IR 05000266/1999301

NRC Operator Licensing Exam Repts 50-266/99-301OL & 50-301/99-301OL (Including Graded Tests) for Tests Administered on 990726-0802

ML20211Q665
Person / Time
Site:	Point Beach
Issue date:	09/08/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20211Q649	List: IR 05000266/1999301 ML20211Q645
References
50-266-99-301OL, 50-301-99-301OL, NUDOCS 9909150115
Download: ML20211Q665 (150)

Text

.

.

U.S. NUCLEAR REGULATORY COMMISSION REGION lli Docket Nos: 50-266;50-301 License Nos: DPR-24; DPR-27

- Report No: 50-266/99301(OL); 50-301/99301(OL)

Licensee: Wisconsin Electric Power Company Facility: Point Beach Nuclear Plant, Units 1 and 2 Location: 6610 Nuclear Road Two Rivers, WI 54241 Dates: July 26 - August 2,1999 l

Examiners: D. McNeil, Chief Examiner M. Bielby, Rlll Examiner B. Hughes, HOLB Examiner D. Pelton, Rlli Examiner (in training)

D. Smith, Rill Examiner (in training)

Approved by: David E. Hills, Chief, Operations Branch Division of Reactor Safety

/

9909150115 990908 PDR V ADOCK 05000266 pg

.

.

EXECUTIVE SUMMARY Point Beach Nuclear Power Plant NRC Examination Report 50-266/99301(OL); 50-301/99301(OL)

A Nuclear Regulatory Commission developed initial operator licensing examination was administered to four Reactor Operator and five Senior Reactor Operator license applicant The examiners reviewed several station administrative and operating procedures while developing the examinatio Results:

All applicants passed all portions of the examinatio Operations:

Station personnel recently revised many of the station's operating and administrative procedures. Deficiencies in two procedures that had not been revised confused several applicants during administration of the operating job performance measure (Section O3.1)

The facility training staff was well prepared to support the examination process and provided high quality reviews of the written examinations and operating test (Section 05.2)

The applicants were well prepared for the operating test and written examination. In general, they displayed good operating and communicating practices during the operating test. The small number of questions missed by more than 50% of the applicants indicated a comprehensive training program that met the needs of the applicants. (Section 05.3 & Section 05.4)

l

l l m

.

l

.

Reoort Details l 1. Operations l

03 Operations Procedures and Documentation 1 .

0 General Comments Scooe (71707)

The examiners reviewed portions of selected administrative and operating procedures

-

during the initial license examination development using Inspection Procedure 7170 )

l See page 9 for a partial list of procedures reviewed.' i

' Observations and Findinas The NRC exam developers identified several procedures that were not well organized and contained some confusing instructions, notes, precautions or caution ,

Independent of these observations and prior to the examination validation week, station l personnel revised many of the station's operating and administrative procedures. The revised proceduras were an improvement over the procedures provided for examination developmen Several applicants became confused while trying to execute two Job Performance ,

,

Measures (JPMs) that used unrevised procedures. One procedure, Emergency j Contingency Action (ECA)-0.0," Loss of All AC Power," Revision 23, Attachment B,

'

l

! contained in-plant steps that the applicants believed were to be executed in the control ( room. The applicants were unable to complete the assigned task because the 1 procedure provided weak guidance and the applicants lacked the necessary system l i

'

knowledge required to compensate for the poorly worded procedure. A second j procedure, Abnormal Operating Procedure (AOP) 10A, " Safe Shutdown - Local Control," i contained incorrect valve designators for two auxiliary feedwater valves (AOP 10A, l Attachment C, Step 7A, Result Not Obtained (RNO)). The applicants were able to complete the JPM because of their knowledge of the system and the final, desired valve lineup. The examiners referred these procedure weaknesses to the station's training department for possible revisio Conclusions Station personnel recently revised many of the station's operating and administrative procedures. Deficiencies in two procedures that had not been revised confused several'

applicants during administration of the operating JPMs.

I

3

_

.

.

.

05 Operator Training and Qualification )

05.1 General Comments Nuclear Regulator Commission examiners developed and administered operator initial .

license examinations at the Point Beach Nuclear Power Plant to four Reactor Operator (RO) and five Senior Reactor Operator (SRO) applicants during the week of July 16, 1999. All applicants successfully passed all sections of the initial license examinatio The NRC examiners used the guidance prescribed in NUREG 1021, Operator Licensing Examination Standards for Power Reactors, Rev. 8 April 1999, to prepare and administer the operating test and written examinatio .2 Pre-Examination Activities Scope Nuclear Regulatory Commission examiners developed the examination material using the guidance prescribed in NUREG 1021. Point Beach training department instructors reviewed the material for technical accuracy and plant applicabilit b.- Observations and Findinas Written Examination Training department instructors reviewed the written examination and made a significant l . number of comments on the proposed examination. The instructors' review was i thorough and accurate. The comments provided by the instructors enhanced the quality I and discriminatory value of the written examination. The NRC examiners determined that the majority of corrections made to the written examination were a result of the ,

licensee's procedure revision program that occurred during development of the !

examination.

i  !

i Operatina Examination l Training department instructors reviewed the operating test and made several comments on the proposed test. Two administrative JPMs and two operating JPMs were replaced because they were either too lengthy, made poor use of the simulator, or l were not supported by the simulator. The dynamic simulator scenarios were acceptable t with minor modifications.

l The instructors' comments on the operating test enhanced the quality and discriminatory value of the operating test. The instructors were always business-like and professional in their dealings with the NRC examination tea r

I L .

!

. Conclusions

- The facility training staff was well prepared to support the examination process and provided high quality reviews of the written examinations and operating tests.

05.3 ' . Examination Activities l Scope l

i The NRC examiners administered the operating test (JPMs and dynamic scenarios)

,

during the week of July 26,1999, and the written examination on August 2,1999. The

, tests were administered using the guidance prescribed in sections ES-302 and ES-402 l of NUREG 1021.

l Observations and Findinas

'

Operatina Test l

!

! The applicants appeared well prepared for the test. They quickly found appropriate procedures when required and normally executed the procedures correctly. Applicants i

. communicated clearly and accurately with only occasional failures to complete repeat-backs; crew briefs were generally good.~ Applicant diagnostic skills were normally accurate. Control board switch manipulations were normally accurate with only a few isolated errors. Senior Reactor Operator command and control skills and knowledge of l technical specifications were good. Reactor Operator ability to use technical specifications was very good.

l

'

The training staff's support of the examination process was noteworthy. Paperwork to support the administration of the operating test was always correct and detailed. Shift tumovers provided by the instructors during the dynamic scenarios were typical of those provided in the control room and added realism to the examination process. The licensee training staff coordinated the arrival times of the applicants and provided escorts to maintain examination security during the ooerating test. The licensee's daily setup and execution of the operating test during the validation and examination weeks was timely and accurat The following weaknesses were observed by the examiners:

Some applicants were not familiar with certain back panel operations and the 1 l locations of back panelinstruments and control ;

l .

!

i

Several applicants were not familiar with the requirements of locally starting and !

L paralleling an emergency diesel generator to its electrical bu ;

l 1

! *- -

Approximately one-third of the applicants failed to correctly review a proposed ;

!

. tagout that contained pre-inserted error i 5 .

.

.

Written Examination The written examination was administered on August 2,1999, prior to the management exit meeting. All applicants completed the examination within the allotted five hour Conclusions The applicants were well prepared for the operating test and written examination. In general, they displayed good operating and communicating practices during the operating test. The facility training staff was well prepared to support the examination proces .

05.4 Post Examination Activities Examination Scope The NRC examiners evaluated individual applicant performance on the operating test and the written examination. The examiners also reviewed post en mination comments submitted by the licensee. Examiners followed the guidelines contained in sections ES-303, ES-403, and ES-501, of NUREG 102 Observations and Findinas Job Performance Measures Two generic knowledge deficiencies were discovered while grading the candidates'

performance of the operating JPMs:

-

Sevful applicants were not able to complete a task to start and parallel an emergency diesel generator to its electrical bus using emergency operating procedures. Most applicants not properly completing the procedure did not realize certain steps of the procedure were required to be locally execute Several applicants were not familiar with the location of instruments and controls ;

'

located in the control room's back panels. When the applicants were given a cue by a simulator operator that they could not find the instrument or controller in the remote location, the applicants then realized the desired equipment was in the control room back panels and they correctly completed the assigned tas Dynamic Simulator Scenarios l The NRC examiners did not identify any significant or generic weaknesses during the review of the dynamic simulator scenario test result Written Examinatio_q All of the applicants passed the written examination with scores ranging from approximately 85% to 94% There were five questions that were answered incorrectly by more than 50% of the applicants. These questions were considered generic

f

,

. .

l .

knowledge deficiencies and are provided to the Point Beach training staff for consideration and implementation into the Systematic Approach to Training-based

,

ps ogram.

!

,

!

Question # Knowledae Weakness

1. 3(RO) Fuel oil transfer pump P206A manual operation during a loss of off-site power.

l #8(RO) Emergency diesel generator monthly surveillance test procedure j precautions which required the suspension of 345kV and 13.8 kV

'

! work and testing while testing the emergency diesel generator #10(RO) Alternate means of exling the spent fuel pool in accordance with AOP-8 l l'

l #81(RO) Indicated auxiliary feedwater system flow on a loss of instrument l #56(SRO) air to the recirculation valve for 1P29.

L '

1. 97(RO) Shifting instrument supply bus feeders followup action ;

1. 72(SRO)

The licensee submitted four post examination comments which were reviewed by the NRC examiners. The licensee's comments and NRC resolution of the comments are detailed in Enclosure 2," Facility Post Written Examination Comments and NRC Resolution." Three comments were accepted, resulting in two correct answers for one question and answer key changes for two questions. The comment for the remaining J question was accepted; however, the question was deleted from the examinatio , Conclusions The applicants were well prepared for the written examination. The small number of questions missed by more than 50% of the applicants indicated a comprehensive training program that met the needs of the applicant .5 Simulator Fidelity The simulator performed well during the validation week, during the operating JPM test and during the dynamic simulator scenarios with no noted deficiencies. This was documented in Enclosure 3, Simulation Facility Report.

r

-

.

V. Manaaement Meetinas X1 Exit Meetina Summary The chief examiner presented the examination team's observations and findings to members of the licensee's management on August 2,1999. The licensee acknowledged the findings presented and indicated that no proprietary information had been identified during the examination or at the exit meetin ,

i

!

l

.

l l

n

.

l

.

PARTIAL LIST OF PERSONS CONTACTED J. R. Anderson, Manager, Process improvement A. J. Cayia, Manager, Regulatory Services & Licensing F. A. Flentje, Senior Regulation & Compliance Specialist J. E. Knorr, Manager, Regulation & Compliance

,

i R. G. Mende, Plant Manager, Point Beach J. G. O'Grady, Operations Manager C. R. Sizemore, Operations Training Coordinator G. D. Strharsky, Assistant Operations Manager W. P. Walker, Training Manager NRC F. Brown, Senior Resident inspector, Point Beach Nuclear Plant P. Louden, Resident inspector, Point Beach Nuclear Plant INSPECTION PROCEDURES USED IP 71707, " Plant Operations" l

ITEMS OPENED, CLOSED, AND DISCUSSED Opened None Closed None Discussed None

p l -

l

-

l LIST OF ACRONYMS USED CFR Code of Federal Regulations DRS Division of Reactor Safety

'

.

IP inspection Procedure JPM Job Performance Measure NRC Nuclear Regulatory Commission OL Operator Licensing RO Reactor Operator SRO Senior Reactor Operator PARTIAL LIST OF PROCEDURES REVIEWED AOP-1B, *RCP Malfunctions" AOP-3, " Steam Generator Tube Leak" AOP-6A, " Dropped Rod," Revision 10

-

AOP-6E, " Alternate Boration/ Loss of Shutdown Margin," Revision 7 AOP-10A, " Safe Shutdown - Local Control," Revision 25 AOP-13C, " Severe Weather Conditions," Revision 6 AOP-17A, " Rapid Power Reduction" AOP-21, "PPCS Malfunction," Revision 0 with PBF 2512 AOP-24, " Response to Instrument Malfunctions" ARB 1C031E21-4," Steam Generator B High Level Channel Alert"

'

ARB 1C031E21-5, " Steam Generator B Level setpoint Deviation" ARB 1C041C 3-2," Pressurize High Pressure Channel Alert" ARB 1C20 C 1-2, " Unit 1 Steam Line A Radiation High" ECA-0.0," Loss of All AC Power," Revision 23 EOP-0.0, " Reactor Scram" EOP-1.1, "Si Termination" EOP-1.3, " Transfer to Containment Sump Recirculation," Revision 20 EPIP 1.1, * Event Classification," Revision 33 and EPIP 1.2, " General Emergency," Revision 31 HP 2.6, "High Radiation Access Control," Revision 17 NP 1.9.15, " Danger Tag Procedure," Revision 9 Ol 100, " Adjusting SI Accumulator Level and Pressure," Revision 13 OP 1C, " Low Power Operation to Normal Power Operation," Revision 68 OP-5B, " Blender Operation / Dilution /Boration," Revision 14 REl-7," Rod Position Determination" SEP-2.0," Shutdown LOCA Analysis," Revision SEP-2.1," Shutdown LOCA with RHR Aligned for Low Head injection," Revision 6 TS 33, " Containment Accident Recirculation Fan-Cooler Units (Monthly)," Revision 16 1-SOP-CC-001," Component Cooling System," Revision 0

l

!

.

.

Enclosure 2 Facility Post Written Examination Comments and NRC Resolution Question #44 (RO); #19 (SRO)

The following conditions exist on Unit 1:

A reactor trip occurred 1 minute ago from 100% power The main feedwater regulating velves failed in the 100% power position Containment humidity le .weasing

. IR "A" SUR: - 0.00 d?m IR "B" SU.'* - 0.15 dpm

Which one of the following actions is required FIRST following a manual safety injection actuation? ' ASSUME: All other equipment performed as designed.) Transitico to EOP-2," Faulted Steam Generator Isolation."

- Manually opan containment ventilation cooler outlet emergency FCV Transitior tc OSP-S.1," Response to Nuclear Power Generation /ATWS." Trip the main fe.ed pumps, place the condensate pumps in PULL OUT, and stop the heater dab pum ANSWER COMMENT:

The answer key should be changed to only accept answer d as the correct answer. Por EOP-0, " Reactor Trip or Safety injection," actions to verify feedwater isolation occur in Step 5. This is the first step after the immediate actions are complete. The actions referred to in answer b are not complete until Step 12 within EOP-0. The actions referred to in answer e are not required at this time per EOP-0, Step 1. Additionally, even if status tree monitoring were implemented on an earlier transition out of EOP-0, the actions referred to in answer c are not required per CSP-ST.0, " Critical Safety Function Status Trees," Figure 1, "ST-1 Suberiticality," based on the information contained in Questien 1 NRC Resolution:

The comment was accepted. This was a typographical error and was discovered by NRC examiners prior to receiving the licensee's comments. The answer was changed to accept only answer d as the correct answe .

.

2. Question #45 (RO); #20 (SRO) ,

'

With the RCS at normal operating pressure and temperature, what is the condition of the steam entering the PRT if a PORV opens? (ASSUME: PRT is at 100*F, 5 psig; an -

ideal thermodynamic process)

Saturated steam-water mixture at: -

, * ' ' * ANSWER !

I COMMENT:

l The answer key should be changed to only accept answer b as the correct answer. The RO examination question #45 that is the same as SRO question 20 uses answer b as its correct answer. Per the Mollier Diagram, following the constant enthalpy line from 2000 psia saturated conditions to 20 psia (equivalent of 5 psig) and then following up the 20 psia curve to the saturation line, you cross the saturation line at -228' NRC Resolution:

l The comment was accepted. This was a typographical error. This was a typographical error and was discovered by NRC examiners prior to receiving the licensee's comment l The answer was changed to accept only answer b as the correct answe '

l l

l

1 I

-

-

,

, ' Question #59 (RO); #34 (SRO)

Given the following conditions:

,

,

' A large break LOCA has occurred inside containment Containment pressure initially increased to 30 psig, then began trending down and is currently.10 psig The spray addition tank level has decreased by 20% i What is the current status of the containment spray pumps and valves?

PUMPS DISCHARGE VALVES Running Full open Secured Closed Running Mid-position Secured Full open ANSWER d

Comment:

The question as worded is not clear as to whether any assumptions should be made for operator actions being completed, if the examinee assumes that no operator action has occurred, then the spray system would still be in operation and fully aligned, since there are no conditions which would automatically secure the containment spray pumps, even though containment pressure is below the actuation setpoint (25 psig) and the design quantity (12%) of NaOH has been injected. Under these assumptions, both spray pumps would be running and the discharge valves would be full open. The examinee would select answer a. as correct. If the examinee assumes that a procedure guided operator action has occurred, then the spray system would have been secured in accordance with EOP-1.3, " Transfer to Containment Supm Recirculation," Step 3 (a continuous action). EOP-1.3 would be entered based on a large break LOCA. EOP-1.3, Step 3, checks for containment spray pumps running (Substep a), containment pressure less than 15 psig (Substep b) and spray addition tank level being lowered by at least 12% (Substep c). Since the question states that the last two of these conditions are satisfied, then the examinee would complete the remaining substeps, that reset containment spray signal (Substep d), stop both containment spray pumps (Substep e),

shut the suction MOVs (Substep f), and shut the spray additive tank discharge valves

-

(Substep g). Since EOP-1.3 does not direct shutting the discharge valves, the examinee would select answer d. as correct. The examinee could correctly select either answer a. or d. based on the assumptions made and the lack of clarity in the question with respect to the performance of operator actions. Additionally, based on guidance contained in EOP-1.3, answers b. or c. could not be considered as correct answer l l

!

-

'

)

,

NRC Resolution:

The comment was accepted. T he original intent of the question was for an applicant to recognize that after the given conditions were met, an operator would realign the system to meet the desired answer. The answer was changed to accept both a. and d. as correct answer !

4. Question #87 (RO); #62 (SRO)

During a liquid release, which ONE of the following conditions would require compensatory actions to be taken?  ; RE-223 out-of-service while discharging radwast i RE-219 and RE-222 out-of-service while SG blowdown in progres ; RE-229 out-of-service during unit operatio RE-230 out-of-service during a retention pond discharg ANSWER Comment:

There are two correct answers for the question as written. The answer key should be changed to accept both answers c. and d. as correct answer This comment is supported by the " Radiological Effluent Control Manual," (RECM)

Section 3.2.2, that states,"If fewer than the minimum number of radioactive effluent monitoring channels are operable, the action statement listed in either Tabel 3-1 or 3-2 opposite the channel shall be taken." Per RECM Table 3-1, RE-229 (Item #3a) and RE-230 (Item #4) are the only monitors available for certain effluent releases. Therefore, if these monitors are out-of-service, the minimum operable channels are not satisfied; therefore, the action statement is applicable. The same note applies in both these

,

situations. Note #3 states, "... effluent releases via this pathway may continue provided that at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> grab samples are collected and analyzed..." This

'

action satisfies the definition of a compensatory action. With regards to RE-223 (Item

1. 1a), RE-219 and RE-222 (Item #2a), since the question looks at each of the distractors separately, the minimum operable channels are still met. Therefore, answers a. and do not require any compensatory action.

l NRC Resolution

!

The comment is accepted in part. The examiners agreed that there were compensatory actions required for distractors c. and d. However, the examiners determined that there were compensatory actions to be taken when any of the radiation monitors referred to in the question were out of service. Although the compensatory act. ions for a. and b. were not tech spec related, they were compensatory actions, and had to be taken. The

,

,

.

.

examiners determined that there were compensatory actions for a, b, c, and d. The question was deleted because all four of the distractors were correc !

i

!

l t

!

l l

l

..

t

'#*

SIMULATION FACILITY REPORT Facility Licensee: Point Beach Nuclear Power Plant Facility Licensee Docket Nos: 50-266; 50-301 Operating Tests Administered: July 26-30,1999 The following documents observations made by the NRC examination team during the initial license examination. These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative of non-compliance with 10 CFR 55.45(b).

' These observations do not affect NRC certification or approval of the simulation facility other than to provide information which may be used in future evaluations. No licensee action is required in response to these observation During the conduct of the simulator portion of the operating tests, the following items were

- observed:

ITEM DESCRIPTION 1 None 1

y l

L U.S. Nuclear Regulatory Commission Site-Specific Written Examination

,.

Applicant information Name: MASTER EXAMINATION Region: ll1 Date: August 2,1999 Facility / Unit:

License Level: RO Reactor Type: WESTINGHOUSE 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 percen Examination papers will be collected five hours after the examination start Applicant Certification All work done on this examination is my own. I have neither given nor received ai Applicant's Signature Results Examination Value 10 Points Applicant's Score Points Applicant's Grade Percent l

l

{

! REACTOR OPERATOR Page 2

WRITTEN EXAMINATION GUIDELINES After you complete the examination, sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examinatio . To pass the examination, you must achieve a grade of 80.00 percent or greater; grades will not be rounded up to achieve a passing score. Every question is worth one poin . For an initial examination, the time limit for completing the examination is five hour . You may bring pens, pencils, and calculators into the examination room. Dark pencil should be used to facilitate machine gradin . Print your name in the blank provided on the examination cover sheet and on the answer sheet. You may be asked to provide the examiner with some form of positive identificatio . Mark your answers on the answer sheet provided. Use only the scan-tron sheets provided and do not write on the back side of the pages unless instructed. If you change an answer, ensure your erasure is complete. If the grading examiner is unable to determine which answer you have marked, the answer will be marked wron . If you have any questions concerning the intent or the initial conditions of a question, do not hesitate asking them before answering the question. Ask questions of the NRC examiner or the designated facility instructor only. When answering a question, do not make assumptions regarding conditions that are not specified in the question unless they occur as a consequence of other conditions that are stated in the question. For example, you should not assume that any alarm has activated unless the question so states or the alarm is expected to activate as a result of the conditions that are stated in the questio . Restroom trips are permitted, but only one applicant at a time will be allowed to leav Avoid all contact with anyone outside the examination room to eliminate even the appearance or possibility of cheatin . When you complete the examination, assemble a package including the examination questions, examination aids, answer sheets, and scrap paper and give it to the NRC examiner or proctor. Remember to sign the statement on the examination cover sheet indicating that the work is your own and that you have neither given nor received assistance in completing the examination. The scrap paper will be disposed of immediately after the examinatio . After you have turned in your examination, leave the examination area as defined by the proctor or NRC examiner. If you are found in this area while the examination is still in progress, your license may be denied or revoke . Do you have any questions?

REACTOR OPERATOR Page 3 IS PAGE INTENTIONALLY LEFT BLANK, I

I

,

e s

!

i l

l I

i

'$

!

!

J

r

<

REACTOR OPERATOR . Page 4 l

l l

l l

l QUESTION: 001 (1.00)

i The probability of a complete loss of the Unit 1 RED instrument bus is reduced since a failure of the normal inverter supplying power to the bus,1DY-01,.will result in an automatic transfer to: The RED spare inverter; DYO The Unit 2 RED inverter; 2DY-01.

l l A non-safeguards 120 Vac power supply from B09.

I , The 120 Vac non-protection instrument bus; 1Y-0 QUESTION: 002 (1.00)

l A LOCA has occurred and both RCPs have been re-started per CSP-C.1, " Response to l Inadequate Core Cooling." Given the below list of conditions:

1 RCS pressure 200 psig and decreasing l ll Narrow range reactor vessel level is at 27 feet and increasing Ill A and B RCS hot leg loop are at 340*F and decreasing

,

IV Core exit thermocouples are at 1150*F and decreasing Which ONE of the following gives the two conditions from the above list that directs the operator j to secure the RCPs per CSP-C.17 I and Il Il and I and I and Ill.

