as " Low-Low Set".
Which one of the following describes the bases and operation of the " Low-Low Set M Vs"?
A. When the first SRV opens on the relief function, two (2) SRV relief setpoints are lowered. This is done to minimize the cyclic stress on the Containment due to SRV lifting and ensures the Containment design basis is met.
B. As reactor pressure increases above the scram setpoint, the relief setpoints on five (5) SRVs are lowered to start them opening well below the design pressure of the reactor vessel to prevent exceeding reactor design pressure.
C. When the first SRV opens on the relief function, five (5) SRVs are opened automatically and their reset pressures are lowered. This minimizes the number of SRV lifts by extending the length of time they are open.
D. As reactor pressure increases above the scram setpoint, the reset setpoints for five (5) SRVs are lowered. This minimizes the number of SRV lifts by extending the length of time they are open.
QUESTION SRO 14 NRC RECORD # WRI 14 ANSWER: A. SYSTEM # 050 K/A 295025 EK3.09: 3.7/3.7 LP#
!' OBJ. SRO TIER 1 GROUP I / RO TIER 1 GROUP 1 REFERENCE: Tech Spec Bases B3.3.6.4 NEW CLASS MODIFIED BANK l DIFF 3 ( DATE USED: RO ERO BOTH CFR 41.3/41.5 l 41.7/43.2 l
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l U.S. NUCLEAR REGULATORY COMMISSION i WRITTEN EXAMINATION FEBRUARY 1999 l SENIOR REACTOR OPERATOR i
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QUESTION 15 l The following stable conditions exist in the plant:
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Reactor Power 0 % (All Rods In) l Reactor Pressure 150 psig i Reactor Water Level + 4 inches l Drywell Pressure 0.8 psig Main Steam Tunnel Temperature 150 *F 1 Reactor Mode Switch in SHUTDOWN i
Given the above plant conditions, determine which one of the following describes the systems which should have received isolation signals. I i
A. CCP; MSIVs; RCIC; RWCU B. MSIVs; RCIC; RHR to Radwaste; RWCU i
C. CCP; RCIC; Reactor Samplelines; RWCU D. MSIVs; Reactor Sample lines; RHR to Radwaste l
QUESTION SRO 15 NRC RECORD # WRI 15 ANSWER: B. SYSTEM # 058 K/A 223002 A1.02: 3.7/3.7 LP# ;
OBJ. SROTIER 2 GROUP 1/ ROTIER 2 GROUP 1 REFERENCE: AOP-0003 NEW CLASS ;
MODIFIED BANK
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DIFF 2 DATE USED: RO SRO BOTH CFR 41.7/41.9 i
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i U.S. NUCLEAR REGULATORY COMMISSION
. WRITTEN EXAMINATION FEBRUARY 1999 I l
SENIOR REACTOR OPERATOR
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i QUESTION 16 l
The plant is operating at 100 % power. The Feedwater Level Control (FWLC) System is )
in three element control with the "A" Reactor Water Level Channel selected. A rupture 1 occurs on the "A" reference leg causing a level change.
Assuming no other instruments are affected by the rupture, which one of the following describes the required operator action?
The Operator should:
A. transfer the FWLC System to single element control.
B. select the "B" Reactor Water Level Channel.
C. allow the level dominant signal to take control and return level to normal. ,
i D. manually control water level with RCIC and / or HPCS.
' QUESTION SRO 16 NRC RECORD # WRI 16 ANSWER: B. SYSTEM # 107; K/A 295009 AA1.02: 4.0/4.0
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501 LP#
OBJ. SRO TIER I GROUP 1/ ROTIER I GROUP 1 REFERENCE: A OP-0006 NEW CLASS ARP-P680-3A-C08 MODIFIED BANK DIFF 4 NRC 3 DATE USED: RO SRO BOTH CFR 41.7
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 17 The plant is operating at 100 % power. A failure of the in-service Steam Jet Air Ejector has resulted in Condenser vacuum being lost.
Which one of the following describes the sequence of events in response to a loss of Condenser vacuum? ASSUME NO OPERATOR ACTIONS.
A. Main Turbine Trip Reactor scram on Turbine trip MSIV closure Main Steam Bypass Valves close Condensate Pumps continue to operate B. MSIV closure Reactor scram on MSIV closure Main Turbine Trip j Main Steam Bypass Valves close l Condensate Pumps continue to operate I l
C. Reactor scram l Main Turbine Trip l MSIV closure Main Steam Bypass Valves close ;
Condensate Pumps trip on low suction pressure D. Main Turbine Trip MSIV closure Reactor scram on MSIV closure
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Main Steam Bypass Valves close
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Condensate Pumps trip on low suction pressure QUESTION SRO 17 NRC RECORD # WRI 17 '
ANSWER: A. SYSTEM # 104 K/A 295002 2.4.4: 4.0/4.3 LP#
OBJ. SRO TIER 1 GROUP 2 / ROTIER 1 GROUP 2 REFERENCE: AOP-0005 NEW CI. ASS ARP-P680-2A-A01; MODIFIED BANK DIFF 3 2A-B91 DATE USED: RO SRO BOTH CFR 41.4 l
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l U.S. NUCLEAR REGULATORY COMMISSION
- WRITTEN EXAMINATION FEBRUARY 1999
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SENIOR REACTOR OPERATOR f
QUESTION 18 l
l The plant is operating at 100 % power. A rupture of the Instrument Air line entering Containment has caused a loss of air inside the Containment. The Unit Operator has isolated the leak by closing the Containment Instrument Air Isolation Valve (IAS-MOV106)
Which one of the following describes the response of the Control Rod Drive System to a loss ofInstrument Air? ASSUME NO OTHER OPERATOR ACTION.
A. The Scram Discharge Volume Vent and Drain Valves will open, the Scram valves will immediately open on a low air pressure signal, and the CRD Flow Control Valve will fail closed.
B. The Scram Discharge Volume Vent and Drain Valves will close, the Scram valves will immediately open on a low air pressure signal, and the CRD Flow Control Valve will fail open.
C. The Scram Discharge Volume Vent and Drain Valves will open, the Scrata valves will individually drift open as air pressure drops, and the CRD Flow Control Valve will fail open.
D. The Scram Discharge Volume Vent and Drain Valves will close, the Scram valves will individually drift open as air pressure drops, and the CRD Flow Control Valve will fail closed.
QUESTION SRO 18 NRC RECORD # WRI 18 ANSWER.: D. SYSTEM # 122 K/A 295019 AK2.01: 3.F3.9 LP# RBS-1-LEC-LP-Il052 OBJ. Sg; 6b SROTIER 1 GROUP 2 / ROTIER 1 GROUP 2 REFERENCE: AOP-0008 NEW CLASS MODIFIED BANK DIFF 4 NRC 3 DATE USED: RO SRO BOTH CFR 41.4/41.7 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 19 The plant was operating at 100 % power. A rupture of the Instrument Air line on the main header downstream of the Air Compressors has resulted in a complete loss of instrument air to the plant. The At-The-Controls operator has manually scrammed the !
reactor.
Which one of the following describes the response of the Condensate and Reactor Feedwater System to a loss ofInstrument Air and the ability of the plant to feed the Reactor Vessel? ASSUME NO OTHER OPERATOR ACTION.
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A. The Condensate demin inlet and outlet valves will fail open and the feed pump !
min flow valves will fail closed. The Feed Reg Valves fail as is allowing i continued feeding of the reactor.
B. The Condensate and Reactor Feedwater system min flow valves fail open, l however, the lines are sized to allow continued operation of the systems, Feed l Reg valves fail open to allow continued injection to the Reactor. l C. The Reactor Feedwater pumps will trip on low suction due to min flow valves failing open. Makeup to the reactor will come from HPCS and RCIC when they auto start.
D. The Condensate and Reactor Feedwater flow to the reactor will stop due to the Feed Reg Valves failing closed. Makeup to the reactor will come from HPCS and RCIC when they auto start.
- QUESTION SRO 19 NRC RECORD # WRI 19 ANSWER: C. SYSTEM # 122; K/A 300000 K3.02: 3.3/3.4 104;107 LP# HLO-030 OBJ. . 8 SRO TIER 2 GROUP 2 / ROTIER 2 GROUP 2 REFERENCE: AOP-0008 NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO ROTH CFR 41.4 I
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l U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 20 I
The plant is in a refueling outage with the reactor vessel disassembled. The Reactor cavity is filled to 23 feet above the flange. Fuel movement is in progress. The refueling cavity bellows ruptures.
In accordance with AOP-0027 FUEL HANDLING MISHAPS which one of the following is NOT an allowed safe position for an irradiated fuel bundle?
A. The Upper Containment Fuel Pool Fuel Rack.
B. The Cattle Chute hanging on the fuel grapple.
C. The Fuel Transfer Mechanism carriage rack.
D. The Reactor Vessel in an area of the core which has no fuel, i
QUESTION GRO 20 NRC RECORD # WRI 20 l ANSWER: B. SYSTEM # 055 K/A 295023 AK3.01: 3.6/4.3 LP# ;
OBJ. SRO TIER I GROUP i / ROTIER I GROUP 3 )
REFERENCE: AOP-0027 NEW CLASS FIIP-0003 MODIFIED BANK DIFF 2 ,
DATE USED: RO SRO BOTH CFR 41.2/41.10/ l
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41.12/43.4/43.5/
43.6/43.7
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! U.S. NUCLEAR REGULATORY COMMISSION I WRITTEN EXAMINATION FEBRUA~RY 1999 SENIOR REACTOR OPERATOR QUESTION 21 The plant is in an ATWS with reactor level being maintained low.
The following parameters are indicated in the Main Control Room:
Reactor Pressure 450 psig Reactor Level Wide Range - 135 inches
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Reactor Level Upset Range + 6 inches ,
Reactor Level Shutdown Range + 10 inches Reactor Level Narrow Range 0 inches Reactor Level Fuel Zone - 180 inches Drywell Temperature 145 ft 310 F Containment Temperature 119 ft 165 *F 2 SRVs open I
Which one of the following describes the Reactor Level instruments allowed to be used? !
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A. Fuel Zone and Upset Range only.
- B. Upset Range and Wide Range only.
C. Fuel Zone and Wide Range only.
D. All Reactor Level instruments are invalid.
I QUESTION SRO 21 NRC RECORD # WRI 21 ANSWER: C. SYSTEM # 051 K/A 295027 EKI.02: 3.0/3.2 LP#
OBJ. SROTIER I GROUP 1/ RO TIER I GROUP 2 REFERENCE: EOP-0001 NEW CLASS Caution 1 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.5/41.9/
41.10/43.5 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 j SENIOR REACTOR OPERATOR QUESTION 22 The plant is in a LOCA with attempts being made to restore reactor water level. I.evel is offscale low on all level instruments except Fuel Zone.
The following parameters are indicated in the Main Control Room:
l Reactor Pressure 50 psig Reactor Level Fuel Zone - 220 inches Drywell Temperature 145 ft 310 F Containment Temperature 119 ft 165 F Suppression Pool Level 15.0 ft ;
7 SRVs open
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Which one of the following describes the method to be used for determining Suppression j Pool Temperature? l l
l A.' Suppression Pool Temperature indicators on H13*P808. l B. Safety Parameter Display System (SPDS) computer.
C. Remote Shutdown Panel Suppression Pool Temperature indicators.
D. RHR temperature recorder with RHR operating.
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QUESTION SRO 22 NRC RECORD # WRI 22 ANSWER: D. SYSTEM # 057 K/A 295026 EA2.01: 4.1/4.2 LP#
OBJ. SROTIER 1 GROUP 1/ RO TIER I GROUP 2 REFERENCE: EOP-0001 NEW CLASS Caution 9 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.7/41.9/
41.10/43.5
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR
' QUESTION 23 The plant is operating at 4 % power in a reactor startup. The B CRD pump is tagged out with the oil sump drained for maintenance.
The A CRD pump trips. The CRS dispatches a SNEO to investigate the pump circuit breaker. The SNEO reports that the breaker has over current trip flags and the lockout device is tripped.
Electrical Maintenance is called to investigate.
The following parameters are indicated in the Main Control Room:
Reactor Pressure 450 psig ;
Reactor Water Level + 34 inches Main Steam Bypass valves are fully closed.
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With present plant conditions, which one of the following describes the actions to be taken?
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A. Increase reactor pressure to > 600 psig and wait for electrical maintenance to j repair the CRD Pump.
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i B. If two or more control rod accumulator faults exist on withdrawn control rods, l fully insert the control rods within 20 minutes or place the reactor mode switch in i SHUTDOWN. !
C. If one or more control rod accumulator faults exist on withdrawn control rods, which cannot be inserted, immediately place the reactor mode switch in SHUTDOWN. l D. Increase reactor pressure to > 600 psig, and restore charging water header pressure to >l520 psig within 20 minutes or place the reactor mode switch in SHUTDOWN.
QUESTION SRO 23 NRC RECORD # WRI 23 ANSWER: C. SYSTEM # 052 K/A 295022 AK3.01: 3.7/3.9 LP#
OBJ. SROTIER 1 GROUP 2 / RO TIER 1 GROUP 2 REFERENCE: ARP-P60122A-A01 NEW CLASS ( Yech Specs 3.1.5 MODIFIED BANK
!- DIFF 3 DATE USED: RO SRO BOTH CFR 41.5/41.6/
6 43.2
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U.S. NUCLEAR REGULATORY COMMISSION c WRITTEN EXAMINATION FEBRUARY 1999 l SENIOR REACTOR OPERATOR
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QUESTION 24 The plant is operating at 90 % power.
