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NRC Operator Licensing Exam Rept 50-282/97-306OL & 50-306/97-306OL (Including Completed & Graded Tests) for Tests Administered on 970616-20.Exam Results:All License Applicants Passed Respective Initial Exams
	ML20210K685
Person / Time
Site:	Prairie Island
Issue date:	08/13/1997
From:	; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:	;
Shared Package
ML20210K614	List: IR 05000282/1997306; ML20210K608; ML20210K629;
References
	50-282-97-306OL, 50-306-97-306OL, NUDOCS 9708200029
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. . U. S. NUCLEAR REGULATORY COMMISSION REGION lli Docket Nos: 50 282:50 306 Licensos No: NPF 42; NPF 60 Reports No: 50 282/97306(OL); 50 300/97306(OL) Licensoo: Northorn States Power Company Facility: Pralrlo Island Nuclear Plant, Units 1 & 2 Location: Wolch, Minnesota Datos: June 10 20,1997 Examinors: R. Dailey, Examinor RU: E. Plottner, Examinor Rlli J. Hansen, Resident inspector /Examinor Rlll Approved by: M. Leach, Chief, Operator Licensing Branch Division of Reactur Safety , g?'200029 970813 ' ADOCK 05000282 V M .l,

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. EXECUTIVE SUMMARY Prairie Island Nuclear Plant, Units 1 & 2 NRC Examination Reports 50 282/97306;50 306/97300 A licensee developed and NRC approved initial operator licensing examination was administered by the NRC, with the exception of the written portion, to four applicants who applied for Senior Reactor Operator (SRO) licenses. This examination process was composed of a two week period involving the validation and administration of the written and operating examination to each applicant.

Examination Results: Alllicense applicants passed their respective initial examinations and were issued

licenses to rgcrate the Prairie Island Nuclear Plant, Units 1 & 2. Two were issued at the Reactor Operator level and two were issued at the Sonict Reactor Operator level.

The license applicants performed wellin the simulator, addressing normal and

abnormal situations in a competent manner. However, communication between operators in the simulator did not include proper confirmation of orders given or information received. (Section 05.3) The examiners identified some operating procedure deficiencies which were

promptly addrassed by the licensee. (Section 03.2) Examination Matprials: The dmam!c simulator examinations and the administrative section of the

examination were at the appropriate difficulty level and required little modification.

The job performance measures were of an appropriate difficulty level but did require some clarification. The written examination required significant modification to bring the examination to an acceptable level. All administcred examinations were of an appropriato quality. (Section 05.2)

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, - ! , a Report Details I.. Optfallent

Conduct of Operations

01.1 General Comments The examiners reviewed selected operations and administrative procedures during the review, validation, and administration phases of the examination process (See Section 03 for additional comments). Additionally, the examiners observed . licensed operator performance during steady state power operations on Units 1 and 2.

The examiners observed an appropriate level of attention and detail during a control . board walkdown and parameter logging evolution. Operator actions were consistent with licersee expectations as outlined in their administrative procedures.

Operations Procedures 6nd Documentation 03.1 Ooeratina Procedure Adeauaev - , a.

Intpection Scope l The examiners reviewed selected operating procedures during the examination development and administration process. The examiners also reviewed administrativo procedure H14. " Procedure Writer's Guide," Revision 6.

b.

Observations and Findinot The examiners noted itolated procedural guidance deficiencies during examination material review'and validation. Four operating procedures were compared to the standards preacribed in administrative procedure H14. The cross section of procedures included emergency operations, abnormal operations, and refu61ing operations. Examples of the deficiencies found included: A lack of component nomenclature (i.e. valve name and number designator)

to assist personnelin identifying the correct one to operate.

A lack of sufficient information (i.e. panel location) to successfully complete

the task withoet assistance from more experienced personnel.

Multiple action items contained within one step with a single sign off.

c.

Conclusions The examiners discussed these procedural guidance deficiencier. with licensee management and provided specific exermples as noted, The licensee took prompt action to generate procedural change requests, None of these deficiencies adversely affected the applicants' performances during the licensing examination.

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Operator Training and Qualification 05.1 General Comments Operator initiallicensing examinations were administered at the Prairie Island Nuclear Plant to four Senior Reactor Operator (SRO) applicants during the week of June 16,1997.

The initial operator licenso examinations woro developed by the licensee in accordance with guidance prescribed in NUREG 1021, * Operator Licensing Examiner Standards," Revision 7, and as noted in NRC Generic Letter 95 06, " Changes in the Operator Licensing Program," August 15,1995. The NRC reviewed and approved the proposed examination material prior to administration by liennsing examinors, with one exception. The NRC allowed the licensoo staff to administer the written portion of the examination without NRC presence.

The licenseo developed examination material was submitted to the NRC in a timely manner and ahead of the scheduled deadlino as specified in NRC Gl. 95 06. The NRC review determined that the outline and the written and operating examination material met the requiromonts as specified in NRC GL 95 06. However, the quality and lovel of difficulty of the written examination questions required substantial revision to ensure consistency with previously administered examinations. (See Section 05.2).

05.2 Examination Premation and Validation a.

Examination Scong Examination guidance contained in NUREG 1021, Revision 7, and NRC Generic Letter 95-06 was mado available to the licenseo for the development of the written and operating examination material. The examiners were able to review, reviso, and validate the written and operating examination material during the week of June 2,1997. The validation process involved the verification of licenseo administrative and operating procedures to ensure that exam validity and content were consist with operations expectations, b.

Observations and Findinna The review and verification of the examination outline and examination material was f acilitated by early submittal of the required material and by the level of detail contained in the proposed examination naterial.

The operating examination consisted of an administrative section, a job performance measure (JPM) walkthrough section, and an operating (dynamic scenario set) section. The operating examination contained a mixed level of quality and difficulty. The dynamic scenario set was well structured and contained all of the necessary elements to evaluate the applicants at the senior operator level. Both sconarios required very little revision prior to administration. The watkthrough JPMs were appropriately structured but contained some elements that needed clarification or some elements that were omitted. Ono JPM required replacement

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,, , while some other JPMs required a mixed level of revision. The administrative tasks were appropriate for the senior operator position and required little change.

The written examination contained 100 multiple choice questions and was evaluated as meeting the standard as outlined in NUREG 1021, Revision 7.

However, selected questions varied widely in quality and difficulty from straight memory to comprehension / analysis. The quality and difficulty of the overall examination was marginal as compared to previously administered examinations and required significant improvement. The licensee revised the examination and - with NRC oversight was able to develop a high quality examination. The overall difficulty was raised to an acceptable level as indicated by an average score of 84.25E The examiners were afforded the appropriate level of assistance and access to the plant specific simulator during setup and validation of the operating examination in accordance with ES 301 of NUREG 1021. A high level of examination security was maintained during this validation process. Additionally, a portion of the walkthrough JPMs were validated using the site facility. The licensee's assistance during the examination review and validation process provided the NRC examiners with appropriate operational expertise and clarification in technical matters.- The written and operating examinations were verified to be valid for the tasks and conditions specified, c.

Conclusions The dynamic simulator examination was well structured and of the appropriate difficulty level. The job performance measures were of the appropriata difficulty

level but did require some clarification or some additional information. The administrative section was appropriate.

l The written examination was of marginal quality and required significant revision by the NRC.

The licensee demonstrated good support to the examination process by maintaining a high level of examination security and early and complete submittal of l examination materials.

l 05.3 Examination Administration i a.

Ex6mination Scong The written examination was administered on June 16,1997, by the licensee's training staff with approval from the NRC. The examiners administered the operating examination between June 17 19,1997, to four applicants using guidance provided in NUREG 1021i Revision 7. The operational examination involved extensive use of the plant specific simulator during dynamic scenario sets , and control room job performance tasks. Additional testing was conducted at the site facility.

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QhtflyAllens and Find;not r The applicants demonstrated satisfactory knowledge and skills in handling the assigned operational tasks required during normal and abnortnal evolutions. All of j the applicants demonstrated a weakness in their communication skills as demonstrated during the dynamic scenarios. A comnion problem was improper or l incomplete repeat backs and inconsistent acknowledgements to reports (i.e. "Yes' or " Understand *).

, I Additionally, the following individual weaknesses were noted during the operating examination: Procedural guidance usaga during JPMs was not consistent with licensee

expectations. More than one applicant either overlooked a step or performed an evolution on the wrong equipment.

JPM follow up questions identified a lack of understanding of a recent

modification to the auxiliary feedwater motor-driven pump trip circuit, JPM follow up questions idantified a lack of understanding of the effect that + a loss of Bus 111 instrument power would have on the RHR system.

The following common weaknesses were noted on the written examination: ' Knowledge of the electrical power supply to the charging pumps when

cross tied.

Ability to operate or verify proper operation of the auxiliary feedwater

system following an anticipated transient without scram condition.

Knowledge of the process for removing decay heat from the core upon entry

into function recovery procedures.

Abill.y to perform without reference the immediate operator actions required

for a loss of component cooling.

Ability to d,termine the required core exit temperature that must exist

followir.g a reactor coolant system cooldown in response to a steam generator tube rupture event.

These deficiencies are provided as feedback to the licensee's systems approach to ' training process. No response is required, c.

Conclusions Overall, the operator performance demonstrated during the operating examination was appropriate to ensure safe reactor operations. All of the applicants demonstrated appropriate skills required to handle any operational task. Even though poor use of repeat backs and other performance weaknesses were noted, the NRC examiners concluded that no programmatic deficiency was indicated.

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, ( . . 05.4 Epat Examination Activities a.

Examination ScoD2 The examiners reviewed the licensee's formally documented post-examination comments on the written examination. The review and documentation of the resolution of the comments were performed using the guidance in NUREG 1021, " Operator Licensing Examinor Standards," Revision 7.

b.

Qhservations and Findinns The writton tests were considered technically accurate with exceptions noted by the licenson post examination comments included in enclosure 3. The NRC resolutions to the comments are also included in enclosuro 3.

c.

Conghdong The examinors concluded that the identified changes were appropriate.

05.5 Simulator Fidolity a.

Examination Scono The examiners reviewed and noted any simulator fidelity issues, including the offectiveness of the licensee to identify and correct simulator discrepancies. The review and documentation of the simulator fidelity issues were performed using the guidance in NUREG 1021, " Operator Licensing Examiner Standards," Revision 7.

b.

Qbjervations and Firdinag The examiners observed no simulator modeling deficiencies during both the examination validation and administration weeks. The licensoo identified deficiencies did not preclude completion of valid evaluations of licenso applicant performanco, c.

Qpnclusions The examiners concluded that fidelity problems were documented and tracked by the licensoo. The inspectors had no additional simulator fidelity concerns. See attached simulator f acility report (enclosure 2).

V. Manaatment Meetinal X1 Exit Meeting Summary The examiners presented their observations and findings mentioned previously to members of the licensee's management on June 20,1997. The licensee acknowledged the findings presented. No proprietary information was given to the examinors during the examination process or at the exit meeting.