!

l

r'

REACTOR OPERATOR Page 5 l

l l

l QUESTION: 003 (1.00)

l Fuel oil transfer pump P206A is being operated "in-hand" to support filling emergency diesel generator fuel oil day tank T-31 A during the performance of TS 81, " Emergency Diesel Generator G-01 Monthly." While the tank is being filled, Unit i experiences a loss of off-site power. Which of the following best describes the effect on the operation of fuel oil transfer pump 1P-206A?

' Pump 1P-206A will stop and will not restart until off-site power is restored.

, Pump 1P-206A will continue to operate, powered from its rectified DC backup l power supply.

l Pump 1P-206A will continue to operate, powered from emergency diesel generator G-0 Pump 1P-206A will stop initially but will be sequenced on in 20 (-1/+2) seconds, l powered from emergency diesel generator G-0 { I l

QUESTION: 004 (1.00)

l The Unit 2 component cooling water system was being operated in its normal full power line-up with pump 2P-11 A running and pump 2P-11B in standby. Pump 2P-11 A experienced a failure of its outboard pump bearing resulting in rapidly decreasing pump RPM. Given that no operator action was taken, pump 2P-118 will start automatically on: Iow system flow (<1000gpm).

l high system temperature (>200*F). Iow system header pressure (<35 psig). Iow system differential pressure (<25 psid).

!

l

.,

__

II

,

REACTOR OPERATOR Page 6 t

QUESTION: 005 (1.00)

Unit 2 is currently at five percent reactor power, preparing for shift turnover. Which of the following describes a proper shift turnover activity for the " operator at the controls" as described in OM 1.1 and OM 3.g7 One (and only one) of the operator at the controls may be relieved by the DSS short-ter Off-going and on-coming watchstanders shall walk-down the front control boards togethe The off-going and on-coming operator at the controls personnel shall determine which AO watchstations may perform relief activitie The on-coming watchstander shall review the STATION LOG back to the last time he or she stood the watch or for a minimum of three shifts (whichever is of shorter duration).

QUESTION: 006 (1.00)

The refueling unit control operator in the control room during fuel movement is responsible for all of the following activities with the exception of . Maintaining the ICRR (1/M) plot (s) during fuel loadin Maintaining the fuel loading sequence document and core computer datsh:: Relieving the core load supervisor for short periods of time when core #ierations are NOT in progres Monitoring source range nuclear instrumentation indication and RHR/RCS heat exchanger inlet temperatur I

! l i

l I

REACTOR OPERATOR Page 7 j

-

i 1 i

I QUESTION: 007 (1.00)

i

!

A nineteen (19) year old new employee received 360 mrem during the current quarter (2250 l mrem for the calendar year) at the Monticello Nuclear Generating Station before being hired 1 here. Which one of the following is the MAXIMUM additional exposure the new employee may l receive throughout the remainder of the calender year at PBNP with an ADMINISTRATIVE annual uose level extension?

1 No additional exposure is permitte l mre mre mrem.

! QUESTION: 008 (1.00)

l

'

The emergency diesel generator monthly surveillance test procedures for G-01, G-02, G-03, and G-04 (TS 81, TS 82, TS 83, and TS 84 respectively) each contain a precaution which requires the suspension of 345kV and 13.8 kV work and testing while testing the diesel generators. This precaution is: Only applicable when 345kV and 13.8 kV work and testing is performed in the vicinity of energized buses, Only applicable when paralleling diesel generators with offsite power when offsite power is unreliable due to weather conditions or other factor Necessary to reduce the probability of a loss of electrical load trip of either unit; the associated 4.16kV fsst-bus-transfer may challenge the overload protectio Necessary to reduce the probability of inadvertently opening the output breaker of the associated diesel generator while an auto-closure signal is present and having to reset the breaker's anti-pumping circuit.

l l

.,

i L

r f

REACTOR OPERATOR Page 8

!

QUESTION: 009 (1.00)

i During a surveillance test on the Unit 1 train A Si sequencer, service water (SW) pump 1P32A failed to receive an auto-start signal. The pump started normally h: manual. Which one of the following describes the operability and Technical Specification (TS) applicability associated with SW pump P32A7

, SW pump P32A is still operable because it can still be manually started and a

!

service water TS LCO action statement would not be entere SW pump P32A is inoperable and a service water TS LCO action statement would be entered only if an additional two pumps fail since 4 SW pumps are required to be operabl SW pump P32A is inoperable and a service water TS LCO action statement would be entered because 6 SW pumps are required to be operabl SW pump P32A is still operable because it will start automatically if the pump's discharge pressure falls below 88 psig and a service water TS LCO action statement would not be entered.

,

o k

~

REACTOR OPERATOR Page 9 QUESTION: 010 (1.00)

Annunciator " SPENT FUEL POOL TEMP HIGH LEVEL HIGH OR LOW' was received on Unit The DSS dispatched an operator to the spent fuel pool (SFP). The operator reported that pool temperature is 130*F and rising slowly. The operator also reported that no spent fuel pool cooling water pumps were running. He attempted to restart the pumps, but the attempt faile What alternate means of cooling the SFP would be used in accordance with AOP-8F7 Maintain spent fuel pool level via the Unit 1 blender and allow losses to ambient to cool the poo Cross-tie the spent fuel pool cooling system with the residual heat removal (RHR)

system and utilize a RHR heat exchanger to cool the poo Open one transfer canal door, line up the HUT recirculation pump to recirculate water between the pool and the transfer canal, allow losses to ambient to cool the poo Open one transfer canal door, line up the Unit 1 RWST recirculation pump to j recirculate water between the pool and the transfer canal, allow losses to ambient to cool the poo l

,

QUESTION: 011 (1.00)

There is thick smoke in the main control room along with a heavy acrid smell. Several control room operators are having trouble seeing and breathing. The DSS gave the order to " evacuate the control room." The control room operator's actions include: Manually initiate safety injectio Take possession of the AOP-10A packs and exit the control roo Notify plant personnel of a fire in the control room utilizing the Gal-Tronics system.

! Secure the reactor coolant pumps, main feed pumps, and component cooling water pumps by momentarily taking the applicable control switches to TRI ., F .

I

,

i i REACTOR OPERATOR - Page 10 l

!

QUESTION: 012 (1.00) '

A RED light is illuminated on the EBERLINE radiation monitoring system operator control panel in the control room. The illuminated RED light indicates which of the following? A source check is in progres l l An external failure caused by high sample flow has occurre A detector is sensing a high alar A detectoris sensing a trend alar QUESTION: 013 (1.00)

Unit 1 is at 40% rated thermal power. Condensate pump 1P25A is running, condensate pump 1P25B is idle with it's control switch in AUTO. Hotwell level control is in AUTO. The following plant conditions are observed:

-

1A steam generator levelis 64%.

-

Main feed pump suction header pressure is at 179 psi Unit 1 condenser hotwell level is at 20 inche The effect of the above plant conditions on the condensate system would be? Automatic trip of pump 1P-25 I Automatic start of pump 1P-25 l l Hotwell make-up valve 1CV-2125 opens to increase leve i Hotwell make-up valve 1CV-2125 closes to decrease leve j l

'

!

f . REACTOR OPERATOR . Page 11 QUESTlON: 014 (1.00)

Unit 2 is at 5% rated thermal power. Feedwater is being supplied to the steam generators via the feed bypass valves and the controllers are in AUTO. Steam generator B level is at 80%.

Which of the following actions must be taken in order to reestablish SG B level control via

- feedwater bypass valve 2CV-481B7 Place bypass valve controller LC-481 in manua Steam generator B level must be reduced to below 60%. Locally reset the closing solenoid valves once the closure signal has been cleared and rese Momentarily depress the respective feedwater control valve bypass reset pushbutton (FWCV BYPASS) once the closure signal has been cleare QUESTION: 015 (1.00)

Given the following Unit 1 plant conditions:

• The plant is at 100% powe *

Pressurizer pressure is in AUTOMATIC contro '

The Pressurizer Pressure Channel Defeat switch is in the normal positio No operator action is take .

i Which ONE of the following actions occurs when pressurizer pressure transmitter PT-431 (blue) l fails LOW 7 l 1 Pressurizer pressure INCREASES, resulting in a HIGH pressure reactor tri Pressurizer pressure DECREASES, resulting in a LOW pressure reactor tri l Pressurizer PORV (PCV-430) cycles to maintain pressure below the reactor trip setpoint.

Pressurizer heaters and spray valves operate normally to maintain pressurizer l- pressur ,

l  ;

t 1

-

'

REACTOR OPERATOR Page 12 OUESTlON:016 (1.00)

required to be operated has had it's ,

identification enh e ocalequipment an installed red loc being red color-code placing yellow and black tape on the equipmen placing yellow and magenta tape on the equipmen QUESTION: 017 (1.00)

With the plant operating normally at 70% power, the following symptoms o ccurred:

• Increasing steam generator pressures.

a Increasing pressurizer pressure

TAVG greater than TRE Annunciator"TAVG STEAM DUMP CHANNEL ALERT" illumin Which of the following would cause the above symptoms? Excessive boratio Inadvertent AFW actuatio .

Uncontrolled rod withdrawa Inadvertent steam dump actuatio I l

l

. REACTOR OPERATOR Page 13 l l I l

l QUESTION: 018 (1.00) -

!

With the Pressurizer Level Control Selector Switch in the NORMAL position, a pressurizer level instrument failure caused the following SEQUENTIAL plant events:

- Charging flow was reduced to minimu Pressurizer level decrease I

-

Letdown flow was secured and heaters turned of Pressurizer level increased until a high level trip occurre Which one of the following instrument failures occurred? (Assume NO operator action) Pressurizer level channel 428 (blue) failed low, Pressurizer level channel 428 (blue) failed hig Pressurizer level channel 427 (white) failed lo Pressurizer level channel 427 (white) failed hig QUESTION: 019 (1.00) l l

Which ONE of the following is the ALTERNATE power supply for the 2P-2A Charging Pump? B08 B09 ,

l B03

..

n REACTOR OPERATOR Page 14 l

l

'

QUESTION: 020 (1.00)

Given the following plant conditions for Unit 2:

Unit 2 is at 100% power

Annunciator 2P-1B, RCP LEVEL STANDPIPE HIGH is alarming

-

Seal injection flow to 2P-1B RCP is 8 gpm l

-

1. 1 Seal leakoff flow has decreased to 1.3 gpm from 1.8 gpm l

=

2P-1B pump vibration is approximately 1 mil above normal l

Which one of the following would cause the above conditions for Unit 2?

)

Increasing No. 3 sealleakag I Increasing No. 2 sealleakag c. - No.1 Seal bypass valve (CV-386) leaking b No. 2 seal leakoff standpipe drain valve (556A) leaking b QUESTlON: 021 (1.00) ,

Given the following Unit 2 plant conditions:

• A large break LOCA has occurre At t=0, an Si signal was generate *

At t=30 seconds, a containment spray signal was generate Which ONE of the following describes the sequence of operation of containment spray components? Spray pumps A and B start Spray discharge valves 860A, B, C, D open NaOH addition valves 836A and B ope NaOH addition valves 836A and B open Spray discharge valves 860A, B, C, D open Spray pumps A and B star Spray discharge valves 860A, B, C, D open Spray pumps A and B start NaOH addition valves 836A and B ope Spray pumps A and B start NaOH addition valves 836A and B open Spray discharge valves 860A, B, C, D open.

,

-

REACTOR OPERATOR Page 15 l

l QUESTION: 022 (1.00)

i Due to a leak in the fire suppression header, system pressure has dropped rapidly to 75 psi With no suppression signals to the fire system present, which of the following describes automatic fire pump operation? Only the electric motor driven fire pump will start to restore system pressure, The jockey pump and the electric motor driven fire pump will start and automatically secure once system pressure is restore !

i The electric motor driven fire pump, and the diesel engine driven fire pump will )

l start but must be manually secured once system pressure is restore The system air compressor will start, precluding the start of the automatic fire pumps, and combine with the hydropneumatic tank to restore system pressur !

! QUESTION: 023 (1.00) '

Given the following Unit 2 plant conditions:

l

-

RHR is in servic RCS pressure is 320 psig and INCREASING.

,

-

RCS temperature is 340 degrees F and INCREASING.

!

-

ALL systems are in a normal shutdown configuration for solid plant operatio Which one of the following will act FIRST to prevent overpressurizing the RHR System? Pressurizer PORVs will ope RHR return isolation valve will auto close.

l RHR pump hot leg relief valve RH-PCV-861C will open.

l RHR pumps discharge relief valve RH-PCV-861 A will ope :

l REACTOR OPERATOR Page 16 l

l QUESTION: 024 (1.00)

The reactor was operating norma ly at 100% power when a "PRT PRESS Hi TEMP HI LEVEL HI OR LO" alarm activated on the control board. The pressurizer PORVs and safety valves have been verified shut. The following are indications for the PRT:

PRT pressurs: 4.0 PSIG

.

PRT temperature: 100*F

PRT1evel: 69%

Which of the following would cause this alarm condition? 4 Seat leakage .from PRT drain valve RC-596 to the RCD i Backleakage from the reactor coolant system into the PR RHR suction relief valve (RH-861B) seat leakage into the PR Leakage of nitrogen past valve RC-555 into the PRT (assuming the nitrogen pressure regulator valve has been fully opened).

QUESTION: 025 (1.00)

Given the following Unit 2 plant conditions:

-

Unit is at 50% powe Control rods are in AUTOMATI Which ONE of the following instrument malfunctions would result in a CONTINUOUS rod withdrawal? PT-485 fails HIG Loop "A" Tsar RTD fails LO Loop "B" Teoto RTD fails HIG Power range channel N-42 fails LOW.

l t

-

I REACTOR OPERATOR Page 17 QUESTlON: 026- (1.00)

During a natural circulation cooldown on Unit 1, a void was created in the reactor vessel. Given

. that no RCPs are available and RCS pressure is approximately 1500 psig, which one of the following actions is used to collapse the void in accordance with EOP-0.2, " Natural Circulation Cooldown?" - Decrease RCS temperature while maintaining RCS pressure constan Fill the pressurizer solid and ventilate via the post accident ventilation syste Start an SI pump to increase RCS pressure while maintaining temperature constan Increase RCS pressure using pressurizer heaters while maintaining RCS pressure / temperature within the acceptable regio .

i i

L

r 4 REACTOR OPERATOR Page-18 QUESTION: 027 (1.00{

Given the following Unit 2 plant conditions:

a The Unit is in Cold Shutdow .

The RCS is soli *

RHR flow has been lost and CANNOT be restore Wide range water level for both steam generators (SGs) is 250 inches and stead *

All other systems and components are available, in accordance with SEP 1.1, which one of the following methods of cooling is the preferred method of removing the are's decay heat? Establish RCS cooling by aligning AFW system flow to at least one SG and bleeding steam through the respective S/G Atmospheric Steam Dump Valve, Establish feed and bleed through the RCS by aligning the RWST to the suction of the charging pumps, starting a charging pump, and venting through the pressurizer PORV Establish feed and bleed through the RCS by aligning RWST to the suction of the safety injection pumps, starting a safety injection pump, and venting through the pressurizer PORV Establish feed and bleed through the RCS by aligning RWST to the RCS through an RHR heat exchanger, gravity draining the RWST to the RCS, and venting through the pressurizer PORV l k

\

i l

l REACTOR OPERATOR Page 19 l

i QUESTION: 028 (1.00)

EOP-0 has been entered on Unit i due to an automatic initiation of safety injection and a reactor trip. The following plant conditions were observed:

Reactor coolant system pressure is 1500 psig and decreasing slowl *

Radiation monitor RE-113, " Elevation -19' PAB area monitor"is alarming on the RMS panel; all other RMS indications are norma The safety injection pump suctions are aligned to the RWS Residual heat removal pump suctions are aligned to RWS *

RWSTlevelis 95%

-

Annunciator C01 A1-11, " AUX BLDG 19 FT SUMP LEVEL HIGH is li The above plant conditions would require transition from EOP-0 into which of the following procedures? EOP-1.1; " Unit 1 SI Termination."

, EOP-0.1; " Reactor trip response." ECA-1.2; "LOCA Outside Containment." ECA-1.1; " Loss of Containment Sump recirculation."  ;

QUESTION: 029 (1.00)

Which of the following is the reason for depressurizing the Steam Generators at the maximum rate during ECA-0.0," Loss of All AC Power?" To prevent lifting PZR PORVs.

l To minimize RCS inventory los To prevent inadvertent reactor re-star To enhance restoration of SG level from the TD AFW Pump.

t

r

,

' REACTOR OPERATOR Page 20

, - QUESTION: 030 (1.00)

' Which of the following is the reason for the RCP tripping criteria during the performance of EOP-0, " Reactor Trip or Safety injection?" To prevent exceeding containment pressure limitctions during a large break LOCA even To minimize the effects of RCS cooldown in the event of a major steam line break '

for PTS consideration To limit the rate of RCS depressurization in the event of a large break LOCA and reduce the amount of voiding in the cor To limit the RCS inventory depletion through a small break LOCA leading to a more severe core uncovery if RCP's were tripped some time late QUESTION: 031 (1.00)

Which one of the following is the basis for terminating Si flow when the criteria are satisfied during the performance of EOP-3," Steam Generator Tube Rupture?"

! Prevent overcooling the RCS.

, Prevent solid plant operation Prevent exhausting RWST level, Prevent overfilling the ruptured S ..

L

i l REACTOR OPERATOR Page 21 g

'

l l

QUESTION: 032 (1.00)

According to AOP-0.0," Vital DC System Malfunction," ALL of the following are associated with loss of DC control power to an AC bus, EXCEPT the: associated breaker positions remain "as is." undervoltage stripping of the associated bus remains operable, associated breakers cannot be electrically operated from the control roo i

associated breakers cannot be electrically operated from local control stations. i QUESTION

033 (1.00)

The following conditions exist: l

Reactor Bypass breaker "A" racked in and closed for testin Reactor Trip breaker "A" ope Which of following describes the response if bypass breaker "B"is racked in? Both bypass breakers (A and B) will trip ope Bypass breaker "A" and trip breaker "B" will trip open, Bypass breaker "A" and trip breaker "B" will remain shu Bypass breaker "A" will remain shut and trip breaker "B" will trip open.

I l

l l'

p:

t l

REACTOR OPERATOR Page 22 l

l l . QUESTION: 034 (1.00)

. .

Which one of the following is the reason for the Engineering Safety Features Actuation System precaution and limitation which states," Do not exceed 1700 psig in the primary system during heat up until the reactor coolant temperature is at least 480*F7" l

' To ensure Tavg is above the Low-Tavg setpoint preventing an inadvertent S To avoid pressurized thermal shock concems in the event of an inadvertent SI.

l To ensure the Si termination criteria are satisfied before automatic unblocking of

SI.

! ,

l

.

.., To ensure steam generator pressure is greater than 530 psig before automatic

. unblocking of S {

L l

,

- QUESTION: 035 (1.00)

l .

l Which of the following is an automatic action directly associated with a Safety injection Actuation l only on Unit 17 l Electric *ie nump trips.

' Charging pump, P-2A trip Instrument air compressor, K2B trips.

l Pressurizer heater backup groups A and B trip.

i l

l

'

REACTOR OPERATOR Page 23 QUESTION: 036 -(1.00)

. Which one of the following describes the automatic response / sequence of events of the j Containment Air Recirculation System upon an Sl actuation signal? (assume off-site power is ;

'

available) Both the large and small fans receive start signals for each fan coole The small fans trip in all fan cooler units and the large fans receive a start signa Both the large and small fans trip and then the large fans start for each fan cooler l uni l 1 Both the large and small fans trip and then the small fans start for each fan cooler uni !

QUESTION: 037 (1.00) .j The following Unit 1 plant conditions exist:

,

-

Reactor power: 100% steady state

• All control systems in automatic

-

T temperature transmitter TE-401 B failed low Which of the following describes the immediate effects of this failure on the corresponding Tavg and Delta-T indications?

Tavg Delta-T INCREASES INCREASES DECREASES DECREASES INCREASES DECREASES DECREASES INCREASES ,

l

.

e ,

i l

h REACTOR OPERATOR Page_24 l

t

QUESTION: 038 (1.00)

! ; With Unit 1 operating normally at 100% power, two out of four OT/AT channels rise to within 3

!

degrees of the trip setpoint for 100 seconds. How much will turbine load be reduced? j l

! %- l

%

i

, % % l l

' QUESTION: 039 (1.00)

!  !

'

With the plant at 98% reactor power and Bank D control rods at 200 steps, which of the l following is a symptom of a stuck control rod that would require entry into AOP-6B, " Stuck Rod l Or Malfunctioning Position Indication", following a transient?

! a.- A variation in NIS instrumentation resulting in a quadrant tilt of 1.1%. I I An individual RPI with a 13 step disagreement with the bank demand locatio ;

l l A variation in core outlet thermocouples of 8% relative to symmetric .

thermocouple !

l . A variation in axial flux of 1.1% of axial peak at any location relative to l l

symmetrical trace.

l t

i i

.p r

r REACTOR OPERATOR Page 25 QUESTION: 040 (1.00)

- Given the following Unit 1 plant conditions:

The Unit has tripped from 100% due to a small break LOC *

. Conditions have stabilized and operators are evaluating the criteria for terminating S *

Adverse Containment conditions do NOT exist.

!

Which ONE of the following conditions would PREVENT Si termination per EOP-1.2, "Small Break LOCA Cooldown and Depressurization?" RCS subcooling is 40* Pressurizer levelindicates 9%.

! Pressurizer pressure is 2050 psig.

i- Both steam generator levels indicate 40% NR.

l QUESTION: 041 (1.00)

During a normal reduction in power using boration, which one of the following is the reason that

-

additional pressurizer heaters should be energized?

. a. ' Allow an increased ramp rate for the power chang b.- Equalize the reactor coolant system and Pressurizer boron concentrations, Maintain PZR pressure in normal operating range during the power chang Ensure positive pressurizer pressure control is established prior to starting the power chang REACTOR ' OPERATOR Page 26 QUESTION: 042 (1.00)

Which ONE of the following RVLIS readings indicates the highest probability of core voiding? W

, ide Range reading 98 ft. with NO RCPs runnin Wide Range reading 120 ft. with ONE RCP runnin Narrow Range reading 35 ft. with NO RCPs runnin Wide Range reading 140 ft. with BOTH RCPs runnin QUESTION: 043 (1.00)

Which ONE of the following describes the purpose of the back draft dampers installed in the Containment Air Recirculation System? Prevent backflow in a cooling unit in the event of fire in Containmen Serve as explosion dampers preventing duct work collapse during an acciden Prevent unit air backflow when the accident fan is running and the cooling fan is no Serve as a system air backflow damper in idle cooling units (both accident and cooling fans secured.

,

I )

'

i REACTOR OPERATOR Page 27 i

QUESTION: 044 (1.00)

The following conditions exist on Unit 1:

A reactor trip occurred 1 minute ago from 100% power

.

The main feedwater regulating valves failed in the 100% power position a

Containment humidity is increasing

IR "A" SUR: 0.00 dpm 1

-

IR "B" SUR: - 0.15 dpm Which one of the following actions is required FIRST following a manual safety injection actuation? (ASSUME: All other equipment performed as designed.) Transition to EOP-2, " Faulted Steam Generator Isolation."