Which one of the following descriptions of plant conditions will result in a Main Turbine !
Trip and descr bes the basis for the trip?
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A. The Main Turbine will trip when the selected Reactor Narrow Range Level Instruments has level at + 51 inches. This is to prevent the erosion of the Main !
Steam piping and Main Control Valves' seats, from moisture carryover. l l
l B. The Main Turbire will trip when tv o of the Reactor Narrow Range Lc el Instruments have level at + 51 inchea. This is to prevent the erosion of the Main i Steam piping and Main Control V6ves' seats, from moisture carryover. )
C. The Main Turbine ivill trip when two of the Reactor Narrow Range Level Instruments have level at 4 51 inches. This is to prevent the erosion of the Main Turbine blades, from moisture carryover.
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D. The Main Turbine will trip when the selected Reactor Narrow Range Level Instruments has level at + 51 inches. This is to prevent the erosion of the Main ;
Turbine blades, from moisture carryover. !
QUESTION SRO 24 NRC RECORD # WRI 24 ANSWER: C. SYSTEM # 110 K/A 295008 AKl.01: 3.0/3.2 LP# 295005 AA2.07: 3.5/3.6 OBJ. SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 REFERENCE: A OP-0002 NEW CLASS TRM 3.3.7.3 MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.5 i
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 25 A plant transient has caused Suppression Pool level to reach 20 ft. 8 in. The HPCS suction valves have transferred to the Suppression Pool suction alignment.
The plant is operating at 90 % power.
With present plant conditions, which one of the following describes the actions to be taken?
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A. Leave HPCS suction aligned to the Suppression Pool until the Suppression Pool level can be lowered and the transfer logic reset. ,
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B. If desired, transfer suction back to the CST since a low level in the CST will transfer suctions back to the ' Suppression Pool if required.
C. Transfer the HPCS suction back to the CST since the CST is the required suction source for HPCS to remain operable.
D. Leave HPCS suction aligned to the Suppression Pool since HPCS will transfer to j the CST on a Low Suppression Pool Level to ensure an adequate source.
QUESTION SRO 25 NRC RECORD # WRI 25 ANSWER: A. SYSTEM # 203 K/A 295029 EK2.03: 3.5/3.3 LP#
OBJ. SRO TIER 1 GROUP 2 / ROTIER 1 GROUP 2 REFERENCE: ARP-P60116A-C04; C05 NEW CLASS 16A-F03 MODIFIED BANK DIFF 3 Tech Spec 3.5.1 & bases DATE USED: RO SRO BOTH CFR 41.7
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 26
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The plant is in mode 2. The following parameters are indicated in the Main Control I
Room:
APRMs I
A B C D E F G Il 11 % 12 % 13 % 11 % 14 % 15 % 13 % 13 %
IRMs (range / reading)
A B C D E F G 11 R9/ 36 R9/ 39 R9/ 37 R9/ 37 R9/ 36 R9/ 36 R9/ 37 R9/ 37 Reactor Water Level + 36 inches Rs actor Pressure 950 psig Main Turbine speed 1800 rpm-With present plant conditions, which one of the following is correct with regard to the status of the Reactor?
A. No RPS actuation.
B. Half scram on Division I. ,
C. Half scram on Division II.
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D. Full scram.
QUESTION SRO 26 NRC RECORD # WRI 26 ANSWER: C. SYSTEM # 508 K/A 212000 K5.02: 3.3/3.4 LP# HLO-061 OBJ. 5 SROTIER 2 GROUP 1/ ROTIER 2 GROUP 1 REFERENCE: A0P-0001 NEW CLASS Tech Specs 3.3.1.1 MODIFIED BANK l DIFF 3 l
DATE USED: RO SRO BOTH CFR 41.6 l l
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l 1 U.S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR ;
QUESTION 27 ;
The plant is in mode 2 after a normal refueling outage. The following parameters are indicated in the Main Control Room:
IRMs (range / reading) !
! I A B C D E F G H R2/100 R3/ 39 R2/ 75 R3/ 30 R2/ 80 R3/15 R3/18 R3/ 36 SRMs (cps)
A B C D i l
3 2 5 2.0 x 10 3.0 x 10 2.5 x 10 Bypassed i
With present plant conditions, which one of the following is correct with regard to the status of the Reactor?
A. No RPS actuation and no Control Rod Blocks I B. Control Rod Block only.
C. Half scram and Control Rod Block.
D. Full scram and Control Rod Block.
QUESTION SRO 27 NRC RECORD # WRI 27 ANSWER: C. SYSTEM # 500; K/A 215004 A3.04: 3.6/3.6 !
i 508; 503; 504 LP# RBS-1-LEC-GPST-A0503 l OBJ. 4; 13 l LP# 11L0-061
!- OBJ. ?; 5 SROTIER 2 GROUP 1/ ROTIER 2 GROUP 1 l REFERENCE: A0P-0001 NEW CLASS I
Tech Specs 3.3.1.1; MODIFIED BANK DIFF 2 3.3.2.1 DATE USED: RO SRO BOTH CFR 41.6
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 28 The plant .N 'n mode 3. The Control Room personnel are in the process of cooling down the reactor from rated conditions.
Which one of the following is used to reduce the possibility of erroneous Reactor Vessel Level indications and trips?
A. Temperature / Pressure compensation inputs from the ERIS computer.
B. Continuous reference leg fill to each of the reference legs from CRD.
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C. Reference leg purge to a common line from Reactor Recirculation loops.
l D. Manual Temperature compensation by I&C via resetting calibration conditions.
QUESTION SRO 28 NRC RECORD # WRI 28 ANSWER: B. SYSTEM # 051; K/A 216000 K5.06: 3.4/3.6 052 LP# RBS-1-LEC-LP-H052 -
OBJ. Ig SROTIER 2 GROUP 1/ ROTIER 2 GROUP 1 REFERENCE: SOP-0001 NEW CLASS SOP-0002 MODIFIED BANK DIFF 2 I DATE USED: RO SRO BOTH CFR 41.2
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! U.S. NUCLEAR REGULATORY COMMISSION ,
l WRITTEN EXAMINATION FEBRUARY 1999 l SENIOR REACTOR OPERATOR QUESTION 29 l Which one of the following describes the hazards associated with the injection of RCIC l into the Reactor head region?
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A. RCIC injection will cause pressure fluctuations on all of the Reactor Vessel Level reference legs.
B. RCIC injection will cause a downshift of the Reactor Recirculation Pumps due to a change ofinlet subcooling.
C. RCIC injection will cause a rapid decrease in Reactor Pressure due to the collapse of the steam bubble thus causing a reactivity excursion.
D. RCIC injection will result in excessive carryover of moisture into the steam lines that in turn will cause impingement on the main turbine blading.
i QUESTION SRO 29 NRC RECORD # WRI 29 ANSWER: D. SYSTEM # 209 K/A 217000 2.1.32: 3.4/3.8 )
l LP#
OBJ. SROTIER 2 GROUP 1/ ROTIER 2 GROUP 1 REFERENCE: SOP-0035 NEW CLASS MODIFIED BANK DIFF 2 l DATE USED: RO SRO BOTH CFR 41.7/41.10 l i 1 l
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U.S. NUCLEAR REGULATORY COMMISSION
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WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR l QUESTIP./. 30 Whid one of the following is NOT a basis for the normal Suppression Pool Level?
A. Prevent th .. ;ntroduction of steam into the suction of the ECCS pumps by providing subcooling for ECCS suctions.
B. Provide a mechanism for the limiting of fission product release in a LOCA condition. ,
i C. Prevent excessive clearing loads (Level in SRV Tailpipe) on the SRVs that could i result in tail pipe damage.
D. Provide sufficient water for the absorption of steam energy before heating up excessively.
QUESTION SRO 30 NRC RECORD # WR130 ANSWER: A. SYSTEM # 057 K/A 223001 K4.02: 3.6/3.7 K4.01: 3.7/3.8 LP# 2.2.25: 2.5/3.7 OBJ. SRO TIER 2 GROUP t / ROTIER 2 GROUP 1 ;
REFERENCE: Tech Spec 3.6.2.2 bases NEW CLASS USAR 6.2.1 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.7/43.2 l-
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l U.S. NUCLEAR REGULATORY COMMISSION l l l WRITTEN EXAMINATION FEBRUARY 1999 i l SENIOR REACTOR OPERATOR l l 1 QUESTION 31
During an ATWS, with stable conditions, the Control Room Supervisor has ordered Standby Liquid Controlinitiated.
Which one of the following would be an indication that Standby Liquid Control is injecting into the core? (CONSIDER EACH ANSWER SEPARATELY.) l A. Reactor pressure is rising.
B. Main Steam Bypass valves are throttling closed. l C. Reactor levelis lowering.
D. Feedwater flow is rising.
QUESTION SRO 31 NRC RECORD # WRI 31 ANSWER: B. SYSTEM # 201 K/A 211000 A1.09: 4.0/4.1 A1.01: 3.6/3.7 A1.02: 3.8/3.9 !
A1.03: 3.6/3.6 LP# i;LO-016 A1.04: 3.6/3.7 OBJ. 6 SROTIER 2 GROUP I / ROTIER 2 GROUP 1 REFERENCE: SOP-028 NEW CLASS MODIFIED BANK DIFF 4 NRC 3 DATE USED: RO SRO ROTH CFR 41.6
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 32 The plant was operating at 80 % power. A transient in the Fancy Point Switchyard resulted in a Main Turbine trip. All control rods did not fully insert. The following are the plant parameters at present:
Reactor Pressure 900 psig Reactor Level - 50 inches wide range Reactor Power 22.5 %
Suppression Pool Temperature 112 F Feedwater flow is stable 2.79 Mlbm/hr.
IRMs and SRMs detectors have been inserted.
All Main Steam Bypass valves are fully open. j
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Two (2) SRVs are open.
To avoid exceeding the Heat Capacity Temperature Limit (HCTL) curve, the Control Room Supervisor has ordered Reactor Pressure lowered to 700 psig using SRVs.
Which one of the following describes the reaction ofindicated Reactor Power !
immediately following the opening of the SRVs, and why?
A. Reactor power will rise due to the lowering of the reactor coolant temperature adding positive reactivity.
B. Reactor power will rise due to the water level inside the core rising causing more moderation of neutrons.
C. Reactor power will drop due to the voiding of the water in the core as it flashes to steam.
D. Reactor power will drop due to the o aderator temperature rising with the low flow through the core.
QUESTION SRO 32 NRC RECORD # WRI 32 ANSWER: C. SYSTEM # K/A 295037 EK1.01: 4.1/4.3 ATWS LP#
OBJ. SROTIER 1 GROUP 1/ ROTIER 1 GROUP 1 REFERENCE: Applied Reactor Theory NEW CLASS EOP Bases MODIFIED BANK DIFF 3 i DATE USED: RO SRO BOTH CFR 41.1/41.2/41.6 !
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I U.S. NUCLEAR REGULATORY COMMISSION I WRITTEN EXAMINATION FEBRUARY 1999 I
, SENIOR REACTOR OPERATOR t
QUESTION 33 I
Due to an ATWS, Standby Liquid Control has been initiated from an initial tank level of l l 1985 gallons.
Which one of the following describes when Standby Liquid Control Injection can be terminated? .
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A. Reactor Temperature has been lowered to 68 F with no control rod motion. )
I B. All Control Rods have been inserted fully with the exception of two control rods l on cpposite sides of the core.
C. Reactor Pressure has been lowered to 0 psig and RHR Shutdown Cooling interlocks have been met.
D. Standby Liquid Control Boron tank level has lowered from its initial level to 441 gallons.
l QUESTION SRO 33 NRC RECORD # WRI 33 l ANSWER: D. SYSTEM # 201; K/A 295037 EA2.03: 4.3/4.4 j ATWS EK1.04: 3.4/3.6 i LP# EK1.05: 3.4/3.6 l OBJ. SRO TIER 1 GROUP 1/ ROTIER 1 GROUP 1 REFERENCE: EOP-0005 Encl.15 NEW CLASS EOP-1A step RQA-22 MODIFIED BANK DIFF 3 I DATE USED: RO SRO BOTH CFR 41.6/41.10 l 43.5/43.6
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U.S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION FEBRUARY 1999 l SENIOR REACTOR OPERATOR QUESTION 34 l
The plant is performing a reactor startup from cold shutdown. The reactor isjust at the point of adding heat. The CRS instructed the operators to stop the startup to perform a surveillance. During this time, the reactor went suberitical and power had dropped to range 3 of the IRMs. The At-The-Controls operator noting power selected the next control rod and withdrew the control rod from 00 to 48 with continuous motion, resulting !
in a sustained 20 second period. The following are the plant parameters at present: l l
Reactor Pressure 80 psig !