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. .__. _ -_..._ _ _. _ _ _ _ _. _ _ _ _ _. _ _ ._. .__._ . . i . l' . , , -, P PARTIAL LIST OF PERSONS CONTACTED . I ' t Lkt0119 ' i K. Albrecht, General Superintendent of Engineering M. Gardzinski, Senior Technicat instructor / Simulator > ' M. Ladd, Training issues Manager J. Lash, Senior TechnicalInstructor/ Licensed Operators l P. Valtakis, Shif t Manager / Operations l D. Westphal, Superintendent of Operations Training l l g ! S. Ray, Senior Resident inspector

P. Krohn, Resident inspector , ITEMS OPENED, CLOSED, AND DISCUSSED NONE l

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, I I ! . ! Enclosure 2 , i - SIMUL ATION FACILITY REPORT , i facility Licensee: Prairie Island Nuclear Plant, Units 1 and 2 l

Facility Licensee Docket No: 50 282:50 306 ' Operating Tests Administered: June 17 19,1997 , ! The following documents observations made by the NRC examination team during the ' June 1997,initiallicense 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 f acility other than to provide information which may be used in future evaluations. No licensee action is regulrad in response to these observations.

During the conduct of the simulator portion of the operating tests, the following item was ! observed: . ITEM DESCRIPTION . NONE NONE , l __.

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Enclosure 3 l

Written Examination Post Review Comments ' l.

QUESTION No. 4: ! t Given the following Unit 2 plant conditions: Normal 100% power lineup , 21 CC pump is supplying all CC loads l 21 RCP seallenkoff is 5.7 gpm 22 RCP sealleskoff is 2.7 gpm t

A small thermal barrier HX leak is believed to have developed and the engineer I would like to sequential lv close the CC outlet from the thermal barrier heat ! exchangers to attempt to isolate the leak. Which ONE of the following actions is correct for this situation while still maintaining 100% power? . . . I a.

21 and 22 thermal barrier HX can both be isolated, b.

21 and 22 thermal barrier HX must both remain in service.

c.

21 thermal barrier HX can be isolated and 22 thermal barrier HX must remain in service.

i d.

21 thermal barrier HX must remain in service and 22 thermal barrier HX can ' be isolated.

. , ANS: d.

, REF: C3 Precautions / C14 AOP2 Subsequent Actions

P8170L 002 "RCP's" Rev 2, p. 27/Obj. 5 K/A: 003000K404 12.8/3.1) i Licensee Recommendation ' Accept only a. as the correct answer Licensee Comments , , , The question proposes a situation in which RCP thermal barrier leakage is - suspected. This action is supported by Attachment #1, IC14 AOP2 " Leakage into , the Component Cooling System," Step 2.4.3, which directs the operator to close ' the affected thermal barrier coolant outlet valves if flow or temperature is higher . than normal. 'RCP sealinjection will provide adequate seal cooling for this situation ! since #1 seat leakoff flows are less than the normal sealinjection value of 6 to [ 10gpm.

- Answer a. was the answer on the original version of this question. Changes to the question were made during the review procen* and the answer did not get changed.

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. l . NRC Resolution A review of the provided documentation support the discussion on expected operator actions. Based upon NRC review of this and other references, the NRC - will change the correct answer from "d" to "a.' 11.

QUESTION No.12: Given the following plant conditions: ' The unit has just increased power from 50 to 100% + An IRPI in indicates a misaligned rod.

An actual rod misalignment would be determined by comparing core exit thermocouple between... a.

Adjacent core quadrants before and after the power change.

b.

Symmetric core quadrants before and af ter the power change.

c.

Adjacent core quadrants af ter the power change.

d.

Symmetric core quadrants after the power change.

ANS: b REF: P8170L 001 " Reactor & Internals" Rev.1; p.16/Obj. 2; SP 1319 ' K/A: 017020K402 13.1/3.6) i Licensee Recommendation Accept only a. or b. as the correct answer ' . Licensee Comments , Refer to Attachment #2, SP 1319 ' Rod Position Verification," C10 "Incore instrumentation" Tables 1 and 2, and B101 Distribution of Flux Thimbles and Thermocouple."

SP 1319 provides direction for using incore thermocouple to determine if a rod is misaligned. The surveillance directs the operator to obtain a thermocouple map for before and af ter the load reduction or rod deviation. C10 Table 1 provides a list of thermocouple which are adjacent to the rod in que6 tion. The before and af ter values for these thermocouple are compared to those of thermocouple which are symmetric locations in the core to determine if the rod is misaligned. C10 Table 2 provides a list of symmetric thermocouple. Some of these are located in adjacent co;e quadrants,'and some are in opposite core quadrants. For example, , thermocouple B5 has three symmetric thermocouple + 12. E12 and L9. While L9 is located in the opposite quadrant, the other two are in adjacent quadrants.

Therefore either answer a or b. can be considered as correct.

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. NRC Resolution I A review of the provided documentation supports your discussion on expected I operator actions. Based upon NRC review of this and other references, the NRC will change the correct answer from "b" to "a or b."

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QUESTION No. 54: Given the following Unit 1 plant conditions: , ! Unit was initially at 100% power and has been rnanually tripped.

! - Tave is 542*F on all channels.

- "A" Condenser vacuum is 14" vacuum -

B" Condenser vacuum is 18" vacuum

- 11 & 12 Circ water pumps ele running j -

Which ONE of the following describes steam dump availability? . a.

Steam dump is NOT available.

' b.

Only the atmospheric dumps are available, c.

Only the condenser dump is available.

d.

Both atmospheric and condenser dumps are available.

ANS: c REF: P8197L 002 " Steam Dump Control" Rev. 3; pp.17-18 /Obj. 7 , K/A: 000051K301 12.8/3.1] , , Licenses Recommendation Accept only b. as the correct answer < Licensee Comments Refer to Attachment 3, P8174L 002 " Steam Dump Control System" pages 26 - 27, , ' and C47014 annunciator location 47014 0502, CDSR STEAM DUMP PERMISSIVE.

Atmospheric dumps are available since Tave is above 540*F on all channels and the trip will provide the required " Loss of Load" arming signal for steam dump operation.

The condenser steam dump would not be availablo since "A" condenser vacuum is 14" hg. Greater than 16" bg vacuum is required in both condensers to satisfy the " Condenser Steam Dump Permissive."

Answer b. was the answer on the original version of this question. Changes to the question were made during the review process and the answer did not get changed.

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_ ',.,. . . NRC Resolution A review of the provided documentation supports your discussion on expected plant response. Based upon NRC review of this and other references, the NRC will change the correct answer from "c" to "b."

IV.

QUESTION No. 70: S/G Tube Leak,1C4 AOP2 was implemented yesterday for tube leakage in 11 S/G.

Leakage was measured at 120 GPD. In the last half hour R 15 counts have increased and the calculated leak rate is 100 GPD, Which ONE of the following actions should be taken? a.

Stop S/G blowdown to the river.

b.

Align the S/G Safety Relief header drains to the aerated sump tank, c.

Commence a controlled shutdown.

d.

Roset R 15 and R 19 alarm setpoints to the shutdown action level.

ANS: d REF: C4 AOP2 P8170L 003 "RCS" Rev. 4; p. 44/Obj.12 K/A: 000037G011 13.9/4.1) Licensee Recommendation Accept only c. as the correct answer Licensee Comments Por Attachment 4,1C4 AOP2 " Steam Generator Tube Leak," Section 1.4, tube leakage greater than 150 GPD places the plant in action level 2. Step 2.7.3 requires the operator to commence a controlled shutdown as quickly as safe plant operation permits.

The initial tube leakage of 120 GPD placed the plant in action level 1. During this action level several measures are taken. These include distractors s., b and d.

Therefore these would not be correct answers for the situation given.

Answer c. was the answer on the original version of this question. Changes to the question were made during the review process and the answer did not get changed.

NRC Resolution A review of the provided documentation supports your discussion on expected operator actions. Based upon NRC review of this and other references, the NRC will change the correct answer from "d" to "c."

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_ _ _ _ _. _ _ _ _. _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .. _ _. , - U.S. NUCLEAR REGULATORY COMMISSION SITE-SPECIFIC WRITTEN EXAMINATION APPLICANT INFORMATION Namo: Region 111 MASTER EXAMINATION Date Administered: 6/16/97 Prairio Island 1 & 2 ' Sonlor Reactor Operator License PWR-WEC2 INSTRUCTIONS Uso the answer sheets provided to document your answers. Staple this cover shoot on ' top of the answer sheets. Each question is worth one (1) point. The passing grade requires a final grade of at least 80%. Examination papers will be picked up 4 hours after the examination starts.

All work dono on this examination is my own I have neither given nor received aid.

Applicant's Signature RESULTS Examination Value 100 Points Applicant's Score Points Applicant's Grado Percent _ _ - _ - _ - - _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____ _ _ _ _ _ _ _ _ _ _ _ _ _

-- . ... SEllIOR REACTOR OPERATOR Page 2 of 75 llame Date A 11 S W E R S 11 E E T For multiple choice questions, circle or X your choice.

If you change your answer, write your selection in the blank.

MULTIPLE CliOICE 001 a b c d 026 a b c d 002 a b c d 027 a b c d 003 a b c d 028 a b c d 004 a b c d 029 a b c d 005 a b c d 030 a b c d 006 a b c d 031 a b c d 007 a b c d 032 a b c d 008 a b c d 033 a b c d 009 a b c d 034 a b c d 010 a b c d 035 a b c d 011 a b c d 036 a b c d 012 a b c d 037 a b c d 013 a b c d 038 a b c d 014 a b c d 039 a b c d 015 a b c d 040 a b c d 016 a b c d 041 a b c d 017 a b c d 042 a b c d 018 a b c d 043 a b c d _ 019 a b c d 044 a b c d 020 a b c d 045 a b c d 021 a b c d 046 a b c d 022 a b c d 047 a b c d 023 a b c d 048 a b c d 024 a b c d 049 a b c d 025 a b c d 050 a b c d

SEllIOR REACTOR OPERATOR Page 3 of 75 llame Date A 11 S W E R SHEET For multiple choice questions, circle or X your choice.

If you change your answer, write your selection in the blank.

051 a b c d 076 a b c d 052 a b c d 077 a b c d 053 a b c d 078 a b c d 054 a b c d 079 a b c d 055 a b c d 080 a b c d 056 a b c d 081 a b c d 057 a b c d 082 a b c d 058 a b c d 083 a b c d 059 a b c d 084 a b c d , 060 a b c d 085 a b c d 061 a b c d 086 a b c d 062 a b c d 087 a b c d 063 a b c d 080 a b c d . 064 a b c d 089 a b c d 065 a b c d 090 a b c d 066 a b c d 091 a b c d 067 a b c d 092 a b c d 068 a b c d 093 a b c d 069 a b c d 094 a b c d 070 a b c d 095 a b c d 071 a b c d 096 a b c d 072 a b c d 097 a b c d , 073 a b c d 098 a b c d 074 a b c d 099 a b c d 075 a b c d 100 a b c d

_ SENIOR REACTOR OPERATOR Pago 4 of 75 , NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules coply: - 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.

2. After the examination has been completed, you must sign the statement on the cover sheet ,u.

Indicating that the work is your own and you have not received or given assistance in y completing the examination. This must be done after you complete the examination.

>: l 3. Restroom trips are to be limited and only one applicant at a time may leave. You must avoid all contacts wi3. a.-ane outside the examination room to avoid even the appearance or possibility ] of craating.

Y 4. Use black ink or dark pencil ONLY to facilitate legibie reproductions.

5. Print your name in the blank provided in the upper portion of the exa'. nation cover sheet and e

each answer sheet.

G. Mark your answers on the answer sheet provided. USE ONLY THE PAPER PROVIDED AND ' J DO NOT WRITE ON THE BACK SIDE OF THE PAGE.

7. The point value for each question is one (1) point.

8. Partial credit will not be given. Therefcre, ANSWER ALL OF THE QUESTIONS AND DO NOT LEAVE ANY ANSWER BLANK. If you change your answer, write your selvetion in the blank.