I Manually open containment ventilation cooler outlet emergency FCV j l Transition to CSP-S.1, " Response to Nuclear Power Generation /ATWS." ' Trip the main feed pumps, place the condensate pumps in PULL OUT, and stop ;

the heater drain pum !

l

QUESTION: 045 (1.00)

,

With the RCS at normal operating pressure and temperature, what is the condition of the steam entering the PRT if a PORV opens? (ASSUME: PRT is at 100*F,5 psig; an ideal

" thermodynamic process) Saturated steam-water mixture at: ' I ' *F.

L *F.

t

_

REACTOR OPERATOR Page'28

- QUESTION: 046 (1.00)-

The following conditions exist on Unit 1:

]

A LOCA has occurred

-

Transfer to containment sump recirculation is required ' i

. RCS pressure is approximately 50 psig What is the expected Si pump TOTAL flow indicated on the main control board prior to entering 1 EOP 1.3, and how will this value change following transfer of BOTH trains of ECCS to containment sump recirculation?

Total Flow Flow Change gpm Decrease ' 700 gpm Increase gpm . Decrease l gpm ~lncrease QUESTION: 047 (1.00)

Given the following Unit 1 plant conditions:

A Control Bank D rod was dropped and recovere *

The Pulse to Analog Converter was NOT reset as required by AOP-6A, " Dropped Rod."

Which one of the following will occur on the next rod movement? If control rods are withdrawn, OT/AT will NOT stop Control Bank D withdrawal when require ' If control rods are withdrawn, OP/AT will NOT stop Control Bank D withdrawal when require If control rods are inserted, the Rod Insertion Limit Alarm will be received at a lower rod position than require ., If control rods are inserted, Bank C control rods will begin insertion at a lower value of Control Bank D position.

n

p  !

l i

REACTOR OPERATOR Page 29

)

QUESTION: 048 (1.00)-

in procedure OP-5A, " Reactor Coolant Volume Control," there is a PRECAUTION that states l

"Do not secure letdown flow without also securing charging flow ..." Which ONE of the following

.

' statements describes why charging flow is required to be isolated? (ASSUME: All systems are in a normal at power lineup.) Reduce thermal shock on the charging penetration into the RC Reduce thermal shock on the Non-Regenerative Heat Exchange VCT level will decrease until charging pump suction shifts to the RWS VCT level will decrease causing possible damage to the charging pumps.

I QUESTION: 04g (1.00)

l Given the following Unit 1 plant conditions: [ .

-

!

'

The Unit was operating at 75% steady state powe All systems were in automatic contro The "A" SG atmospheric steam dumps failed open. Main turbine control is in IMP IN with the valve position limiter set at 95%.

Which one of the following describes the plant response to this condition? (Assume NO operator l l action is taken.) Turbine load decreases by 5%, reactor power remains stable at 75%. Control rods initially insert then withdraw to maintain reactor power at 75%. Control rods withdraw and reactor power increases to 80% where it stabilize Turbine governor valves open in response to lower steam header pressure to l increase turbine load to 80%.

b

REACTOR OPERATOR Page 30 l

QUESTION

050 (1.00)

The following conditions exist on Unit 1:

l

• .

Reactor power 80%

The lower detector for N-43 is open circuited (due to a broken cable) 1 Which one of the following is acceptable for determining core quadrant power tilt under the 4 above conditions? l l Plant Process Computer.

l Movable incore Detector Manual calculations using operable excore detector Manual calculations using estimated current for N-43 lower detecto QUESTION: 051 (1.00)

Performance of EOP-1.2,"Small Break LOCA Cooldown and Depressurization,"is in progres What is the reason for starting both control rod shroud fans after depressurizing the RCS? To provide adequate cooling for the CRDM To reduce containment pressure and humidit To provide adequate cooling for the Nis detector To cool down the upper head region of the reactor vessel.

!

l

,

I

..

.

l

..

- REACTOR OPERATOR Page 31 QUESTION: 052 (1.00)

The following plant conditions exist:

.

Steam generator A steam flow: 0.0005 E6 lbm/hr and stable

-

Steam generator B steam flow: 0.6 E6 lbm/hr and decreasing

.

Steam generator A level: 70% and stable

• Steam generator B level: 10% and decreasing

-

Turbine driven AFW pump: running

-

Motor driven AFW pumps: running

. Tavg: 516*F and decreasing

-

Containment pressure: 5 psig and increasing

-

SI: actuated if no operator action has been taken, which ONE of the following indicates the status of the main steam isolation valves?

MSIV A MSIV B

, open open l open shut shut open shut shut l

,

!

I

i jl REACTOR OPERATOR Page 32

QUESTION: 053 (1.00)

During operation at 100% power, impulse pressure channel (PT-486) failed LOW. Which one of the following describes the response of the condenser steam dump control system to this failure and why? . The steam dump valves remain closed but are " armed" due to a loss of lead l condition being sensed.

f l The dump valves modulate open due to a Tavg/ Tref deviation generated by the t

loss ofimpulse pressure.

l - The steam dump valves trip open on a turbine trip signal being generated by the loss ofimpulse pressure.

L l The loss of impulse pressure would only have an affect on the steam dump if it j was operating in the " pressure" mode.

!

l

'

QUESTION: 054 (1.00)

During refueling operations inside containment, the control room receives an RE-211/212,

'

" Containment Air Particulate / Noble Gas Monitor," alarm. Which one of the following actions is REQUIRED to be performed?

a.- Verify Containment purge supply fans are running.

i Suspend all refueling operations inside containment.

l l Notify the Emergency Plan Coordinator to implement emergency plan.

i

.. Notify the Plant Manager of the need to perform a full plant evacuatio *4

,

'

i

REACTOR OPERATOP Page 33

<

QUESTION: 055 (1.00)

, RHR shutdown cooling was in progress when the 1P-10A RHR pump had to be secured due to L

mechanical failure. The remaining RHR pump (1P-108) is indicating erratic flow characteristics.

, Which one of the following actions would be the first operator action taken in accordance with

' SEP-1, " Degraded RHR System Capability,"?

.

!

l vent the operating RHR pump as necessary, i check for completion of RHR suction line refloo stop the RHR pump and isolate the RCS drain path adjust RHR system flow to betweeri 1300 gpm and 1500 gpm.

j QUESTION: 056 (1.00)

! Which ONE of the following is the basis for isolating the feedwater to a faulted SG7 l

' To maximize the energy release from the faulted S ' To maximize the cooldown capability from the non-faulted SG and minimizing RCS cooldow To prevent overflowing the faulted SG with feedwater and thus minimizing the cooldown for PTS concern To prevent thermal shock to the faulted SG "U" tubes and thus minimizing the potential of rupturing a SG tube with subsequent off-site release.

!

REACTOR OPERATOR Page 34 QUESTION: 057 (1.00)

The plant was in a normal configuration at 100% power when the following occurred:

• '

Supply breaker,1 A52-77,1 A-04 to 1 A-06 Bus Tie Breaker tripped.

Supply breaker,1 A52-76, 2A-03 to 2A-05 Bus Tie Breaker tripped.

Which of the following statements correctly identifies the subsequent electrical lineup? Only G01 and G04 start and supply buses 2A-05 and 1A- 06, respectively, Only G02 and G03 start and supply buses 2A-05 and 1 A-06 respectivel All four EDGs start with G01 supplying 2A-05 and G04 supplying 1 A-06 All four EDGs start with G02 supplying 2A-05 and G03 supplying 1 A-0 QUESTION: 056 (1.00)

Given that the Service water pump (P32A) and component cooling water pump (P11 A) are running:

At T=0 Containment pressure increased to 6 psig.

At T = +10 seconds (Ten seconds later) the supply breaker to 1B-03 tripped open.

As a result of the above conditions, what is the starting order of the service water pump (P32A)

and compo_nent cooling water pump (P11 A)?

P32A P11A  ! Restart at same time as P11A Restarted at same time as P32A j

.

I Restarted before P11A Restarted after P32A itestorted after P11A Restarted before P32A I Does not restart Does not restart l

'

<

!

!

!

!

l-l l REACTOR OPERATOR Page 35 l ' QUESTION: 059 (1.00)

Given the following conditions:

-

A large break LOCA has occurred inside containment .

-

Containment pressure initially increased to 30 psig, then began trending down and is currently 10 psi The spray addition tank level has decreased by 20%

.

What is the current status of the containment spray pumps and valves.

,

PUMPS DISCHARGE VALVES Running Full open L b.' Secured Closed Running Mid-position

!

di Secured Full open QUESTION: 060 (1.00)

If a safety injection signal occurred coincident with an undervoltage condition on bus 1-A05 which of the following correctly describes the response of P-38A, Motor Driven Aux Feedwater Pump, assuming all system operated as expecte P38A trips and must be manually starte P38A trips and sequences on (starts) approximately 10 seconds after 1 A-05 l- reenergizes.

i

' P38A trips and sequences on (starts) approximately 20 seconds after 1 A-05 ,

reenergizes.

!

L P38A continues to ru ,

!

!

l n REACTOR OPERATOR Page 36 QUESTION: 061 (1.00) >

Given the following plants conditions:

Unit 1 is operating at 100%, steady state powe *

CCW surge tank level has increased since the last log reading

High CCW pump inlet temperature alarms exist

High CCW pump inlet radiation alarms exist i These conditions describe a/an: RCP thermal barrier cooling coil leak. .

RifS. residual heat exchanger tube-to-shell side lea primary-to-secondary steam generator tube lea seal return heat exchanger tube-to-shell side lea QUESTION: 062 (1.00)

Given the following plants conditions:

• The reactor was at 80% powe * Circulating water pump, P30A was out of servic At 0900, a down power was commenced at 0.5%/ min for the hour.

l At 1003 excessive vibrations caused the P308 pump to trip. Which one of the following describes the expected response and the correct reason for thi1 respons The reactor ... trips; the P-9 permissive was active.

1 does not trip; the P-10 permissive was activ , trips; reactor power was greater than 55% powe does not trip; the P-9 permissive was automatically blocke .,

,

e I l l REACTOR OPERATOR Page 37 l

!

QUESTION: 063 (1.00)  !

You have completed a hand frisk which found no contamination after you alarmed the PCM-1B !

at the RCA exit. Which of the following actions are you now required to take? Proceed directly to the portal monitors since the hand frisk. indicated no contaminatio i Recount once in PCM-1B, if no alarms occur during this recount, then proceed to the portal monitor l

. Contact your supervisor to resolve the discrepancy between the alarming PCM-1B and non-alarming hand fris : Recount twice in the PCM-1B, if no alarms occur during both of these recounts, then proceed to the portal monitor !

i i

QUESTION: 064 (1.00) .

The heater drain tank was vented, depressurized, and opened for personnel access. You are -

the attendant at the personnel access point for the confined space. A short time after a mechanic entered the tank, the atmosphere sampling instrument alarmed and you noticed that the mechanic in the confined space was unconscious. In accordance with NP 1.9.4," Confined Spaces Procedure," what is your required action?

a, contact the security coordinator to perform a rescue of the mechani call for a backup attendant, and attempt to rescue the mechanic yoursel contact the fire protection and safety coordinator to sample the confined space, contact the control room operator to summon rescue and other emergency services.

!

'

L l REACTOR OPERATOR- Page 38

!

QUESTION:065 (1.00)

~ You are performing a monthly surveillance on the AFW system for the 'A' steam generator.

L Under which one of the following conditions should you initiate a temporary change? It becomes necessary to abort the AFW surveillance procedure prior to completio The AFW surveillance procedure incorrectly directs you to manipulate switch 1-P38B instead of 1-P38 You have completed section A and B of the AFW surveillance procedure and are starting Section C. Step C.1 directs you to record bearing temperatures that are ONLY required for the quarterly tes You have 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> before having to enter TS 15.3.0.8 due to the improper performance of the previously performed AFW surveillance and a one-time change to the AFW surveillance procedure is required which does not change the intent of the procedur QUESTION: 066 (1.00)

.Which one of the following describes how you would perform an independent verification on an OPEN manual valve? tum the valve handwheel in the open direction unit there is no more valve movemen tum the valve handwheel in the closed direction enough to observe valve movement, then tum the valve in the open direction until there is no more valve movemen tum the valve handwheel in the closed direction two full turns, then tum the valve in the open direction until there is no more valve movemen tum the valve handwheel in the closed direction until the control room receives a dual position indication for the valve, then tum the valve in the open direction until

. there is no more valve movemen fi

!

!

REACTOR OPERATOR Page 39 QUESTION: 067 (1.00)

Your current annual exposure is 1750 mrem. If you were assigned a task in an area where the dose rate is 80 mrem /hr, what is the maximum number of hours you could spend in the area without exceeding your ANNUAL administrative limit? (You do not have any authorized extensions.) hours hours hours hours QUESTION: 068 (1.00)

Given the following conditions on Unit 1:

-

Tm= 558'F and slowly decreasing )

Power = 45% and stable

'

=

-

Pressurizer pressure = 2000 psig and stable a

Condenser vacuum has degraded to 19" Hg With no operator action, what is the plant response? Turbine trip ONL Annunciator low vacuum ONL The turbine will trip, and then the reactor will trip, The reactor will trip, and then the turbine will tri *

,

!

E

,

,..

! REACTOR OPERATOR Page 40 i

l- )

,

_ QUESTION: 069 (1.00),

Given the following plant conditions on Unit 1:

• -

' A LOCA has occurred

The reactor has tripped Pressurizer pressure is 1685 psig and stable 1 minute after the above conditions occur,

' service water pumps P-32A, P-328, and P-32D have not started.

..

Which of the following valves is in its correct position? SW-2880, unit 1 turbine building feeder valve is OPE SW-2816,- service building air conditioning supply valve is OPE SW-2907, containment fan service water outlet valve is CLOSE SW-2930A, spent fuel pool heat exchanger outlet valve is CLOSE QUESTION: 070 (1.00)

A fast spreading fire exists in the unit i turbine lube oil system reservoir. Which one of the following automatic fire suppression systems will deploy? Halon' system.

l Deluge sprinkler syste Dry pipe sprinkler system.

l Wet pipe sprinkler system.

l'

!

-.

REACTOR OPERATOR Page 41 l

l QUESTION: 071 (1.00)

l Given the following plant conditions on Unit 2:

-

Unit 2 is in hot shutdown following a reactor trip 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ag Unit 2, A steam generator is at 60% level and increasing

-

Unit 2, B steam generator is at 25% level and decreasing

-

Both motor driven AFW pumps (P38A and P388) have started After the AFW pumps started, a pipe rupture occurred dodtream of the suction check valve to the AFW pump, P38B. How will the Unit 2,,1B steam generator level be affected?

q j

The Unit 2, B steam generator level will ..... j decrease until operator action is taken to start the steam-driven auxiliary feedwater pump,2P2 increase due to flow from automatic starting of both steam driven feedwater pumps,1P29 and 2P2 ]

j

! decrease until service water valve AF-4016 is manually opened by the unit control operator in the control roo decrease until service water valve,2AF_-40_06 automatically opens when the Unit .

2, B S/G level reaches 20%.  !

..

l

l- REACTOR OPERATOR Page 42 l

QUESTION: 072 (1.00)

Given the following plant conditions on Unit 1:

S/G A has a partially stuck open Atmospheric Steam Dump Valve

Both motor-driven AFW pumps (P38A and P38B) started when level decreased to 25% i in the Unit 1, A S/G After the start of the AFW pumps, annunciator C01 A 4-9, " AUX FEED PUMP SUCTION i

'

PRESSURE LOW," alarmed. The B AFW pump (P38B) tripped due to a failed low AFW pump suction pressure switch. What actions are required to override the B AFW pump (P388) low suction pressure trip? Select a different pressure switch for the 'B' AFW pump and take the switch to the auto positio I Take the_ pump control switch for the 'B' AFW pump to start and return the switch to the auto positio Take the pump control switch for the 'B' AFW pump to pullout, and then place the i switch in the auto positio ] Wait for a 30 second time delay, pump will then automatically restart after placing the pump control switch for the 'B' AFW pump to auto position, QUESTION: 073 (1.00) ,

Which of the following conditions will result in the automatic closure of waste distillate overboard valve, FCV-1.W-157 waste distillate pump trip, high levelin the waste holdup tan high alarm on RE-218, waste disposal system liquid monitor.

l high alarm on RE-223, waste distillate tank overboard monitor.

l l

.

I.

l

I l

REACTOR OPERATOR Page 43 l

l l QUESTION: 074 (1.00)

It's 1530 and a release of the "A" monitor tank is in progress.

l

'

You receive a high alarm on RE-218, waste disposal system liquid monitor. In accordance with RMSASRB Cl RE-218 "Radiatiore Monitoring System Alarm Setpoint & Response Book," you MUST perform all the following actions EXCEPT: re-check release calcolatio notify RP supervision (Duty and Call). recommence discharge orice the high alarm clear verify shut RCV-018, waste I: quid overboard valv QUESTION: 075 (1.00)

Degassing of the reactor coolant system has comraenced. Gas Decay Tank #1 is in service and at 80 psig and increasing at 4 psig/ hour. Four houro after commencing degassing, you note the following valve alignment:

1. 1 GDT inlet valve is closed

=

1. 2 GDT inlet valve is open What caused the valve alignment? #1 GDT tank pressure increased to 95 psig, b. _ receipt of a high alarm on RE-114, El. 26' PAB West area monito back pressure on the waste gas compressor decreased to 35 psi RE-214, auxiliary building vent exhaust gas monitor, came in violet on RMS displa I

!

i I

l

!

r REACTOR OPERATOR Page 44 QUESTION: 076 (1.00)

Control room ventilation has re-aligned to Mode 4. The following conditions are indicated on the RMS display:

RE-235, control' room noble gas monitor is YELLOW

• _ RE-101, control room monitor is RED

-

RE-234, control room iodine monitor is RED

-

RE-214, auxiliary building vent exhaust gas monitor is RED What caused the Mode 4 realignment of CR ventilation? RE-235 being in aler RE-214 being in high alar RE-101 being in high alar RE-234 being in high alar QUESTION: 077 (1.00)

Given the following indications on the Unit 1 RMS display

-

RE-116, Demineralizer Valve Gallery Area monitor, is RE *

1RE-102, EL. 66' containment low range monitor, is LIGHT BLU New fuel receipt inspection is in progress in addition, the following conditions exist:

Annunciator,1C20 B 1-9," COMMON AREA RADIATION MONITOR HIGH"is lit

Containment sump level is zero and stable What is the cause for the Common Area Radiation Monitor High annunciator being lit?

- having the gas stripper onlin ' loss of reactor coolant in containmen high pressure in the letdown line causing a diaphragm valve to lea > fuel was inadvertently dropped while being placed in the new fuel vaul . ._

- ___-

.

. .

.

REACTOR OPERATOR Page 45 QUESTION: 078 (1.00)

During natural circulation cooldown, it is possible for a bubble to form in the reactor vessel hea The existence of a bubble in the reactor vessel is normally indicated by the Reactor Vessel Level Indication System (RVLIS). If RVLIS is not available, which one of the following post accident monitoring instruments would provide the first indication nf the existence of a bubble in the reactor vessel head? Reactor Coolant System (RCS) pressure (wide range), Core exit thermocouple Pressurizer water level (narrow range). RCS hot leg temperature (wide range).

QUESTION: 079 (1.00) .

Given the following plant conditions:

-

100% power

-

RCP A and B #1 seal return flow is returning to the VCT

Instrument air pressure has dropped to 75 psig and is still rapidly decreasing Without operator action, what is the tesponse of RCP seal return flow? RCP seal return flow will fill and pressurize the: PRT, and cause a relief valve (CV-314) to PRT to lift, sending the water in the PRT to the RCD b, seal return line and cause a relief valve (CV-314) to PRT to lift, sending the water in the seal retum line to the PR RCDT, and cause a relief valve (CV-314) to PRT to lift, sending the water in the RCDT to the 'B' containment sump.

, seal return line, and cause a relief valve (CV-314) to PRT to lift, sending the water in the seal retum line to the RCDT.

!

._

p

!-

l l REACTOR OPERATOR Page 46 l

,

QUESTION: 080 (1.00)

Following a reactor power increase from 80% to 90% power, the following events occur on Unit 2:

'

• Unit 2 Loop B S/G level channel, L471 failed low

-

Main feedwater pump suction pressure dropped to 180 psig and is stable

-

Pressurizer pressure dropped to 1910 psig and is stable Based on the above conditions, without operator action, the plant will experience which of the following transients: a safety injection due to low pressurizer pressur a feedwater isolation on high level in the Unit 2 B S/ a feedwater isolation on high level on the Unit 2 A S/ a reactor trip due to main feedwater pumps tripping on low suction pressur l QUESTION: 081 (1.00)

l Given the following initial conditions:

! 1

-

iP29, Unit 1 turbine-driven AFW pump, received an automatic start signal and is currently injecting at 50 gpm into each S/G on Unit 1

+

TDAFW mini-recirculation flow is 135 gpm What will the effect be on indicated AFW system flow if a loss of instrument air to the recirculation valve for 1P29 occurred? remain unchanged.

l increase to new valu decrease to new valu , Initially decrease then return to original valu .

REACTOR OPERATOR Page 47 QUESTION: 082 (1.00)

An RCS cooldown is in progress. RHR has been placed in service for shutdown cooling. The following plant conditions exist:

j

RCS pressure: 380 psig and slowly decreasing

-

RCS wide range temperature - Hot Leg 340*F and slowly decreasing

• -

Flow controller RHR-626 in automatic: 10% open Local RHR Heat Exchanger outlet temp:

-

320*F and slowly decreasing

.

If the RHR return header flow transmitter (FT-626) failed LOW under these conditions, which one of the following would correctly describe the plant's response? RHR Heat Exchanger flow control valves (1FCV-624/625) will automatically close to attempt to prevent exceeding RCS cooldown rate of 100*F/h ) RHR retum header isolation valve (1RH-720) will automatically open to attempt to maintain RHR pump discharge pressure and flow constant to the reactor vesse RHR Heat Exchanger bypass valve (1FCV-626) will position to 100% full open to ,

attempt to maintain RHR desired system flow rate and RCS cooldown rate will 1 decreas I

! RHR Heat Exchanger bypass valve (1FCV-626) position will change to 100% full I close to attempt to maintain RHR desired system flow rate and RCS cooldown rate will decreas i l

l i-

REACTOR OPERATOR Page 48 QUESTION: 083 (1.00)

How will the AFW system respond if Unit 1, steam generator A, level lowers to 25% due to a partially opened S/G Atmospheric Steam Dump Valve when the plant is at 100% power?

.(Assume B S/G level remains at 64%.) .

I Only the P38A motor driven AFW pump will start; only discharge valve 4022 discharge to Unit 2, A S/G will ope i Only the P38B motor driven AFW pump will start; only discharge valve 4023 discharge to Unit 1, A S/G will ope I Both turbine-driven AFW pumps (1P29 and 2P29) will start; both discharge valves 1MOV-4001 and 2MOV- 4001, to Unit 2 A S/G will ope Both the P38A and P38B motor drive AFW pump will start; both discharge 4 valves: 4021 discharge to Unit 1, B S/G and 4023 discharge to Unit 1, A S/G will ope QUESTION: 084 (1.00)

You are an extra operator filling in for a sick non-licensed operator. While performing your plant rounds (first of shift), a LIMITED plant evacuation was announced due to a dropped fuel assembly. You may reenter the area and resume conducting your rounds after approval is obtained from the .... Plant Manager and the Duty Shift Superintenden Duty Shift Superintendent and the RP superviso RP Supervisor and the Shift Technical Adviso Duty and Call Superintendent and the Plant Manage p.-

REACTOR OPERATOR Page 49 QUESTION: 085 (1.00)

A Unit 1 plant cooldown is in progress in accordance with EOP 0.2, " Natural Circubtion Cooldown." The following plant conditions exist:

Both CR shroud fans: running

-

RCS hot leg temperatures: 446*F and decreasing

RCS cold leg temperatures: 430*F and decreasing a

RCS pressure: 900 psig and decreasing in accordance with EOP 0.2, " Natural Circulation Cooldown," the operator is required at this time to: verify safety injection unblocke establish nitrogen backup to the PORV Isolate reactor coolant pump seal injectio temporarily restore power to both accumulator discharge valves.

l QUESTION: 086 (1.00)

l Following a Unit 1 plant trip at 0900, the following plant conditions exist:

l -

Reactor Trip Breakers indicate open in the control roo *

Three (3) control rods are not fully inserted.