Reactor Level + 40 inches l I
Which one of the following describes the action the At-The-Controls operator should l take? '
A. Monitor IRMs and range them according to the power increase to keep them on scale.
B. Perfomt the coupling checks ior the Control Rod, and inform the Reactor Engineer of an increase in power rise,
C. Insert the Control Rod to a position which causes Reactor Periad to be > 30 seconds.
D. Withdraw the next in sequence Control Rod to maintain the power increase to achieve the point of adding heat.
QUESTION SRO 34 NRC RECORD # WRI 34 ANSWER: C. SYSTEM # 500; K/A 295014 A1.04: 3.2/3.3 503; 504 LP# .
05LI. SRO TIER I GROUP 1/ RO TIER 1 GROUP 1 REFERENCE: GOP-0001 Att 1 step 3.3 NEW CLASS Susquehanna reactivity MODIFIED BANK 1 DIFF 3 Event 7/98 DATE USED: RO SRO BOTH CFR 41.1/41.2 41.6/43.6
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U.S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION FEBRUARY 1999 l SENIOR REACTOR OPERATOR
QUESTION 35
l The plant was operating at 100 % power at the end of cycle when a failure of the A Reactor Feedwater line in the Turbine Building has caused reactor water level to drop. ,
l The reactor scrammed however all control rods did not insert. The CRS is directing l actions per EOP-0001 A - RPV Control - ATWS. l The following are the plant parameters at present:
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Reactor Power 25 %
Reactor Pressure 940 psig Reactor Level - 180 inches Fuel Zone Range l l HPCS injection is overridden The MSIVs are closed.
RCIC, CRD, and SLC are injecting into the Reactor.
Four(4) SRVs are open.
Which one of the following describes the method of Adequate Core Cooling being employed?
A. Core submergence.
B. Steam Cooling with injection.
C. Steam Cooling without injection.
D. Adequate Core Cooling is NOT being assured such that Emergency Depressurization is required.
QUESTION SRO 35 NRC RECORD # WRI 35 ANSWER: B. SYSTEM # RPV K/A 295031 EA2.04: 4.6/4.8 Control -
LP#
. OBJ. SRO TIER 1 GROUP 1/ RO TIER I GROUP 1 REFERENCE: EOP-0001A RLA-12 NEW CLASS EOP Bases MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.10/43.5 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 36 Which one of the following describes the basis for the maximum design internal pressure
- of the Drywell?
A. - Maximum Drywell Pressure is + 20 psid based on a double-ended shear of a l Recirculation Pump discharge pipe.
B. Maximum Drywell Pressure is + 20 psid based on a double-ended shear of a Main Steam Line upstream of the MSIVs.
C. Maximum Drywell Pressure is + 25 psid based en a double-ended shear of a Recirculation Pump discharge pipe.
D. Maximum Drywell Pressure is + 25 psid based on a double-ended shear of a Main Steam Line upstream of the MSIVs.
QUESTION SRO 36 NRC RECORD # WRI 36 AMWER: D. SYSTEM # 057 K/A 295024 EK1.01: 4.6/4.2 LP# llLO-013 (057)
OBJ. 3; 4 SRO TIER I GROUP 1/ RO TIER 1 GROUP 1 REFERENCE: Tech Spec Bases 3.6.5.4 NEW CLASS USAR 6.2.1.1.1 MODIFIED BANK DIFF 3 '
NRC 2 DATE USED: RO SRO BOTH CFR 41.9 l
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U.S. NUCLEAR I1EGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 37 i The plant is operating at 100 % power. A leak on the Service Water Header in the !
Drywell requires the isolation of the Service Water piping inside the Drywell.
Which one of the following describes the reaction of the plant to this isolation? l A. Drywell temperature will rise along with Drywell pressure such that eventually the scram and isolation setpoint for Drywell pressure will be reached. .
l t B. Drywell temperature will remain stable due to the evaporation of water inside the i Drywell sumps absorbing heat energy.
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C. Drywell temperature will rise and stabilize at the point where evaporation of the l
water in the Drywell will absorb the heat and Drywell pressure will stabilize 4
< l.68 psig. l D. Drywell temperature will remain stable due to the continued circulation of the Drywell atmosphere through the Drywell Coolers and the transfer of heat to any residual water remaining in the Service Water piping.
QUESTION SRO 37 NRC RECORD # WRI 37 ANSWER: A. SYSTEM # 404; K/A 295010 AK2.05: 3.7/3.8 ;
118; 057 j LP# RBS-I-LEC-GPST-A0118 '
OBJ, 7; 8 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 ;
REFERENCE: SOP-0060 step 2.2 NEW CLASS l A OP-0009 MODIFIED BANK DIFF. 3 DATE USED: RO SRO BOTH CFR 41.4/41.9
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 38 l l
The plant is at 5 % power. Chemistry samples taken indicate that fuel damage is present in the core. Radiation levels in Offgas and the Main Steam lines have risen drastically.
Which one of the following describes the reaction of the plant if the Main Steam Line Radiation Levels reach 3 times the normal background readings?
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A. The Reactor will scram, the Main Steam I ines and the Reactor Sample Lines will isolate, and the Condenser Air Removal Pumps will trip.
B. Initia* ion of Standby Gas Treatment and Annulus Mixing, and an isolation of the Main Steam Lines and Reactor Sample Lines.
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C. The Reactor Sample Lines will isolate and the Condenser Air Removal Pumps will trip and isolate.
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D. The Reactor will scram, Standby Gas Treatment and Annulus Mixing will initiate, I and the Condenser Air Removal Pumps will isolate.
QUESTION SRO 38 NRC RECORD # WRI 38 )
ANSWER: C. SYSTEM # 058; K/A 295033 EK3.03: 3.8/3.9 511 LP# j OBJ. SROTIER 1 GROUP 2 / ROTIER 1 GROUP 2 REFERENCE: AOP-0003 Att 1 E NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.11/41.12 43.4 l
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U.S. NUCLEAR REGULATORY COMMISSION
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WRITTEN EXAMINATION FEBRUARY 1999
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SENIOR REACTOR OPERATOR
QUESTION 39 l
The plant is at 5 % power. Chemistry samples taken indicate that fuel damage is present
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in the core.
l Which one of the following will NOT automatically initiate measures to control an l
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Offsite Radiation release?
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l A. Fuel Building Ventilation Radiation Monitors.
B. Control Room Ventilation Radiation Monitors.
C. Offgas Post-Treatment Radiation Monitor.
D. Reactor Building Annulus Ventilation Exhaust Radiation Monitor.
QUESTION SRO 39 NRC RECORD # WRI 39 ANSWER: B. SYSTEM # 058; K/A 295034 EKl.02: 4.1/4.4 511; 606; 402; 403 LP# j OBJ. SRO TIER 1 GROUP 2 / ROTIER 1 GROUP 2 REFERENCE: AOP-0003 Att 1 E; Z; AA NEW CLASS BB; CC; FF MODIFIED BANK DIFF 3 AOP-0039 i DATE USED: RO SRO BOTH CFR 41.7/41.11 l 41.13/43.4 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 40 The plant is in a reactor startup Mode 2. Due to an I&C error, HPCS initiated. The CRS directed the Control Room Operator to override close the HPCS injection Valve E22*F004. The Control Room Operator closed E22*F004. The maximum Reactor Water Level reached was + 56 inches. I&C can NOT reset the trip unit that caused the Initiation Signal.
Reactor Water Level is now + 36 inches and lowering.
Which one of the following describes the operation of E22*F004 HPCS INJECT ISOL VALVE with water level now in the normal band?
A. The valve will automatically open on receipt of a Reactor Water Level - Low Level 2 signal.
B. The valve can only be opened using the valve hand switch in the open position.
C. The valve can only be opened if the HPCS High Reactor Water Level signal is )
manually reset, and then the valve hand switch is taken to the OPEN position.
D. The valve will automatically reopen if the HPCS Manual Initiation Pushbutton is l
depressed. l l
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QUESTION SRO 40 NRC RECORD # WRI 40 ANSWER: C. SYSTEM # 203 K/A 209002 A4.03: 3.8/3.8 j LP#
OBJ. SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 I
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REFEPENCE: SOP-0030 NEW CLASS ARP-P601 16A-B04 MODIFIED BANK DIFF 3 16A-F02 !
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DATE USED: RO SRO BOTH CFR 41.7/41.8 l
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I l U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 41 l The plant is in a LOCA condition. The leak is inside the Drywell from the "A" Main Steam Line.
The following plant parameters exist:
Reactor Power 0 % (All Rods In)
Reactor Level +15 inches Reactor Pressure 600 psig I Drywell Pressure 10 psig j Drywell Temperature 230 'F i Containment Temperature 102 F Containment Pressure 3.2 psig Suppression Pool Level 21.0 feet l Suppression Pool Temperature 102 *F Which one of the following describes the method that should be employed to control Containment Pressure?
l A. Align the Contaimnent Ventilation Systems for Normal Containment Venting.
B. Operate all available Containment Cooling fans and coolers.
C. Use Emergency Containment Venting, if radioactive release rates are acceptable.
D. Perform an Emergency Containment Venting irrespective of radioactive release.
QUESTION SRO 41 NRC RECORD # WRI 41 ANSWER: B. SYSTEM # 057; K/A 226001 2.1.27: 2.8/2.9 403 LP#
OBJ. SROTIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: EOP-0002 CP-2,3,4; NEW CLASS CT-2,3,4 MODIFIED BANK DIFF 3 DATE USED: RO SRO ROTH CFR 41.9
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 42 The plant is eperating at 100 % power. The CCP line inside Containment going to the RWCU Non-Regenerative Heat Exchangers has ruptured. An operator in the area has manually isolated CCP to the Non-Regenerative Heat Exchangers.
Which one of the following describes the plant response with NO further operator actions?
A. The RWCU Filter Demins will isolate and go into hold due to Low CCP Flow through the Non-Regenerative Heat Exchangers.
B. The RWCU Filter Demins bypass valve will open and the Filter Demins will go into Hold due to High Filter Demin Inlet temperature.
C. The RWCU pumps will immediately trip on High Filter Demin Inlet Temperature and G33*MOVF004, RWCU PUMPS OUTBD SUCTION VALVE will isolate to protect the Filter Demins.
D. G33*MOVF004, RWCU PUMPS OUTBD SUCTION VALVE will isolate on High Filter Demin Inlet temperature causing the Icl/CU pumps to trip on low flow.
l QUESTION SRO 42 NRC RECORD # WRI 42 ANSWER: D. SYSTEM # 601; K/A 204000 K4.04: 3.5/3.6 115 LP#
OBJ. SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2 REFERENCE: ARP-P680 01 A-B01 NEW CLASS 01A-A01 MODIFIED BANK DIFF 3 SOP-0090 DATE USED: AOP-0011 RO SRO BOTH CFR 41.4 i
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j U.S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION FEBRUARY 1999 l SENIOR REACTOR OPERATOR {
QUESTION 43 I I
The plant is operating at 30 % power. ;
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Met Tower Data indicates 40 * F. l The Circulating Water Pumps IB and IC are tagged out for repairs. !
Suddenly the l A Circulating Water Pump trips due to a phase to phase short. .
l Which one of the following describes the expected response of the Main Condensers? l i
(Assume NO operator actions.) l l
A. Main Condenser vacuum will initially decrease then return to original value due to the one remaining Circulating Water Pump.
B. Main Condenser vacuum will decrease and stabilize above the turbine trip setpoint, since power is within the capabilities of one Circulating Water Pump.
C. Main Condenser vacuum will increase due to the increased flow rate of the remaining Circulating Water Pump. j D. Main Condenser vacuum will remain stable at its present value, as the Steam Jet Air Ejectors will control Main Condenser Vacuum. i i
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l QUESTION SRO 43 NRC RECORD # WRI 43 )
ANSWER: B. SYSTEM # 104; K/A 256000 K6.02: 3.1/3.1 ]
103 J LP# j OBJ. SROTIER 2 GROUP 3 / ROTIER 2 GROUP 2 l REFERENCE: AOP-0005 NEW CLASS I MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.4 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999
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SENIOR REACTOR OPERATOR l QUESTION 44 The plant is in a Refueling outage with the Refueling Platform located over the Cattle Chute area.
Which one of the following WILL ALLOW movement of the Refueling platform over the Reactor Vessel core?
A. The Main Holst unloaded, Control Rod 24-37 is selected at position 24 on H13*P680, and the Reactor Mode Switch is in REFUEL.
B. The Main Hoist loaded, Control Rod 24-37 is at position 24 on H13*P680, and the Reactor Mode Switch is in REFUEL.
C. The Main Hoist is unloaded, no Control Rods are selected on H13*P680, and the Reactor Mode Switch is in STARTUP.
D. The Main Hoist is loaded, no Control Rods are withdrawn as indicated on H13*P680, and the Reactor Mode Switch is in STARTUP.
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QUESTION SRO 44 NRC RECORD # WRI 44 ANSWER: A. SYSTEM # 055 K/A 234000 K5.02: 3.1/3.7 LP# HLO-022 OBJ. 2 SROTIER 2 GROUP 2 / RO TIER 2 GROUP 3 REFERENCE: FHP-0003 NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.4/43.7 l
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( U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 45 The plant hasjust returned to 100 % power following completion of Refueling Outage 7.
The Refueling Outage began 45 days ago.
l The Fuel Pool Cooling Pump SFC-PI A seal blew out and started dropping level in the Lower Fuel Pool, the pump has been secured and isolated.
The Fuel Pool Cooling Pump SFC-P1B is tagged out for seal replacement.