9. If the intent of e question in unclear, ask questions of the examiner only.

10.When turning in your examination, assemble tna completed examination with examination questions, examination aids and answer sheets. In addition, turn in all scrap paper.

11, Ensure all information you wish to have evaluated as part of your answer is on your answer sheet. Scrap paper will be disposed of im,nediately following the examination.

12.To pass the examination, you must achieve a grade of 80% or greater.

, 13.There is a time limit of four (4) hours for completion of the examination.

1,4,._When you" are done and have turned in your examination, leave the examination area (EXAMINER WILL DEFINE THE AREA). If you are found in this area while the examination is still in progress, your license may be denied or revoked.

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SENIOR REACTOR OPERATOR Page 5 of 75 1.

Which ONE of the following combinations d inoperability, would place the plant in a 1 nour action statement to correct or be in hot shutdown in the next 6 hours? a.

21 safety injection pump AND 21 containment spray pump, b.

21 RHR pump AND 22 containment spray pump.

c.

21 containment fan coil unit AND 21 containment spray pump.

] I d.

21 containment fan coil unit AND 22 safety injection pump.

ANS: c

REF: TS 3.3.B.2 [ P8180L-002 " Containment Spray" Rev 4; p.32/Obj.12 KA & Importance: 026.000.G0.05 [3.3/3.9) Pedigroo info: New question (The Tech Spec has recently changed) . d!? . '$

. - 2.

Given the following conditions: - Unit is at 100% steady state operation.

- Two charging pumps and two 40 gpm orifices are in service.

- The letdown divert valve to the hold-up tank fails to the divert position.

Which ONE of the following describes the effects? a.

Charging pump suction willbe lost.

b.

Charging pump suction will swap to the RWST.

c.

Automatic rnakeup will maintain VCT level between 17% and 28%. d.

The divert valve will maintain VCT level between 56% and 78%. ANS: b REF: NF-40784-5 P8172L-001a "CVCS" Rev 3; p. 31/Obj. 3 KA & Importance: 004.010.A1.05 [3.0/3.2) Pedigree Info: Pl Bank Question 1294

SENIOR REACTOR OPERATOR Pago 6 of 75 3.

A cooldown on Unit 2 is in progress with Tavg above 350*F. Which ONE of the following alignments would be considered inoperable as per the Steam and Power Conversion System section of Tech Specs? 21 and 22 Main Feed Pumps are OFF and the turbine driven AFW pump is RUNNING a.

with its selector switch in MANUAL.

b.

21 and 22 Main Feed Pumps are OFF and the turbine driven AFW pump is OFF with its selector switch in SHUTDOWN AUTO.

21 and 22 Main Feed Pumps are ON and the motor driven AFW pump is RUNN'NG c.

with its selector switch in AUTO.

d.

21 and 22 Main Feed Pumps are ON and the motor driven AFW pump is OFF with its selector switch in AUTO.

s ANS: a REF: 2C28.1 section 5.11 (Rev.1); Tech Spec section 3.4 l P8180L-007 "AFW" Rev 4; p.37/Obj.10 i KA & Importance: 061.000 GO.05 [3.3/4.0) ) Pedigree Info: New question ,,, _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ __

- _ _. -- -. . .- SENIOR REACTOR OPERATOR Pago 7 of 75 i 4.

.Given the following Unit 2 plant conditions: - Normal 100% power lineup - 21 CC pump is supplying all CC loads - 21 RCP sealleakoff is 5.7 gpm - 22 RCP sealleakoff is 2.7 gpm . A small thermal barrier HX leak is believed to have developed and the engineer would like to sequentially close the CC outlet from the thermal barrier heet exchangers to attempt to isolate the leak. Which ONE of the following actions is correct for this situation while still maintaining 100% power? a.

21 and 22 thermal barrier HX can both be isolated, b.

21 and 22 thermal barrier HX must both remain in service.

21 thermal barrier HX can be isolated and 22 thermal barrier HX must remain in c.

service.

d.

21 thermal barrier HX must remain in service and 22 thermal barrier HX can be isolated.

ANS: g /1 l REF: C3 Precautions / C14 AOP2 Subsequent Actions P8170L-002 "RCP's" Rev 2; p. 27/Obj. 5

KA & importance: 003.000.K4.04 [2.8/3.1] Pedigree Info: New question i . . L J , , -. -. ....v.--. ..,,- - -... y --n y ., - _ .s s .. i - -- - ---

lI SENIOR REACTOR OPERATOR Pagg 8 of 75 5.

Given the following conditions: - Unit 1 is in a 100% normal lineup.

- A loss of 4160V safeguards bus results in a loss of power to 11 charging pump.

- 480V safeguards power cross-tie is performed.

Which ONE of the following 480V busses is supplying power to 11 charging pump? a.

Bus 111 b.

Bus 121 c.

Bus 211 d.

Bus 221 ANS: b REF: P8186L-008 " Safeguards Electrical' P8172L-001a "CVCS' Rev 3; p.19/ Obj. 3 KA & Importance: 004.000.K2.03 [3.3/3.5] Pedigree Info: New question 6.

Which ONE of the following will occur following a manual containment spray actuation but NOT from an automatic containment spray actuation signal? a.

Open caustic additive isolation vaives.

b.

Open containment spray pump discharge isolation valves.

c.

Actuate safety injection.

' d.

Actuate containment ventilation isolation.

ANS: d REF: NF-40762 P8180L-006 " Engineered Safeguards" Rev 3; p. 21/Obj. 4 KA & importance: 013.000. A3.02 [4.1/4.2] Pedigree Info: 1995 Audit Exam question #2 _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _

_ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - --. SENIOR REACTOR OPERATOR Pcg3 9 of 75 , 7.

Given the following plant conditions: - Rod Channel Pressurir.or Pressure has been removed from service per C51.

- A plant cooldown is in progress.

- RCS Tavg at 500*F.

- RCS Pressure is 1400 psig.

- White Channel Pressurizer Pressure fails high.

Which ONE of the following is true? a.

A safety injection signal will occur on low steamline pressure.

b.

A safety injection signal will occur on low pressurizer pressure, c.

No Safety injection signal will occur, but will occur later during the plant cooldown.

d.

No Safety injection signal will occur, but will occur if bistables for the White channel are tripped.

ANS: b REF: C1.3 " Unit Shutdown" p.25; Steam tables P8180L-006 " Engineered Safeguards" Rev 3; p. 27/Obj. 5 KA & Importance: 013.000.A4.02 [4.3/4.4j Pedigree Info: New question based on PB QUESTION: 9000220 8.

During a unit shutdown, the " Rod at Bottom" annunciator will alarm when: a.

Control bank A reaches 20 steps.

b.

Control bank A reaches 35 steps.

c.

Control bank D reaches 20 steps.

d.

Control bank D reaches 35 steps.

ANS: a REF: P8184L-005 " Rod Control & RPl" Rev 2; p. 55/Obj. 8 KA & Importance: 014.000.K4.03 [3.2/3.4] Pedigreo Info: Pl question 1612 (Used on Review Phase Quiz 4 - May 97) \\ _ _ - - _ _ - _ _ _ - _ _ _ _ - _ _ _ - _ _ _ - _ _ _ _ _. _ _ _ _ _ _ _

- .. - _. = . .- - = .- . - _- . .- - -. - - -. - SENIOR REACTOR OPERATOR Pcg310 of 75 9.

Given the following conditions: - Control rods are being inserted during a unit shutdown.

- N-35 is reading 2E-11 amps.

_N-36 is reading 3E-9 amps.

- Operator went to reset on both manual block switches for source ranges which resulted in a SR HIGH FLUX reactor trip.

Which ONE of the following situations could have caused the reactor trip? a.

N-35 is overcompensated, b.

N-35 is undercompensated.

c.

N-36 is overcompensated.

~ . d.

N-36 is undercompensated.

ANS: a REF: P8184L-002 "NIS' Rev 4; p. 51/Obj.11 KA & Importance: 015.000.K6.01 [2.9/3.2) Pedigree Info: Pl Bank Question 2201 . s

- . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ SENIOR REACTOR OPERATOR Pago 11 of 75 10.

Given the following Unit 2 plant conditions: - Unit is at power with rod control in automatic and bank D rods at 208 steps.

- Power Range channel N41 failed high and was taken OOS.

- The operator mispositioned the POWER MISMATCH BYPASS switch to the N43 position.

- l&C is attempting to repair N41.

Which ONE of the following describes the response of the control rods when fuses are installed? a.

Rods initially drive IN when the instrument power fuses are installed.

s b.

Rods initially drive IN when either the control or instrument power fuses are installed.

c.

Rods initially drive OUT when the instrument power fuses are installed, d.

Rods initially drive OUT when either the control or instrument power fuses are installed.

ANS: a REF: C51 P8184L-002 "NIS' Rev 4; p. 36/Obj.13 KA & Importance: 015.000.K3.02 [3.3/3.5) Pedigree info: New question i - _ _ - _ - _ _ _ - _ _ _ _ - _ - _ _ _. _ _ _ _. _ _ _ _ _ _ _ _ _ _

_ _ - _ - _ ___ - _ SENIOR REACTOR OPERATOR Pago 12 of 75 11.

Given the following Unit 1 plant conditions: Reactor is shutdown.

- Reactor decay heat is being removed by natural circulation.

- RCS pressure is 1550 psig.

- Average core thermocouple temperature is 402 degrees F.

- Which ONE of the following is the maximum amount of subcooling that exists in the RCS? a.

50 deg. f. b.

100 deg. F.

c.

200 deg. F.

d.

300 deg. F.

ANS: c REF: Steam Tables P8159L-052 " Properties of Water" Rev 2; pp. 22-24/Obj.10 KA & Importance: 017.020.K5.02 [3.7/4.0) Pedigree info: New question PB QUESTION: 9000224 _ _ _ __ - ______ _________

SENIOR REACTOR OPERATOR Pago 13 of 75 12.

Given the following plant conditions: - The unit has just increased power from 50 to 100%. - An IRPI in indicates a misaligned rod.

An actual rod misalignment would be determined by comparing core exit thermocouple between... a.

Adjacent core quadrants before and after the power change, b.

Symmetric core quadrants before and after the power change, c.

Adjacent core quadrants after the power change.

d.

Symmetric core quadrants after the power change.

ANS: b ca 4 REF: P8170L-001 ' Reactor & Internals" Rev 1; p.16/Obj. 2 ; SP 1319 KA & Importance: 017.020.K4.02 [3.1/3.6] Pedigree Info: New question . - -

- . -... . SENIOR REACTOR OPERATOR Peg)14 of 75 13.

Which ONE of the following conditions states the reason that the control rod drive fans are required during a natural circulation cooldown, in accordance with ES-0.3A, " Natural Circulation Cooldown?" To prevent... . a.

CRDM damage due to excessive heat stress, b.

Formation of a steam bubble in the reactor vessel head.

c.

Catastrophic failure of the reactor vessel flange O-rings.

d.

Damage to the Core Exit Thermocouple's electrical circuitry from overheating.

ANS: b REF: ES-0.3A * Natural Circulation Cooldown with CRDM Fans" P8197L-011 "E-0 Review" Rev 2; pp.17-18/Obj.16 KA & Importance: 022,000 K4.04 [2.8/3.1) Pedigree Infoi Significantly altered form of PI question 1764 PB QUESTION: 9000226 , 14.