=

'

Emergency boration was commenced in EOP 0.1, " Reactor Trip Response."

+

The Boric Acid Storage Tank level was at 90%.

l -

Both Boric Acid Transfer pumps,1-P4A and I-P4B tripped and cannot be restarted after BAST TBA decreased by 47%

What action is required in accordance with AOP-6E, "Altemate Boration/ Loss of Shutdown Margin", after both of the Unit 1 BAT pumps tripped? de-energize rod drive MGs Inject 6,000 gallons from RWST inject 1,000 gallons from Boric Acid Storage Tank, T6C dispatch operator to locally open Reactor Trip Breakers .

REACTOR OPERATOR Page 50 QUESTION: 087 (1.00)

During a liquid release, which ONE of the following conditions would require compensator actions to be taken? . RE-223, Waste Distillate Tank Discharge, out-of-service while discharging radwaste, RE-219, S/G Blowdown Liquid Discharge, and RE-222, Blowdown Tank Monitor, out-of-service while SG blowdown in progress, RE-229, Service Water Discharge, out-of-service during unit operatio . RE-230, Retention Pond Discharge, out-of-service during a retention pond discharg QUESTION: 088 ' (1.00)

The plant is recovering from a reactor trip. The following plant conditions exist:

-

RCP seal water outlet temperature: 190*F and rising

+ RCS pressure: 1500 psig and stable

-

iP-1 A RCP No.1 seat leakoff: 0.60 gpm and stable

-

iP-1 A RCP No.1 seal leakoff valve (MOV-270A): open

-

Sealinjection from the 1P-1A RCP: 8 gpm and stable

• iP-1 A RCP shaft vibration: 13 mils and stable Annunciator 1C031D 3-1, "1P-1A OR B RCP NO 1 SEAL WATER OUTLET TEMPERATURE  !

'

HIGH," eriergized. Which of the following actions is required in response to these conditions? open 1CV0386,1 A and 1B RCP seal water bypass control valv ' increase CCW flow through the A RCP thermal barrier heat exchange increase seal injection flow to 12 gpm to the A RCP by throttling open 1CV-300 trip the 1P-1 A RCP and shut 1CV-270A, RCP Number 1 Seal Return isolation valve after 3 minute .,

p l l

l l REACTOR OPERATOR Page 51 QUESTION: 089 (1.00)

Concerning the Radiation Monitoring System, which one of the following indicates normal operation of the Control Room and the TSC Control Terminals (cts)? All six (6) control terminal lights are O l

' All six (6) control terminal lights are OF Control Terminal " monitor" and " control" lights are CYCLING. O N 4eb ofF Control Terminal " monitor" light is OFF, and " control" light is O QUESTION: 090 (1.00)

After operating at 30% power for four days, the plant was ramped to 40% power over a one hour period. Power was stabilized at 40% power and cuatrol rods were manipulated to maintain Tavg. If RCS boron concentration and Tavg are maintained constant, which one of the following describes which direction control rods will need to be moved during the next hour in order to maintain Tavg constant? OUT, because Xenon production by direct fission yield will be temporarily dominan OUT, because Xenon production by decay of krypton will be temporarily dominan , IN, because Xenon removal by decay will be dominan IN, because Xenon removal by bumout (neutron absorption) will be temporarily i dominan j i

I I

l I l

l f

!

!

n

~

i l

REACTOR OPERATOR - Page 52 j l

l QUESTION: 091 (1.00) l

!

Which one of the below listed DC control power alignments will result in. overloading a DC power !

supply if EDGs G03 and G04 fields' flashed at the same time? j control power for Bus 1 A06 in normal, control power for Bus 2A06 in norma control power for Bus 1 A06 in altemate, control power for Bus 2A06 in norma l control power for Bus 1 A05 in normal, control power for Bus 2A05 in norma l l control power for Bus 1 A05 in alternate, control power for Bus 2A05 in norma j QUESTION: 092 (1.00)

!

Which of the following valves will fail open on a loss of instrument air? l I RCP Therma' harrier Isolation, CC-761 I L'? 2 radw_sts ^.CW supply, CC-LW-6 l Si accumulator ventiN? supply, SI-834A, Turbine-driven auxiliary feeuc'ater pump recirc valve, AF-400 ]

.

,

REACTOR OPERATOR Page 53

.

QUESTION: 093 (1.00)

l With the plant in solid operations, the following plant conditions exist: '

RCS temperature: 365'F and decreasing

'

RCS pressure: 375 psig and decretssing

,

!

Low temperature over pressure mitigation system is in service PORV 431C control switch is in the auto position PORV 430 control switch is in the auto position LTOP enable keyswitch #1 is in the ON position LTOP enable keyswitch #2 is in the ON position  ;

What is the plant response if pressurizer pressure instrument, PT-493 fails HIGH7 ONLY Pressurizer PORV FCV-431C ope ONLY Pressurizer PORV FCV-431C opens and all pressurizer heaters de-energiz ONLY Pressurizer PORV FCV-431C opens and both pressurizer spray valves ope ONLY Pressurizer PORV FCV-431C opens, all pressurizer heaters de-energize, and both pressurizer spray valves open.

!~

l

.. . . . _ - _ _ - _ _

. . . . . , , ,

REACTOR OPERATOR Page 54

,

QUESTION: 094 (1.00)

Given the following Unit 2 initial conditions:

.

The Unit is at 70% power and stable

.

Rod controlis in MANUAL

.

Pressurizer levelis at 38% and stable

.

Tavg is at 563*F and stable Without operator action, what will be'the effect on pressurizer level if Loop B cold leg temperature detector TE-4018 fails low?

Pressurizer level will ....... decrease to 20% and stabiliz remain unchanged and stabl Increase to 46% and stabiliz increase to 100% and stabiliz QUESTION: 095 (1.00)

Given the following Unit 1 plant conditions:

-

The reactor is shutdown a

Decay heat is being removed by natural circulation

-

RCS pressure is 1050 psig

-

Average core thermocouple temperature is 450*F How much subcooling exists in the RCS at the above conditions? * ' * ' g )

i . REACTOR OPERATOR Page 55

' QUESTION: 096 (1.00)

Given the following initial plant conditions:

-

Normal CCW lineup with P-11 A CCW pump running

-

Letdown flow temperature high -

-

Letdown high temperature divert valve,1CV-145, is bypassing demineralizers

'

Which one of the following are you required to perform after entering AOP-98 due to a pipe rupture of CCW at the inlet to the non-regenerative heat exchanger? Tum off all pressurizer heaters, Start the standby component cooling water pump, P-11 Isolate RCP thermal barrier cooling retum valves,1CC- 761 A, Start a reactor makeup service pump and open 1CC-815, " Emergency Makeup Valve."

QUESTION: 097 (1.00)

Given the following:

-

Inverter 1DYO3 is to be taken OOS

Operators have shifted WHITE buses 1Y03, "120VAC" and 1Y103, "120VAC" to the attemate inverter DYOC, "120VAC" In accordance with 01-37, " Shifting of instrument Supply Bus Feeders," after the shift to the alternate inverter (DYOC) which one of the following is a required action? Verify the hydrogen computer's trouble alarm is ILLUMINATED on panel 1C-20 ASI Install magnetic administrative control tag for UNIT 1 buses 1YO3, "120VAC" and 1Y103, "120VAC."

,

- Open the electrical interlock between Unit 1 (1YO3 and 1Y103) and Unit 2 (2Y03 I

and 2Y103) instrument buses.

l Establish a fire watch inspection twice per shift of auxiliary feed pump room, fire Zone 304 and remote shutdown panel area (W46 and W-46A).

i l

REACTOR OPERATOR Page 56 )

QUESTION: 098 (1.00) l The following conditions exist on Unit 1: )

Turbine load is 15%

• Reactor power is 15% .

PZR pressure is 1985 psig

PZR levelis 24%

-

'A' Train of Secondary equipment is aligned for operation

- CO2 D 3-4, "4.16KV BUS LOCKOUT UNIT 1" was received

-

All three RCS Loop A flows (FT-411,412,413) are significantly less than all three RCS i Loop B flow l Which one of the following MANUAL actions is required?

! initiate safety injectio start the third charging pum place the reactor in hot shutdow reduce turbine load to 10% and trip the turbin .,

REACTOR OPERATOR Page 57 QUESTION: 099 (1,00)

Given the following PR NI readings Ni-41 NI-42 NI-43 NI-44 10% 8% 9% 8%

Subsequently, a slight transient occurs, resulting in the following readings:

NI-41 NI-42 Ni-43 NI-44 11 % 8% 10% 8%

No additional operator action is taken. What is the status of the IR high flux reactor trip and the PR high flux low range reactor trip?

IR HIGH FLUX REACTOR TRIP PR HIGH FLUX LOW RANGE REACTOR TRIP Blocked Blocked Unblocked Unblocked Blocked Unblocked Unblocked Blocked i

l

I

'

REACTOR OPERATOR Page 58 QUESTION: 100 (1.00)

While refueling operations are in progress, a package is submitted which would de-energize 120Vac buses: (1YO1 and 1Y101) and (1Y03 and 1Y103), so that work can be performed on PRN141 and 427 How would this affect refueling operations if the IN40 gammametric detector is out of service? fuel moves can not continue, because only one source range channel is

. energize fuel moves can continue, since power range indication is not needed during -

refuelin ]

i fuel moves can not continue, because no source range channels are energize fuel moves can continue, as long as P-10 bistables are tripped pt r applicable '

I & C procedure !

l (*""""* END OF EXAMINATION """*"*)

.

l

REACTOR OPERATOR Page 59 l

ANSWER: 001 (1.00) ANSWER: 006 (1.00) ANSWER: 011 (1.00) REFERENCE: REFERENCE: REFERENCE:

OM 3.10, Revision 10; RP AOP-10A, Revision 23 j 000057A219 ..(KA's) 1C, Revision 41 000068G ..(KA's)

ANSWER: 002 (1.00) 194000K ..(KA's) REFERENCE: ANSWER: 007 (1.00) ANSWER: 012 (1.00)

CSP-C.1, Rev. DRAFT, " Response to inadequate REFERENCE:

Core Cooling" NP 4.2.14, Revision 1 REFERENCE:

000074K201 ..(KA's) 194001K ..(KA's) 000061G ..(KA's)

ANSWER: 003 (1.00) ANSWER: 008 (1.00) ANSWER: 013 (1.00)

REFERENCE: !

REFERENCE: REFERENCE: l MDB 3.2.5, U1,1830 TS 81, Revision 55; TS 82, TRHB 11.2, Rev. 6 Revision 1; PB FSAR Chapter Revision 51; TS 83, Revision Modified, , TS 84, Revision 7 056000A204 ..(KA's) )

064000K202 ..(KA's) 194001K ..(KA's)

ANSWER: 014 (1.00) ,

'

ANSWER: 004 (1.00) ANSWER: 009 (1.00) REFERENCE:

REFERENCE: REFERENCE: TRHB 13.7, Revision 2; TRHB 11.8, Revision 6; TRHB 11.3, Revision 3 -

1-SOP-CC-001, Revision 0; Technical Specification Modified,1996 PB Exam, P&lD West 110E01 .3.3. QNUM 47520 008000K409 ..(KA's) 076000G ..(KA's) 059000K419 ..(KA's)

ANSWER: 005 (1.00) ANSWER: 010 (1.00) ANSWER: 015 (1.00) REFERENCE: REFERENCE: REFERENCE:

AOP-8F, Revision 5 TRHB 10.3, Revision 3; LP OM 1.1, Attachment 4, 2438, revision 0, Logic Sheet Revision 2; OM 3.9, Rev 5 033000A203 ..(KA's) 18 Direct,1996 PB Exam, 194000K ..(KA's) QNUM 47593 000027A101 ..(KA's)

<

-

l

( REACTOR OPERATOR Page 60 l

l ANSWER: 016 (1.00) ANSWER: 020 (1.00) ANSWER: 024 (1.00)

a.

i REFERENCE: REFERENCE: REFERENCE:

! AOP-10A, Attachment C and AOP-18, Revision 10 TRHB 10.3, Revision 3; J

D, Revision 23 Direct,1994 PB Exam, OP-4C, Revision 13; ARB Direct,1994 PB Exam, QNUM 38657 1C1-1, Revision 4 QNUM 38727 Modified,1995 PB Exam, 003000A201 ..(KA's) QNUM 45738 000068K203 ..(KA's)

,

]

ANSWER: 021 (1.00)

ANSWER: 017 (1.00) c. _ REFERENCE: ANSWER: 025 (1.00)

REFERENCE: TRHB 10.12, Revision 4; AOP 6C, Revision 8 STPT, Section 2.4, Revision REFERENCE:

Direct,1995 PB Exam, 2; Logic Sheet 9 TRHB 13.8, Revision 1; LP QNUM 45759 Direct,1996 PB Exam, 2441, Revision 0; Logic Sheet

)

QNUM 47961 16 )

000001A205 ..(KA's) Direct,1996 PB Exam, I 026000A301 ..(KA's) QNUM 47605 l

ANSWER: 018 (1.00) 000001K105 ..(KA's) ANSWER: 022 (1.00)

'

REFERENCE: TRHB 13.6, Revision 2; REFERENCE: ANSWER: 026 (1.00)

Logic Sheet 18 TRHB 11.14, Revision 8 Direct,1996 PB Exam, Direct,1995 PB Exam, REFERENCE:

QNUM 47562 QNUM 45773 CSP-l.3, Revision 7; BG CSP-l.3, Revision 7 000028K202 ..(KA's) 086000K402 ..(KA's) Modified,1994 PB Exam, QNUM 38735 ANSWER: 019 (1.00) ANSWER: 023 (1.00)

b, ANSWER: 027 (1.00)

REFERENCE: REFERENCE: TRHB 10.7, Revision 9; REFERENCE:

TRHB 10.6, Revision 6;MDB STPT 2.3, Revision 3; STPT SEP-1.1, Revision !

3.2.4, 809, Revision 4 8.2, Revision 6, STPT 11.2, Modified,1996 PB Exam, !

Direct,1996 PB Exam, Revision 3 QNUM 47555 I QNUM 47510 Direct,1996 PB Exam, )

j QNUM 47535 000025K101 ..(KA's)

!

-

004000K203 ..(KA's) l l

005000K109 ..(KA's)

..

.

I

I REACTOR OPERATOR Page 61 l

l ANSWER: 028 (1.00) ANSWER: 032 (1.00) ANSWEP.: 036 (1.00)

l b b I

REFERENCE: REFERENCE: REFERENCE:

EOP-0, Revision 27 AOP-0.0, " Vital DC System TRHB 10.1, " Primary New Malfunction," Rev 13, Step System Description

Reactor i 2.5, p Containment," Rev j ANSWER: 029 (1.00) 6, p 15; i b 000058A203 ..(KA's) TRHB 10.16, " Primary l REFERENCE: System Description:

BG ECA-0.0," Loss of All AC Engineered Safeguard Power, " Rev. DRAFT, p 57 ANSWER: 033 (1.00) Systems," Rev. 4, p.21 a

000055K302 ..(KA's) REFERENCE: 022000A301 ..(KA's)

TRHB 13.3, " Instrument and Control Systems Description:

ANSWER: 030 (1.00) Reactor ANSWER: 037 (1.00)

d Protection (Reactor Trip)," d REFERENCE: Rev. 4, p 3 REFERENCE:

EOP-0, " Reactor Trip or TRHB 13.3, " Instrument and Safety injection," Rev 001000K603 ..(KA's) Control Systems Description:

DRAFT, Step 24, p Reactor 21; BG EOP-0, " Reactor Protection (Reactor Trip),"

Trip or Safety injection," ANSWER: 034 (1.00) Rev. 4, FIG 13.3.4, " T and DRAFT, p. 43 d Tavg instrumentation" REFERENCE:

000009K323 ..(KA's) TRHB 13.4, " Instrumentation 002000K512 ..(KA's)

and Control Systems:

Engineered ANSWER: 038 (1.00)

ANSWER: 031 (1.00) Safety Features Actuation c d instrumentation (Safeguard REFERENCE:

REFERENCE: system)," Rev 4, p 3 TRHB 13.3, " Instrument and EOP-3, " Steam Generator Control Systems Description:

Tube Rupture," Rev DRAFT, 013000A101 ..(KA's) Reactor Protection (Reactor Caution before Trip)," Rev. 4, p. 36; stpt 4.2, step 31, p 2 Re Background Document ANSWER: 035 (1.00)

EOP-3," Steam Generator K106 ..(KA's)

Tube Rupture," Rev REFERENCE:

DRAFT, p 6 ANSWER: 039 (1.00)

MDB 3.2.3, Rev 8 K309 ..(KA's) WEST 883D195 REFERENCE:

AOP-6B, " Stuck Rod or 013000A403 ..(KA's) Malfunctioning Position Indication Unit 1,"

Rev.10,p A201 ..(KA's)

,

I l l

l REACTOR OPERATOR Page 62 ANSWER: 040 (1.00) ANSWER: 044 (1.00) ANSWER: 048 (1.00)

l a.

l REFERENCE: REFERENCE: REFERENCE:

EOP-1.2, "Small Break OP-5A, " Reactor Coolant LOCA Cooldown and EOP-0, " Reactor Trip or Volume Control," Rev. 30. Depressurization Unit 1," Safety injection," Re Rev. DRAFT, Foldout Page DRAFT, pg 7 and 8 -THRB 10.6, " Chemical and Volume Control System."

000009A234 ..(KA's) 000007A201 ..(KA's) Rev. 6, K307 ..(KA's)

ANSWER: 045 (1.00)

ANSWER: 041 (1.00) REFERENCE: ANSWER: 049 (1.00)

REFERENCE: OP-3A, " Normal Power Steam Tables (Material REFERENCE:

Operation to Low Power required for the examination) TRHB 13.8, " Rod Speed Operation," Rev. 43, and Direct Control." Rev.1, p K101 ..(KA's) pgs. 2, 3, and 4 004000K601 ..(KA's) 035000K501 ..(KA's)

ANSWER: 046 (1.00) ANSWER: 042 (1.00) REFERENCE: ANSWER: 050 (1.00) TRHB 10.8, " Safety REFERENCE: Injection System." Rev. 6, REFERENCE:

BG CSP-l.3, " Response to pgs. 5,10, and 17 REI 13.0, " Quadrant Power volds in Reactor Vessel," Tilt," Rev.14, Rev.7,p K603 ..(KA's) AOP-6H, " Quadrant Pow-er Tilt," Step 2 - RNO 016000G ..(KA's) I ANSWER: 047 (1.00) 015000K516 ..(KA's) ] ANSWER: 043 (1.00) REFERENCE: TRHB 13.13, " Rod Insertion ANSWER: 051 (1.00)

REFERENCE: and Delta T Deviation TRHB 10.1, " Primary Alarms." Rev.1, REFERENCE:

System Description: Reactor pgs. 2,3, and 4 BG EOP-1.2, "Small break Containment," Rev. 6, pgs.14 LOCA Cooldown and and 15 000003A101 ..(KA's) Depressurization," Rev.18, p.50 022000G2,1 ..(KA's)

022000K405 ..(KA's)

.,

V

r REACTOR OPERATOR Page 63 ANSWER: 052 (1.00) ANSWER: 056 (1.00) ANSWER: 061 (1.00) b a REFERENCE: REFERENCE: REFERENCE:

TRHB 13.4,"ESF Actuation BG EOP-2, " Faulted Steam FSAR, Section 6.5-8, page Instrumentation (Safeguard Generator isolation," Re .5-8 System), DRAFT, p.14 Rev. 4, pgs. 2,5, and 6 003000A408 ..(KA's)

000040K304 ..(KA's)

039000K408 ..(KA's)

' ANSWER: 062 (1.00)

ANSWER: 057 (1.00) ANSWER: 053 (1.00) d REFERENCE: REFERENCE: FSAR, Section 7.2, Page REFERENCE: FSAR, chapter 8, page 8.8-3 7.2-16 TRHB 13.9, " Instrumentation and and Control System 075000K307 ..(KA's)

Description: 062000K102 ..(KA's)

Condenser Steam Dump, "

Rev.3, ANSWER: 063 (1.00)

ANSWER: 058 (1.00) d l 04100K603 ..(KA's) d REFERENCE:

REFERENCE HP1.11.1, " Personnel FSAR, chapter 8. Page Contamination Monitor l ANSWER: 054 (1.00) 8.8-6,7 [3.6/3.7). (PCM-1B), i i Contamination Alarm REFERENCE: 008000A308 ..(KA's) Response, Rev.13, p 4,5 AOP-8B, " Irradiated Fuel Handling Accident in 194000G ..(KA's)

Containment." Rev 6, pgs 1-4 ANSWER: 059 (1.00)

a or d 000036G ..(KA's) REFERENCE: ANSWER: 064 (1.00) !