Using the Decay Heat curves, determine the present conditions of the Spent Fuel Pool.
Time to Boil Decay Heat Heat-up rate A. 33 hrs 6.4 Mbtu/hr 2.0 *F/hr B. 48 hrs 8.5 Mbtu/hr 2.0 F/hr C. 33 hrs 8.5 Mbtu/hr 1.8 F/hr D. 48 hrs 6.4 Mbtu/hr 1.8 F/hr l
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QUESTION SRO 45 NRC RECORD # WRI 45 ANSWER: D. SYSTEM # 602 K/A 233000 A4.05: 2.7/3.1 A2.07: 3.0/3.2 LP# 2.1.25: 2.8/3.1 OBJ. SROTIER 2 GROUP 3 / RO TIER 2 GROUP 3 l REFERENCE: OSP-0037 Att.9 NEW CLASS AOP-0051 Att.1-5 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.10/43.5 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 46 The plant is shutting down for a refueling outage.
The following are the Temperature and Pressure parameters as the plant was cooled -
down.
Time Rx Press Rx Temp.
1000 103 psig 340 F 1030 74 psig 320 F 1100 45 psig 292 F 1130 18 psig 256 *F 1200 2 psig 220 "F 1230 0 psig 182 *F 1300 0 psig 145 F 1330 0 psig 107 F !
1400 0 psig 69 F
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Which of the following statements is correct concerning the Reactor Coolant System?
A. All RPV pressure and temperature limits are within specifications )
B. RPV pressure vs. temperature limits are satisfied, but the cooldown rate has been exceeded.
C. RPV pressure vs. temperature limits have been violated, but the cooldown rate is satisfactory.
D. RPV pressure vs. temperature limits and the cooldown rate have been exceeded.
QUESTION SRO 46 NRC RECORD # WRI 46 I ANSWER: C. SYSTEM # 050 K/A 280002 K5.05: 3.1/3.3 LP# 2.1.25: 2.8/3.1 OBJ. SRO TIER 2 GROUP 3 / RO TF.R 2 GROUP 3 REFERENCE: Tech Specs 3.4.11 NEW CLASS Steam Tables MODIFIED BANK DIFF 3 STP-050-0700 DATE USED: GOP-0002 RO SRO BOTH CFR 41.3/43.2 i
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 I SENFOR REACTOR OPERATOR l l
QUESTION 47 The plant was scrammed at 28 % power as part of a normal shutdown.
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Which one of the following statements is true regarding the ability to reset RPS?
A. The scram signal is unable to be reset, until RPV water level is restored to below Level 8.
B. The scram signal can be reset by taking the RPS reset switches to RESET.
C. The scram signal can be reset by resetting the ARI logic channels then taking the RPS reset switches to RESET. j D. The scram signal can be reset by taking the SDV Bypass switches to BYPASS and then placing the RPS reset switches to RESET. '
l QUESTION SRO 47 NRC RECORD # WRI 47 ANSWER: D. SYSTEM # 508 K/A 212000 A4.14: 3.8/3.8 LP#
OBJ. SROTIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: Tech Specs 3.3.1.1 NEW CLASS q AOP-0001 MODIFIED BANK I DIFF 3 DATE USED: RO SRO BOTH CFR 41.6/41.7
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U.S. NUCLEAR REGULATORY COMMISSION ;
WRITTEN EXAMINATION FEBRUARY 1999 l SENIOR REACTOR OPERATOR l
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j QUESTION 48 l l
The plant has experienced an ATWS condition. I The following parameters exist at the present time:
Reactor Mode Switch is in Shutdown Reactor Pressure is 600 psig ,
Reactor Water Levelis - 120 inches I
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Reactor Power is <1 %
Suppression Pool Temperature is 130 F MSIVs are closed.
Under which one of the following coaditions would it be appropriate to exit EOP-1 A -
'A. Standby Liquid Control has injected Cold Shutdown Boron Weight into the reactor with an additional 25 % injected to allow for imperfect mixing and leakage. l B. Standby Liquid Control has injected sufficient quantities to allow the SLC pumps to be secured as directed by EOP-1 A.
C. The Reactor Engineer has performed shutdown margin determinations and has determined that adequate shutdown margin exists for all conditions.
D. Chemistry and the STA have determined that the combination of Boron and Control Rods has brought the reactor subcritical for all conditions.
QUESTION SRO 48 NRC RECORD # WRI 48 ANSWER: C. SYSTEM # K/A 295015 AA2.02: 4.1/4.2 EOP-0001A LP#
OBJ, SROTIER 1 GROUP 1/ ROTIER 1 GROUP 1 REFERENCE: Tech Specs 3.1.1 bases NEW CLASS EOP-0001 A step RCA-1 MODIFIED BANK-DIFF 3 EPSTG*0002 step RCA-1
[ DATE USED: Tech Spec 3.1.7 bases RO SRO ROTH CFR 41.1/41.2/41.6 l U.6 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR j QUESTION 49 The Design Basis Accident for Maximum Drywell Atmosphere Temperature is . .
A. SRV tai pipe failure within the Drywell. j B. A smal'. steam leak break in the Drywell. l C. A double ended shear of the Recirculation System discharge piping D. A double ended shear of a Main Steam pipe upstream of the Inboard MSIV.
l QUESTION SRO 49 NRC RECORD # WRI 49 ANSWER: B. SYSTEM # 057 K/A 295012 2.4.18: 2.7/3.6 LP#
OBJ. SROTIER I GROUP 2 / RO TIER I GROUP 1 REFERENCE: Tech Specs 3.6.5.5 bases NEW CLASS USAR 6.2.1.1 MODIFIED BANK DIFF 3 6.2.1.1.3.1.7 NRC 2 i DATE USED: 6.2.1.1.3.1.7.4 RO SRO BOTH CFR 41.9 l l
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l l U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 50 The temperatures in the Auxiliary Building (Secondary Containment) are increasing.
The Control Room Supervisor has entered EOP-0003 Secondary Containment and Radioactive Release Control, and has determined a leak exists.
Which one of the following would require the Emergency Depressurization of the Reactor and Why?
l A. If the temperature in the RCIC Room exceeds 200 *F, Emergency Depressurization is required to protect the equipment in the area from being l exposed to extreme conditions. I B. If the temperature in the RCIC Room exceeds 200 F and the RCIC Room Sump Level is five (5) inches over the floor of the RCIC Room, Emergency Depressurization is required to limit the extreme energy release into the RCIC Room.
C. If the temperature in the RCIC Room exceeds 144 *F and the Main Steam Line Tunnel exceeds 135 *F, an Emergency Depressurization i an option to remove the driving head of the only system which could cause the extreme conditions.
D. If the temperature in the RCIC Room and Main Steam Line Tun 1el both exceed 200 F, Emergency Depressurization is required to reduce the flow from the break by reducing the thermal driving head.
QUESTION SRO 50 NRC RECORD # WRI 50 ANSWER: D. SYSTEM # EOP-3 K/A 295032 EK3.01: 3.5/3.8 LP#
OBJ. SROTIER 1 GROUP 2 / RO TIER 1 GROUP 3 REFERENCE: EOP-0003 STEPS SC-19; NEW CLASS SC-20 MODIFIED BANK DIFF 3 EPSTG*0002 SC-19; DATE USED: SC-20 RO SRO BOTH CFR 41.4/41.9 41.10/43.5 w
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t U.S. NUCLEAR REGULATORY COMMISSION WRflTEN EXAMINATION FEBRUARY 1999 l SENIOR REACTOR OPERATOR
QUESTION 51 l
l All Power to the Division I DC bus has been lost.
i Concerning the Low Pressure Core Spray System operation, which one of the following statements is true?
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A. In the event of an actual LOCA condition, LPCS will NOT operate automatically, however, the system can be manually initiated from the Main Control Room and inject into the Reactor. I B. In the event of an actual LOCA condition, LPCS will automatically start, however, the injection valve must be manually opened due to the loss of the
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l automatic opening feature of the pressure permissive.
C. Low Pressure Core Spray is unable to be initiated manually or automatically, ;
however, the LPCS pump can be manually started from the Main Control Room I and placed on minimum flow or can be aligned for injection.
l D. Low Pressure Core Spray is unable to be initiated manually or automatically, and the LPCS pump will not operate from the Main Control Room, if the pump is !
started locally, it will operate on minimum flow.
QUESTION SRO 51 NRC RECORD # WRI 51 ANSWER: D. SYSTEM # 205 K/A 209001 K2.03: 2.9/3.1 LP#
OBJ. SROTIER 2 GROUP 1/ ROTIER 2 GROUP 1 REFERENCE: ARP-P60121 A-H08 NEW CLASS Electrical Dwgs MODIFIED BANK-DIFF 4 828ES35AA sh 3,4,6,10 NRC 3 DATE USED: ESK5CSL01 RO SRO ROTH CFR 41.7/41.8 ESK6CSL01 i
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 52 During maintenance in the H13*P628, an I&C Technician shorted ajumper causing a loss of DC power to the Division I SRV solenoids.
l Concerning the operation of the Automatic Depressurization System, which one of the
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following statements is true? ;
i A. The ADS bRVs have opened on a logia initiation signal and must have their ;
handswitches taken to OFF to close the ADS SRVs.
B. The ADS SRVs are unable to be opened using manual or automatic initiation I signals, however, they will still function by placing either division handswitch in I the OPEN position.
C. The ADS SRVs will operate in automatic for all modes of operation using the Division II logic system or the Division II handswitch in the OPEN position.
D. The ADS SRVs are unable to be opened using manual and automatic initiation logic, however, the valves may be opened using the Division II handwitches.
QUESTION SRO52 NRC RECORD # WRI 52 ANSWER: C. SYSTEM # 202; K/A 218000 K3.02: 4.5/4.6 305 K4.03: 3.8/4.0 K5.01: 3.8/3.8 LP# A2.05: 3.4/3.6 OBJ. SRO TIER 2 GROUP 1/ ROTIER 2 GROUP 1 REFERENCE: ARP-P60119A-A07; B08 NEW CLASS 19A-B11; E08; H08 MODIFIED BANK DIFF 4 Elect DWGs 851E225AA NRC 3 DATE USED: RO SRO BOTH CFR 41.7/41.8 P
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l l U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR
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QUESTION 53
Which one of the following is used to operate the Safety Relief Valves?
The SRVs are operated by:
A. at least one solenoid valve must energize admitting air pressure OR Reactor pressure overcomes spring pressure.
B. two solenoids must energize for the air valve to admit air pressure OR Reactor pressure overcomes spring pressure.
C. at least one solenoid must de-energize to close the air admitting valve which allows Reactor pressure to open the SRV.
D. two solenoids must de-energize to open the valve admitting air pressure to open the SRV OR Reactor pressure overcomes spring pressure.
QUESTION SRO53 NRC RECORD # WRI 53 ANSWER: A. SYSTEM # 202; K/A 239002 K2.01: 2.8/3.2 305; 109 LP#
OBJ. SROTIER 2 GROUP 1/ ROTIER 2 GROUP 1 REFERENCE: Elect Dwgs 851E225AA NEW CLASS P&lD 3-1C Syst.109 MODIFIED BANK DIFF 3 ADS Logic DATE USED: RO SRO BOTH CFR 41.3 l
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U.S. NUCLEAR REGULATORY COMMISSION I WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 54
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l Standby Gas Treatment has started on a high Drywell pressure. The Unit Operator has placed the "B" Standby Gas Train in standby. j Which one of the following describes the response of the f 'andby Gas Treatment System i to a High-High Annulus Exhaust Radiation signal on both divisions?
A. The "B" Standby Gas Treatment Train will automatically restart from standby,
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B. The "A" Standby Gas Treatment Train will shutdown, then both Standby Gas
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Treatment Trains will re-initiate.
C. The "A" Standby Gas Treatment Train will remain operating and the "B" Standby I Gas Treatment Train will remain in standby.
D. Both Standby Gas Treatment Trains shutdown and isolate awaiting further i operator action.
QUESTION SRO54 NRC RECORD # WRI 54 ANSWER: C. SYSTEM # 257; K/A 261000 K1.08: 2.8/3.1 403 K4.01: 3.7/3.8 i LP#
OBJ, SROTIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: Elect Dwgs NEW CLASS l ESK-06GTS01 sh 1,2 MODIFIED BANK DIFF 3 ESK-06GTS02 sh 1,2 NRC 2 DATE USED: SOP-0043 & 0059 RO SRO BOTH CFR 41.? 4 ARP-P863 71 A-C07; G07 73A-C04; D05; E05; F04 i i
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 55 ACB28 for E12-C002C RHR Purnp "C" Motor on bus ENS-SWGlB (4160 volt) has to be racked out to allow an electrical surveillance to be completed.
Which one of the following describes the safety materials required to be utilized when racking out the circuit breaker?
A. Face shield; approved eye protection; switchingjacket; and high voltage rubber gloves only.
B. Face shield; approved eye protection; switching jacket; and leather gloves only.
l C. Face shield; switchingjacket and leather gloves only.