Which ONE of the following conditions would result in an automatic trip of all operating Condensate pumps? (Asud ett fwm.4 Au c A wdica r.6)

a.

Main feed pump suction pressure at 179 psig, b.

Containment pressure at 5.2 psig.

c.

Steam generator level at 73%. d.

Condenser hotwelllevel at inches.

ANS: b REF: E-0 Auto Actions Table EO-1 P8174L-003 " Condensate & Feedwater" Rev 3; p.17/Obj. 4 . KA & importance: ' C56.000.A2.04 [2.6/2.8] Pedigree Info: New question PB QUESTION: 9000228 .. . -. . .-. -

.__ - _ _ ____ - - - f SENIOR REACTOR OPERATOR Paga 15 of 75 q 15.

Which ONE of the following will allow opening a bypass feed regulating valve from the - control room, following a valid automatic isolation due to Hi-Hi S/G level? a.

Depress the A and B feedwater bypass reset pushbuttons.

b.

Go to manual on the feedwa. i control valve controller and open the yelve.

c.

The condition which caused the isolation signal is restored to its pre-trip value, d.

Locally reset the closing solenoid valves once the closure signal has been cleared.

ANS: a REF: P8180L-006 " Engineered Safeguards" Rev 3; p. 31/Obj. 6 KA & Importance: 059.000.K4.19 [3.2/3.4) Pedigree Info: New question PB QUESTION: 9000229 16.

During normal plant operation, a short circuit causes a loss of power to DC panel 17. Which ONE of the following will not operate as designed? a.

D1 Diesel Generator b.

D2 Diesel Generator c.

12 Diesel Cooling Water Pump d.

22 Diesel Cooling Water Pump ANS: c REF: 820.9 and C20.9 P8186L-005 "DC Distribution" Rev 2; Obj. 3 KA & Importance: 063.000.K3.02 [3.5/3.7) Pedigree info: New question - - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _______ __ _

... .- - -.. - . - .. .. . -- -. -- t , JSENIOR REACTOR OPERATOR Pcg316 of 75 _With _the plant in a normal full power lineup, an increase in RCDT level is identified. Which - 17e _ ONE of the following describes the cause of this level increase? , -

a.

RCP #1 sealleakoff high.- b. - Reactor head to vessel inner "O" ring seal failure.

. c.

Attempting to ressat an' accumulator check valve.

' d.

Pressurizer PORV leak-by with its block valve open.

ANS: b REF: _C47012-0603; X-HIAW-1-7 P8182L-001a ' Rad Waste-Liquid' Rev 1; p. 29/Obj. 2-

KA & Importance: 068.000.K1.07 [2.7/2.9) b ' Pedigree Info: New question ! d 18.

The plant is at 100% power when it becomes necessary to switch a CFCU from Slow to Fast speed. Which ONE of the following describes the reason for pausing for 1 second during the switching? a.

Allows the magnetic field in the motor to collapse, b.

Allows the supply witage to re-zero prior to starting, c.

Prevents a momentary loss of flow to vessel support equipment.

d.

Prevents the downstream ductwork from being damaged.

ANS: a REF: -C19.2 Limitation 4.3.5 P8180L-009H _' Containment Air Handling' Rev 0; p. 23/Obj. 4 KA & Importance:

022.000.GO.10 [3.2/3.4)- Pedigree Info: New question 19.

Which ONE of the following completes the statement to explain why the steam generator level program is reduced at low power? . -. - - ... .--

A -~ SENIOR REACTOR OPERATOR P go 17 of 75 To reduce... a.

time delays due to " thermal lag" during a loss of forced circulation.

b.

thermal stratification above the U-tubes in the event of a SGTR.

c.

mass inventory available to boil off in the event of a steam break.

d.

reactor power level oscillations due to xenon oscillations at 800.

ANS: c REF: P8174L-006 " Steam Generator Level Control" Rev 3; p.10/Obj. 3 KA & Importance: 059.000.A3.02 [2.9/3.1) Pedigree Info: New question PB QUESTION: 38159 20.

Fuel handling is in progress in containment. The operator mistakenly lowers a control rod into the East RCC change fixture instead of the West RCC change fixture. No fuel assembly is in the East RCC change fixture. Which ONE of the following will result when the Gripper Engage / Disengage Valve is positioned to DISENGAGE.

The control rod will... a.

fallinto the refueling pool, b, drop into the change fixture.

c.

not release due to an electrical interlock.

d.

not release due to the mechanical support of the change fixture.

ANS:. b REF: C17 & ERTF report 96-04 (Recent PINGP event) P8182L-003 " Fuel Handling" Rev 3; Obj. A; KA & Importance: 034.000.A2.03 [3.3/4.0] Pedigree Info: New question L - - - - - - - - - - - - - - - - -. - - -.

.. - . . _ . _ _ _ _ _ _ _ _. _... _ _ _ _. _ _ _ _. _ _ _ _ . _ SENIOR REACTOR OPERATOR Pag 318 of 75 -- - 21.

Given the following Unit 2 plant conditions: - Reactor power is 1215 MWt.

- RCS pressure is in the norme! operating band.

Which ONE of the following RCS average temperatures would FIRST result in exceeding a safety limit? a.

574 degroes F b.

594 degrees F . c.

616 decrees F d.

621 degrees F ANS: c 121SA M REF: T.S. 2.1-1; 1650hM 60AT= 44 AT a. if 100* FAT / b.if 100%AT d. If 2385 line

P8170L-003 "RCS' Rev 4; p. 46/Obj.15 (Tech Spec Figura 2.1-1 attached).- KA & Importance: 002.000.G0.05 [3.6/4.1) Pedigree Info: New question ' PB QUESTION: 9000321 . J

f

1 ., -- , . .

I SENIOR REACTOR OPERATOR Page 19 of 75 22.

Given the following Unit 1 conditions: - Unit 1 is in an outage.

- Containment in service purge in ope.ation.

- Spent Fuel Pool Ventilation, R-25 high alarm has occurred.

Which ONE of the following describes the effects on the containment in-service purge lineup? a.

Supply and exhaust valves in containment isolate.

b.

Supply from spent fuel pool ventilation isolates.

c.

Discharge aligns to containment.

d.

Discharge aligns to aux building special ventilatlun.

ANG: a REF: C47047 R-25; P8180L-009E " Containment Purge..." Rev 0; p.18/Obj.14 KA & Importance: 029.000.K4.03 (3.2/3.5) Pedigree Info: New question _ _ _ _ _ _ _ _ _ _ _

_- _ - - , ,_ ' SENIOR REACTOR OPERATOR Pcg3 20 of 75 , ' 23.

Unit 1 is in a normal at power lineup when the high pressure sensing line on the RCS loop B flow instrument FT-414 breaks. Which ONE of the following describes the resulting RCS flow indication? , a.

Only one (1) RCS loop B flow transmitter is affected (FT-414) with HIGH flow indication, b.

Only one (1) RCS loop B flow transmitter is affected (FT-414) with LOW flow indication.

c.

All RCS loop B flow transmitters are affected (FT-414, FT-415, FT-416) with HIGH flow indication.

d.

All RCS loop B flow transmitters are affected (FT-414, FT-415, FT-416) with LOW . flow indication.

ANS: d , REF: P8170L-003 "RCS' Rev 4; p. 32/Obj.12 . KA & Importance: 002.000.K6.06 [2.5/2.8] Pedigree Info: Significantly altered form of Pl question 1217 PB QUESTION: 9000233 24.

Which ONE of the following conditions could be identified by an ILLUMINATED light on the Unit 1 Si Not Ready Status Panel? a.

12 Si pump is not running.

b.

22 CL pump is it. manual.

c.

11 CS pump has lost DC centrol power.

d.

121 CR chiller is running.

ANS: b REF: P8180L-005 *ECCS' Rev 2; p.14/ Obj. 4 NF-40315-2 logic diagram KA & Importance: 006.030.A4.01 (4.4/4.4] Pedigree info: New question PB QUESTION: 9000234

SENIOR REACTOR OPERATOR Pcg3 21 of 75 25.

Which ONE of the following 21 SI Accumulator parameters needs to be corrected for the 21 SI Accumulator to be OPERABLE while the reactor is operating in hot shutdown conditions? a.

Water Volume is 1130 cubic feet.

b.

Pressure is 730 psig.

c.

Boric acid concentration is 1910 ppm.

d.

Outlet isolation valve is OPEN with its control switch RED position indicating light NOT LIT.

ANS: a REF: T.S. 3.3.A.1 P8180L-004 "SI & Accumulators" Rev 3; p. 45/Obj.11&13 KA & Importance: 006.000.GO.05 [3.5/4.2] Pedigree Info: Significantly altered form of Pl question 2094 PB QUESTlON: 9000235 26.

Given the following Unit 1 plant conditions: Pressurizer pressure channel selector switch is in the white-blue position.

- PT-449 (Yellow Channel), Pressurizer Pressure, has just failed LOW.

- Which ONE of the following describes the response of the pressurizer pressure control system to this failure? a.

PORV 431C will not operate automatically but, PORV 430 is operable.

b.

Spray valve 431 A closes but,431B continues to modulate.

c.

Both spray valves are prevented from automatic operation.

d.

Both PORVs are prevented from automatic operation.

ANS: a REF: C51.4 P8170L-005 " Pressurizer Press. Control" Rev 4; p. 26/Obj. 4 KA & Importance: 010.000.A4.03 [4.0/3.8) Pedigree Info: Significantly altered form of Pl question 1260 PB QUESTION: 9000236

_..____. _ _ _ - _ _ _ _ _ _-- SENIOR REACTOR OPERATOR Pcga 22 of 75 i 27.

The master pressurizer pressure controller output has failed to minimum Assuming no operator ection, which ONE of the following describes the effect on the rsact 3r protection system? e.

OPAT reactor trip setpoints increase.

b.

OPAT reactor trip setpoints decrease.

c.

OTAT reactor trip setpoints increase, d.

OTAT reactor trip setpoints decrease.

ANS: c REF: C51; P8170L-005 'Pzr Pressure Control' Rev 3; pp. 22-23/Obj. 4; P8184L-003 " Reactor Protection" Rev 4; p. 20 KA & Importance: 010.000.K3.02 [4.0/4.1) Pedigree Info: New question . -ri - . _ _, _ _ _ _ _m.

___ ___

SENIOR REACTOR OPERATOR Pego 23 of 75 28.

Given the following Unit 2 plant conditions: - Unit is at 33% power.

- Rod control is in MANUAL.

- Loop B cold leg temperature detector TE-401 fails low.

- No operator action is taken.

Which ONE of the following will be the steady-state pressurizer level? a.

15% b.

25% c.

33% d.

55% ANS: b REF: 2C51.1 P8170L-006 "Pzr Level Control" Rev 3; p.20/Obj. 2 KA & Importance: 011.000.K6.04 [3.1/3.1] Pedigree Info: New question. based on PB QUESTION: 9000238 29.

Which ONE of the following conditions on LT-428 would result in an increase in indicated pressurizer level? a.

Reference leg leak develops.

b.

Pressurizer liquid temperature increases to 652 F.

c.

Containment temperature decreases to 65*F.

d.

Containment pressure increases to 15.5 psia.

ANS: a REF: P8159L-062 " Fluid Statics" Rev 2; p.18/Obj. 6 KA & Importance: 011.000.A1.01 [3.5/3.6] Pedigree Info: New question PB QUESTION: 9000239

. . . - . - ... . .. .- _ - -.. SENIOR REACTOR OPERATOR Paga 24 of 75 30.