EOP 1.3, Rev. 20, Step d REFERENCE:

ANSWER: 055 (1.00) 026000A208 ..(KA's) NP 1.9.4, " Confined Spaces Procedure," Rev. 3, p. 5,10 REFERENCE:

j'

SEP-1, " Degraded RHR ANSWER: 060 (1.00) 194000G ..(KA's)

System Capability," Rev.1, b ,

p.3,5 REFERENCE- -

FSAR, Chapter 8, page ANSWER: 065 (1.00)

000025A103 ..(KA's) 8.8-6 d REFERENCE:

064000A307 ..(KA's) NP 1.2.3, " temporary '

changes," Rev. 6, p.2-5 194000G ..(KA's)

E REACTOR OPERATOR Page 64

- ANSWER: 066 (1.00) ANSWER: 070 (1.00) ANSWER: 074 (1.00)

l b c l REFERENCE: REFERENCE: REFERENCE:

NP 2.1.2, " Independent . TRHB 11.14, " Secondary RMSASRB Cl RE 218 Verification and Concurrent Systems Descriptions: Fire " Radiation Monitoring system Checks," Re Protection System," Rev. 8, Alarm Setpoint &

0, p. 8 pgs.3,4,,20 Response book Channel ,

Information Sheets," Rev. 2 194000G ..(KA's) 000067K102 ..(KA's) ' Waste Disposal System Liquid Monitor ANSWER: 067 (1.00) ANSWER: 071 (1.00) 194000G ..(KA's) , REFERENCE: REFERENCE:

Site NGET training TRHB 11.4, " Secondary ANSWER: 075 (1.00)

Systems Descriptions: a 194000G ..(KA's) auxiliary Feedwater REFERENCE:

System," Rev. 6, p.11 TRHB 10.15, " Primary Systems Descriptions: waste l ANSWER: 068 (1.00) 061000K302 ..(KA's) Disposal j c System," Rev. 7, p.22 REFERENCE:

Setpoint 3.1,1 ANSWER: 072 (1.00) 071000A416 ..(KA's)

Logic diagram 883D195, b ,

Sheet 12 REFERENCE: 1 Ol 62A, " Motor-driven ANSWER: 076 (1.00)

000051A202 ..(KA's) Auxiliary Feedwater System (P-38A & P-38B)," Rev.16, REFERENCE:

pg.9,20,21 TRHB 13.12, ANSWER: 069 (1.00) " Instrumentation and Control d 061000A204 ..(KA's) System Description:

REFERENCE: Process and Area Radiation TRHB 10.16, " Primary Monitoring System," Rev. 3, Systems Descriptions: ANSWER: 073 (1.00) pg.15,16, 36 l Engineered Safeguards d RMSASRB Cl RE-235, pg. 2 l Systems," Rev. 4. p.20,21 REFERENCE:

'

l EOP 0, psge 12 RMSASRB Cl RE-223, 072000K104 ..(KA's)

l " Radiation Monitoring system

000062A102 ..(KA's) Alarm Setpoint &  ;

Response book Channel information Sheets," Rev. 2 i

" Waste Distillate Tank I ( Overboard Monitor 000059K201 ..(KA's) 4 l

l l

( _

,

REACTOR OPERATOR Page 65 ANSWER: 077 (1.00) ANSWER: 081 (1.00) ANSWER: 085 (1.00) b d REFERENCE: REFERENCE: REFERENCE:

TRHB 13.12, " Instrument OI-62A, " Motor-driven EOP-0.2, " Natural circulation and Control system Auxiliary Feedwater System Cooldown," Rev. DRAFT, Description: Process (P-38A & P-38B)" pgs. 8, 9,12,17,18,20 and Area Radiation Rev.16, p 4 Monitoring System," Rev. 3, TRHB 11.4," Secondary pgs. 5,15,16 systems Descriptions: ANSWER: 086 (1.00)

RMSASRB Cl RE-116 Auxiliary Feedwater b

" Radiation Monitoring system System, " Rev. 6, p 6 REFERENCE:

Alarm Setpoint & AOP-58, Attachment R TRHB 10.6, " Primary Response book Channel Systems Description:

Information Sheets," Rev. 2 061000A207 ..(KA's) chemical and Volume Demineralizer Valve Gallery Control Systern" Rev. 6, pg ,17 072000A101 ..(KA's) ANSWER: 082 (1.00) in accordance with EOP c O.2, " Natural Circulation REFERENCE: Cooldown" Rev. 7, pg. 3 ANSWER: 078 (1.00) TRHB 10.7, " Primary EOP-0.1, " Reactor Trip Systems Descriptions: Response," Rev. DRAFT, p. 9 REFERENCE: Residual Heat Removal AOP-6E, " Alternate System," Rev. 9, p. 3 Boration/ Loss of Shutdown 194000G ..(KA's) Figure Margin," Rev. 7, pgs. 3 005000K403 ..(KA's) 000024K302 ..(KA's)

ANSWER: 079 (1.00)

b REFERENCE: ANSWER: 083 (1.00) ANSWER: 087 (1.00)

AOP-58, " Loss of d deleted Instrument Air," Rev.12, p REFERENCE: REFERENCE:

21 TRHB 11.4, " Secondary P&lD PB 01MCVK00000804 Systems Descriptions: 072000G ..(KA's)

Auxiliary Feedwater 000065A208 ..(KA's) System," Rev. 6, p 3,13 l Setpoint ANSWER: 088 (1.00) j d i ANSWER: 080 (1.00) ANSWER: 084 (1.00) REFERENCE:

'

b b ARB 1C031D 3-1," 1P-1 A REFERENCE: REFERENCE: or B RCP No.1 Seal Water TRHB 13.7, " Instrument and EPIP 6.1, " Limited Plant Outlet Control systems Descriptions: Evacuation," Rev.15, pg. 3 Temperature High, Rev. 4 Feedwater Control System," AOP-1B, Reactor Coolant Rev. 2, pg. 3,4,23 194000G ..(KA's) Pump Malfunction Unit 1, l Setpoint 1.4, 2.1, 4.2, Rev.10, pgs. 3,5,6 Setpoint A306 ..(KA's) 000015G ..(KA's) j

I l

l l

REACTOR OPERATOR Page 66 l

ANSWER: 089 (1.00) ANSWER: 093 (1.00) ANSWER: 097 (1.00) a d REFERENCE: REFERENCE: REFERENCE:

RMSASRB 1.0, 01-80 TRHB 13.5, Rev.3, pgs. 21, 01-37, " Shifting of 25,28,29, Figure 13.5.13, instrument Supply Bus 073000A401 ..(KA's) " Low Temperature Feeders," Rev. 30, pg Overpressure Protection 3,11,13, and 20 Logic," Logic Sheet 18 ANSWER: 090 (1.00) 194000G ..(KA's) K301 ..(KA's)

REFERENCE:

OP-3A, " Normal Power ANSWER: 098 (1.00)

Operation to Low Power ANSWER: 094 (1.00) c Operation," Rev. 43, pg. 2 b REFERENCE:

REFERENCE: TRHB 13.3, " Instrument and 001000K513 ..(KA's) TRHB 13.6, " Instrument and Control System Descriptions:

Con +rol systems Description: Reactor Pressurizer Level Control Protection (Reactor Trip),"

ANSWER: 091 (1.00) System," Rev. 2 , Logic Sheet Rev. 4, pgs. 6, 9, 27,34,38, b 18 REFERENCE: 003000K201 ..(KA's)

AOP 0.0, " Vital DC System 011000K604 ..(KA's)

Malfunction, Rev. 4, p. 9 ANSWER: 099 (1.00)

063000G ..(KA's) ANSWER: 095 (1.00) b c REFERENCE:

REFERENCE: TRHB, " Instrument and ANSWER: 092 (1.00) Control System Description: <

I a Steam Table Nuclear Instrumentation REFERENCE: System (Excore AOP-58, " Loss of Instrument 017000K502 ..(KA's) Instrumentation), Rev. 3, Air, " Rev.12, pgs, 24, 28, 39, Setpoint and 40 ANSWER: 096 (1.00) 012000K604 ..(KA's) 1 078000K302 ..(KA's) d REFERENCE:

AOP-98, '" Component ANSWER: 100 (1.00)

Cooling System Malfunction," c Rev.12, pgs. 3, 4, 5, 8,12, REFERENCE and 16 TRHB 13.1, Rev. 3, p. 2

,

000026K303 ..(KA's) 000032G ..(KA's)

{

.

("*""*" END OF EXAMINATION *""""*)

F REACTOR OPERATOR Page 67 ANSWER KEY I

001 c 021 c 041 b 061 a 081 b 002 c 022 c 042 c 062 a 082 c 003 c 023 a 043 c 063 d 083 d 004 c 024 a 044 d 064 d 084 b 005 b 025 a 045 b 065 d 085 d 006 c 026 d 046 d 066 b 086 b 007 c 027 a 047 c 067 d 087 deleted 008 c 028 c 048 a 068 c 088 d 009 c 029 b 049 c 069 d 089 c 010 c 030 d 050 b 070 b 090 d 011 b 031 d 051 d 071 a 091 b j 012 c 032 b 052 b 072 b 092 a 013 b 033 a 053 a 073 d 093 a 014 d 034 d 054 b 074 c 094 b 015 a 035 a 055 c 075 a 095 c 016 b 036 b 056 b 076 c 096 d 017 c 037 d 057 d 077 c 097 d 018 b 038 c 058 d 078 c 098 c 019 b 039 b 059 a or d 079 b 099 b 020 b 040 b 060 b 080 b 100 c (*""*"" END OF EXAMINATION "*"*"")

!

U.S. Nuclear Regulatory Commission Site-Specific Written Examination Applicant information Name: MASTER EXAMINATION Region: 111 Date: August 2,1999 Facility / Unit:

License Level: SRO Reactor Type: WESTINGHOUSE Start Time: Finish Time:

Instructions j Use the answer sheets provided to document your answers. Staple this cover sheet on top i of the answer sheets. The passing grade requires a final grade of at least 80.00 percen Examination papers will be collected five hours after the examination start l Applicant Certification All work done on this examination is my own. I have neither given nor received ai Applicant's Signature Results l

Examination Value 10 Points j Applicant's Score Points Applicant's Grade Percent l

.

SENIOR REACTOR OPERATOR Page 2 WRITTEN EXAMINATION GUIDELINES After you complete the examination, sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examinatio . To pass the examination, you must achieve a grade of 80.00 percent or greater; grades will not be rounded up to achieve a passing score. Every question is worth one poin . For an initial examination, the time limit for completing the examination is five hour . You may bring pens, pencils, and calculators into the examination room. Dark pencil should be used to facilitate machine gradin . Print your name in the blank provided on the examination cover sheet and on the answer sheet. You may be asked to provide the examiner with some form of positive identificatio . Mark your answers on the answer sheet provided. Use only the scan-tron sheets provided and do not write on the back side of the pages unless instructed. If you change an answer, ensure your erasure is complete. If the grading examiner is unable to determine which answer you have marked, the answer will be marked wron . If you have any questions concerning the intent or the initial conditions of a question, do not hesitate asking them before answering the question. Ask questions of the NRC examiner or the designated facility instructor only. When answering a question, do not make assumptions regarding conditions that are not specified in the question unless they occur as a consequence of other conditions that are stated in the question. For example, you should not assume that any alarm has activated unless the question co states or the alarm is expected to activate as a result of the conditions that are stated in the questio . Restroom trips are permitted, but only one applicant at a time will be allowed to leav Avoid all contact with anyone outside the examination room to eliminate even the appearance or possibility of cheatin . When you complete the examination, assemble a package including the examination questions, examination aids, answer sheets, and scrap paper and give it to the NRC examiner or proctor. Remember to sign the statement on the examination cover sheet indicating that the work is your own and that you have neither given nor received assistance in completing the examination. The scrap paper will be disposed of immediately after the examinatio . After you have turned in your examination, leave the examination area as defined by the proctor or NRC examiner, if you are found in this area while the examination is still in progress, your license may be denied or revoke . Do you have any questions?

SENIOR REACTOR OPERATOR Page 3 THIS PAGE INTENTIoyALLY LEFT m q ,

l

r

SENIOR REACTOR OPERATOR Page 4 QUESTION: 001' (1.00)

'

During a natural circulation cooldown on Unit 1, a void was created in the reactor vessel. Given that no RCPs are available and RCS pressure is approximately 1500 psig, which one of the following actions is used to' collapse the void in accordance with EOP-0.2, " Natural Circulation Cooldown?" Decrease RCS temperature while maintaining RCS pressure constan b.' Fill the pressurizer solid and ventilate via the post accident ventilation syste Start an Si pump to increase RCS pressure while maintaining temperature constant, Increase RCS pressure using pressurizer heaters while maintaining RCS pressure / temperature within the acceptable regio i

!

t l

'

.

I i

!

l l

I

SENIOR REACTOR OPERATOR Page 5 QUESTION: 002 (1.00)

'

Given the following Unit 2 plant conditions:

The Unit is in Cold Shutdown.

The RCS is solid.

-

RHR flow has been lost and CANNOT be restored.

-

Wide range water level for both steam generators (SGs) is 250 inches and steady.

-

All other systems and components are available.

In accordance with SEP 1.1, which one of the following methods of cooling is the preferred method of removing the core's decay heat? Establish RCS cooling by aligning AFW system flow to at least one SG and bleeding steam through the respective S/G Atmospheric Steam Dump Valv l l Establish feed and bleed through the RCS by aligning the RWST to the suction of the charging pumps, starting a charging pump, and venting through the pressurizer PORV Establish feed and bleed through the RCS by aligning RWST to the suction of the safety injection pumps, starting a safety injection pump, and venting through the pressurizer PORV Establish feed and bleed through the RCS by aligning RWST to the RCS through an RHR heat exchanger, gravity draining the RWST to the RCS, and venting through the pressurizer PORV !

p p'

SENIOR REACTOR OPERATOR .Page 6 i

,

QUESTION: 003 (1.00)

EOP-0 has been entered on Unit 1 due to an automatic initiation of safety injection and a reactor

- trip. The following plant conditions were observed:

-

Reactor coolant system pressure is 1500 psig and decreasing slowl .

- Radiation monitor RE-113, " Elevation -19' PAB area monitor"is alarming on the RMS panel; all other RMS indications are norma The safety injection pump suctions are aligned to the RWS :

Residual heat removal pump suctions are aligned to RWS .

RWST levelis 95%.

• Annunciator C01 A1-11, " AUX BLDG 19 FT SUMP LEVEL HIGH is li The above plant conditions would require transition from EOP-0 into which of the following procedures? EOP-1.1; " Unit 1 SI Termination." EOP-0.1; " Reactor trip response."

c, . ECA-1.2; "LOCA Outside Contain lent." ECA-1.1; " Loss of Containment Sump recirculation."

QUESTION: 004 (1.00)

Which of the following is the reason for depressurizing the Steam Generators at the maximum rate during ECA-0.0, " Loss of All AC Power?" To prevent lifting PZR PORV To minimize RCS inventory los . ' To prevent inadvertent reactor re-star To enhance restoration of SG level from the TD AFW Pum I

SENIOR REACTOR OPERATOR Page 7 QUESTION: 005 (1.00)

'

Which of the following is the reason for the RCP tripping criteria during the performance of EOP-0, " Reactor Trip or Safety injection?" To prevent exceeding containment pressure limitations during a large break LOCA even To minimize the effects of RCS cooldown in the event of a major steam line break for PTS consideration j i

I To limit the rate of RCS depressurization in the event of a large break LOCA and reduce the amount of voiding in the core, To limit the RCS inventory depletion through a small break LOCA leading to a more severe core uncovery if RCP's were tripped some time late QUESTION: 006 (1.00) l Which one of the following is the basis for terminating Si flow when the criteria are satisfied during the performance of EOP-3," Steam Generator Tube Rupture?" Prevent overcooling the RC Prevent solid plant operations, Prevent exhausting RWST leve Prevent overfilling the ruptured S I i

SEN!OR REACTOR OPERATOR Page 8 l

QUESTION: 007 (1.00)

According to AOP-0.0, ' Vital DC System Malfunction," ALL of the following are associated with loss of DC control power to an AC bus, EXCEPT the: associated breaker positions remain "as is." undervoltage stripping of the associated bus remains operabl I associated breakers cannot be electrically operated from the control roo associated breakers cannot be electrically operated from local control station QUESTION: 008 (1.00)

The following conditions exist:

-

Reactor Bypass breaker "A" racked in and closed for testin .

Reactor Trip breaker "A" ope Which of following describes the response if bypass breaker "B"is racked in? Both bypass breakers (A and B) will trip ope Bypass breaker "A" and trip breaker "B" will trip ope Bypass breaker "A" and trip breaker "B" will remain shu Bypass breaker "A" will remain shut and trip breaker "B" will trip open.

SENIOR REACTOR OPERATOR Page 9 0.UESTION: 009 (1.00)

'

Which one of the following is the reason for the Engineering Safety Features Actuation System precaution and limitation which states," Do not exceed 1700 psig in the primary system during heat up until the reactor coolant temperature is at least 480*F?" To ensure T,,, is above the Low-T , setpoint preventing an inadvertent S To avoid pressurized thermal shock concerns in the event of an inadvertent S To ensure the Si termination criteria are satisfied before automatic unblocking of S To ensure steam generator pressure is greater than 530 psig before automatic unblocking of S QUESTION: 010 (1.00)

Which of the following is an automatic action directly associated with a Safety injection Actuation only on Unit 1? Electric fire pump trip Charging pump, P-2A trip Instrument air compressor, K28 trips, Pressurizer heater backup groups A and B trip.

I

, I

~

SENIOR REACTOR OPERATOR ~ Page 10 QUESTION: 011 (1.00)

'

Which one of the following describes the automatic response / sequence of events of the Containment Air Recirculation System upon an Si actuation signal? (assume off-site power is available)_- Both the large and small fans receive start signals for each fan coole I The small fans trip in all fan cooler units and the large fans receive a start signa ) Both the large and small fans trip and then the large fans start for each fan cooler unit.- I Both the large and small fans trip and then the small fans start for each fan cooler uni QUESTION: 012 (1.00)

The following Unit 1 plant conditions exist:

-

Reactor power: 100% steady state j l

-

All control systems In automatic  !

'

• T-cold temperature transmitter TE-401 B failed low j Which of the following describes the immediate effects of this failure on the corresponding Tavg ]

and Delta-T indications? j Tav Delta-T i INCREASES INCREASES

' DECREASES DECREASES INCREASES DECREASES DECREASES INCREASES

.

'

n

I i

L SENIOR REACTOR OPERATOR Page 11 QUESTION: 013 (1.00)

With Unit 1 operating normally at 100% power, two out of four OT/AT channels rise to within 3

'

degrees of the' trip setpoint for 100 seconds. How much will turbine load be reduced? % % % %

QUESTION: 014 (1.00)

With the plant at 98% reactor power and Bank D control rods at 200 steps, which of the following is a symptom of a stuck control rod that would require entry into AOP-68, " Stuck Rod Or Malfunctioning Position Indication," following a transient? A variation in NIS instrumentation resulting in a quadrant tilt of 1.1%. An individual RPI with a 13 step disagreement with the bank demand locatio A variation in core outlet thermocouples of 8% relative to symmetric thermocouples, A variation in axial flux of 1.1% of axial peak at any location relative to symmetrical trac .

_- __ _ _ _ _ _ _ - _ - _ _ - _ _ .

.

.

SENIOR REACTOR OPERATOR Page 12 QUESTION: 015 (1.00)

Given the following Unit 1 plant conditions: '

-

The Unit has tripped from 100% due to a small break LOC Conditions have stabilized and operators are evaluating the criteria for terminating S Adverse Containment conditions do NOT exis Which ONE of the following conditions would PREVENT Si termination per EOP-1.2, "Small Break LOCA Cooldown and Depressurization?" RCS subcooling is 40* Pressurizer levelindicates 9%. Pressurizer pressure is 2050 psi Both steam generator levels indicate 40% N QUESTION: 016 (1.00)

During a normal reduction in power using boration, which one of the following is the reason that additional pressurizer heaters should be encrgized? Allow an increased ramp rate for the power change, Equalize the reactor coolant system and Pressurizer boron concentration Maintain PZR pressure in normal operating range during the power change, Ensure positive pressurizer pressure control ls d51ab' lished prior to starting the power chang __

(.

i ,

!

l L SENIOR REACTOR OPERATOR Page 13 l

l QUESTION: 017 (1.00)

'

Which ONE of the following RVLIS readings indicates the highest probability of core voiding? Wide Range reading 98 ft. with NO RCPs runnin Wide Range reading 120 ft. with ONE RCP runnin Narrow Range reading 35 ft. with NO RCPs runnin Wide Range reading 140 ft. with BOTH RCPs runnin QUESTION: 018 (1.00)

Which ONE of the following describes the purpose of the back draft dampers installed in the Containment Air Recirculation System? Prevent backflow Fn a cooling unit in the event of fire in Containmen Serve as explosion dampers preventing duct work collapse during an acciden Prevent unit air backflow when the accident fan is running and the cooling fan is not, Serve as a system air backflow damper in idle cooling units (both accident ahd i cooling fans secure I J

l l

!

l

!

!

l

l

.

.

l

SENIOR REACTOR OPERATOR Page 14 QUESTION: 019 (1.00)

'

The following conditions exist on Unit 1:

-* A reactor trip occurred 1 minute ago from 100% power

.

The main feedwater regulating valves failed in the 100% power position

.

Containment humidity is increasing

. IR "A" SUR: 0.00 dpm

.

IR "B" SUR:' - 0.15 dpm Which ONE of the following actions is required FIRST following a manual safety injection actuation? (ASSUME: All other equipment performed as designed.) Transition to EOP-2, " Faulted Steam Generator Isolation." Manually open containment ventilation cooler outlet emergency FCV Transition to CSP-S.1, " Response to Nuclear Power Generation /ATWS." Trip the main feed pumps, placo the condensate pumps in PULL OUT, and stop the heater drain pum !

l QUESTION: 020 (1.00) ji With the RCS at normal operating pressure and temperature, what is the condition of the steam entering the PRT if a PORV opens? (ASSUME: PRT is at 100*F, 5 psig; an ideal .

thermodynamic process) Saturated steam-water mixture at:

I * ' c.- 235* *F.

h

.

J

- - - - - - -

i SENIOR REACTOR OPERATOR Page 15 l QUESTION: 021 (1.00)

The following conditions exist on Unit 1:

,

A LOCA has occurred

-

Transfer to containment sump recirculation is required

-

RCS pressure is approximately 50 psig What is the expected SI pump TOTAL flow indicated on the main control board prior to entering EOP 1.3, and how will this value change following transfer of BOTH trains of ECCS to containment sump recirculation?

Total Flow Flow Change gpm Decrease gpm increase gpm Decrease gpm increase QUESTION: 022 (1.00)

Given the following Unit 1 plant conditions:

-

A Control Bank D rod was dropped and recovered.

-

The Pulse to Analog Converter was NOT reset as required by AOP-6A, " Dropped Rod."

Which one of the following will occur on the next rod movement? if control rods are withdrawn, OT/AT will NOT stop Control Bank D withdrawal when require If control rods are withdrawn, OP/AT will NOT stop Control Bank D withdrawal when require If control rods are inserted, the Rod insertion Limit Alarm will be received at a lower rod position than require If control rods are inserted, Bank C control rods will begin insertion at a lower value of Control Bank D positio _

f-

, SENIOR REACTOR OPERATOR Page 16 l i

!

'

l

QUESTION: 023 (1.00)

'

In procedure OP-5A, " Reactor Coolant Volume Control," there is a PRECAUTION that states:

"Do not secure letdown flow without also securing charging flow ..." Which ONE of the following statements describes why charging flow is required to be isolated? (ASSUME: All systems are in a normal at power lineup.) Reduce thermal shock on the charging penetration into the RC Reduce thermal shock on the Non-Regenerative Heat Exchange VCT level will decrease until charging pump suction shifts to the RWS VCT level will decrease causing possible damage to the charging pump QUESTION: 024 (1.00)

Given the following Unit 1 plant conditions:

- The Unit was operating at 75% steady state powe All systems were in automatic contro ~

j

• - The "A" SG atmospheric steam dumps failed ope !

• Main turbine control is in IMP IN with the valve position limiter set at 95%.

~ Which ONE of the following describes the plant response to this condition? (Assume NO operator action is taken.) Turbine load decreases by 5%, reactor power remains stable at 75%. i Control rods initially insert then withdraw to maintain reactor power at 75%. Control rods withdraw and reactor power increases to 80% where it stabilize d.' Turbine govemor valves open in response to lower steam header pressure to increase turbine load to 80%. .

__ _ _ _ _ _ _ - _ _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - .

SENIOR REACTOR OPERATOR Page 17 l QUESTION: 025 (1.00)

The following conditions exist on Unit 1: '

-

Reactor power 80%

-

The lower detector for N-43 is open circuited (due to a broken cable)

Which ONE of the following is acceptable for determining core quadrant power tilt under the above conditions? Plant Process Computer, Movable incore Detector Manual calculations using operable excore detector Manual calculations using estimated current for N-43 lower detecto QUESTION: 026 (1.00)

Performance of EOP-1.2,"Small Break LOCA Cooldown and Depressurization,"is in progres What is the reason for starting both control rod shroud fans after depressurizing the RCS? To provide adequate cooling for the CRDM To reduce containment pressure and humidit To provide adequate cooling for the NIS detector To cool down the upper head region of the reactor vessel.

.