D. Face shield; switching jacket; and high voltage rubber gloves only.
QUESTION SRO 55 NRC RECORD # WRI 55 ANSWER: B. SYSTEM # 303 K/A 262001 2.1.30: 3.9/3.4 LP# 2.1.26: 2.2/2.6 OBJ. SROTIER 2 GROUP 1/ RO TIER 2 GROUP 2 j REFERENCE: SOP-0046 Att.5 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.4 l l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR l l
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I QUESTION 56 l
l l The plant electrical distribution system is in a normal lineup (on the preferred I transformers). i The plant has scrammed. j Reactor Water Level dropped to - 82 inches and is rising.
The Entergy Grid has experienced transients due to severe weather in the area.
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Plant 4160 Volt bus voltages DROPPED to 3500 volts for 10 seconds.
Which one of the following statements is the condition of the Division I, II, III buses after this voltage transient?
A. Division I bus is being supplied from 1RTX-XSRIC Division II bus is being supplied from IRTX-XSRID Division III bus is 1 emg supplied from 1RTX-XSRID B. Division I bus is being supplied from Div I D/G Division II bus is being supplied from Div II D/G Division III bus is being supplied from Div III D/G C. Division I bus is being supplied from 1RTX-XSRIC Division II bus is being supplied from 1RTX-XSR1D Division III bus is being supplied from Div III D/G D. Division I bus is being supplied from Div I D/G Division II bus is being supplied from Div II D/G Division III bus is being supplied from 1RTX-XSRID QUESTION SRO 56 NRC RECORD # WRI 56 ANSWER: C. SYSTEM # 309 K/A 264000 A3.01: 3.0/3.1 LP# HLO-037 OBJ. 3- SROTIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: Tech Specs 3.3.8.1 NEW CLASS Tech Specs TR3.3.8.1 MODIFIED BANK DIFF 4 Tech Specs TR3.3.5.1
U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 57 The p! ant is operating at 70 % power. The Master Level Co itrol System is selected for three (3) element control and Rx Level A. j
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The D004A Condensing pot has ruptured.
Which one of the following statements defines the response of the Feedwater System and Reactor Level?
l A. The Feedwater Level Control System will close the Feed Reg Valves to a lower l position and actual Reactor Level will lower and stabilize at a point just above the scram setpoint.
B. The Feedwater Level Control System will open the Feed Reg Valves to full open and actual Reactor Level will rise to the point to pick up the Main Turbine trip on i
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high level.
C. The Feedwater Level Control System will c?ose the Feed Reg Valves to a lower position and actual Reactor Level will lower to below the Reactor scram setpoint ,
resulting in a Reactor scram.
D. The Feedwater Level Control System will shift the Master Level Controller to manual and lock up the Feed Reg Valves at their present position and actual Reactor level will stabilize at a slightly higher level.
QUESTION SRO 57 NRC RECORD # WRI 57 ANSWER: C. SYSTEM # 107 K/A 259002 K5.01: 3.1/3.1 '
LP# K5.03: 3.1/3.2 ;
OBJ. SROTIER 2 GROUP 1/ ROTIER 2 GROUP 1 j REFERENCE: ARP-H13-P680 03A-C08 NEW CLASS BANK
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MODIFIED I DIFF 3 l
DATE USED: RO SRO BOTH CFR 41.5 j
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 58 l
The plant is operating at 100 % power. Condensate Pump CNM -PIB has tripped on over-current.
The pressure at the suction of the Reactor Feed Pumps has dropped to 200 psig for 12 seconds.
Which one of the following would be the response of the Reactor Feed Water System?
A. The Feed Reg Valves will throttle back to increase suction pressure resulting in a low Reactor Level. )
B. The "A" Reactor Feed Pump will trip and cause Reactor Feed Pump suction pressure rise. l l
C. The "A and B" Reactor Feed Pumps will trip and cause a Reactor scram on low Reactor Level.
D. All three Reactor Feed Pumps will trip causing Reactor Level to lower and result in a Reactor scram.
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QUESTION SRO 58 NRC RECOPD # WRI 58 l ANSWER: B. SYSTEM # 107; K/A 259001 A3.10: 3.4/3.4 104 LP#
OBJ. SRO TIER 2 GROUP 2 / ROTIER 2 GROUP 1 l
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REFERENCE: ARP-II13-P680 03A-A01 NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.4
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 59 The plant is Shutdown in Mode 4.
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The Main Condensers are drained and open for maintenance. l
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RHR "A" is in Shutdown Cooling with Service Water at 82 *F.
I HPCS, RHR "B", and RHR "C" pump motors are tagged out for repairs. ;
I Reactor Recirculation Pump "A" is operating in Slow Speed. Reactor Recirculation Pump "B" is tagged out for seal replacement.
Reactor Engineering has calculated decay heat as 17 x10 Btu /hr at 130 *F.
RHR "A" Pump has tripped due to unknown reasons. i Based on present plant conditions, which one of the following would be the minimum Alternate Shutdown Cooling Methods required to maintain present plant conditions?
(OSP-0041 is available.)
B. MSL Flooding ONLY.
C. CRD and SFC ONLY.
D. ADHR ONLY.
QUESTION SRO 59 NRC RECORD # WRI 59 ANSWER: D. SYSTEM # 204 K/A 295021 AA1.04: 3.7/3.7 LP#
OBJ. SROTIER I GROUP 2 / RO TIER 1 GROUP 3 REFERENCE: AOP-0051 NEW CLASS OSP-0641 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.5/43.5 i
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i U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 l
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SENIOR REACTOR OPERATOR QUESTION 60 i Plant conditions are as follows:
Reactor Power 42 %
Actual Turbine Generator Load 420 MWe l Turbine Generator Load Set 1040 MWe I The At-The-Controls Operator withdraws a control rod which causes Reactor thermal l power to change by + 30 MWth.
Which one of the following would occur based on the changes made?
A. The Turbine Control Valves will open as required to maintain Reactor pressure as Reactor power increases.
R The Turbine Control Valves will close as required to maintain Reactor pressure as Reactor power increases.
C. The Turbine Control Valves will open as required to turn Reactor power by lowering Reactor pressure.
D. The Turbine Control Valves will close as required to turn Reactor power by raising Reactor pressure.
QUESTION SRO 60 NRC RECORD # WRI 60 ANSWER: A. SYSTEM # 110 K/A 241000 K5.04: 3.3/3.3 LP# K5.03: 3.5/3.6 OBJ. SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: EHC Functional Diagram NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.5 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 61 The plant is operating at 100 % power.
A leak in the RWCU pump room caused an isolation of the RWCU System.
G33*MOVF040, RWCU RETURN TO FW failed to close.
Which one of the following actions is REQUIRED to be taken?
(Tech Specs attached, if needed.)
A. The penetration is allowed to remain unisolated if the remainder ofisolation valves in the rest of the RWCU system have isolated. j l
B. Verify another valve in the associated penetration is closed and is also de-activated.
C. The penetration is allowed to be unisolated dtuing present conditions as long as the RWCU pumps have tripped.
D. The plant must shutdown to cold shutdown and shutdown the RWCU system.
QUESTION SRO 61 NRC RECORD # WRI 61 ANSWER: B. SYSTEM # 110 K/A 290001 A2.06: 3.7/4.0 I LP#
OBJ. SROTIER 2 GROUP 1/ RO TIER 2 GROUP 2 REFERENCE: Tech Specs 3.6.1.3 NEW CLASS TR3.6.1.3-1 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.9
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 62 The plant is operating at 100 % power when B21-AOVF028B, an outboard MSIV fails closed due to a rupture of the valve actuator air supply.
Which one of the following describes the response of the reactor?
ASSUME NO OPERATOR ACTION.
A. RPV pressure will increase and stabilize at a higher pressure.
Reactor power will increase and stabilize at a higher power.
RPV water level will decrease and then retum to normal level.
B. RPV pressure will increase and then decrease folowing the scram.
Reactor power will increase and cause a reactor scram on power.
RPV water level will decrease and then stabilize at a lower level.
C. RPV pressure will decrease and stabilize at a lower pressure.
Reactor power will decrease and stabilize at a lower power.
RPV water level will increase and then retum to normal level. ,
D. RPV pressure will d: crease and stabilize at a lower pressure.
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- Reactor power will increase and retum to the original power. j RPV water level will increase and then retum to normal level. !
QUESTION SRO 62 NRC RECORD # WRI 62 ANSWER: B. SYSTEM # 109; 1 K/A 295007 AK1.03: 3.8/3.9 LP# 107 OBJ. SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 REFERENCE: USAR 15.2.4.1.2.2 NEW CLASS MODIFIED BANK DIFF 3 ID 308 i DATE USED: RO SRO BOTH CFR 41.5/41.14 i
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l U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR i QUESTION 63 The plant was operating at 100 % power when a tornado caused a complete loss of Offsite Power.
l Division I, II, and III diesel generators failed to start resulting in a Station Blackout.
! Which one of the following describes the response of the RCIC system?
ASSUME NO OPERATOR ACTION.
A. RCIC steam supply willisolate.
RCIC will be unable to be unisolated until AC power is restored.
B. All AC powered components of RCIC will fail as-is. l RCIC will respond to all initiation signals and inject to the reactor.
C. All RCIC isolations are still available.
RCIC will respond to all initiation signals and inject to the reactor. i D. All AC powered components of RCIC will fail as-is. l l RCIC will ONLY start on manual initiation and alignment signals.
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QUESTION SRO 63 NRC RECORD # WRI 63 ANSWER: B. SYSTEM # 303 K/A 295003 AA1.03: 4.4/4.4 LP# 209 OBJ. SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 2 REFERENCE: SOP-0035 Attachment 3 NEW CLASS MODIFIED BANK DIFF 3 ID 287 l l DATE USED: RO SRO BOTII CFR 41.7 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR
QUESTION 64 j
l The plant is operating at 100 % power when a leak causes reactor water level to lower to j
- 50 inches. i l
Which one of the following will be the affects on the Containment and Containment l
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Cooling System?
ASSUME NO OPERATOR ACTION.
l A. Standby Gas Treatment System and Annulus Mixing will initiate.
Containment Cooling System will operate without cooling water causing Containment temperatures to rise. 1 l
B. Standby Gas Treatment System and Annulus Mixing will initiate. I Containment Cooling System will shutdown all fans causing Containment temperatures to rise.
C. Containment Cooling System will operate without cooling water causing l Containment temperatures to rise. !
Containment Ventilation will align to purge the Containment.
D. Containment Cooling System Mll shutdown all Containment Cooling fans causing Containment temperatures to rise.
Containment Ventilation will align to purge the Containment.
QUESTION SRO 64 NRC RECORD # WRI 64 ANSWER: A. SYSTEM # 403 K/A 295020 AA1.03: 2.9/3.1 LP#
OBJ. SRO TIER I GROUP 2 / RO TIER I Gv .UP 2 REFERENCE: AOP-0003 NEW CLASS MODIFIED BANK DIFF 3 ID 518 DATE USED: RO SRO BOTH CFR 41.9
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U.S. NUCLEAR REGULATORY COMMISSION i WRITTEN EXAMINATION FEBRUARY 1999 l SENIOR REACTOR OPERATOR l I
QUESTION 65 l
The Control Room Supervisor has entered EOP-0003 Radioactive Release Control.
Turbine Building Ventilation has shutdown. !
Why is the Control Room Supervisor directing an operator to restart Turbine Building Ventilation?
A. Provide for the filtration of the Turbine Building atmosphere to prevent the release of any radioactive material from the Turbine Building.
B. Provide for the filtration of the Turbine Building Exhaust to ensure radioactive releases do not exceed General Emergency Levels. l C. Provide for the monitoring of the air released from the Turbine Building instead of unmonitored release to the environment.
D. Provide for the condensing of steam in the Turbine Building which may contain radioactive particulate thus preventing any release.
QUESTION SRO 65 NRC RECORD # WRI 65 ANSWER: C. SYSTEM # 408 K/A 295038 EK2.04: 3.9/4.2 LP#
OBJ. SRO TIER 1 GROUP 1/ RO TIER I GROUP 2 l REFERENCE: EPSTG*0002 NEW CLASS l EOP-0003 RR-1 MODIFIED BANK DIFF 2 ID 311 DATE USED: RO SRO BOTH CFR 41.13/43.4
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WRITTEN EXAMINATION FEBRUARY 1999 l SENIOR REACTOR OPERATOR l
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QUESTION 66 The plant is operating at 100% power.
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! I&C is performing a surveillance on the APRMs and causes an inadvertent Reactor Scram.
Which one of the following describes the response of Reactor Water Level on the Reactor l Scram? (The Scram was NOT caused by Reactor Water Level.)
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A. Level will rise tojust above the High level alarm then drop back to the normal I level setpoint.
B. Level will rise to above level 8 tripping all Reactor Feed Pumps causing level to drop.
C. Level will drop to level 3 causing Setpoint Setdown to take effect and return level to 18 inches.
l D. Level will drop to just below the Low level alarm and then rise back to the normal level setpoint.
QUESTION SRO 66 NRC RECORD # WRI 66 ANSWER: C. SYSTEM # 501 K/A 295006 AK3.01: 3.8/3.9 LP# HLO-060 OBJ. 4 SRO TIER 1 GROUP 1/ RO TIER I GROUP 1 REFERENCE: AOP-0001 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.5/41.14
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR
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QUESTION 67 The plant has scrammed due to an isolation of the MSIVs.
l Suppression Pool Temperature is 95 *F.
j Which one of the following describes the approved method to prevent localized heating of the Suppression Pool?