Given the following unit 2 conditions: - Load reduction to hot shutdown is in progress.

- At 40% power, Turbine 1st Stage Pressure PI-486 fails low.

- Actions for C51 including blstable tripping is completed.

- Plant shutdown is continued.

- Which ONE of the following describes the effects on the reactor protection system when power goes less than 10%? , a.

No effect, all protection signals will work as designed.

' b.

Intermediate and Power Range low power trips will not be enabled, c.

Turbine trip and single-loop loss of flow trips will not be bypassed, d.

Low pressurizer pressure, high pressurizer level and two-loop loss of flow trips will not be bypassed.

ANS: d REF: C51 P8184L-004 ' Reactor Protection" Rev 4; pp. 23-24/Obj. 7 ' KA & Importance: 012.000.K4,00 [3.2/3.5] Pedigree Info: New question _

_ _ _________ __ - _____ _ _ __- - - SENIOR REACTOR OPERATOR Page 25 of 75 31.

Unit 2 has experienced an accident and the following conditions exist: - E-0 directs a transition to E-1 " Loss of Reactor or Secondary Cooling" - RCS pressure is 840 psig - Both RCP's are running.

- PR nuclear instruments are reading 3%. - Core exit thermocouple temperatures are 750*F.

- RVLIS Dynamic Range levelis 38%. - Total AFW flow is 150 gpm.

- Both SG WR levels are 45% - Both SG pressures are 800 psig.

- Containment pressure is 24 psig.

,,, , Which ONE of the following procedures should be implemented? a.

FR-S.1.

b.

FR-Z.1 c.

FR-C.1.

d.

F R-H.1.

, ANS: d REF: E-0 Information page P8197L-014 "F/FR Review" Rev 2; p.15/Obj. 9 KA & Importance: 016.000.G0.15 (3.6/3.8] Pedigree Info: New question . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

_ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ -__- . .. . .. SENIOR REACTOR OPERATOR Pags 26 of 75 32.

Given the following Unit 1 plant conditions: - Unit is operating at 75% steady state power in imp Out turbine control.

- All systems are in automatic control.

- The "A" SIG PORV fails open.

Which ONE of the following describes the plant response to this condition? (Assume no operator action is taken.)

a.

Control rods insert and reactor power remains at 75%. b.

Turbine load decreases by 5% and reactor power remains at 75%. c.

Control rods withdraw and reactor power rises to 80% where it stabilizes.

d.

Turbine control valves open in resporise to lower steam header pressure to increase turbine load to 80% where it stabilizes.

ANS: c REF: P8174L-001 " Main & Aux Steam" Rev 2; p.15/Obj. 3 KA & Importance: 035.010.K5.01 [3.4/3.9) Pedigree Info: New question -Similar to Pl question 1338 PB QUESTION: 9000242 33.

Which ONE of the following describes, in order of the load sequencer selection of available sources, the voltage restoration sequence for Safeguards 4.16 KV bus 257 , a.

2RY CT11 EDG D5 b.

CT11 2RY EDG D6 c.

2RY CT12 EDGD5 d.

CT12 2RY EDG D6 ANS: c REF: P8186L-008 "Sofeguard Distribution" Rev 3; p. 30/Obj. 6 KA & Importance: 062.000.K2.01 [3.3/3.4] Pedigree Info: Significantly altered form of Pl question 2140 _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _

- -.. _ - . - - -- -. - .. - -.. ~ - ... - - .- SENIOR REACTOR OPERATOR - Pego 27 of 75 34.

Unit 2 is operating at full power. During SP-2093, D5 Diesel Generator Slow Start Test, the D5 EMERGENCY GENERATOR LOCAL ALARM comes in. The outplant operator determines the alarm is due to a low fuel oil day tank level. Investigation shows 21 fuel oil transfer pump, which is selected as preferred, is tripped due to a fault. Which ONE of the following describes D5 operability? D5 diesel generator is...

' a.

operable because fuel oil can be manually transferred from D6 fuel oil storage tanks.

b.

operable because the back-up fuel oil transfer pump is available.

c.

NOT operable because all support systems are required to be available.

l d.

NOT operable because the Unit 2 fuel oil supply is inadequate.

ANS: b REF: 2C20.7 page 7

P8186L-014 "DS/D6 Generator Set" Rev 1; p. 45/Obj. 9 KA & Importance: 064.000.K1,03 [3.6/4.0)

Pedigree Info: Pl Requal Bank question 2709 i

, ) i l

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f>A+p <#, < s 7/ , emexe.,=gggg o.1,es , ' WEBSTER NE YORK 14580 (716) 265-1600 - - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _

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<$ kN? 3,*' 't<?:?k% .$.<! o '" ' *"""E !!a"nLI "' "*" " WEBSTER NE YORK 14580 (716) 265-1600 _ - _ _ _ - _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _

_ SENIOR REACTOR OPERATOR Prga 28 of 75 , . 35.

D1 diesel generator is loaded onto its respectiva bus for testing following an overhaul. The following conditions were just established: - Generator power is 3100 KW.

' - Reactive load is 1400 KVAR delivered.

Which ONE of the following ls the LONGEST amount of time the generator can remain at the above coaditions without exceeding the machine ratings? a.

0.25 hours b.

0.75 hours c.

990 hours d.

1100 hours ANS: a REF: (1C20.7 Figure 1 provided) Placard [2750- Continuous; 2750-3000- 1000 hrs; 3000-3250- 1/2 hrJ P8186L-004 " Diesel Generators" Rev 3; obj. 6 KA & Importance: 064.000.A2.06 [2.9/3.3] Pedigree Info: New question PB QUESTION: 9000245

- - . - __ - ._ ~ _. _ . _- .. - - _.. - - c SENIOR REACTOR OPERATOR Pego 29 of 75 36.

' Both units are in a normal at power lineup and the following occurs: -A unit 1 LOCA to the Aux Building occurs.

-Safety injection is actuated.

-All normal safeguards equipment operates as expected.

Which ONE of the following pairs of radiation monitors would detect that a radioactive _ , release to atmosphere from the Aux Buildi.ig is in progress? a.

1R30 and 1R37; Unit 1 Auxiliary Building Vent b.

2R30 and 2R37; Unit 2 Auxiliary Building Vent c.

1R48 and 1R49; Unit 1 Containment Hi-Range

d.

1R22 and 1R50; Unit i Shield Building Vent LANS: d REF:. P8182L@2 ' Radiation Monitoring System" Rev 4; pp.12 & 17/ Obj. 6 KA & importance: 073.000.A4.01 (3.9/3.9) Pedigree Info: New question , < w +-

,- -+.,,r-e.- y -

.. _ - ---. ... . - -. -. - - -. - -.. - - - -. .- - .. -.. SENIOR REACTOR OPERATOR Pcgo 30 of 75' '

37.

Given the following Unit 2 plant conditions: , - Operating 100% power with all systems operable.

- Rod control is in AUTO.

- Control Bank "D" starts stepping in slowly, but at a noticeable rate.

A tube leak. i the which ONE of the following will cause this response? a.

Letdown Heat Exchanger,

b.

Seal Water Heat Exchan0er.

c.

Regenerative Heat Exchanger.

- d.

Excess Letdown Heat Exchanger.

ANS: b

REF: C14 AOP2 ' Leakage into the CC System" P8170L-003 "RCS' Rev 4;- p. 44/Obj.12 KA & Importance: 008.010.A3.02 [3.0/3.1) , Pedigree Info: Pl question 2580 PB QUESTION: 9000318 . W , . ,w- -,, ,. = c - ,. ----- - u - - -. -.. - -. -. _. -.. -. ... - --.- --._-.. -....- -- ' SENIOR ' REACTOR OPERATOR Paga 31 of 75 - - 38.

Unit 1 is in cold shutdown with the following conditions: ' - 11 Component Cooling pump is running.

' - 11 RHR pump is running.. i ' - CC inlet motor valve to 11 RHR heat exchanger is open.

- CL inlet motor valve to 11 CC heat exchanger is open.

. Which ONE of the following should te performed NEXT in order to place the standby train of > ' RHR in service? a.

Start 12 CC pump, . ] b.

Start 12 RHR pump, s c.

- Open 12 CC heat exchanger CL inlet valve.

' d.

Open 12 RHR heat exchanger CC inlet valve.

, ' ANS: a

REF: C15 Precautions and section 5.2, ' P8180L-003 'RHR' Rev, 3; Obj. 8 KA & Importance: 005.000.K1.01 [3.2/3.4] Pedigree Info: New question ' ,

4 -

i e 'D

F , a - =---+-m--- - - ~ - - - -. - --va,,e w rs , a-w w --r m--,.n, ev-, a eem-e- -- +r,w .

. . __ .. . _ _ _. -. _ _ _ .. _ _ _ _ _ _ _ _ _ .. _ _ _ ._ _ _. - _ _ _ _ _. SENIOR REACTOR. OPERATOR . Pcgs 32 of 75 - 39; Given the following plant. conditions: <- - Unit 1 was at 100% power.

- Rod Controlis in Manual.

- A load reduction to 70% occurred fast enough to actuate steam dump.

Which ONE of the following desenbes the operation of the steam dumps with no operator action? ' Steam dumps open... a.

to reduce Tavg to 560*F and then modulate close.

b, to reduce Tavg to 556*F and then modulate close, c.

and stay open to control Tavg at 547'F.

- d.

and stay open to control Tavg at 556'F.

ANS: d

REF: P8174L-002 " Steam Dump" Rev 3; pp. 24-26/Obj. 4 KA & importance: 041.020.K6.03 [2.7/2.9] Pedigree Info: New question

,

1 i

.- r - , ,, -.,w -.-

_ _ _______ - _ _ _ _ _ SENIOR REACTOR OPERATOR Page 33 of 75 40.

Given the following Unit 2 initial conaitions: - Cooldown is in progress using steam dump to the condenser.

- Condenser steam dump valve positioner faitt, causing the valve to open fully.

Which ONE of the following actions is required to close the valve?

a.

Set the steam dump controller to 0% output in MANUAL.

b.

Set the steam dump controller to 100% output in MANUAL.

c.

Place a bypass inter'ock switch to OFF/ RESET.

d.

Place the mode solcetor switch to RESET.

ANS: c REF: P8174L402 * Steam Dump" Rev 3; pp.16-17/Obj. 5 KA & Importance: 041.020.A2.02 [3.6/3.9) Pedigree Info-New question

. . -. . -. . - ... _. _ _ _ _ = - - . - - - - - _- - SENIOR REACTOR OPERATOR Pcg3 34 of 75 41, Given the following Unit 2 plant conditions: Unit is at 50% power.

- Control rods are in AUTOMATIC.

- Which ONE of the following instrument malfunctions would result in a CONTINUOUS rod withdrawal? ' a.

Loop "A" Thot RTD fails LOW.

b.

Power range channel N-42 fails LOW.

c.

PT-485 falls HIGH.

. d.

Loop "B" Tcold RTD fails HIGH.

. ANS: c REF: C51.2 P8184L-003 " Reactor Process Instrumentation' Rev 3; pp. 'C35/Obj.11 ' KA & Importance: 000.001.K1.05 [3.5/3.8)_ Pedigree Info: New question PB QUESTION: 9000323 !

a it I f i ..<m.

. .... - - , _,.. . ., _ -

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

_ __ _.___ , SENIOR REACTOR OPERATOR Paga 35 of 75 42.

Given the following Unit 1 plant conditions: - Unit 1 is at 16% power.

- Bank D rods are at 55 steps.