(

SENIOR REACTOR OPERATOR Page 18 l

l

'

QUESTION: 027 (1.00)

The following plant conditions exist:

'

-

Steam generator A steam flow: 0.0005x108 lbm/hr and stable

- Steam generator B steam flow: 0.6x10' Ibm /hr and decreasing

- Steam generator A level: 70% and stable

'

Steam generator B level: 10% and decreasing

-

Turbine driven AFW pump: running

- Motor driven AFW pumps: running

-

,Tavg: 516*F and decreasing

- Containment pressure: 5 psig and increasing

-

SI: actuated

.

If no operator action has been taken, which ONE of the following indicates the status of the main l steam isolation valves?

MSIV A MSIV B open open b, ope shut shut open shut shut '  ;

)

..

l l~

l l

i

. SENIOR REACTOR OPERATOR Page 19 l

l l

!

QUESTION: 028 (1.00)  !

During operation at 100% power, impulse pressure channel (FT-486) failed LOW. Which one of '

' the following describes the response of the condenser steam dump control system to this failure and why?

) The steam dump valves remain closed but are " armed" due to a loss of lead j condition being sense I The dump valves modulate open due to a Tg,,, deviation generated by the l loss ofimpulse pressur The steam dump valves trip open on a turbine trip signal being generated by the i loss of impulse pressur The loss of impulse pressure would only have an affect on the steam dump if it was operating in the " pressure" mod QUESTION: 029 (1.00)

During refueling operations inside containment, the control room receives an RE-211/212,

" Containment Air Particulate / Noble Gas Monitor," alarm. Which one of the following actions is REQUIRED to be performed? Verify Containment purge supply fans are runnin Suspend all refueling operations inside containmen Notify the Emergency Plan Coordinator to implement emergency pla Notify the Plant Manager of the need to perform a fu!! plant evacuatio SENIOR REACTOR OPERATOR Page 20

!

QOESTION: 030 (1.00)

'

l 4 RHR shutdown cooling was in progress when the 1P-10A RHR pump had to be secured due to mechanical failure. The remaining RHR pump (1P-10B) is indicating erratic flow characteristics.

l Which one of the following actions would be the first operator action taken in accordance with

'

SEP-1, " Degraded RHR System Capability?" vent the operating RHR pump as necessary, c check for completion of RHR suction line reflood, stop the RHR pump and isolate the RCS drain path adjust RHR system flow to between 1300 gpm and 1500 gp QUESTION: 031 (1.00)-

, Which one of the following is the basis for isolating the feedwater to a faulted SG7 To maximize the energy release from the faulted SG.

l- .. To maximize the cooldown capability from the non-faulted SG and minimizing l f

RCS cooldow To prevent overflowing the faulted SG with feedwater and thus minimizing the cooldown for PTS concems.

, To prevent thermal shock to the faulted SG "U" tubes and thus minimizing the

'

potential of rupturing a SG tube with subsequent off-site releas I

.

. . . . . . . .

SENIOR REACTOR OPERATOR Page 21 QUESTION:032 (1.00)

The plant was in a normal configuration at 100% power when the following occurred:

-

- S,upply breaker,1A52-77,1A-04 to 1A-06 Bus Tie Breaker trippe Supply breaker,1A52-76,2A-03 to 2A-05 Bus Tie Breaker trippe Which of the following statements correctly identifies the subsequent electricallineu Only G01 and G04 start and supply buses 2A-05 and 1A-06, respectivel .

Only G02 and G03 start and supply buses 2A-05 and 1 A-06 respectivel All four EDGs start with G01 supplying 2A-05 and G04 supplying 1A-06 All four EDGs start with G02 supplying 2A-05 and G03 supplying 1A-0 __ _ _ _ _ - - - -

..

- - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

SENIOR REACTOR OPERATOR Page 22 QUESTION: 033 (1.00)

'

Given that the Service water pump (P32A) and component cooling water pump (P11 A) are running:

-

At T=0 Containment pressure increased to 6 psi *

At T = +10 seconds (Ten seconds later) the supply breaker to 1B-03 tripped ope As a result of the above conditions, what is the starting order of the service water pump (P32A)

and component cooling water pump (P11 A)?

P32A P11A Restart at same time as P11 A Restarted at same time as P32A Restarted before P11A Restarted after P32A Restarted after P11 A Restarted before P32A Does not restart Does not restart t

. __

F SEN!OR REACTOR OPERATOR Page 23 QUESTION: 034 (1.00)

Given the following conditions: '

-

A large break LOCA has occurred inside containment

-

Containment pressure initially increased to 30 psig, then began trending down and is current'y 10 psig

-

The spray addition tank level has decreased by 20%

What is the current status of the containment spray pumps and valves?

PUMPS DISCHARGE VALVES Running Full oren Secured Closed Running Mid-position Secured Full open QUESTION: 035 (1.00)

If a safety injection signal occurred coincident with an undervoltage condition on bus 1-A05 which of the following correctly describes the response of P-38A, Motor Driven Aux Feedwater ,

Pump, assuming all system operated as expected, P38A trips and must be manually starte P38A trips and sequences on (starts) approximately 10 seconds after 1 A-05 reenergize P38A trips and sequences on (starts) approximately 20 seconds after 1 A-05 reenergize P38A continues to ru . . .. .

-

. .. .. - _

. . . , . _

-

. . . .

i

SENIOR REACTOR OPERATOR Page 24 QUESTION: 036 (1.00)

' Given the following plants conditions:

Unit 1 is operating at 100%, steady state power

-

CCW surge tank level has increased since the last log reading

-

High CCW pump inlet temperature alarms exist a High CCW pump inlet radiation alarms exist These conditions describe a/an: RCP thermal barrier cooling coil lea RHR heat exchanger tube-to-shell side lea primary-to-secondary steam generator tube lea seat return heat exchanger tube-to-shell side lea l

.

r-SENIOR REACTOR OPERATOR Page 25

- QUESTION: 037 (1.00)

'

Given the following plants conditions:

. The reactor was at 80% powe . Circulating water pump, P30A was out of servic At 0900, a down power was commenced at 0.5%/ min for the hou At 1003 excessive vibrations caused the P30B pump to tri Which one of the following describes the expected response and the correct reason for this response?

The reactor ... trips; the P-9 permissive was activ does not trip; the P-10 permissive was activ trips; reactor power was greater than 55% powe does not trip; the P-9 permissive was automatically blocke l

!

l

,

(

i

SENIOR REACTOR OPERATOR Page 26

!

!

I QUESTION: 038 (1.00)

'

You have completed a hand frisk which found no contamination after you alarmed the PCM-1B at the RCA exit. Which of the following actions are you now required to take? Proceed directly to the portal monitors since the hand frisk indicated no ,

contaminatio Recount once in PCM-18, if no alarms cccur during this recount, then proceed to the portal monitor l

' Contact your supervisor to resolve the' discrepancy between the alarming PCM-1B and non-alarming hand fris Recount twice in the PCM-1B, if no alarms occur dunng both of these recounts, ;

then proceed to the portal monitor i

!

SENIOR REACTOR OPERATOR Page 27 QUESTION: 039 (1.00)

'

Given the following initial conditions:

.

The heater drain tank has been vented, depressurized, and opened for personnel i access.

=

You are the attendant at the personnel access point for the confined space.

A short time after a mechanic entered the tank, the atmosphere sampling instrument alarmed .

and you noticed that the mechanic in the confined space is uncor.scious. In accordance with I NP 1.9.4, " Confined Spaces Procedure," what is your required ac! ion? l l contact the security coordinator to perform a rescue of the mechani ' call for a backup attendant, and attempt to rescue the mechanic yoursel I I

, contact the fire protection and safety coordinator to sample the confined spac contact the control room operator to summon rescue and other emergency :

service i

R i

SENIOR REACTOR OPERATOR Page 28 l

- QUESTION: 040 (1.00)

'

You are performing a monthly surveillance on the AFW system for the 'A' steam generato Under which one of the following conditions should you initiate a temporary change? It becomes necessary to abort the AFW surveillance procedure prior to completio The AFW surveillance procedure incorrectly directs you to manipulate switch 1-P38B instead of 1-P38 You have completed section A and 8 of the AFW surveillance procedure and are starting Section C. Step C.1 directs you to record bearing temperatures that are ONLY required for the quarterly tes . You have 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> before having to enter TS 15.3.0.B due to the improper performance of the previously performed AFW surveillance and a one-time change to the AFW surveillance procedure is required which does not change the intent of the procedure.

a

_

I r SENIOR REACTOR OPERATOR Page 29 QUESTION: 041 ' (1.00)

'

Which one of the following describes how you would perform an independent verification on an OPEN manual valve? tum the valve handwheel in the open direction unit there is no more valve movemen tum the valve handwheel in the closed direction enough to observe valve movement, then turn the valve in the open direction until there is no more valve movemen turn the valve handwheel in the closed direction two full tums, then tum the valve in the open direction until there is no more valve movemen tum the valve handwheel in the closed direction until the control room receives a dual position indication for the valve, then tum the valve in the open direction until there is no more valve movemen QUESTlON: 042 (1.00)

Your current annual exposure is 1750 mrem. If you were assigned a task in an area where the dose rate is 80 mrem /hr, what is the maximum number of hours you could spend in the area without exceeding your ANNUAL administrative limit? (You do not have any authorized extensions.) hours hours hours d.- 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />

p ,

c .c

$

i SENIOR REACTOR OPERATOR Page 30 i

l QUESTION: 043 (1.00)

. Given the following conditions on Unit 1:

=-

Tm = 558'F and slowly decreasing

< -

Power = 45% and stable Pressurizer pressure = 2000 psig and stable

-

! -

Condenser vacuum has degraded to 19" Hg (- , With no operator action, what is the plant response?

L'

a.' Turbine trip ONL Annunciator low vacuum ONL The turbine will trip, and then the reactor will trip, The reactor will trip, and then the turbine will trip.

l l QUESTION: 044 (1.00)

Given the following plant conditions on Unit 1:

-

A LOCA has occurred L

-

The reactor has tripped

-

Pressurizer pressure is 1685 psig and stable

- One minute after the above conditions occur, service water pumps P-32A, P-328, and P-32D have not starte Which of the following valves is in its correct position? SW-2880, unit 1 turbine building feeder valve is OPE i l  !

I

~ SW-2816, service building air conditioning supply valve is OPEN. SW-2907, containment fan service water outlet valve is CLOSE : SW-2930A, spent fuel pool heat exchanger outlet valve is CLOSE .

t L=

, . , , ..

!

SENIOR REACTOR OPERATOR Page 31 QUESTION: 045 (1.00)

'

. A fast spreading fire exists in the unit 1 turbine lube oil system reservoir. Which one of the following automatic fire suppression systems will deploy? Halon syste Deluge sprinkler syste Dry pipe sprinkler syste Wet pipe sprinkler system.

. -

.

..

-.

.

. . .

. . . . . . _ _ _ _ _ _

r

{

'

l SENIOR REACTOR OPERATOR Page 32 l

{

QUESTION: 046 (1.00)

'

Given the following plant conditions on Unit 2:

,

.

. Unit 2 is in hot shutdown following a reactor trip 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ago l * Unit 2, A steam generator is at 60% level and increasing i

-

Unit 2, B steam generator is at 25% level and decreasing i

-

Both motor driven AFW pumps (P38A and P38B) have started After the AFW pumps started, a pipe rupture occurred downstream of the suction check valve to the AFW pump, P388. How will the Unit 2,1B steam generator level be affected?

' The Unit 2, B steam generator level will .. .. decrease until operator action is taken to start the steam-driven auxiliary l feedwater pump, 2P2 I increase due to flow from automatic starting of both steam driven feedwater pumps,1P29 and 2P2 decrease until service water valve AF-4016 is manually opened by the unit !

control operator in the control roo I decrease until service water valve,2AF-4006 automatically opens when the Unit 2, B S/G level reaches 20%

i i

l

!

I

SENIOR REACTOR OPERATOR Page 33 )

QUESTION: 047 (1.00)

'

Given the following plant conditions on Unit 1:

.

S/G A has a partially stuck open Atmospheric Steam Dump Va!ve i

• Both motor-driven AFW pumps (P38A and P388) started when level decreased to 25%

l in the Unit 1, A S/G i After the start of the AFW pumps, annunciator C01A 4-9," AUX FEED PUMP SUCTION PRESSURE LOW," alarmed. The B AFW pump (P38B) tripped due to a failed low AFW pump ,

suction pressure switch. What actions are required to override the B AFW pump (P388) low l suction pressure trip? Select a different pressure switch for the 'B' AFW pump and take the switch to the auto positio ! Take the pump control switch for the 'B' AFW pump to start and return the switch to the auto position, Take the pump control switch for the 'B' AFW pump to pullout, and then place the switch in the auto positio Wait for a 30 second time delay, pump will then automatically restart after placing the pump control switch for the 'B' AFW pump to auto positio l l

l I

,

!

l SENIOR REACTOR OPERATOR Page 34 l

QUESTION: 048 (1.00)

'

Which of the following conditions will result in the automatic closure of waste distillate overboard valve, FCV-LW-157 waste distillate pump tri high levelin the waste holdup tan I high alarm on RE-218, waste disposal system liquid monito i high alarm on RE-223, waste distillate tank overboard monito i QUESTION: 049 (1.00)

It's 1530 and a release of the "A" monitor tank is in progres You receive a high alarm on RE-218, waste disposal system liquid monitor. in accordance with RMSASRB Cl RE-218 " Radiation Monitoring System Alarm Setpoint & Response Book," you MUST perform all the following actions EXCEPT: re-check release calculatio l notify RP supervision (Duty and Call). recommence discharge once the high alarm clear verify shut RCV-018, waste liquid overboard valv l SENIOR REACTOR OPERATOR Page 35 QUESTION: 050 (1.00)

Degassing of the reactor coolant system has commenced. Gas Decay Tank #1 is in service and at 80 psig and increasing at 4 psig/ hour. Four hours after commencing degassing, you note the j following valve alignment: I l

1. 1 GDTinlet valve is closed I

- #2 GDT inlet valve is open -

!

What caused the valve alignment?

' #1 GDT tank pressure increased to 95 psi receipt of a high alarm on RE-114, El. 26' PAB West area monito back pressure on the waste gas compressor decreased to 35 psi I

' RE-214, auxiliary building vent exhaust gas monitor, came in violet on RMS displa QUESTION: 051 (1.00) i I

Control room ventilation has re-aligned to Mode 4. The following conditions are indicated on the'

RMS display:

• RE-235, control room noble gas monitor is YELLOW

- RE-101, control room monitor is RED a RE-234, control room iodine monitor is RED a RE-214, auxiliary building vent exhaust gas monitor is RED What caused the Mode 4 realignment of CR ventilation? RE-235 being in aler RE-214 being in high alarm, RE-101 being in high alar RE-234 being in high alar c:

I

!

l l

. SENIOR REACTOR OPERATOR Page 36 o

l QUESTION: 052 (1.00) q l

'

Given the following indications on the Unit 1 RMS display:

RE-116, Demineralizer Valve Gallery Area monitor, is RE RE-102, EL. 66'. containment low range monitor, is LIGHT BLU New fuel receipt inspection is in progress in addition, the following conditions exist:

- ' Annunciator,1C20 B 1-9, " COMMON AREA RADIATION MONITOR HIGH"is lit

-

Containment sump level is zero and stable What is the cause for the Common Area Radiation Monitor High annunciator being lit? having the gas stripper online.

! loss of reactor coolant in containmen high pressure in the letdown line causing a diaphragm valve to lea fuel was inadvertently dropped while being placed in the new fuel vaul ,

l

!

l l

!

l

SENIOR REACTOR OPERATOR Page 37

' QUESTION: 053 (1.00)

l During natural circulation cooldown, it is possible for a bubble to form in the reactor vessel hea The existence of a bubble in the reactor vessel is normally indicated by the Reactor Vessel Level Indication System (RVLIS). If RVLIS is not available, which one of the following post accident monitoring instruments would provide the first indication of the existence of a bubble in the reactor vessel head?

l Reactor Coolant System (RCS) pressure (wide range). Core exit thermocouple Pressurizer water level (narrow range). RCS hot leg temperature (wide range).

i

)

i

'

.

'*

i

<

k

e

!.

f

'

.

!

SENIOR REACTOR OPERATOR Page 38 QUESTION: 054 (1.00) ,

'

Given the following plant conditions:

. 100% power

.

RCP A and B #1 seal return flow is returning to the VCT .

.

Instrument air pressure has dropped to 75 psig and is still rapidly decreasing Without operator action, what is the response of RCP seal return flow? RCP seal return flow will fill and pressurize the: PRT, and cause a relief valve (CV-314) to PRT to lift, sending the water in the PRT to the RCD seat return line and cause a relief valve (CV-314) to PRT to lift, sending the water in the seal return line to the PR RCDT, and cause a relief valve (CV-314) to PRT to lift, sending the water in the RCDT to the 'B' containment sum seal return line, and cause a relief valve (CV-314) to PRT to lift, sending the water in the seat return line to the RCDT.

I

,

i SENIOR REACTOR OPERATOR Page 39 QUESTION: 055 (1.00)

'

Following a reactor power increase from 80% to 90% power, the following events occur on Unit 2:

. Unit 2 Loop B S/G level channel, L471 failed low

-

Main feedwater pump suction pressure dropped to 180 psig and is stable

Pressurizer pressure dropped to 1910 psig and is stable Based on the above conditions, without operator action, the plant will experience which of the

- following transients: a safety injection due to low pressurizer pressur a feedwater isolation on high level in the Unit 2 B S/ a feedwater isolation on high level on the Unit 2 A S/ a reactor trip due to main feedwater pumps tripping on low suction pressure.

t ..

. ..

.. ..

. ..

. . _ _

_ __

, . . .

.

.. . . .

SENIOR REACTOR OPERATO Page 40 QUESTION: 056 (1.00)

'

Given the following initial conditions:

-

1P29, Unit 1 turbine-driven AFW pump received an automatic start signal and is

'

currently injecting at 50 gpm into each S/G on Unit 1

=

TDAFW mini-recirculation flow is 135 gpm

.What will the effect be on indicated AFW system flow if a loss of instrument air to the recirculation valve for 1P29 occurred? remain unchange ~ increase to new valu decrease to new valu initially decrease then return to original value.

. . . . .. .. .

.

. ..

.__ _

SENIOR REACTOR OPERATOR Page 41 QUESTlON: 057 (1.00)

'

An RCS cooldown is in progress. RHR has been placed in service for shutdown cooling. The following plant conditions exist:

• RCS pressure: 380 psig and slowly decreasing

.

RCS wide range temperature - Hot Leg 340*F and slowly decreasing a

Flow controller RHR-626 in automatic: 10% open

-

Local RHR Heat Exchanger outlet temp: 320*F and slowly decreasing if the RHR retum header flow transmitter (FT-626) failed LOW under these conditions, which one of the following would correctly describe the plant's response? RHR Heat Exchanger flow control valves (1FCV-624/625) will automatically close

- to attempt to prevent exceeding RCS cooldown rate of 100*F/h RHR retum header isolation valve (1RH-720) will automatically open to attempt to maintain RHR pump discharge pressure and flow constant to the reactor vesse RHR Heat Exchanger bypass valve (1FCV-626) will position to 100% full open to f attempt to maintain RHR desired system flow rate and RCS cooldown rate will decreas RHR Heat Exchanger bypass valve (1FCV-626) position will change to 100% full close to attempt to maintain RHR desired system flow rate and RCS cooldown rate will decreas I

!

-

l

y

' SENIOR REACTOR OPERATOR Page 42

l l QUESTION: 058 (1.00)

'

How will the AFW system respond if Unit 1, steam generator A, level lowers to 25% due to a partially opened S/G Atmospheric Steam Dump Valve when the plant is at 100% power?

(Assume B S/G level remains at 64%.) Only the P38A motor driven AFW pump will start; only discharge valve 4022 discharge to Unit 2, A S/G will ope Only the P38B motor driven AFW pump will start; only dischar0e valve 4023 discharge to Unit 1, A S/G will ope Both turbine-driven AFW pumps (1P29 and 2P29) will start; both discharge valves 1MOV-4001 and 2MOV- 4001, to Unit 2 A S/G will open, Both the P38A and P388 motor drive AFW pump will start; both discharge valves: 4021 discharge to Unit 1, B S/G and 4023 discharge to Unit 1, A S/G will

.ope QUESTION: 059 (1.00)

You are an extra operator filling in for a sick non-licensed operator. While performing your plant rounds (first of shift), a LIMITED plant evacuation was announced due to a dropped fuel assembly. You may reenter the area and resume conducting your rounds after approvalis obtained from the . .. Plant Manager and the Duty Shift Superintenden Duty Shift Superintendent and the RP superviso RP Supervisor and the Shift Technical Adviso Duty and Call Superintendent and the Plant Manage .

g- -- -

i SENIOR REACTOR OPERATOR Page 43 QUESTION: 060 (1.00)

A Unit 1 plant cooldown is in progress in accordance with EOP 0.2, " Natural Circulation Cooldown." The following plant conditions exist:

.-

Both CR shroud fans: running

.- RCS hot leg temperatures: 446*F and decreasing

RCS cold leg temperatures: 430*F and decreasing

- RCS pressure: 900 psig and decreasing in accordance with EOP 0.2, " Natural Circulation Cooldown," the operator is required at this time to: verify safety injection unblocke establish nitrogen backup to the PORV isolate reactor coolant pump seal injectio temporarily restore power to both accumulator discharge valve I i

I i

'

l SENIOR REACTOR OPERATOR Page 44

!

l QUESTION: 061 (1.00)

'

Following a Unit 1 plant trip at 0900, the following plant conditions exist:

-

Reactor Trip Breakers indicate open in the control room

-

Three (3) control rods are not fully inserted

-

Emergency boration was commenced in EOP 0.1, " Reactor Trip Response"

-

The Boric Acid Storage Tank level was at 90%

-

Both Boric Acid Transfer pumps,1-P4A and I-P4B tripped and cannot be restarted after I I

BAST T6A decreased by 47%

What action is required in accordance with AOP-6E, "Altemate Boration/ Loss of Shutdown !

Margin," after both of the Unit 1 BAT pumps tripped? de-energize rod drive MG Inject 6,000 gallons from RWS inject 1,000 gallons from Boric Acid Storage Tank, T6 dispatch operator to locally open Reactor Trip Breaker !

QUESTION: 062 (1.00)

l During a liquid release, which one of the following conditions would require compensatory actions to be taken? RE-223, Waste Distillate Tank Discharge, out-of-service while discharging radwast RE-219, S/G Blowdown Liquid Discharge, and RE-222, Blowdown Tank Monitor, out-of-service while SG blowdown in progres RE-229, Service Water Discharge, out-of-selvice during unit operation.

. RE-230, Retention Pond Discharge, out-of-service during a retention pond

! discharge.

i

!

r SENIOR REACTOR OPERATOR Page 45 QUESTION: 063 (1.00)

'

The plant is recovering from a reactor trip. The following plant conditions exist:

RCP seal water outlet temperature: 190*F and rising a RCS pressure: 1500 psig and stable

-

1P-1A RCP No.1 sealleakoff: 0.60 gpm and stable

-

1P-1 A RCP No.1 seal leakoff valve (MOV-270A): open

-

Sealinjection from the 1P-1A RCP: 8 gpm and stable

-

1P-1 A RCP shaft vibration: 13 mils and stable Annunciator 1C031D 3-1, "1P-1 A OR B RCP NO 1 SEAL WATER OUTLET TEMPERATURE HIGH," energized. Which of the following actions is required in response to these comditions? open 1CV0386,1 A and 1B RCP seal water bypass control valve increase CCW flow through the A RCP thermal barrier heat exchanger increase seal injection flow to 12 gpm to the A RCP by throttling open 1CV-300 trip the 1P-1A RCP and shut 1CV-270A, RCP Number i Seal Return isolation valve after 3 minute ,

i

_^

[.

g SENIOR REACTOR OPERATOR Page 46-l I l QUESTlON: 064 (1.00).