A. Cycle all SRVs going Main Steam Line to Main Steam Line in order.
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B. Cycle the Low-Low Set SRVs in a specified order.
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l C. Cycle non-Low-Low Set SRVs in any order and start RHR C in Suppression Pool l
Cooling.
l D. Use SRV B21 *F051D to control reactor pressure and start Suppression Pool Cleanup to circulate water.
QUESTION SRO 67 NRC RECORD # WRI 67 ANSWER: B. SYSTEM # 057; K/A 295013 AK1.03: 3.0/3.3 LP# 109 OBJ. SROTIER I GROUP 1/ RO TIER 1 GROUP 2 REFERENCE: Operator Aid II13-P601 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.9/41.10 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SEN!DR REACTOR OPERATOR QUESTION 68 ,
The Suppression Pool is leaking into the crescent area of the Auxiliary Building.
Suppression Pool Level has dropped to 12 ft 6 inches.
Concerning Emergency Depressurization, which one of the following describes the significance of this Suppression Pool Level?
A. A vortex will be formed when an SRV is opened drawing air into the SRV tailpipe causing water hammer.
B. If an SRV is opened while any ECCS pump is drawing a suction from the Suppression Pool the pump will draw in steam.
C. The SRV discharge quencher may not be covered, such that SRV operation may directly pressurize Containment with steam.
D. The horizontal vents are not covered such that SRV operation will admit steam into the Drywell.
QUESTION SRO 68 NRC RECORD # WRI 68 ANSWER: C. SYSTEM # 057; K/A 295030 EK2.08: 3.5/3.8 LP# 109 OBJ. SROTIER I GROUP 1/ RO TIER I GROUP 2 REFERENCE: EPSTG*0002 NEW CLASS EOP-0004 ED-3; AED-2 MODIFIED BANK DIFF 3 ID23 DATE USED: RO SRO BOTH CFR 41.9
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR Ol"ERATOR QUESTION 69 l
l The Unit Operator notices the AUX BLDG FL DRAIN SUMP LEVEL EXTREME HIGH/ LOW ANNUNCIATOR (P870-51 A-G3) is in alann.
The Auxiliary Building SNEO has reported the sump level in the HPCS Pump Room is overflowing onto the floor.
Which one of the following describes the expected equipment operation and procedural requirements?
A. Both sump pumps should be operating and the Control Room Supervisor should be entering EOP-0003.
B. Only one of the sump pumps should be operating and the Auxiliary Building Operator should be locating the source of the leakage.
C. Both sump pumps should be operating and the Control Room Supervisor should evacuate the Auxiliary Building.
D. Only one of the sump pumps should be operating and the Control Room Supervisor should be entering EOP-0003.
QUESTION SRO 69 NRC RECORD # WRI 69 ANSWER: A. SYSTEM # 604 K/A 295036 EK3.04: 3.4/3.8 LP#
OBJ. SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 3 REFERENCE: ARP-P870 51 A-G03 NEW CLASS EOP-0003 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.4 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 70 i
The plant is operating at 100 % power.
Both Offgas Post Treatment Radiation monitors have alarmed on a High-High-High Radiation signal.
Which one of the following describes the effects on the Offgas System and the Main Condenser?
A. Offgas will shift into a bypass mode of operation causing Main Condenser vacuum to be lost.
B. Offgas will isolate only the charcoal adsorbers inlet and outlet valves causing Main Condenser vacuum to be lost.
C. Offgas will continue operation allowing Main Condenser vacuum to remain constant.
D. The Offgas System will isolate causing Main Condenser vacuum to be lost.
QUESTION SRO 70 NRC RECORD # WRI 70 ANSWER: D. SYSTEM # 606 K/A 271000 K3.01: 3.5/3.5 LP#
OBJ. SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2 REFERENCE: ARP-P60122A-A03 NEW CLASS AOP-0005 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.13
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 71 The Main Control Room has been evacuated due to a fire.
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All Immediate Operator Actions have been completed perAOP-0031. )
l Control of the plant has been established at the Remote Shutdown Panel. ,
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Which one of the following systems could have the possibility of overfilling AND over-pressurizing the Reactor?
A. RHR "A" i l
B. RCIC C. HPCS l D. LPCS QUESTION SRO 71 NRC RECORD # WRI 71 ANSWER: C. SYSTEM # 200 K/A 295016 AA1.06: 4.0/4.1 LP#
OBJ. SRO TIER 1 GROUP 1/ ROTIER 1 GROUP 2 REFERENCE: AOP-0031 Caution NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.7
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 72 The plant is operating at 100 % power.
Radwaste is discharging a Recovery Sample Tank.
Circulating Water blowdown is stopped due to a failure of the Circulating Water Blowdown Isolation Valve CWS-MOV104.
The Aux Control Room Operator notifies the Main Control Room of alarm LWS-PNL187 4-A4 BLOWDOWN WATER FLOW LOW has been received.
Which one of the following actions should occur?
A. LWS-AOV257, RCVY SAMPLE DISCH V TO CRCLT WATER BLWDN will auto close, and LWS-AOV258, RCVY SAMPLE DISCHARGE DIVERTING will auto open.
B. The Aux Control Room Operator should manually secure the discharge by closing LWS-AOV257, RCVY SAMPLE DISCH V TO CRCLT WATER BLWDN, and securing the discharge lineup.
C. Continue the discharge, and monitor the Recovery Sample Process Radiation Monitor, and only secure the discharge if the radiation levels reach the High alarm setpoints.
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l D. Continue the discharge, and inform Chemistry to take grab samples, and secure the discharge only if the Radiation levels are above the limits of the Discharge Permit. '
QUESTION SRO 72 NRC RECORD # WRI 72 I ANSWER: B. SYSTEM # 603; K/A 172000 A3.03: 3.1/3.5 l 103 LP#
OBJ. SRO TIER 2 GROUP 2 / ROTIER 2 GROUP 2 REFERENCE: ARP-LWS-PNL187 4-A4 NEW CLASS 4-C4 MODIFIED BANK DIFF 3 SOP-0113 DATE USED: RO SRO BOTH CFR 41.11/41.13/
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I U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 73 The plant is operating at 100 % power.
A fire erupts in the Division 1 Diesel Generator room causing the sprinkler system to initiate.
Fire Water header pressure has dropped to 115 psig.
Which one of the following actions would be expected to occur? !
l A. The Motor Driven Fire Pump will auto start and the Diesel Driven Fire Pump "A" and "B" will start immediately if the Motor Driven Fire Pump fails to start.
B. The Motor Driven Fire Pump will auto start, if header pressure is still at 115 psig afterl5 seconds AND the Motor Driven Fire Pump failed to start, then the Diesel l Driven Fire Pump "A" will start. I C. The Diesel Driven Fire Pump "A" will auto start, if fire water header pressure remains at 115 psig for 10 seconds, whether the Motor Driven Fire Pump starts or NOT.
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D. The Diesel Driven Fire Pump "A" will auto start, if fire water header pressure remains below 140 psig for 10 seconds and the Motor Driven Fire Pump is running.
QUESTION SRO 73 NRC TECORD # WRI 73 ANSWER: D. SYSTEM # 251 K/A 286000 K4.03: 3.3/3.4 LP# l OBJ. SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2 REFERENCE: SOP-0037 NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.4 i
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 74 The plant is operating at 100 % power.
The following annunciators have been received in the Main Control Room:
H13-P870 55A-B01, TURB CMPNT CLG WTR SYS SURGE TK LOW LEVEL H13-P870 55A-B02, TURB CMPNT CLG WATER SYSTEM LOW PRESSURE H13-P870 55A-E01, TURB CMPNT CLG WATER PUMP BRKR AUTO TRIP H13-P870 55A-E02, TURB CMPNT CLG WATER PUMP LOW DISCH PRESS H13-P870 55A-G02, TURB CMPNT CLG WATER PUMP IB OVERLOAD ,
The Turbine Building SNEO reports that there is a large leak on the common discharge piping of the TPCCW pumps which is spraying the IB and IC TPCCW Pumps. The SNEO reports the leak is unisolable.
When checked, TPCCW Surge Tank level is offscale low.
Which one of the following actions is required to take place under these conditions?
A. Verify TPCCW Pump IC has started and MWS-AOV132, TPCCW SURGE TK MAKE-UP valve is full open to add water to the TPCCW Surge Tank.
B. Lower power to 45 Mlbm/hr Core Flow, verify TPCCW Pump 1C has started, and MWS-AOV132, TPCCW SURGE TK MAKE-UP valve is full open to add water to the TPCCW Surge Tank.
C. Manually scram the reactor, shutdown the equipment which is being supplied by TPCCW, and secure the remaining TPCCW pumps and isolate the leak.
D. Verify TPCCW Pump IC has started, and MWS-AOV132, TPCCW SURGE TK
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MAKE-UP valve is full open, and shutdown equipment as necessary on TPCCW.
QUESTION SRO 74 NRC RECORD # WRI 74 ANSWER: C. SYSTEM # 116 K/A 400000 K6.04: 3.0/3.1; K4.01: 3.3/3.9 LP# A2.01: 3.3/3.4; A2.02: 2.8/3.0 OBJ. SROTIER 2 GROUP 2 / ROTIER 2 GROUP 2 REFERENCE: AOP-0012 NEW CLASS ARP P870 55A-B01; CO2 MODIFIED BANK
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DIFF 2 55A-E01; E02; G02 l DATE USED: RO SRO BOTH CFR 41.4
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 75 The plant is operating at 40 % power.
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l H13-P680 07A-B01, ROD CONTROL AND INFO SYS INOPERATIVE l
I&C has determined the power supply to the Division 2 RACS cabinet has failed. I Which one of the following describes the At-The-Controls Operators ability to move control rods with present plant conditions?
l A. Control rods may be inserted or withdrawn normally.
B. Control rods may be inserted normally however, are unable to be withdrawn.
C. Control rods may be inserted or withdrawn by bypassing the control rod in the Division 2 RACS Cabinet.
D. Control rods are unable to be inserted or withdrawn normally, the only control rod movement is by scram.
' QUESTION SRO 75 NRC RECORD # WRI 75 ANSWER: D. SYSTEM # 500 K/A 262002 K3.07: 2.6/2.8 LP#
OBJ. SROTIER 2 GROUP 2 / ROTIER 2 GROUP 2 REFERENCE: SOP-0071 NEW CLASS ARP P680 07A-B01 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.4
U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 76 The plant is in a LOCA condition such that Containment parameters are elevated.
Which one of the following describes plant conditions and the basis requiring Emergency Depressurization of the RPV7 A. Containment pressure is 2.0 psig and Containment Temperature of 187 *F because Containment conditions with a LOCA could result in a failure of the i Containment.
B. Containment pressure is 3.0 psig and Containment Temperature is 175 F because the saturation temperature of the Suppression Pool cotdd be exceeded during a blowdown from rated RPV pressure.
C. Containment pressure is 3.0 psig and Containment Temperature of 187 F because ;
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the saturation temperature of the Suppression Pool could be exceeded during a l blowdown from sted RPV pressure. l D. Containment pressure is 2.0 psig and Containment Temperature of 175 F because Containment conditions with a LOCA could result in a failure of the Containment. l
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l QUESTION SRO 76 NRC RECORD # WRI 101 i ANSWER: A. SYSTEM # 057 K/A 215003 AKl.01: 4.1 l LP# 2.4.18: 3.6
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OBJ. SROTIER I GROUP 2 / ROTIER GROUP REFERENCE: EOP-0002 step CT-6 NEW CLASS EPSTG*0002 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.9 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 i
SENIOR REACTOR OPERATOR QUESTION 77 The plant haa evacuated the Main Control Room and established control at the Remote Shutdown Panels.
Which one of the following methods is NOT allowed to be used for control of Reactor Cooldown while attempting to reach Cold Shutdown?
A. Main Steam Bypass Valves B. RCIC C. Safety Relief Valves I
D. RHR "A" or "B" l QUESTION SRO 77 NRC RECORD # WRI 102 ANSWER: A. SYSTEM # 200 K/A 295016 AK2.06: 3.5 LP#
OBJ. SRO TIER 1 GROUP 1/ ROTIER GROUP q REFERENCE: AOP-0031 NEW CLASS :
MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.9
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US NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 78 The Operations Shift Superintendent has declared a General Emergency due to EIP-2-001 Attachment 5 EAL 1.2. Field monitoring teams and Chemistry have reported a 5450 mrem Thyroid CDE dose commitment at five (5) miles from the plant.
Which one of the following is the Protective A tion Recommendation to be issued to the State and Local Agencies?
Evacuate 2 mile radius of the plant, and:
A. evacuate the 5 mile down wind sectors of the plant, and shelter the remainder of the 10 mile radius, and evacuate schoolr.. institutions and recreation areas 5 mile radius.
. B. evacuate the 10 mile down wind sectors of the plant, and shelter the remainder of j the 10 mile radius, and evacuate schools, institutions and recreation areas 10 mile l radius.
C. evacuate the 5 mile radius of the plant, and evacuate the 10 mile down wind I sectors, and shelter 10 mile radius, evacuatc schools, institutions and recreation areas 10 mile radius.
l D. evacuate the 5 mile radius of the plant, and shelter the 10 mile radius, evacuate l
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schools, institutions and recreation areas 5 mile radius.