- A Control Bank D rod was dropped and recovered.

- The Pulse-to-Analog Converter was NOT reset per C5 AOP4, " Dropped RCCA".

What effect will these events have on continued rod control system operation? - As control rods are... a.

Inserted, the Rod Insertion Limit Alarm will be received at a lower rod position than required.

b.

withdrawn, Overtemperature Delta T will NOT stop Control Bank D withdrawal when required.

c.

Inserted, Bank C control rods will begin insertion at a lower value of Control Bank D , position.

d.

withdrawn, the Bank D Rod Withdrawal Hi Limit Alarm is will NOT alarm before ' Control Bank D is fully withdrawn.

ANS: a REF: C5 AOP4 P8184L-005 " Rod Controf and RPl" Rev 2; p. 84/Obj.11 KA & Importance: 000.003.A1.01 [2.9/2.9] . Pedigree Info: PB QUESTION: 9000324 i

. - .es.

.m.s-a, w.

_, -i_y - + - w- - y...

. _.. ._ ._ __.

.__ . - ~ _. _ . SENIOR REACTOR OPERATOR Paga 36 of 75_ 43.

In accordance with E-1, " Loss of Reactor or Secondary Coolant" the reactor coolant pumps are required to be tripped when RCS pressure decreases sufficiently. This is done to.... a.

prevent excessive cooldown during a steam break.

b.

. minimize RCP heat input.

, c.

prevent cavitation of the RCPs.

. d.

minimize mass loss out the break.

ANS: d REF: E-1 Step 2 basis P8197L-012 "E-1/E-2 Review" Rev 1; p. 27/Obj. 8 KA & importance: 000.011.K3.14 [4.1/4.2] Pedigree Info: Pi question 1811 Sim initial Quiz #4 question #6 PB QUESTION: 9000325 44.

Which ONE of the following is the MAXIMUM acceptable time to actuate AFW following an ATWS condition in order to limit the consequences of an RCS overpressurization.

a.

30 seconds b, 60 seconds c.

120 seconds d.

180 seconds ANS: b REF: P8197L-014 "F/FR Review" Rev 2; p.12/Obj. 5 KA & Importance: 000.029.A1.15 [4.1/3.9] Pedigree Info: Significantly altered form of Pl question 2601 PB QUESTION: 9000326

CENIOR REACTOR OPERATOR Page 37 cf 75 45.

Wr.ach ONE of the following detects infrared radiation produced in the combustion process in order to alert the plant to a fire? a.

Thermal detector b.

Flame detector c.

valzation detector d.

Heat actuated detector ANS: b REF: P8178L-003 " Fire Detection' Rev 1; pp. 7-8/Obj.1 KA & Importance: 000.067.K2.01 [2.3/2.5] Pedigree Info: Significantly altered form of PI question 1414 46.

Which ONE of the following is the proper sequence of major actions for removing decay heat from the core per FR-C.1, " Response to inadequate Core Cooling?" a.

Reinitiation of high pressure safety injection; open RCS vent paths; RCP restart; rapid secondary depressurization, b.

Rapid secondary depressurization; reinitiation of high pressure safety injection; open PORVs; RCP restart.

c.

RCP restart; reinitiation of high pressure safety injection; rapid secondary depressurization; open PORVs.

d, Reinitiation of high pressure safety injection; rapid secondary depressurization; RCP restart; open RCS vent paths.

ANS: d REF: FR-C.1 P8197L-014 "F/FR Review" Rev 2; pp.19-20/Obj.11 KA & Importance: 000.074.K1.03 [4.5/4.9] Pedigree Info: Significantly altered form of Pl question 2458 PB QUESTION: 9000329

. .. . - - ---. . - - - - -. _ _ _ .... --- . -.- - .- SENIOR REACTOR OPERATOR Pags 38 of 75 47.- If Reactor Coolant activity is excessive,- Tavg is reduced to below 500*F in order to... a.

reduce the severity of pressurized thermal shock from a steam line break accident.

b.

minimize potential for containment contamination from inadvertent PORV operation, c.

prevent uncontrolled release of radioactivity if a steam generator tube ruptures, i ' d.: . inhibit the release of f;ssion products from a loss of coolant accident.

ANS: c REF: T.S. 3.1.D P8!70L-003 "RCS' Rev 4; p.46/Obj.15 KA & Importance: 000.076.K3.05 -[2.9/3.6) Peaigiss info: Significantly altered form of Pl question 1232 PB QUESTION: 9000330

48.

During implementation of FR-S.1, " Response to Nuclear Power Generation / ATWS", which one of the following requires MSIV closure? a.

Steam generator PORV is open with Tavg stable at 547 F.

b.

Steam generator tube is ruptured with Tavg stable at 527*F, - c.

Steam dump to condenser is open with Tavg at 557'F and decreasing.

. d.

- Steam dump to atmospheric is open with Tavg at 537'F and decreasing.

ANS: d REF: FR-S.1 P8197L-014 "F/FR Review" Rev 2; p.14/Obj. 7 KA & Importance: 000.029.K3.08 [3.6/3.8) - Pedigree Info: Significantly altered form of Pl question 1859 , , I f , -. - , . - ... . -. - - -. - - - . --

. _ ._.

. _.

... _. _.

_ _ _. _ _ _ _. _ _ - _. _ _ _. _. _ _ _.. _ SENIOR REACTOR OPERATOR Peg)39 of 75 49.

After a plant transient the position indicator for rod G-11 was found to be 15 steps above the. Bank D step. counter. Actions to determine actual rod position are in inconclusive. While. attempting to move the rod to determine the cause, the following indications are observed: ' 20 steo insertion 20 steo withdrawal Tavg decrease - Increase . Calorimetric power constant constant Bank D Step counter - decrease increase , RPI constant constant Based on the above information, rod G-11... a.

has a failed RPl.

b.

has a failed step counter.

c.

is misaligned.

d.

is stuck.

ANS: a REF: C5 AOPS " Misaligned Rod, Stuck Rod and/or RPI Failure" section 2.4 , P8184L-005 " Rod Control & RPl" Rev 2; p. 56/Obj.12 , .

KA & Importance: 000.005.A2.01 (3.3/4.1] Pedigree Info: New question ! 4-a ., _ -m . _. - ~.. , ,. ..- _ ,.,

- _ _ - _. -. - - _.. _.- _ ...._ _ _ _ _ _. _ _. _ _. _. _. _. _. _. .. _. _ _ _.. _... _ _. _. _. _.. - SENIOR REACTOR OPERATOR Pago 40 of 75 _ 50.

Given the following Unit 1 plant condition:' ~ A loss of all AC power has occurred.

- - Seal injection to the RCP's has been isolated.

- CC return from the RCP's has been isolated.

- Both RCP seal leakoff flow recorders indicate 7.99 gpm.

- Both RCP seal leakoff flows indicate Fall on ERCS - - - Power has been restored; - - - The crew has transitioned to ECA 0.1.

< - ECA 0.1 directs restoration of RCP seal cooling.

- Which ONE of the following should be done regarding RCP seal cooling? . a.

- Seal injection flow is initiated to cool the seals. Once temperatures are trending ' down, then CC flow is established.

b.

CC flow is initleted to cool the seals. Once temperatures are trending down, then seal-injection flow is established, c.

Both seal injection flow and CC flow to the RCP should not be established.

d.

Both seal injection flow and CC flow to the RCP may be c' ablished in any order.

ANS: c REF: ECA-0.1 Step 12 caution and basis; 1C3 AOP4 section 2.4.2 Rev. 2 P8197L-011 "E-0 Review" Rev 2; p. 25/Obj. 20 KA & Importance: 000.015.GO.07 [3.1/3.2] o Pedigree Info: New question __

a d n

,,,.+a-m-m-n-

- -- ,-ww.- - - -,n m,. ,,-m ,+, - n-- - --

. .. - - - - -.. -... ... - . -. - I SENIOR REACTOR OPERATOR Pcgs 41 of 75 51.

Given the following Unit 1 plant conditions: , A plant trip has just occurred.

_ ._ - 2 control rods are stuck out of the core following the trip.

-- An emergency boration has besn initiated by the reactor operator in accordance with - ES-0.1, " Reactor Trip Recovery".

Which ONE of the following lists the MINIMUM amount of boric acid required to be added? a.

250 gallons b.

500 gallons c.

750 gallons d.

1000 gallons ANS: b

REF: ES-0.1 P8197L-011 "E-0 Review" Rev 2; p 73/Obj.10 KA & importance: 000.024.A2.05 [3.3/3.9) Pedigree Info: Significantly altered form of Pl question 1610 PB QUESTION: 9000251, Revised to PI numbers 't . k . - - -, - -, c-- . + - ,, y

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

- . _. l SENIOR REACTOR OPERATOR Pcg3 42 of 75 52.

Given the following Unit 1 plant conditions: - Steam generator "A" is faulted due to a steam line break outside of containment, i - The crew is performing actions of E-2, " Faulted Steam Generator".

J - The AFW system is in operation.

- Steam generator 'A' narrow range level is 0%. - Steam generator "B' narrow range level is 20%. Whlch ONE of the following actions concerning the AFW pumps is required?

a.

Maintain at least 40 gpm AFW flow to each SIG.

b.

Maintain at least 200 gpm total AFW flow to both S/Gs.

c.

Isolate the AFW pumps from SIG 'A' (steam and AFW flow).

d.

Isolate the AFW pumps from both S/Gs (steam flow and AFW flow).

ANS: c - REF: E-2 P3197L-012 'E-1/E-2 Review' Rev 1; p. 50/Obj. 21 . ' KA & Importance: 000.040.A1.02 [4.5/4.5) Pedigree Info: New question

PB QUESTION: 9000253

.,

.,( .s, ,.-r.

-.* m - -w-rt w-w - f

- - , . .-. - . -. i LSENIOR REACTOR. OPERATOR Pag)43 of 75 , - 53.

In'accordance with E-1, " Loss of Reactor or Secondary Coolant,"which ONE of the following - groups of parameters is iequired to be verified, in addition to pressurizer level, prior to terminating SI flow?. a.

RCS subcooling, secondary heat sink, and containment pressure, b.

RCS pressure, RCS subcooling, and secondary heat sink.

c.

Containment pressure, RCS pressure, and RCS subcooling, j d.

Secondary heat sink, containment pressure, and RCS pressure.

ANS: b > REF: P8197L-012 *E-1/E-2 Review" Rev 1; p. 21/Obj.6 KA & importance: 000.040.A2.05 [4.1/4.5] Pedigree Info: New question - PB QUESTION: 9000254

. 54.

Given the following Unit 1 plant conditions: - Unit was initially at 100% power and has been manually tripped.

- Tave is 542'F on all channels.

- A" Condenser vacuum is 14" vacuum " - B" Condenser vacuum is 18" vacuum " - 11 & 12 Circ water pumps are running Which ONE of the following describes steam dump availability? a.

Steam dump is NOT available, b.

Only the atmospheric dumps are available.

_ Only the condenser dump is available, c.

d.

Both atmospheric and condenser dumps are available, ' _.ANS: # B . REF: P8197L-002 " Steam Dump Control" Rev 3; pp.17-18 /Obj. 7 KA & Importance: 000.051.K3.01 [2.8/3.1] . Pedigree Info: New question . . . . __, ,

.. _. -_.. _ .. . - _ .. _ _... _ _ _ ~ _ _ _ _ _ _ _ _ _ _. _ _ _ -. _ _. -. SENIOR REACTOR OPERATOR Pegs 44 of 75 i55, Given the following Unit 2 plant conditions: - Reactor Power is 40%- - 1st Stage Pressure is 270 psig - - A" Condenser Vacuum is 24' Hg _ " - B" Condenser Vacuum is 26' Hg " - - 1- - Barometric Pressure is 29' Hg

Which ONE of the' following actions and reasons apply? a, Trip the turbine due to high back pressure.

b.