'

Conceming the Radiation Monitoring System, which one of the following indicates normal operation of the Control Room and the TSC Control Terminals (cts)?'

' All six'(6) control terminal lights are O All six (6) control terminal lights are OFF.

, Control Terminal " monitor" and " control" lights are CYCLIN ' Control Terminal " monitor" light is OFF, and " control" light is ON.

I

. s i'

- QUESTION: 065 (1.00)

' After operating at 30% power for four days, the plant was ramped to 40% power over a one hour period. Power was stabilized at 40% power and control rods were manipulated to maintain

.

Tavg. If RCS boron concentration and Tavg are maintained constant, which one of the following describes which direction control rods will need to be moved during the next hour in

, order to maintain Tavg constant? . OUT, becauso Xenon production by direct fission yield will be temporarily dominant. ,

i OUT, because Xenon production by decay of krypton will be temporarily L >

dominan IN, because Xenon removal by decay will be dominant.

I IN, because Xenon removal by burnout (neutron absorption) will be temporarily dominant, i

l-  ;

l'

!' i

)

_

.

SENIOR REACTOR OPERATOR Page 47 QUESTION: 066 (1.00)

- Which one of the below listed DC control power alignments will result in overloading a DC power '

supply if EDGs G03 and G04 fields' flashed at the same time? control power for Bus 1 A06 in normal, control power for Bus 2A06 in normal.-

b.' conNol power for Bus 1 A06 in attemate, control power.for Bus 2A06 in norma control power for Bus 1 A05 in normal, control power for Bus 2A05 in norma d.- control power for Bus 1 A05 in attemate, control power for Bus 2A05 in norma QUESTION: 067 (1.00)

Which of the following valves will fail open on a loss of instrument air? RCP Thermal Barrier isolation, CC-761 Unit 2 radwaste CCW supply, CC-LW-63.~ Si accumulator vent /N2 supply, SI-834A, B Turbine-driven auxiliary feedwater pump recirc valve, AF-400 !

l i

l

!

,

... _ _ da_

h

' SENIOR REACTOR OPERATOR Page 48 l ,

L QUESTION: 068 (1.00):

'

With the plant in solid operations, the following plant conditions exist:

- RCS temperature: 365*F and decreasing

• RCS pressure: 375 psig and decreasing

-

Low temperature over pressure mitigation system is in service

=-

PORV 431C control switch is in the auto position

-

PORV 430 control switch is in the auto position

-

LTOP enable keyswitch #1 is in the ON position

LTOP enable keyswitch #2 is in the ON position What is the plant response if pressurizer pressure instrument, PT-493 fails HIGH7

- . ONLY Pressurizer PORV FCV-431C ope ONLY Pressurizer PORV FCV-431C opens and all pressurizer heatets de-energiz ONLY Pressurizer PORV FCV-431C opens and both pressurizer spray valves ope ONLY Pressurizer PORV FCV-431C opens, all pressurizer heaters de-energize, and both pressurizer spray valves ope )

l l

l

..

SENIOR REACTOR OPERATOR Page 49 '

QUESTION: 069 (1.00)

'

Given the following Unit 2 initial conditions:

.

The Unit is at 70% power and stabl *

Rod controlis in MANUA *

Pressurizer level is at 38% and stabl Tavg is at 563*F and stabl Without operator action, what will be the effects on pressurizer level if Loop B cold leg ,

temperature detector TE-4018 falis low? Pressurizer level will .. . decrease to 20% and stabiliz remain unchanged and stable, increase to 46% and stabiliz increase to 100% and stabiliz QUESTION: 070 (1.00)

Given the following Unit 1 plant conditions:

• The reactor is shutdow Decay heat is being removed by natural circulatio RCS pressure is 1050 psig

- Average core thermocouple temperature is 450*F How much subcooling exists in the RCS at the above conditions? * ' * {

I 'F.

t

!

I

'

- <

-

.

.,

SENIOR REACTOR OPERATOR Page 50 QUESTION: 071 (1.00)

'

Given the following initial plant conditions:

-

Normal CCW lineup with P-11 A CCW pump running

-

Letdown flow temperature high

-

Letdown high temperature divert valve,1CV-145, is bypassing demineralizers Which one of the following are you required to perform after entering AOP-98 due to a pipe rupture of CCW at the inlet to the non-regenerative heat exchanger? Turn off all pressurizer heaters, Start the standby component cooling water pump, P-11 Isolate RCP thermal barrier cooling return valves,1CC- 761 A, Start a reactor makeup service pump and open 1CC-815, " Emergency Makeup Valve."

!

.. ..

. .. .

.

.

l

SENIOR REACTOR OPERATOR ' Page 51 QUESTION: 072 (1.00)

Given the following: >

+ Inverter 1DY03 is to be taken OOS

. Operators have shifted WHITE buses 1Y03, "120VAC" and 1Y103, "120VAC" to the alternate inverter DYOC, "120VAC."

In accordance with 01-37, " Shifting of Instrument Supply Bus Feeders," after the shift to the alternate inverter (DYOC) which one of the following is a required action? Verify the hydrogen computer's trouble alarm is ILLUMINATED on panel 1C-20 l ASI I Instal! magnetic administrative control tag for UNIT 1 buses 1YO3, "120VAC" and 1 Y103, "120VAC." l I Open the electrical interlock between Unit 1 (1YO3 and 1Y103) and Unit 2 (2YO3 and 2Y103) instrument buse Establish a fire watch inspection twice per shift of auxiliary feed pump room, fire Zone 304 and remote shutdown panel area (W46 and W-46A).

l

)

%.

,

SENIOR REACTOR OPERATOR Page 52 l

l QUESTION: 073 (1.00)

'

The following conditions exist on Unit 1:

.

Turbine load is 15%

• Reactor power is 15%

PZR pressure is 1985 psig

-

PZR levelis 24%

-

. 'A' Train of Secondary equipment is aligned for operation

. CO2 D 3-4, "4.16KV BUS LOCKOUT UNIT 1" was received

• ' All three RCS Loop A flows (FT-411,412,413) are significantly less than all three RCS Loop B flow Which one of the following MANUAL actions is required? initiate safety injectio start the third charging pump, place the reactor in hot shutdow reduce turbine load to 10% and trip the turbin .

.~.

--

.

, SENIOR REACTOR OPERATOR Page 53 QUESTION: 074 (1.00)

'

Given the following PR NI readings NI-41 - - NI-42 NI-43 NI-44 10% 8% 9% 8%

Subsequently, a slight transient occurs, resulting in the following readings:

NI-41 NI-42 NI-43 NI-44 11 % 8% 10% 8% .,

I No additional operator action is taken. What is the status of the IR high flux reactor trip and the PR high flux low range reactor trip?

IR HIGH FLUX REACTOR TRIP

'

PR HIGH FLUX LOW RANGE REACTOR TRIP i

, Blocked Blocked Unblocked Unblocked Blocked Unblocked Unblocked Blocked l

I

, ,

L

'

.

('

SENIOR REACTOR OPERATOR Page 54

- QUESTION: 075 (1.00)

Following a reactor trip on Unit 1, the following plant conditions were observed:

-

RCS pressure was 1985 and stabl * Boric acid flow control valve,1CV110A, was out of service CLOSE Emergency boration valve,1CV350, was out of service CLOSE .

.Two control rods were stuck out at 180 step Which one of the following boration flow paths, in accordance with AOP-6E, is immediately available to reestablish shutdown margin? Borate using a charging pump and the RWS Borate using the centrifugal charging pumps and the blende Borate using thi c::fety injection pumps and normal pressurizer spray Borate usinp the boric acid tanks, boric acid transfer pumps, and charging pumps.

l l

i

t I

l l ,

l

.

i

p

!

l SENIOR REACTOR OPERATOR Page 55 )

I QUESTION: 076 (1.00)

'

The DOS was notified that the HALON gaseous suppression system in the vital switchgear and battery rooms has being taken out of service for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in support of making modifications to HALON release assemblies. In accordance with OM 3.27, what actions must be taken to

,

maintain the operability of the equipment within the effected rooms?

!

l Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> establish an HOURLY fire watch inspection and provide backup fire l suppression capability for areas where redundant systems or components could l be damage Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> establish fire hose station suppression capability for the affected area, and restrict activities in the affected area to those necessary for continued operation of the equipmen Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> establish a CONTINUOUS fire watch inspection and provide

backup fire suppression capability for areas where redundant systems or components could be damaged.

' Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> inspect the affected area to assure tnat potential fire hazards are

,

minimized and restrict activities in the affected area to those necessary for L continued operation of the equipment.

l l

l I

.

l

,

I l

!

l

.

.

-

SENIOR REACTOR OPERATOR Page 56 l

t-

!

l- QUESTION: 077 (1.00)

'

Unit 2 pressurizer pressure channel PT-431 (BLUE) has been removed from service IAW

! ICP-10.2 in support of channel maintenance. The channel will be unavailable for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. One hour into maintenance testing, RCS temperature channel TE-401 (RED) failed j low. Which one of the following describes the impact on continued operation of Unit 2 and i

actions that must be taken? Unit 2 must be in HOT SHUTDOWN within eight hours.

l

, Unit 2 must be in COLD SHUTDOWN within forty eight hour Unit 2 full power operation may continue, Channel TE-401 (RED) must be placed in TRIP within one hour, Unit 2 operation may continue provided reactor power is reduced to a power level

. below that required to support low power physics testing.

!

l l

l l

l l .

l

!

n

-

i l

[. SENIOR REACTOR OPERATOR Page 57

,

i l

! QUESTION: 078 (1.00)-

!

'

Given the following conditions:

l

-

A General Emergency has been declare +- A radioactive release is occurring from the Auxiliary Buildin *

A radiochemist has injured her leg and cannot exit the Auxiliary Buildin .

An " Emergency Plan Dose Authorization" form (EPIP 5.1a) has been completed.

l Which of the following best describes the requirement for assigning an emergency worker to l assist the radiochemist:

!

l The emergency worker must be a male and have his dose limited to 75 rem TED b.' The emergency worker must be a volunteer and have his or her dose limited to 75 rem TED The emergency worker can be male or female and have his or her dose limited to 25 rem TED The emergency worker can be male or female and his or her exposure may not exceed the annual 10CFR20 exposure limi . SENIOR REACTOR' OPERATOR'. Page 58 QUESTION: 079 (1.00)

'

While performing independent verification (IV) of valves listed in a system checklist, the assigned individuals report four conditions that they believed meet station requirements for designating a valve as " inaccessible." Which one of the following reported conditions would NOT satisfy station requirements for designating a valve as " inaccessible?"

a. - A valve requiring IV is located in the overhead approximately 25 feet above the floor, A valve requiring IV is a normally locked valve and is on a valve checklist to be performed, A valve requiring IV is located within a confined space although no forced ventilation is required for entry, A valve requiring IV is located in an airborne contamination area that would result in a total exposure of eight (8) mrem to perform the I I

)

!

i i

,

l l

x

_

SENIOR REACTOR OPERATOR Page 59

!

l QUESTION: 080 (1.00)

'

Unit 1 is operating at 100% rated thermal power. Containment pressure has increased to a value at which a 30cfm forced containment vent is desired. In accordance with OP 9C,

" Containment Venting and Purging," which one of the following describes an(the) action (s) :i requiring coordination between operations department personnel and health physics department l personnel? Preparing a containment vent permit, Setting up air samplers with air particulate and charcoal filters at selected exhaust fan Breaking Red Locks and opening the containment purge to exhaust filter valve RM-3200T and blower discharge to purge exhaust fan valve RM-3200S as well as shutting and Red Locking these valves once the venting is complete Comparing the containment air particulate monitor RE- 211 and containment noble gas monitor RE-212 INITIAL and FINAL readings to determined if vent termination is required or if containment atmosphere resampling and further venting is require i l

j i

j

SENIOR REACTOR OPERATOR Page 60 QUESTION: 081 (1.00)

The following plant conditions exist on Unit 1:

'

-

Reactor power:' 65% cnd steady

- RCS pressure: 1965 psig and steady )

=

.RCS temperature: 562*F and steady j

-

Instrument Air header pressure: 75 psig and droppin l

-

MSIVs: OPEN

-

"lNSTRUMENT AIR HEADER PRESS LOW' annunciator is energized .

.

"lNSTRUMENT AIR STANDBY COMPRESSOR START" annunciator is energized a

Which of the following actions are required to be performed in accordance with AOP-5B, " Loss ofInstrument Air?"-

._ Attempt to identify and isolate the instrument air system leakag Initiate an SI, and enter EOP-0, " Reactor trip or Safety injection". Trip the reactor, and enter EOP-0, " Reactor Trip or Safety injection."

>. Perform a shutdown in accordance with OP-3A, " Normal Power Operation to Low Power Operation" and AOP-17A, " Rapid Power Reduction."

,

il I

t __ ._ ._ r SENIOR REACTOR OPERATOR Page 61 QUESTION: 082 (1.00)

'

Given that Unit 1 is at 100% rated thermal power and Unit 2 is in REFUELING SHUTDOWN, what is the minimum number of both licensed and non-licensed operators required to ensure that both fire brigade manning and safe shutdown manning requirements are satisfied? QUESTION: 083 (1.00)

Unit 1 is at 100% rated thermal power. It is a VERY hot and humid day. A reactor operator notes that Unit 1 containment intemal pressure has risen to 3.2 psig. Which one of the following describes the MAXIMUM length of time containment pressure can remain at 3.2 psig before actions MUST be taken to place the unit in HOT SHUTDOWN in accordance with Technical ;

Specifications? l i One (1) hour, Six (6) hours, Twenty Four (24) hour ' Seventy two (72) hours.

l u__ ____ - - -

=

SENIOR REACTOR OPERATOR Page 62

. QUESTION: 084 (1.00)

'

Operators are attempting to restore water level in the 28 steam generator in accordance with CSP-H.5, " Response to Steam Generator Low Level." The following conditions are observed:

The 2B steam generator wide range water level is 50 inches.

'

• -

Main feedwater system flow to the 2B steam generator is 40 GP *

AFW flow has not yet been initiate * Centainment pressure is < 10 psi *

Containment radiation levels are < 10E5 R/h Which ONE of the following statements describes the maximum AFW flow rate operators should establish and why? GPM, to avoid exceeding RCS cooldown limit GPM, to minimize the impact on steam generator chemistr GPM, to minimize the effects of water hammer on the steam generator feed rin GPM, to minimize thermal stresses to steam generator components due to a potential dry out condition.

l ,

!

l l

I ,

i

.

SENIOR REACTOR OPERATOR Page 63-l

j- QUESTION: 085 (1.00)

'

l

'

,

- Annunciator " Fire Protection and Smoke Detector Panel" is received on Unit 1. Upon review of l l the C-900 panel, it is determined that the alarm is the result of a fire in the cable spreading j room. in accordance with NP 1.9.14, " Fire Protection Organization," which one of the following I statements correctly describes the control room supervisory response to the fire?

The " duty and call" superintendent reports directly to the scene of the fire, when 1 l on site, to assist in coordination of activities as directed by the DS '

!

.,

l . The DSS reports directly to the scene of the fire to take any action required by !

'

Technical Specifications or plant procedures to maintain control of the plant.

! I

. After assuring that the DSS is in the control room, the duty technical advisor l I

proceeds to the scene of the fire with a portable FM radio and master key on the

- DSS key rin l l After assuring that the duty shift superintendent (DSS) is in the control room, the )

duty operating supervisor (DOS) proceeds to the scene of the fire with a portable l FM radio and master key on the DSS key ring.

I f

i

,-

i SENIOR REACTOR OPERATOR Page 64 QUESTION: 086 (1.00)

'

The following plant conditions exist on Unit 1:

a PZR Level Channel select switch is in its normal position for startu . RCS Temperature is 550* Plant startup is in progres *

Letdown controlis in AUT Given that no operator action is taken, which one of the following statements describes the initial plant response to pressurizer level channel L428 (BLUE) failing high? The Unit will trip on high pressurizer level, Letdown isolation will occur, pressurizer level will slowly increas One of two signals necessary for an automatic safety injection will be satisfie Charging pump speed will decrease to minimum, all back-up heaters energiz .

.

i

,

p SENIOR REACTOR OPERATOR Page 65

'

l l

!

l ' QUESTION: 087 (1.00)L l

Given the following Unit 1 plant conditions: l

- A loss of all AC power has occurred.

' -

The MSIVs indicate SHU Pressurizer PORVs indicate SHUT and letdown has isolate l

• No RMS high alarms are i ECA-0.0, " Loss of All AC Power"is in effec Per ECA-0.0, certain Engineered Safeguards equipment control switches are placed in Pull-out. .

Which ONE of the following events is prevented by this switch alignment? )

a.- An uncontrolled depressurization of the RC ! An uncontrolled start of large loads on safeguards AC buse l An uncontrolled cooldown of the RCS and possible reactor restar ' An uncontrolled use of water that may be needed for long term cooldown.

i

!

-

l L

p L

l SENIOR REACTOR OPERATOR Page 66

l- - QUESTION: 088 (1.00)

'

The following plant conditions exist:

• Procedure in effect EOP- j

-

Containment pressure 65 psig and increasin . As the DOS, you transition to CSP-Z.1, " Response to High Containment Pressure" and upon completion of all steps in CSP-Z.1, you determine that containment pressure is now 61 psig. At this point, you are required to: ,

! exit CSP-Z.1 and enter AOP-1 ) exit CSP-Z.1 and return to EOP-1 at the step in effec reinitiate and remain in CSP-Z.1 until the condition is no longer a RED priorit reinitiate and remain in CSP-Z.1 until the condition is no longer an ORANGE priorit QUESTION: 089 (1.00)

A reactor trip WITHOUT an automatic Si has occurred. Which one of the following conditions ,

'

would require manual Si initiation while performing EOP-0.1, " Reactor Trip Response"? RCS subcooling is 40* Pressurizer levelis 17%.

I Containment pressure 1.5 psi Pressurizer pressure 1700 psi !

l

,.- . - . __

~

-

,

f l

!

[, _

l s i SENIOR REACTOR OPERATOR Page 67

. QUESTION: 090 (1.00)

'

l- Which ONE of the following is the reason the Containment Purge Air Supply and Exhaust l

_ Valves are required to be locked closed during operations at power? The related piping systems outside containment are NOT seismically qualified.

I The valves are NOT seismically qualified to operate during a design basis earthquake.

l The valve actuators do NOT have class 1E penetration conductor overcurrent

! protection devices, The valves capability to close during a design basis loss-of-coolant accident has l

NOT been demonstrate QUESTION: 091 (1.00)

Which one of the following conditions would require a " full" plant evacuation, according to the

" Emergency Plan Implementing Procedures"? A non-scheduled, valid containment evacuation alar Protected area general radiation levels greater than 100 mR/h Unanticipated airborne activity monitor indicating approximately 30 DAC hours, Excessive radioactive surface contamination levels due to a major spill of radioactive material !

l

l

1

_

I

.

i SENIOR REACTOR OPERATOR Page 68

'#  !

,

' QUESTION: 092 (1.00)

'

Unit 1 is at 100% power with no equipment out of service. A diesel generator G02 Technical Specification surveillance test (TS-82) commenced on June 15 at 1000 hr The following time line of events occurred:

-

June 151000 hrs Surveillance test commenced; LCO entered

-

June 151030 hrs DG started for a one hour ru June 151100 hrs DG tripped due to failure of fuel injecto l

.

June 171450 hrs Spare fuelinjector locate i

-

June 210200 hrs Post maintenance test of DG revealed a crack in the governor )

casing. No replacement is available for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, When must the Unit 1 shutdown commence due to required Technical Specification LCO Actions? June 18 at 0200 hrs June 22 at 1000 hrs  ;

i June 22 at 1100 hrs l June 28 at 0200 hrs

!

l I

j

. SENIOR REACTOR OPERATOR Page 69

' QUESTION: 093:(1.00)

'

The actions contained in EOP-3 " Steam Generator Tube Rupture" are being executed due to a

- steam generator tube rupture in the "B" steam generator (SG). Operations personnel have just completed the RCS cooldown portion of the EOP. Due to an erroneous temperature reading, the actual RCS temperature is higher than the target temperature in the EOP. What consequences will result from this condition? The pressurizer will go solid following a subsequent depressurizatio The resulting increase in pressure in the ruptured SG will result in lifting the steam line safety valve (s), Loss of RCS subcooling margin before the RCS pressure will have equalized with the ruptured SG pressure.

I 1 1 The intact SG pressure will be greater than the ruptured SG pressure and will j result in an adequate subcooling margin in the RC j i

!

QUESTION: 094 (1.00)

. Before a release can be made from the post-accident containment ventilation system (i.e., the

'

hydrogen vent path to the vent stack), which of the following equipment must be functioning? The air particulate monitor; RE-212.

, The containment recirculation fans; W1 A, W1B, W1C, and W1D.

! The purge supply and exhaust fans; W2A or W2B and W6A or W6B.

. The auxiliary building exhaust fans; W31 A or W318 and W21 A or W21 l i

r SENIOR REACTOR OPERATOR Page 70 QUESTION: 095 (1.00)

'

A power ascension is being performed on Unit 1 in accordance with OP-1C,' tow Power Operation to Normal Power Operations." With control rods in AUTO, the following conditions

' are observed:

-

Turbine load: Stabl .

Ts 565'F and stabl *

Reactor Power: At approximately 75% and decreasing slowl Control Rods: Stepping into the core slowl RCS T ,: 564*F and decreasing slowl PZR Level: 40% and decreasing slowl *

PZR Pressure: 1935 psig and decreasing slowl Which action MUST be performed? Manually tripi e reacto . Switch the c olling Tavg channel to alternat Place the rod control bank selector switch to MANUA Decrease turbine load to restore T,- T., deviatio f SENIOR REACTOR OPERATOR Page 71 QUESTION: 096 (1.00)

A shutdown was being performed on Unit 1. Reactor power was at 6% when intermediate range '

channel N36 failed high. Which of the following statements best describes how this failure affects the reactor shutdown and subsequent operation of the Nuclear Instrumentation system? The reactor will NOT trip, and purce range NI's will have to be manually re-energize l The reactor will trip on high IR flux, and source range NI's will have to be manually re-energized, The reactor will NOT trip, and source range NI's will re-energize when N35 reaches the proper setpoin The reactor will trip on high IR flux, and source range NI's will re-energize when i N35 reaches the proper setpoin QUESTION: 097 (1.00)

A spent fuel element became detached from the PAB crane fuel handling machine while being moved from the upender to its final location in the spent fuel pool (SFP). Radiation levels, as read on spent fuel pool low range monitor, RE-105, are increasing. Which one of the following is required by AOP-8C, " Fuel Handling Accident in Primary Auxiliary Building?" Maximize SFP purification flo Secure both SFP cooling pump Place both SFP heat exchangers in servic Initiate a limited plant evacuation of the primary a'ixiliary buildin o SENIOR REACTOR OPERATOR - Page 72 QUESTION: 098 '(1.00) '

'

Which of the following conditions would require the suspension of all core alterations per Technical Specifications during refueling operations? ' The primary coolant system boron had an inadvertent dilution to 1900 pp A review of the Unit RO log indicates that the RHR system has been secured for 30 minute . .