QUESTION SRO 78 NRC RECORD # WRI 103 ANSWER: C. SYSTEM # K/A 295017 AK1.02: 4.3 EPP PARS 2.4.44: 4.0 LP#
OBJ. SRO TIER 1 GROUP 1/ ROTIER GROUP REFERENCE: EIP-2-001 NEW CLASS EIP-2-007 Att.1 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.10/41.12 43.5 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999
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SENIOR REACTOR OPERATOR i
QUESTION 79
! The reactor is operating at 100% power and you have three (3) Senior Reactor Operators
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(Operations Shift Superintendent, Control Room Supervisor /STA, Control Room Supervisor) and three (3) licensed Reactor Operators (Nuclear Control Operators) on shift in the Control Room. Four hours after your crew has relieved the shift, two (2) of the i Reactor Operators and the Operations Shift Superintendent fall ill with apparent food j
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poisoning. The EMTs recommend sending the personnel to the hospital. )
Which one of the following actions is allowed for this situation?
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A. You cannot allow the personnel to leave the site because you will not meet Technical Specification manning requirements. l
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l B. You may let the two (2) Reactor Operators leave, however, the Operations Shift
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Superintendent cannot leave.
C. Allow the transport of all of the sick personnel to the hospital only after the unit has been placed in Mode 3 when staffing requirements are less. l l D. Allow the transport of all of the sick personnel to the hospital and immediately call in personnel to meet minimum staffing requirements within two hours.
QUESTION SRO 79 NRC RECORD # WRI 104 ANSWER: D. SYSTEM # Shift K/A Gentrics 2.1.4: 3.4 Staffing LIM OBJ. SRO TIER 3 GROUP / RO TIER GROUP REFERENCE: Tech Specs 5.2.2c NEW CLASS Table 5.2.2-1 MODIFIED BANK DIFF 3 NRC 2 DATE USED: RO SRO BOTH CFR 41.10/43.2 43.5 l
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U.S. NUCLEAR REGULATORY COMMISSION !
WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 80 The plant is raising power from 80 % to 100 % power. The "A" Recirculation Flow !
Control Valve hydraulic actuator sticks resulting in the valve inadvertently opening from l
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60 % to 100 % valve position.
Reactor power stabilizes at 98 % power.
Total Core Flow 84.4 Mlbm/hr Fraction of Core Boiling Boundary 0.62 LHGR 0.974 MAPLHGR 0.982 MCPR 1.31 Core Exposure 1950 mwd /ST Which one of the following identifies the Thermal Operating Limits which have been violated?
Power has NOT been changed since the transients.
CONSIDER Flow Based Limits ONLY.
A. APLHGR B. FCBB C. LHGR D. MCPR l
QUESTION SMO 80 NRC RECORD # WRI 105 ANSWER: D. SYSTEM # K/A Generics 2.2.25: 3.7 Thermal Limits
.LP#
OBJ. SRO TIER 3 GROUP / RO TIER GROUP REFERENCE: Tech Specs 3.2.1; 3.2.2; NEW CLASS 3.2.3 MODIFIED BANK DIFF 4 COLR graphs NRC 3 DATE USED: RO SRO BOTH CFR 43.1/43.2 l
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UoS. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR ,
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QUESTION 81 The plant is operating at 40% power. Plant Chemistry reported to the Main Co'ntrol Room the following chemistry parameters:
Reactor pH 6.5 Reactor Water conductivity 0.95 mho/cm Reactor Water chlorides 150 ppb Which one of the following describes the required actions for these plant conditions?
A. Continue power operations and continue monitorire plant parameters.
B. Restore to within limits within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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C. Restore to within limits within 72 Irurs or be in mode 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.',
D. Restore to within limits within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and mode 4 I
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in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
QUESTION SRO 81 NRC RECORD # WRI 106 ANSWER: A. SYSTEM # K/A Generics 2.4.5: 3.6; 2.4.11: 3.6 Chemistry Limits 2.2.22: 4.1; 2.1.34: 2.9 LP#-
OBJ. SRO TIER 3 GROUP / ROTIER GROUP REFERENCE: Tech Specs TR 3.4.13 NEW CLASS Table 3.4.13-1 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 43.2
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 82 The plant is shutdown following an ATWS in which 6 control rods did not fully insert.
Implementation of EOP Enclosures was required.
The OSS declared a Site Area Emergency.
The 6 control rods have been inserted and the CRS has exited the EOPs and entered GOP-0002, Plant Shutdown.
Reactor level is at +30 inches and Reactor pressure is 600 psig.
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No Radioactive release has occurred. l
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Which one of the following best describes the conditions to *erminate the Site Area Emergency? l
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A. Terminate the Site Area Emergency and notify the State and Local Agencies and the NRC the event is terminated.
B. The plant must be in mode 4 and have the concurrence of the State and Local Agencies and the NRC to terminate the Site Area Emergency.
C. If the State and Local Agencies and the NRC agree the Site Area Emergency may be immediately terminated with NO further actions.
D. The Site Area Emergency will continue until a root cause has been determined and notify the State and Local Agencies that the emergency is terminated.
QUESTION SRO 82 NRC RECORD # WRI 107 ANSWER: B. SYSTEM # EIP K/A Generics 2.4.29: 4.0 LP# 2.4.37: 3.5 OBJ. SROTIER 3 GROUP / RO TIER GROUP REFERENCE: EIP-2-002 Att 3 NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.10/43.5 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 83 Given the following conditions: ,
Reactor Power 45 %
Reactor Level -100 inches Reactor Pressure 850 psig Suppression Pool Temperature 138'F '
Suppression Pool Level 20 feet 5 inches 2 SRVs are OPEN l
Which one of the following describes the required actions to be taken given the above I conditions?
A. Immediately commence an Emergency Depressurization because limits in the l Containment have been exceeded based on Suppression Pool Temperature. l B. Close the two (2) SRVs and increase the Reactor Pressure band to a top end of 1050 psig, to reduce the amount of heat entering the Suppression Pool.
C. Lower Reactor Pressure using cooldown rates that may exceed 100 'F/hr, to avoid exceeding the heat capacity temperature limit of the Suppression Pool.
D. Conditions at present are acceptable, however all pumps taking a suction from the Suppression Pool should be secured.
QUESTION SRO 83 NRC RECORD # WRI 108
' ANSWER: C. SYSTEM # EOP K/A Generics 2.4.20: 4.0 LP#
OBJ. SROTIER 3 GROUP / ROTIER GROUP REFERENCE: EOP-0001A NEW CLASS RPA-7 MODIFIED BANK DIFF 3 Caution 6 DATE USED: RO SRO BOTH CFR 41.10/43.5 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR l
QUESTION 84 A Tanker Truck hauling 97 % Sulfuric Acid lost control at the Circulating Water Structure and has turned over.
Sulfuric Acid is spilling onto the ground and diesel fuel has also tlled. The fire brigade and HazMat Team from the plant are on the scene to contain . pill. WAFB television station is live on the scene to make a 5 p.m. report.
I Which one of the following describes the reportability of this event to the NRC7 A. This is not a reportable event since it did not occur inside the Protected Area.
B. A report must be made within one (1) hour.
C. A report must be made within four (4) hours.
D. A report must be made within twenty-four (24) hours.
QUESTION SRO 84 NRC RECORD # WRI 109 ANSWER: C. SYSTEM # K/A Generics 2.4.30: 3.6 LP# Notifications OBJ. SRO TIER 3 GROUP / RO TIER GROUP REFERENCE: RBNP-004 NEW CLASS ,
MODIFIED BANK DIFF 4 NRC 2 DATE USED: RO SRO BOTH CFR 41.12/43.4 ,
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 85 The plant is operating during an emergency. The shift determines that conditions are such that there is no appropriate action to be taken which would be in compliance with the station operating license.
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Whose permission at a MINIMUM is required to take the required actions to maintain the plant in a safe condition and when must the NRC be notified of such actions?
A. The NRC Senior Resident Inspector; notify the NRC within one (1) hour.
B. General Manager- Operations; notify the NRC within thirty (30) days in a written report.
C. Control Room Supervisor; notify the NRC within one (1) hour.
D. Nuclear Control Operator; notify the NRC within thirty (30) days in a written report.
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QUESTION SRO 85 NRC RECORD # WRI 110 i ANSWER: C. SYSTEM # K/A Generics 2.4.40: 4.0 LP# 10CFR50.54 OBJ. SRO TIER 3 GROUP / RO TIER GROUP REFERENCE: RBNP-004 NEW CLASS 10CFR50.54(x) MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 43.3/43.5 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY ~1999 SENIOR REACTOR OPERATOR QUESTION . 86 The plant is performing the In-Service Leak Test (Vessel Hydro) on the reactor following refueling operations. A miscommunication results in a significant reactor pressure increase. Pressure as read on the Control Room Wide Range Pressure indication on 1H13*P680 is pegged upscale.
The Post Accident Pressure recorders on 1H13*P601 indicate pressure has reached 1350 psig.
Which one of the following is a correct statement with regard to the RBS Safety Limit for Reactor Pressure?
A. Reactor Pressure was greater than the Safety Limit of 1190 psig, as read on the 1H13*P680 Wide Range Instrument for Tech Specs.
B. Reactor Pressure was greater than the Safety Limit of 1325 psig as read on the Post Accident Pressure indication (on lH13*P601), which comes from the Water Level instrument's reference legs.
C. Reactor Pressure was greater than the Safety Limit of 1375 psig as read on the Post Accident Pressure indication (on 1H13*P601), which comes from the Bottom Head pressure tap.
D. Reactor Pressure was within the Safety Limit of 1550 psig.
QUESTION SRO 86 NRC RECORD # WRI 111 ANSWER: B. SYSTEM # Safety K/A Generics 2.1.32: 3.8 LP# Limits OBJ. SROTIER 3 GROUP / RO TIER GROUP REFERENCE: Tech Spec Bases 2.1.2 NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH rFR 43.2 l
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! U.S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 87
- The plant is starting up, reactor power is currently at 2 %.
A Drywell Unit Cooler needs to be isolated and tagged out for replacement of the Normal Service Water supply valves. The valves to be isolated require independent verification upon rc:toration.
Which one of the following methods is allowed to be used to reduce the exposure to operations personnel?
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A. RP will install 16 lead blankets around the area of the Unit Cooler to reduce radiation levels at thejob site.
B. Flush the Service Water piping to the floor drains to reduce the radiation levels at the job site.
C. Radiation protection personnel will perform verification per independent verification.
D. The Operations Shift Superintendent waives the verification requirements for the valves.
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QUESTION SRO 87 NRC RECO.RD # WRI 112 ANSWER: D. SYSTEM # Rad K/A Generics 2.3.10: 3.3 LP# Protection OBJ. SRO TIER 3 GROUP / RO TIER GROUP REFERENCE: ADM-0076 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 43.4
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O.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR
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QUESTION 88 A Loss of Coolant Accident has occurred. The core has been uncovered and reactor
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water level can not be determined.
The following plant parameters exist:
Reactor Pressure 150 psig Containment Pressure + 8.5 psig !
Drywell Pressure + 16.5 psig ;
Drywell Hyd ' gen Concentration 4.5 % '
Containmen' drogen Concentration 5.5 %
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Which one of the following describes the methods to be used to control Hydrogen Concentrations in the Containment and Drywell?
A. Hydrogen Recombiners, Igniters, and Mir.ing Systems
. B. Hydrogen Igniters, and Normal CTMT Vent and Purge C. Normal CTMT Vent and Purge i D. Hydrogen Recombiners, and Igniters I
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l QUESTION SRO 88 NRC RECORD # WRI 113 ANSWER: D. SYSTEM # 254 K/A 500000 EA1.03: 3.2 l LP# 2.4.14: 3.9 Olu. SROTIER I GROUP 1/ RO TIER GROUP
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' REFERENCE: EOP-0002 NEW CLASS Ilydrogen Control MODIFIED BANK DIFF 3 NRC 2 DATE USED: RO SRO BOTH CFR 41.10/43.5 I
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UcS. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 89 The plant is operating at 100% power.
Maintenance has been performed on the RCIC turbine. Mechanical Maintenance has requested that RCIC be operated for an hour before any test data is taken to allow RCIC to fully heat up.
Which one of the following describes the requirements to allow prolonged operation of RCIC with regard to Suppression Pool operation?
l A. Monitor temperature and level of the Suppression Pool every 30 minutes and I
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verify EOP entry conditions.
B. Operate a loop of RHR Suppression Pool Cooling and monitor Suppression Pool Temperature every 5 minutes.
C. Operate Suppression Pool Cleanup to ensure mixing of the Suppression Pool and monitor Suppression Pool Te nperature.
D. RCIC will provide adequate Suppression Pool Mixing to dissipate the heat to the Containment atmosphere.
QUESTION SRO 89 NRC RECORD # WRI 114 ANSWER: B. SYSTEM # 209: K/A 295013 AK3.01: 3.8 057; 204 2.1.11: 3.8 2.4.21: 4.3 LP# 2.4.4: 4.3 OBJ. SROTIER I GROUP 1/ RO TIER GROUP REFERENCE: Tech Specs 3.6.2.1 NEW CLASS SR3.6.2.1.1 MODIFIED BANK DIFF 3 SOP-0035 DATE USED: RO SRO BOTH CFR 41.10/43.5 i
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' WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 90 i
The plant is operating at 100% power.