Trip the turbine due to differential pressure between condensers, , c.

Reduce load at rapid rate to' decrease backpressure.

, d.

Reduce load at a normal rate to differential pressure between condensers.

. ANS: a REF: C22.9; (Figure C1-20 provided) P8176L-001 " Turbine Control' Rev 1; p. 91/Obj. 6 , e KA & importance: 000,051.A2,02 [3.9/4.1) Pedigree Info: New question

-

9 i

N -,.. +,, n-~e e - - - ... - - - -- ,N,-..,,,n,-, ,~ ,,4 ,, -,, . - - w,w-.,~ ,,.,n.-- ,r-n - +- -

. ... ._- .. . . - . _- . _ .. SENIOR REACTOR OPERATOR Pag)45 of 75 56.- Which ONE of the following is the reason for closing the RCP seal injection throttle valves during a loss of all safeguards AC power? To prevent... a.

overfilling the VCT due to bockflow of hot RCS water.

. b. - steam binding the seal injection piping due to backflow of hot RCS water, - c.

runout of the charging pumps when the pumps are restarted following power

restoration.-

d.

thermal shock to the RCP thaft when the charging pumps are started following power restoration.

ANS: d - REF: ECA-0.0 Step 18 P8197L-011 'E-0 Review" Rev 2; /Obj. 20 KA & Importance: 000.055.K3.02 [4.3/4.6) ' Pedigree Info: Pl Bank Question 2011 ,

T.

]- .. -, , -- ,,,

SENIOR REACTOR OPERATOR Pcg3 46 of 75' 57.? Given the following sequence of events: - - A loss of all safeguards AC power has occurred.

- 1ECA-0.0 has been entered.

- Sl has actuated and was reset.

- Bus 16 load rejection (Green) lights are lit.

- Prior to energizing safeguards bus 16 from unit 2, which ONE of the following is required to - be performed? ' a.

Place 12 and 14 Containment Fan Coil Units in OFF.

i b.

' Place 122 Control Room Chiller and Fan in PULLOUT.

c. - Place 12 Safety !njection Pump in OFF.

: d.

Place 12 Residual Heat Removal Pump in PULLOUT.

' . ! ANS: c.

REF: ECA-0.0, Loss of All Safeguards AC Power Step 8 and basis . - P8197L 011 'E-0 Review" Rev 2; p.25 /Obj. 20 F-KA & Importance: 000.055.A1.06 [4.1/4.5] l-Pedigree Info: New question ! t r , I

Y . --.. -. ~. -. - . .... ... , . . - -.. - . -

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

. . -.. . - SENIOR REACTOR OPERATOR-Pags 47 of 75 58.

Given the following Unit 1 plant conditions: - A load incresse is in progress per 101.4 " Power Operation".

- A lockout has occurred on 480V Bus 122.

- No operator actions have been taken.

. Which GNE of the following is the status of pressurizer heater groups? a.

All groups are energized.

, b.

All groups are deenergized, , c.

Groups B, C, D, and E are energized AND group A heaters are deenergized.

d.

Groups A, C, D, and E are energized AND group B heaters are deenergized.

' ANS: a REF: P8170L-005 'Pzr Pressure Control' Rev 4; p.18/Obj. 4 KA & Importance: 000.057.A2.12 [3.5/3.7) ' Pedigree Info: New question , i 59.

While in containment, ycu discover a small fire in an open waste container located within a radioactive materials storage area. A CO2 fire extinguisher is located 10 feet away. Which. ONE of the following actions should you take?

a.

Exit containment immediately, then notify the control room.

b.

Exit containment immediately, then notify Radiation Protection.

c.

Notify the control room, then attempt to put out the fire out, d.

Don self-contained breathing equipment, then attempt to put the fire out.

ANS: c . REF: F5 Section 5.0 Rev 19, Page 14 KA & importance: 000.067.K1.02 [3.1/3.9) Pedigree Info: Pi Bank Question . _.. , _. . _ _ _ . _ _ -___

. .. . -- ,-. .- -- .. -. .. . SENIOR REACTOR OPERATOR Pcgs 48 of 75 60.

Given the following Unit 2 plant conditions ' The unit is at 100% - Containment pressure is 2.2 psig - Shield building air temperature is 76*F - Contsinment air temperature is 115'F - - , . Which ONE of the following states the required actions and bases for these conditions? i Within 8 hours, reduce containment... a.

pressure to preclude entry into an Integrity orange path during a design basis accident.

b, pressure to prevent exceeding the design pressure during a design basis accident.

c.

temperature to prevent excessive inleakage into the shield building annulus during a design basis accident, d.

temperature to ensure adequate operation of the hydrogen recombiners during a design basis accident.

ANS: b . REF: Tech Spec Basis page 3.6-2 P8180L-001 " Containment Spray" Rev 2; p. 21/Obj. 3 KA & Importance: 000.069 G0.04 [2.5/3.8) Pedigree Info: New question .

4

x $ -s-m -~ ,v.

w , e-

-. ... . - .. -. - -. - -- -.-. __. _ - -. -. -.. . _ - .... . SENIOR REACTOR OPERATOR Peg @ 49 of 75 ,61; A LOCA has occurred and FR-C.1, " Response to inadequate Core Cooling" has been ' implemented. Steam Generator depressurization was ineffective in restoring core cooling.

- ln order to provide core cooling, which ONE of the following conditions must be established ' ~ ' ' prior to starting an RCP? a.

Sealinjection flow is 8 gpm.

. b.

The oil lift pump has been on for 4 minutes.

c.

WR S/G level is greater than the Attachment E required level.

. d.

Both pressurizer PORVs are open.

, s-ANS: c , .REF: 1FR-C.1 step 18 P8197L-014 "F/FR Review" Rev 2; p. 20/Obj.14 KA & Importance: 000.074.K3.07 [4.0/4.4) , i Pedigree Info: New question $ e i i !- . = ~ y-w.. ,...,.. -, --..., , - - - ...-% .. -._.-g,,e- - ,

.-. . - - -- -. - -. - -. --- - . L , SENIOR REACTOR OPERATOR Ptg3 50 of 75 62.

Given the following Unit 1 conditions: - Reactor power is 75% with all systems in e normal lineup.

- A CC transfer is in progress from Unit 1 to Unit 2.

- 11 COMPONENT COOLING SURGE TANK LOW OR HIGH LEVEL alarm.

- LTDN FLOW Hi TEMP alarm.

- 11'CC surge tank level is -7' and decreasing.

- MV-32375 and CV-31342, Reactor Makeup to 11 CC Surge Tank valves are open.

- 12 CC pump is running with a discharge pressure of 75 psig Which ONE of the following actions should be taken FIRST? a.

Trip the reactor then trip the RCPs.

b.

Start 11 CC pump and isolate letdown.

c.

Close MV-32120 and MV-3?121, CC HX Outlet XOVR valves, d.

Close CC-27-8, CC Surge Tank X-tie. Then close CC-33-34 and PT-17086 to stop the CC transfer.

ANS: a

REF: C14 AOP1 section 2.4 P8172L-002 ' Component Cooling" Rev 4; pp. 27-28/Obj. 6 KA & Importance: 000.026.GO.10 [3.6/3.5) Pedigree Info: Significantly altered form of Pl question 2271 PB QUESTION: 47273

k

i .,. -

__ SENIOR REACTOR OPERATOR Pcg3 51 of 75 63.

A fire in the control room with heavy smoke requires immediate evacuation of the control room. The Unit 1 Reactor Operator was only able to trip the turbine prior to exiting the control room. Assuming that the plant responds as expected, which ONE of the following local actions needs to be taken to complete the RO's initial evacuation assignments? a.

Open the reactor trip breakers, b.

Remove pressurizer PORV fuses, c.

Check 4 KV safeguards busses energizei d.

Isolate air to and vent the MSIV accumulators.

ANS: d REF: F5 Appendix B Rev 17 p.5 P8197L-009 'F5 Appendix B & C' Rev 3; p.11/Obj. 4 KA & Importance: 000.068. A1.14 (4.2/4.4] Pedigree Info: New question 64.

Following a reactor trip and safety injection, AFW pump discharge valves are adjusted to maintain... a.

a maximum of 800 psig discharge pressure to prevent pump runout.

b.

a minimum of 900 psig discharge pressure to prevent a pump trip.

c.

a maximum AFW flow of 100 gpm to prevent overfilling intact S/G(s).

d.

a minimum AFW flow of 100 gpm to each S/G to ensure an adequate RCS heat sink.

ANS: b REF: E-0 Step 12 Basis; SWl-O-2; C28.1; P8180L-007 "AFW" Rev 4; p. 30/Obj. 7 KA & Importance: 000.040.A1.10 [4.1/4.1] Pedigree Info: New question (Plant priority - see recent safety evaluation)

. - .- =- .. .. . - -. ... _.. g ' , - SENIOR REACTOR OPERATOR Page 52 of 75 65, - Given the following containment history with a small 1.OCA in progress: Cnmt Cnmt-Cnmt Cnmt Time Radiation Temperature Pressure Humidity 0815 8.0E3 R/Hr 178 Deg. F 2 psig 70% 0830-9.1E3 R/Hr 180 Deg. F 4 psig 80% 0845 9.9E3 R/Hr 183 Deg. F 6 psig 90 % 0900 1.1E4 R/Hr 185 Deg. F 8 psig 100 % 0915 1.5E4 R/Hr-188 Deg. F 10 psig 100 % ' . During which ONE of the following time inte vals did adverse containment FIRST exist? a.

0815 to 0830

b.

0830 to 0845 c.

0845 to 0900 d.

0900 to 0915 ANS: b ,

REF: E-0 Information Page; P8197L-0012 "E-1/E-2 Review" Rev 1; p. 23/Obj. 6 KA & Importance: 000.009.A2.11 [3.8/4.1] Pedigree Info: New question; Based on PB QUESTION: 9000332

. -. , ,

. SENIOR REACTOR OPERATOR Paga 53 of 75 66.

The following conditions exist: Tha Control Room is being evacuated due to a fire in the relay room.

- Operators have performed all steps for CR evacuation except the procedural steps - associated with stopping the instrument air compressors and depressurizing air receivers.

Assuming no additional operator actions are taken, which ONE of the following plant conditions could result from this oversight? a.

Letdown NOT isolating, resulting in uncontrolled RCS inventory loss, b.

Air to control systems in containment could result in exceedi.,g RCS design pressure.

c.

S/G PORVs could actuate early resulting in a loss of heat sink.

d.

The instrument air compressors could be damaged from overheating due to loss of cooling water.

ANS: a REF: F5 Fire Fighting App B, Attach R, page 69, Sect. 4.0 P8197L409 'F5 App. B/C Review" Rev 3; p.13/Obj. 6 KA & Importance: 000.065.A2.08 [2.9/3.3) Pedigree Info: Significantly altered form of Pl question from 95 Audit Exam d

.. - . .. _ - -.- . - . .. . - -. .._ - - - SENIOR REACTOR OPERATOR Pcgs 54 of 75 67.

Which ONE of the following describes the response of the Control Room Ventilation system to a HIGH Radiation alarm on R-23, Control Room Air Monitor? a.