I The temporary third door airlock automatic door closer was found broken with two

! personnel doors close l 1 The containment purge and vent isolation system was determined to be inoperable during a routine (once per seven days) surveillance test with the i

purge and vent penetrations close D6 Au 4. A A SLO - ' T~CvT) - TO 4 f $cn.g y-l l

!

l l

l l

I fl l

,

c l

'

- SENIOR REACTOR OPERATOR Page 73

!

QUESTION: 099 (1.00)

'

Following a large steam line rupture, monitoring of Critical Safety Function Status Trees indicates a RED path for CSP-P.1, " Response to Imminent Pressurized Thermal Shock Condition." Which one of the statements below correctly identifies the major component and ;

reason for concern that Pressurized Thermal Shock conditions may result in brittle failure? The RCS piping due to the increased tensile stresses resulting from cooldown of unexpected severit The pressurizer due to increased stresses resulting from a rapid depressurization -

condition at high temperatur The steam generators due to increased stresses resulting from a rapid overpressure condition at low temperatur ; The reactor vessel due to increased stresses resulting from a cooldown of ;

'

unexpected severity or an overpressure condition at low temperatur !

..

.. l l

- SENIOR REACTOR OPERATOR Page 74 i i

i QUESTION: 100 (1.00)

'

While refueling operations are in progress, a package is submitted which would de-energize

120Vac buses: (1Y01 and 1Y101) and (1YO3 and 1Y103), so that work can be performed on PRN141 and 427 - How would this affect refueling operations if the IN40 gammametric detector is

out of service? i l

, fuel moves can not continue, because only one source range channel is i energized.

I

'

! - fuel moves can continue, since power range indication is not needed during refuelin fuel moves can not continue, because no source range channels are energized.

(

fuel moves can continue, as long as P-10 bistables are tripped per applicable I & C procedures.

l

!

i i i i

l L i

i

!

'

!

(""""" END OF EXAMINATION """"**)

!

t=

-

L SENIOR REACTOR OPERATOR Page 75 l

l ANSWER: - 001.(1.00) ANSWER: 006 (1.00)

l d REFERENCE: REFERENCE:

CSP-l.3, Revision 7; BG CSP-l.3, Revision 7 EOP-3, " Steam Generator Tube Rupture,"

l

. Modified,1994 PB Exam, QNUM 38735 Rev DRAFT, Caution before step 31, p 2 Background Document EOP-3," Steam Generator Tube Rupture," Rev DRAFT, p 6 ANSWER: 002 (1.00) K309 ..(KA's)

REFERENCE:

SEP-1.1, Revision Modified,1996 PB Exam, QNUM 47555 ANSWER: 007 (1.00)

b 000025K101 ..(KA's) REFERENCE:

AOP-0.0, " Vital DC System Malfunction," Rev 13, Step 2.5, p ANSWER: 003 (1.00) A203 ..(KA's)

REFERENCE:

EOP-0, Revision 27 New ANSWER: 008 (1.00)

a REFERENCE:

ANSWER: 004 (1.00) TRHB 13.3, " Instrument and Control Systems b Description:. Reactor Protection (Reactor REFERENCE: Trip)," Rev. 4, p 3 .

BG ECA-0.0," Loss of All AC Power," Re I DRAFT, p 57 001000K603 ..(KA's) -

000055K302 ..(KA's)

ANSWER: 009 (1.00)

d ANSWER: 005 (1.00) REFERENCE:

d- TRHB 13.4, " Instrumentation and Control REFERENCE: Systems: Engineered Safety Features EOP-0, " Reactor Trip or Safety injection," Actuation Instrumentation (Safeguard Rev DRAFT, Step 24, p 21 system)," Rev 4, p 3 BG EOP-0, " Reactor Trip or Safety injection,"

DRAFT, p. 43 013000A101 ..(KA's)

000009K323 ..(KA's)

ANSWER: 010 (1,00)

' REFERENCE:

MDB 3.2.3, Rev 8 WEST 883D195 013000A403 ..(KA's)

t SENIOR REACTOR OPERATOR Page 76 ANSWER: 011 (1.00) ANSWER: 015 (1.00)

b REFERENCE: REFERENCE:

TRHB 10.1, " Primary System Description: EOP-1.2, "Small Break LOCA Cooldown and Reactor Containment," Rev 6, p 15; Depressurization Unit 1," Rev. DRAFT,'

TRHB 10.16, " Primary System Description: Foldout Page Engineered Safeguard Systems," Rev. 4, p.21 000009A234 ..(KA's)

022000A301 ..(KA's)

ANSWER: 016 (1.00) ANSWER: 012 (1.00) REFERENCE:

d OP-3A, " Normal Power Operation to Low REFERENCE: Power Operation," Rev. 43, pg. TRHB 13.3, " Instrument and Control Systems Description: Reactor Protection (Reactor 004000K601 ..(KA's)

! Trip)," Rev. 4, FIG 13.3.4, " T and Tavg l Instrumentation" l ANSWER: 017 (1.00)

002000K512 ..(KA's) REFERENCE:

l BG CSP-l.3, " Response to voids in Reactor ANSWER: 013 (1.00) Vessel," Rev. 7, p c

! REFERENCE: 016000G ..(KA's)

TRHB 13.3, " Instrument and Control Systems Description: Reactor Protection (Reactor Trip)," Rev. 4, p. 36; stpt 4.2, Rev 5 ANSWER: 018 (1.00) K106 ..(KA's) REFERENCE:

TRHB 10.1, " Primary System Description:

Reactor Containment," Rev. 6, pgs.14 and ANSWER: 014 (1.00) 15 REFERENCE: 022000G ..(KA's)

l AOP-6B, " Stuck Rod or Malfunctioning Position Indication Unit 1," Rev.10, pg.1

'

i ANSWER: 019 (1.00)

l 000005A201 ..(KA's) REFERENCE:

EOP-0, " Reactor Trip or Safety injection,"

,

Rev. DRAFT, pg 7 and 8 .

000007A201 ..(KA's) l

,

i

c

[ SENIOR REACTOR OPERATOR Page 77 ANSWER: 020 (1.00) ANSWER: 025 (1.00) REFERENCE: REFERENCE:

Steam Tables (Material required for the REl 13.0, " Quadrant Power Tilt," Rev.14, examination) AOP-6H, " Quadrant Power Tilt," Step 2 -

RNO 000008K101 ..(KA's)

015000K516 ..(KA's)

ANSWER: 021 (1.00)

l ANSWER: 026 (1.00)

REFERENCE: TRHB 10.8, " Safety injection System." Rev. REFERENCE:

6, pgs. 5,10, and 17 BG EOP-1.2, "Small break LOCA Cooldown and Depressurization," Rev.18, p.50 006000K603 ..(KA's)

l 022000K405 ..(KA's)

ANSWER: 022 (1.00)

l ANSWER: 027 (1.00)

l REFERENCE: b.

l TRHB 13.13," Rod insertion and Delta T REFERENCE:

t Deviation Alarms." Rev.1, pgs. 2, 3, and 4 TRHB 13.4,"ESF Actuation Instrumentation

! (Safeguard System), Rev. 4, pgs. 2,5, and 6 l 000003A101 ..(KA's)

039000K408 ..(KA's)  ;

I

'

ANSWER: 023 (1.00)

l ANSWER: 028 (1.00)  !

l REFERENCE: OP-5A, " Reactor Coolant Volume Control," REFERENCE: )

'

Rev.30. TRHB 13.9, " Instrumentation and Control THRB 10.6, " Chemical and Volume Control System Description: Condenser Steam System." Rev 6, Dump, " Rev. 3, p. 5 000022K307 ..(KA's) 04100K603 ..(KA's)

ANSWER: 024 (1.00) ANSWER: 029 (1.00) REFERENCE: REFERENCE:

TRHB 13.8, " Rod Speed and Direct Control." AOP-88," Irradiated Fuel Handling Accident Rev.1, pgs. 2, 3, and 4 in Containment." Rev. 6, pgs 1 - 4 035000K501 ..(KA's) 000036G ..(KA's)

i

I

l SENIOR REACTOR OPERATOR Page 78 l

l l ANSWER: 030 (1.00) ANSWER: 036 (1.00)

i a  !

I REFERENCE: REFERENCE: I SEP-1, " Degraded RHR System Capability," FSAR, Section 6.5-8, page 6.5-8 l l

Rev.1, p.3,5 ,

003000A408 ..(KA's)

000025A103 ..(KA's)  ;

ANSWER: 037 (1.00) ,

ANSWER: 031 (1.00) I b REFERENCE:

REFERENCE: FSAR, Section 7.2, Page 7.2-16 l BG EOP-2, " Faulted Steam Generator ,

Isolation," Rev. DRAFT, p.14 075000K307 ..(KA's) I

000040K304 ..(KA's)

'

ANSWER: 038 (1.00)

d ANSWER: 032 (1.00) REFERENCE:

d HP1.11.1, " Personnel Contamination Monitor i REFERENCE: (PCM-1B), Contamination Alarm Response, i FSAR, chapter 8, page 8.8-3 and Rev.13, p 4,5 )

062000K102 ..(KA's) 194000G ..(KA's)  ;

ANSWER: 033 (1.00) ANSWER: 039 (1.00)

d d REFERENCE: REFERENCE:

FSAR, chapter 8. Page 8.8-6,7 [3.6/3.7]. NP 1.9.4, " Confined Spaces Procedure," Re , p. 5,10 008000A308 ..(KA's)

194000G ..(KA's)

ANSWER: 034 (1.00)

a or d ANSWER: 040 (1.00)

REFERENCE: d EOP 1.3, Rev. 20, Step REFERENCE:

NP 1.2.3, " temporary changes," Rev. 6, p.2-5 026000A208 ..(KA's)

194000G ..(KA's)

ANSWER: 035 (1.00)

b REFERENCE:

FSAR, Chapter 8, page 8.8-6 064000A307 ..(KA's)

,

..

I

\

.

SENIOR RE. ACTOR OPERATOR Page 79 ANSWER: 041 (1.00) - ANSWER: 046 (1.00)

b a.

REFERENCE: REFERENCE:

NP 2.1.2, " Independent Verification and TRHB 11.4, " Secondary Systems Concurrent Checks," Rev. O, p. 8 Descriptions: auxiliary Feedwater System,"

Rev.6,p.11 194000G ..(KA's)

061000K302 ..(KA's)

ANSWER: 042 (1.00) ANSWER: 047 (1.00)

REFERENCE: b Site NGET training REFERENCE:

0162A," Motor-driven Auxiliary Feedwater 194000G ..(KA's)' System (P-38A & P-388)," Rev.16, pg. 9,20,

ANSWER: 043 (1.00) 061000A204 ..(KA's)

c REFERENCE:

Setpoint 3.1,1 ANSWER: 048 (1.00)

Logic diagram 883D195, Sheet 12 d REFERENCE:

000051A202 ..(KA's) RMSASRB Cl RE-223, " Radiation Monitoring system Alarm Setpoint & Response book Channel Information Sheets," Rev. 2 ANSWER: 044 (1.00) ' Waste Distillate Tank Overboard Monitor d

REFERENCE: 000059K201 ..(KA's) j TRHB 10.16, " Primary Systems Descriptions:

Engineered Safeguards Systems," Rev. 4.

p.20,21 ANSWER: 049 (1.00)

EOP 0, page 12 c REFERENCE:

000062A102 ..(KA's) RMSASRB Cl RE-218 " Radiation Monitoring system Alarm Setpoint & Response book Channel Information Sheets," Rev. 2 ANSWER: 045 (1.00) ' Waste Disposal System Liquid Monitor b  !

REFERENCE: 194000G ..(KA's) l TRHB 11.14, " Secondary Systems l Descriptions: Fire Protection System," Rev. 8, I pgs. 3,4,,20 000067K102 ..(KA's)

SENIOR REACTOR OPERATOR Page 80 i ANSWER: 050 (1.00) ANSWER: 054 (1.00) 1 l a b REFERENCE: REFERENCE:

TRHB 10.15," Primary Systems Descriptions: AOP-5B, " Loss of instrument Air," Rev.12, l Waste Disposal System," Rev. 7, p.22 pg.21 i

' '

P&lD PB 01MCVK00000804 l 071000A416 ..(KA's)

000065A208 ..(KA's)

ANSWER: 051 (1.00) ANSWER: 055 (1.00)

REFERENCE: b TRHB 13.12, " Instrumentation and Control REFERENCE:

System Description: TRHB 13.7, " Instrument and Control systems Process and Area Radiation Monitoring Descriptions:

,

System," Rev. 3, pg.15,16, 36 Feedwater Control System," Rev. 2, p RMSASRB Cl RE-235, pg. 2 3,4,23 Setpoint 1.4, 2.1, 4.2, 072000K104 ..(KA's)

059000A306 ..(KA's)

ANSWER: 052 (1.00) ANSWER: 056 (1.00)

REFERENCE: b TRHB 13.12, " Instrument and Control system REFERENCE:

Description: Process and Area Radiation OI-62A," Motor-driven Auxiliary Feedwater Monitoring System," Rev. 3, pgs. 5,15,16 System (P-38A & P-38B)" Rev.16, p 4 RMSASRB Cl RE-116 " Radiation Monitoring TRHB 11.4," Secondary systems system Alarm Setpoint & Response book Descriptions: Auxiliary Feedwater System,"

Channel Information Sheets," Rev. 2 Rev. 6, p 6 Demineralizer Valve Gallery AOP-58, Attachment R 072000A101 ..(KA's) 061000A207 ..(KA's)

ANSWER: 053 (1.00) ANSWER: 057 (1.00) c REFERENCE: REFERENCE:

TRHB 10.7, " Primary Systems Descriptions: ?

! 194000G ..(KA's) Residual Heat Removal System," Rev. 9, p. 3 Figure

005000K403 ..(KA's) i

l V

t u l

SENIOR REACTOR OPERATOR Page 81 ANSWER: 058 (1-.00) ANSWER: 063 (1.00)

d d REFERENCE: REFERENCE:

TRHB 11.4, " Secondary Systems ARB 1C031D 3-1," 1P-1 A or B RCP No.1 Descriptions: Auriliary Feedwater System,' Seal Water Outlet Temperature High, Rev. 4 Rev. 6, p 3,13 p.1 Setpoint AOP-1B, Reactor Coolant Pump Malfunction Unit 1, Rev.10, pgs. 3,5,6 Setpoint ANSWER: 059 (1.00)

b 000015G ..(KA's)

REFERENCE: '

EPIP 6.1, " Limited Pbnt Evacuation," Rev.

15,p ANSWER: 064 (1.00)

c.

194000G ..(KA's) REFERENCE:

RMSASRB 1.0, 01-80 ANSWER: 060 (1.00) 073000A401 ..(KA's)

d REFERENCE:

EOP-0.2, " Natural circulation Cooldown," ANSWER: 065 (.'.00) i Rev. DRAFT, pgs. 8,9,12,17,18,20 I REFERENCE:

OP-3A, " Normal Power Operation to Low ANSWER: 061 (1.00) Power Operation," Rev. 43, pg. 2 b ,

REFERENCE: 001000K513 ..(KA's)

TRHB 10.6, " Primary Systems Description: ,

chemical and Volume Control System" Re '

6,pgs.16,17 ANSWER: 066 (1.00)

EOP O.2, ." Natural Circulation Cooldown" b Rev.7,p REFERENCE:

EOP-0.1, " Reactor Trip Response," Re AOP 0.0, " Vital DC System Malfunction, Rev.

DRAFT, p. 9 4,p.9 AOP-6E, "Altemate Boration/ Loss of Shutdown Margin," Rev. 7, pgs. 3 063000G ..(KA's)

000024K302 ..(KA's)

ANSWER: 067 (1.00)

a ANSWER: 062 (1.00) REFERENCE:

deleted AOP-5B, " Loss of Instrument Air, " Rev.12, REFERENCE: pgs, 24, 28, 39, and 40 072000G ..(KA's) 078000K302 ..(KA's)

-

c SENIOR REACTOR OPERATOR Page 82 ANSWER: 068 (1.00) ANSWER: 073 (1.00)

a e REFERENCE: REFERENCE:

TRHB 13.5, Rev.3 , pgs. 21, 25,28,29, Figure TRHB 13.3, " Instrument and Control System 13.5.13, " Low Temperature Overpressure Descriptions: Reactor Protectica (Reactor Protection Logic," Logic Sheet 18 Trip)," Rev. 4, pgs. 6, 9, 2 /.34,38, 010000K301 ..(KA's) 003000K201 ..(KA's)

ANSWER: 069 (1.00) ANSWER: 074 (1.00)

b b REFERENCE: REFERENCE:

TRHB 13.6, " Instrument and Control systems TRHB, " Instrument and Control System Description: Description: Nuclear instrumentation System Pressurizer Level Control System," Rev. 2 , (Excore Instrumentation), Rev. 3, Logic Sheet 18 Setpoint K604 ..(KA's) 012000K604 ..(KA's)

ANSWER: 070 (1.00) ANSWER: 075 (1.00)

c REFERENCE: REFERENCE:

Steam Table AOP-6E, Revision DRAFT (3/19/99); P&lD Chemical and Volume Control (West 017000K502 ..(KA's) 684J701, SH2)

New ANSWER: 071 (1.00) 000024A201 ..(KA's)

d REFERENCE:

AOP-9B, " Component Cooling System ANSWER. 076 (1.00)

Malfunction," Rev.12, pgs. 3, 4, 5, 8,12, and REFERENCE:

OM 3.27 000026K303 ..(KA's) NEW 086000K405 ..(KA's)

ANSWER: 072 (1.00)

d REFERENCE:

- 01-37, " Shifting of instrument Supply Bus Feeders," Rev. 30, pgs. 3,11,13, and 20 194000G ..(Kh's)

~

_ _ _

.. _ ____ ______ _ _ - _ _ - - - - - - - - -

SENIOR REACTOR OPERATOR Page 83 ANSWER: 077 (1.00) ANSWER: 082 (1.00)

) c.

~

REFERENCE: REFERENCE:

Technical Specification 15.3.5, Table 15.3.5-1 TS 15.6.2.2, OM- PBNP Drawing 883D195, Sheets 7 and 1 Modified,1995 PB Exam, ONUM 45748

-

New 194001K ..(KA's)

}+ 194000 ..(KA's)

ANSWER: 083 (1.00)

ANSWER: 078 (1.00) REFERENCE:

REFERENCE: Tech. Spec.15. PB Station GET Training Program Modified,1996 PB Exam, QNUM 47548 194000 . . (KA' - 194001K ..(KA's)

j ANSWER: 079 (1.00) ANSWER: 084 (1.00)

o d.

}

J__ REFERENCE: REFERENCE:

NP 2.1.2, Revision 0 CSP-H.5, Revision 7;BG CSP-H.5, Revision 7 New and DRAFT (3/16/99).

Modified,1996 PB Exam, QNUM 47559

-

194000 ..(KA's)

000054K102 ..(KA's)

"

ANSWER: 080 (1.00) ANSWER: 085 (1.00)

REFERENCE: OP SC, devision 43 REFERENCE:

New NP 1.9.14, Revision 1 Modified,1996 PB Exam, ONUM 47609 194000 ..(KA's)

000067G ..(KA's)

ANSWER: 081 (1.00) ANSWER: 086 (1.00)

REFERENCE: AOP-58, Revision 12 REFERENCE:

Modified,1995 PB Exam, QNUM 45771 TRHB 13.6, Revision 2; AOP-1 A U2, Revision

000065A201 ..(KA's) PBNP Drawing 883D195, Sheet 1 Modified,1998 PB Exam, ONUM 004 000028K305 ..(KA's)

--

... i SENIOR REACTOR OPERATOR Page 84 ANSWER: 087 (1.00) ANSWER: 092 (1.00) REFERENCE: REFERENCE:

BG ECA-0.0, Revision 19 Tech Spec 15.3.7.B. Direct,1996 PB Exam, QNUM 47545 Direct,1998 PB Exam, QNUM 012 000055A106 ..(KA's) 064000G . .(KA's)

ANSWER: 088 (1.00) ANSWER: 093 (1.00) REFERENCE: REFERENCE:

CSP-Z.1, Revision 11, BG CSP-Z.1, Revision EOP-3, Revision 24; EOP Background 10 Document (EOP-3), Revision, DRAFT Direct,1995 PB Exam, ONUM 45762 Direct,1998 PB Exam, QNUM 072 000069G ..(KA's) 000038K306 ..(KA's)

ANSWER: 089 (1.00) ANSWER: 094 (1.00) REFERENCE: REFERENCE:

EOP-0.1, Revision 19 ; Foldout Page for EOP-0 Serie K502 ..(KA's)

Direct,1991 PB Exam, ONUM 27352 000007A202 ..(KA's) ANSWER: 095 (1.00) REFERENCE:

ANSWER: 090 (1.00) AOP-6C U1, Revision 8 Direct,1998 PB Exam, QNUM 086 REFERENCE:

Tech. Spec. Bases 15.3.6.A. K107 ..(KA's)

Direct,1995 PB Exam, QNUM 47600 194001K ..(KA's) ANSWER: 096 (1.00)

.

, REFERENCE:

ANSWER: 091 (1.00) TRHB 13.1, Revision 3; Logic Sheets 11 and REFERENCE: Direct,1991 PB Exam, ONUM 27272 EPIP 6.1, Revision 15; EPIP 6.2, Revision 19 Direct,1991 PB Exam, QNUM 27285 000033A208 ..(KA's)

194001K ..(KA's)

I

_ _ _ _ _ _ _ - _

-

_

SENIOR REACTOR OPERATOR- Page 85

. ANSWER: 097 (1.00)

d.-

1 REFERENCE: -

!

AOP-8C, Revision 6

Direct,1998 PB Exam, QNUM 047

'

(

000036G ..(KA's)

A

' NSWER:_ 098 (1.00)

i

' REFERENCE:

l Tech Spec.15. Direct,1995 P3 Exam, QNUM 4577C 034000G ..(!(A's)

ANSWER: 099 (1.00) REFERENCE:

BGCSP-P.1, Revision 17; CSP-P.1, Revision 17 and DRAFT (3/16/99).

Direct,1991 PB Exam, QNUM 27363 000040K101 ..(KA's)

ANSWER- 100 (1.00)

c REFERENCE

TRHB 13.1, Rev. 3, p. 2 000032G ..(KA's)

I (*""*"" END OF EXAMINATION *""*"")

!

-

dENIOR REACTOR OPERATOR Pa9e 86 ANSWER KEY 001 d '021 d 041 b 061 b 081 a 002 a 022 c 042 d 062 deleted 082 c

003 c 023 a 043 c 063 d 083 a-004 b 024 c 044 d 064 c 064 d 005 d 025 b 045 b 065 d 035 d 006 d 026 ' d 046 a 060 b 086 d 007 b 027 b' 047 b 067 a 087 b 008 a 028 a 048 d 068 a 088 b i 009 d CL' b 043 c 069 b 089 d 010 a 030 c 050 a 070 c 090 d 1 011 b 031 b 051 c 071 d 091 b 012 d 032. d 052 c 072 d 092 b 013 c 033 d 053 c 073 c 093 c 014 b 034 a or d 054 b 074 b 094 d 015 b 035 b 055 b 075 a 095 c 016 b 036 a 056 b 076 a 096 b 017 c 037 a 057 c 077 a 097 d 018 c 026 d- 058 d 078 c 098 a I 019 d 039 d 059 b 079 b 099 d ,

3 020 b 040 d 060 d 080 b 100 c l (""""" END OF EXAMINATION ****""")

t

I

-