Standby Gas Treatment Filter Train A is being operated for surveillance testing.
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The Auxiliary Building Operator reports smoke coming from the "A" Standby Gas l Treatment Filter Train and the filter train case is glowing red. l Which one of the following describes the method to combat a fire in the Standby Gas Treatment Filter Train? l A. The Fire Protection System will initiate the automatic deluge system and fill the filter train with water.
B. The Fire Protectica System Deluge Valve will automatically open, however, the isolation valves must be opened manually to admit water to the filter train.
i C. The Fire Protection System Deluge Valve will have to be manually initiated via l the pull station to admit water to the filter train.
D. The Fire Protection System at the filter train must be manually valved into the deluge system to admit water to the filter train.
l QUESTION SRO 90 NRC RECORD # WRI 115 ANSWER: D. SYSTEM # 251 K/A 600000 AK1.02: 3.1 LP#
OBJ. SROTIER I GROUP 2 / RO TIER GROUP REFERENCE: SOP-0037 NEW CLASS MODIFIED BANK DIFF 3 NRC 2 DATE USED: RO SRO BOTH CFR 41.10/43.5
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 91 Which one of ths following describes the function of the Control Room HVAC System?
A. Maintains a positive pressure in the Main Control Room while maintaining a l habitable environment during all possible -lant conditions.
B. Maintains a negative pressure in the Main Control Room while maintaining a habitable environment during all possible plant conditions.
C. Maintains a positive pressuie in the Main Control Room and totally isolate the Control Room envelope in the event of accident conditions.
D. Maintains a negative pressure in the Main Control Room and totally isolate the l
Control Room envelope in the event of accident conditions. !
QUESTION SRO 91 NRC RECORD # WRI 116 l ANSWER: A. SYSTEM # 402 K/A Generics 2.1.27: 2.9 LP# HLO-049 290003 A1.04: 2.8 A2.01: 3.2 OBJ. 1 SRO TIER 3 GROUP / RO TIER GROUP REFERENCE: Tech Spec Bases 3.7.2 NEW CLASS 3.7.3 MODIFIED BANK DIFF 3 NRC 2 DATE USED: RO SRO BOTH CFR 41.4 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 t SENIOR REACTOR OPERATOR QUESTION 92 l-The plant is in a refueling outage performing core alterations.
The Fuel Movement Supervisor, discovered that a step involving movement of a fuel l bundle out of the core. had been missed.
The Fuel Movement Supervisor wants to change the Special Nuclear Movement (SNM) ,
I Tracking Sheets to correct the error and get the fuel in the proper locations for future moves.
WhMn 2.ne of the following identifies WHO at a MINIMUM must approve the changes to the " OFFICIAL COPY" of the SNM Tracking Sheet? j A. Fuel Movement Supervisor and Reactor Engineer Superintendent only.
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B. Operations Shin Superintendent and Fuel Movement Supervisor only.
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C. Refueling Floor Senior Reactor Operator only.
D. Fuel Movement Supervisor and Refueling Floor Senior Reactor Operator only.
QUESTION SRO 92 NRC RECORD # WRI 117 ANSWER: D. SYSTEM # K/A Generics 2.2.28: 3.5 LP# Refueling OBJ. SROTIER 3 GROUP / RO TIER GROUP REFERENCE: REP-0029 NEW CLASS MODIFIED BANK DIFF 3 NRC 2 DATE USED: RO SRO BOTH CFR 41.10/43.7 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 l SENIOR REACTOR OPERATOR l
QUESTION 93
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The plant is at 90 % power, on the way to 100% power, when a loss of Feedwater Heating transient occurs.
The STA runs a Thermal Limits calculation and determines that MCPR is Ll1.
Which one of the following describes the actions to be taken for this condition?
A Maintain current power level and monitor feedwater temperature and reactor thermal power.
B. Reduce power by 10 % power and reevaluate the thermal limits for continued power operations.
C. Reduce power to within thermal limits and insert all insertable control rods.
D. Continue power ascension to 100 % power and notify Reactor Engineering.
QUESTION SRO 93 NRC RECORD # WRI 118 ANSWER: C. SYSTEM # Safety K/A Generics 2.2.22: 4.1 i Llw Limits ;
OBJ. SROTIER 3 GROUP / RO TIER GROUP REFERENCE: Tech Specs 2.1.1.2 NEW CLASS 2.2.2 MODIFIED BANK DIFF 4 NRC 3 DATE USED: RO SRO BOTH CFR. 41.10/43.2
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 94 An annunciator on H13*P601 for RCIC has been intermittently alarming.
I&C has determined the alarm transmitter will not work in the RCIC Room environment and recommends a new style of transmitter be installed to clear the problem.
The Unit Operator has requested something be done about the nuisance annunciator.
Which one of the following describes the actions that can be taken to remove the nuisance annunciator?
A Write a work order to wire the contacts of the transmitter closed and remove the bulbs from the annunciator window until 1&C determines a repair.
B. Write a CR to determine a course of action to replace the equipment and initiate a Temporary Alteration to disable the annunciator.
C. Log the annunciator in the Annunciator Tracking Log and place two (2) red stripes across the window.
D. Pull the bulbs from the annunciator window and place two (2) red stripes across the window.
QUESTION SRO 94 NRC RECORD # WRI 119 ANSWER: H. SYSTEM # K/A Generics 2.4.33: 2.8 LP# Annunciator OBJ. SRO TIER 3 GROUP / RO TIER GROUP REFERENCE: OSP-0015 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTII CFR 41.10/43.5 c
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR l
QUESTION 95 l
Which one of the following describes maintenance that would be considered a trouble l shooting evolution per ADM-0082 CONTROL OF TROUBLESHOOTING? )
A The Hydrogen Water Chemistry modification is being installed on the Condensate system. The engineer finds a solenoid valve installed per the modification fails to I operate, I&C is called to investigate.
l B. The System Engineer has requested a Visacorder be installed in the Control Room to document the thermal performance cf the Feedwater Heaters to allow verification of thermal efficiency.
C. The Control Room is attempting to start a Reactor Recirculation Pump with the plant in Cold Shutdown and jumpers are needed to bypass the Thermal Shock Interlocks.
D, The annunciator for the RCIC Turbine Exhaust Drain Trap level high is intermittently alarming. Maintenance and the System Engineer have been called i for assistance.
QUESTION SRO 95 NRC RECORD # WRI 120 ANSWER: D. SYSTEM # K/A Generics 2.2.20: 3.3 LP# Maintenance OBJ. SROTIER 3 GROUP / ROTIER GROUP REFERENCE: ADM-0082 NEW CLASS MODIFIED BANK DIFF 3.
NRC 2 DATE USED: RO SRO BOTH CFR 41.10/43.5
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U.S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR l
l QUESTION 96 A Loss of Coolant Accident has caused Containment Hydrogen Concentration to reach 2%.
Both trains of Standby Gas Treatment will NOT operate EOP 2 PRIMARY CONTAINMENT CONTROL has directed that Containment Purge be placed in service.
Chemistry has notified the Control Room that Containment Radiation Levels are above I
the threshold that if released would exceed Offsite Dose Calculation Manual (ODCM)
levels.
Which one of the following describes the ability to purge the Containment?
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A Purge the Containment irrespective of Offsite Radiation Release. ,
l B. Start Hydrogen Igniters, Hydrogen Recombiners, and isolate Containment Purge System.
C. Purge Containment through the SBGT Filter Train using the Containment Purge Fans as a motive force.
D. Vent the Containment to the Annulus and start the Annulus Mixing System.
QUESTION SRO 96 NRC RECORD # WRI 121 ANSWER: B. SYSTEM # K/A Generics 2.3.9: 3.4 LP# Rad Con OBJ. SROTIER 3 GROUP / RO TIER GROUP REFERENCE: EOP-0002 HC-7 NEW CLASS IIC-8 MODIFIED BANK DIFF 3 Tech Spec Bases 3.6.4.3 DATE USED: RO SRO BOTH CFR 43.2
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERA *f0R QUESTION 97 The plant is operating at 100 % power with a severe thunderstorm in progress. l The National Weather Service has issued a Tornado Watch for West Feliciana Parish.
The Operating Shift has Minimum Shift Staffing per Technical Specifications.
i A fire breaks out in the Standby Cooling Tower Switchgear Room A, and all operators !
are dispatched to support extinguishing the fire. An operator is stationed at the Division I and 11 Diesel Generators per Standing Order # 154.
Which one of the following describes actions to be taken by the Operations Shift l Superintendent with regard to using the Diesel Operator to fight the fire?
A The OSS may revise the Standing Order and dispatch the operator to assist with the fire.
B. There is NO deviation allowed from the Standing Order.
C. The OSS may deviate from the Standing Order, provided he notify the General Manager, Plant Operations as soon as possible.
D. After notification of the NRC Operations Center, the OSS may deviate from the Standing Order.
QUESTION SRO 97 NRC RECORD # WRI 122 ANSWER: A. SYSTEM # K/A Generics 2.1.15: 3.0 LIM Standing Orders OBJ, SROTIER 3 GROUP / RO TIER GROUP REFERENCE: NEW CLASS ADM-0022 MODIFIED BANK DIFF 3 AOP-0029 Standing Order 154 DATE USED: OSP-0003 RO SRO BOTH CFR 41.10/43.5
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 98 An accident has occurred and radioactive fission products have been detected in the Auxiliary Building, Fuel Building, Containment, Drywell, and the Annulus.
Chemistry has performed and verified an Offsite Dose Calculation using Plant Ventilation Radiation Monitor Data and reported the following site boundary calculations:
25 mrem /hr DDE 1.lE -9 pCi/cc I-131 Offsite Monitoring Teams, downwind of the plant have reported and verified the l following readings at the site boundary: )
i 60 mrem /hr DDE 1.95E -7 Ci/cc I-131 1
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Which one of the following best explains the higher readings reported by the Offsite Monitoring Teams ?
A The Offsite Monitoring Teams are near the center of the plume.
B. The wind speed has increased causing higher reading at the site boundary.
C. The Stability Class has changed, it hasbecome more unstable.
D. An unmonitored release is in progress.
QUESTION SRO 98 NRC RECORD # WRI 123 ANSWER: D. SYSTEM # K/A 295038 EA2.04: 4.5 LP#- Offsite Release OBJ. SROTIER 1 GROUP 1/ ROTIER GROUP REFERENCE: ElP-2-001 NEW CLASS MODIFIED BANK l DIFF 3 l DATE USED: RO SRO BOTH CFR 41.11/43.4 i
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l U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 99 The plant has experienced a Loss of Coolant Accident due to a complete break of the Recirculation System piping.
Which one of the following describes the initial response, of Drywell and Containment Pressure?
A. Drywell pressure will rise to a maximum value clearing all Drywell to Containment Vents releasing steam directly into the Containment pressurizing Containment to a maximum value.
B. Drywell pressure will rise to a maximum value clearing all Drywell to Contaimnent Vents causing a rise in Containment Pressure followed by a lowering of Drywell pressure and recovering of the Drywell vents.
C. Drywell pressure will rise to greater than the ECCS and ADS initiation setpoints causing ECCS and ADS depressurization of the reactor to the Suppression Pool, resulting in a slight rise of Containment Pressure.
D. Drywell pressure will rise to greater than the ECCS initiation setpoints causing ECCS injection and collapse of the steam bubble, removing the driving head of Reactor pressure, resulting in a turn of Drywell pressure and a slight rise in Containment Pressure.
QUESTION SRO 99 NRC RECORD # WRI 124 ANSWER: B. SYSTEM # 057 K/A 295024 EA2.09: 4.1 EA2.01: 4.4 ,
LP# HLO-013 EA2.03: 3.8 OnJ. 3d SRO TIER I GROUP 1 / ROTIER GROUP REFERENCE: USAR 6.2.1.1.3.1.5.2 NEW CLASS Table 6.2-7 MODIFIED BANK DIFF 3 Table 6.2-11 DATE USED: RO SRO BOTH CFR 41.9 l
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U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 1999 SENIOR REACTOR OPERATOR QUESTION 100 Given the following condition:
Reactor startup in progress Mode 2 Reactor power 2 %
Pressure set 250 psig Reactor pressure 200 psig (steady)
Turbine Shell warming is in progress.
Turbine 1st stage pressure 10 psig i The Unit Operator closes all the steam drains. Determine the response of reactor pressure, and the reactor / turbine pressure regulating system.
A. Reactor pressure remaining constant at 200 psig, and turbine warming rate and 1st stage pressure will automatically increase. )
B. Reactor pressure remaining constant at 200 psig. and turbine bypass valves will automatically open.
C. Reactor pressure will rise until vessel heat loss is 2% (~920 psig.), and turbine warming rate and 1st stage pressure will automatically hold constant.
D. Reactor pressure will increase to approximately 250 psig, and turbine bypass valves will automatically open.
QUESTION SRO-100 NRC RECORD # WRI-125 ANSWER: D. SYSTEM # 509 K/A 295007 AK2.01: 3.7 LP#
OBJ. SROTIER 1 GROUP 1/ RO TIER GROUP REFERENCE: EllC Functional Diagram (VE"V CLASS MODIFIED BANK DIFF 4 NRC'2 DATE USED: RO SRO BOTH CFR 41.5 l
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