No automatic response occurs until R 24, Control Room Air Monitor has a HIGH radiation alarm.

b.

BOTH control room outside air dampers close, control room exhaust ' steam exclusion damper closes, and BOTH control room cleanup fans start, t c.

Train A control room outside air dampers close, control room exhaust steam exclusion damper closes, and 121 control room cleanup fan starts, d.

Train B control room outside air dampers close, control room exhaust steam exclusion damper closes, and 122 control room cleanup fan starts.

ANS: c REF: P8182L-002 " Rad Monitor" Rev 4; p.16/Obj. 5 Radiation Monitor list from LP KA & Importance: 000.061.A1.01 [3.6/3.6) Pedigree Info: PI bank question 2774 . i 68.

E-3, Steam Generator Tube Rupture has been completed and the post-SGTR cooldown method of backfill was selected. Using this method minimizes... a.

radiological releases but may cause unexpected RCS dilution.

b, dilution of the RCS, while maintaining secondary chemistry stable.

c.

both primary and secondary makeup water requirements during recidown and depressurization.

d.

secondary system contamination, while decreasing the time required to reach cold shutdown.

ANS: a . REF: E-3, Step 40 Basis P8197L-013 "E-3 Review" Rev 1; p. 24/Obj.11 KA & Importance: 000.038.K3.06 [4.2/4.5) Pedigree Info: New Question - _-

.. SENIOR REACTOR OPERATOR Pcg2 55 of 75 69.

Given the following Unit 1 plant conditions: - The reactor tripped from 100% power.

- E-0, " Reactor Trip or Safety injection," has been entered.

- The main turbine did NOT trip as expected.

- MANUAL turbine trip is unsuccessful.

Which ONE of the following is the NEXT action required per E-07 a.

Locally trip the main turbine, b.

Locally close the turbine stop valves.

c.

Manually close turbino control valves, d.

Manually close MSIVs and bypass valves.

ANS: c REF: E-0 step 2 RNO P8197L-011 *E-0 Review" Rev 2; p.11/Obj. 3 KA & Importance: 000.007.A2.02 [4.3/4.6] Pedigree Info: PI OUESTION: 2869 70.

Main feedwater automatically isolates following a reactor trip to prevent... a.

overfilling of the S/Gs.

b.

overpressurization of the S/Gs.

c.

excessive cooldown of the,RCS.

d.

thermal shock to the feedwater nozzles.

ANS;c REF: ES-0.1 Step 6 basis P8197L-011 "E-0 Review" Rev 2; p.13/Obj. 6 KA & Importan::e: 000.007.K3.01 (4.0/4.6] Pedigree Info: New question PB QUESTION: 9000265

SENIOR REACTOR OPERATOR Pcga 56 of 75 71, Given the following Unit 1 plant conditions: - A reactor trip and safety injection have occurred from a normal 100% lineup.

- Pressurizer PORV PUV-430 is closed.

- Pressurizer PORV PCV-431C is open and will not close.

- Pressurizer pressure is 1500 psig.

. - A' Pressurizer spray valve is open.

' 'B' Pressurizer spray valve is closed.

Which ONE of the following actions is required for these conditions as per EOP's? a.

Stop 11 RCP AND Close BOTH PORV block valves.

b.

Stop 11 RCP AND Close PORV; PCV-4310 block valve, Stop BOTH RCP's AND Close BOTH PORV block valves, c.

d.

Stop BOTH RCP's AND Close PORV: PCV-4310 block valve.

ANS: b REF: E-0 Step 16 RNO P8197L-011 'E-0 Rev;ew" Rev 2; pp.11 12/Obj. 5-6 KA & Importance: OW18.A1.01 (4.2/4.0] Pedigree Info: New question based on PB QUESTION: 9000266;

i ! r h r

i . .. . _ _ _. - -.. _ - _. _ -.,... - _ - _... _ _... _. _ _ _. _. _, _ _ .,. _ _. - - - -,. -.. -,_.--.~ . _ _ .

..- .-.... .--._. _._.-_ - - -- . -. -.. _ _ _ - _. - -. -. - -.. SENIOR REACTOR OPERATOR Pcg3 57 of 75 ! 72.

RCS pressure is stable at 1500 psig prior to stopping the first SI pump in ES 1.1 ' Post LOCA _ Cooldown and Depressurization". Which ONE of the following describes the expected trend of RCS subcooling and pressurizer level after stopping the Si pump? Subcoolina Pressurizer Level a.

Decreases Decreases , b.

Increases Decreases c.

Decreases increases d.

Increases increases ANS: a REF: 1E-1, " Loss Of Reactor Or Secondary Coolant", Information Page.

P8197L-012,"E-1/E 2 Review", Objective 5.

KA & importance: 000.009.K3.21 [4.2/4.5) ' Pedigree Info: Pl Bank question 1803

73.

Unit 2 was in a normal at power lineup when CVCS letdown was isolated because of a leak in the letdown heat exchanger. Which ONE of the following conditions would result if charging line flow is not isolated? a.

Increased flashing at the letdown orifice isolation valves.

' b.

Increased thermal stresses at the charging penetration into the RCS.

c.

VCT level will decrease causing a loss of suction to the charging pumps.

' d.

VCT level will decrease until charging pump suction shifts to the RWST.

ANS: b ' REF: 012.1 Precaution 3.2 P8172L-001a 'CVCS' Rev 3; p. 36/Obj. 8 KA & Importance: 000.022.K3.07 [3,0/3.2) Pedigree Info: New question . -r4-t-t

-ir+=-g-tT-t-rww+--

g-wdmg,t7-@ esm+1+v', tem N Nm-w+P--*rr* h-9-*-r--tg 7-gp tr-%-w. v p%- --w6>-+~ * *~ m+=~~ W-

-ha*^--

- -*~-6-- P " 's-" Fe- -

.. - -.. - _. - -. - . - _ -.. . -. - - -. _. - - - SENIOR REACTOR OPERATOR Pcgo 58 of 75 74.

Given the following Unit i plant conditions: Unit is in cold shutdown (Shutdown for 3 days) with RHR cooling in progress.

- Pressurizer level is 30% with the RCS Intact.

- Both S/G WR levels are 65%. - RCS is depressurized.

- Busses 11 and 12 are deenergized for maintenance.

) - RHR flow is lost and CANNOT be restored. Which ONE of the following methods of cooling will be utilized to remove the core decay heat? . a.

Feed SIG(s) with AFW and open S/G PORV(s) to remove decay heat.

, b.

Start a RCP and open S/G PORV(s) to remove decay heat.

c.

Maximize charging flow to the RCS and use letdown to remove decay heat.

d.

Align an SI pump to inject into the vessel and open a PZR PORV for heat removal.

ANS: a REF: E-4 step 15 and C15 AOP1 P8180L-003 "RHR' Rev 3; p. 46/Obj. 8 KA & Importance: 000.025.K1.01 [3.9/4.3) Pedigree Info: New question; Based on PB QUESTION: 9000269 - . . . -.. . - -.. .... -

SENIOR REACTOR OPERATOR Pcg3 59 of 75 75.

During a startup with reactor power at 1E 7 amps in the intermodlato range, a fault causes N 31 SR to fall high. Which ONE of the following describos the expected responso for this failure? a.

Hi Flux At Shutdown alarm, b.

Increaso in audible count rato.

c.

Rod status light for SR reactor trip, d.

SR Hi Flux Reactor Trip first out alarm.

ANS: c REF: C51 P8184L-002 *NIS' Rev 4; p. 44/Obj.17 KA & Importance: 000.032.K1.01 [2.5/3.1) Podigroo Info: Pl question 2316 7G.

S/G Tubo Loak,1C4 AOP2 was implemented yesterday for tubo leakage in 11S/G. Lookage was measured at 120 GPD In the last half hour R-15 counts have increased and the calculated leak rato is 160 GPD Which ONE of the following actions should be taken? a.

Stop S/G blowdown to the river, b.

Align the SIG Safoty Rollef header drains to the aorated sump tank.

c.

Commence a controlled shutdown.

d.

Resot R-15 and R-19 alarm sotpoints to the shutdown action lovel.

ANS: A c.

REF: C4 AOP2 P8170L-003 'RCS' Rov 4; p. 44/Obj.12 KA & Importance: 000.037.GO.11 [3.9/4.1) Podigree Info: New question

- _ _ -. _.. _ - _ _ _. _ _. _ _ _.. _ _ _ _. _ _ _. _... _ _ _ _ _ _ - _ . ... _. -_ -.... _ - SENIOR REACTOR OPERATOR Pcg3 60 of 75 ' 77.

The crew is responding to a ruptured tube in 12 steam generator (SIG) using E 3, " Steam Generator Tube Rupture" with the following conditions: 11 S/G pressure is 500 psig.

-

12 S/G pressure is 700 psig.

- , - RCS cooldown is in progress.- - Both SI pumps are running.

What is the HIGHEST indicated core exit temperature that assures 20*F subcooling will exist, including instrument inaccuracy, after the subsequent RCS depressurization? a.

485'F j b.

465'F { c.

450'F d.

430'F

- ANS: b

REF: E-3 page 8 and steam tables l P8197L-013 'E-3 Review" Rev 1; p.18/Obj. 6 , KA & Importance: 000.038.A1.36 [4.3/4.5) Pedigree Info: Significantly altered PI Ouestion 2885 . i T h

'T

5 ,w.-# r+.%-. w.

w- ,..:,,,-,---,--~.-we,r .,-..,-,w,- , .-.n---,w--~ - .,,., -w.--,.+,-- .-.,, -..,,.- ,.,.-,.,,v, ,,.c,.ye---_.,.-,, c,-,, +.,- -r --.. _ _. -. _.. -_.

.. -. - -- __- --. _._---- i SENIOR REACTOR OPERATOR Pcg)61 of 75 ' O.

9:hich ONE of the following statements explains why AFW fiowrate is procedurally restricted 6 is?;A when recovering steam generator (S/G) level if the level has fallen below 7% on the wide, age indicetion? , I i W mnimize water hamraer to the S/G feed ring.

! h To minimize thermal stresses to S/G components.

j c.

To prevent reactor restart from an excessive cooldown.

d.

To prevent exceeding reactor vessel cooldown rate limit.

! ANS: b REF: FR H.1 Step 20 note P8197L-014 'F/FR Review" Rev 2; p. 28/Obj.19 KA & Irnportance: 000.054.K1.02 [3.6/4.2] Pedigree Info: Significantly altered form of question #3 on 1995 NRC Exam

79.

Which ONE of the following is the basis for stopping all RCPs in FR H.1, " Response to Loss of Secondary Heat Sink?" a.

It extends the time available to restore feed flow before bleed and feed criteria is met, b.

It ostablishes natural circulation to enhance the bleed and feed capability of safety injection.

c.

It anticipates an RCS pressure decrease caused by opening PZR PORVs during bleed and feed.

d.

It anticipates an RCS pressure decrease caused by spray valves opening when air is restored to containment, ANS: a REF: FR H,1 P8197L-014 "F/FR Review" Rev 2; p 27/Obj.19 KA & Importance: 000.054.K3.04 [4.4/4.6) Pedigree Info: Significantly altered form of Pl question 1850 PB QUESTION: 9000274 . u r . .. - -~ , . -..,.., -, .~.-n,,----- -.---nn-v, < r.

-

- _- ) SENIOR REACTOR OPERATOR Page 62 of 75 80.

A loss of Unit 1 train B DC has occurred from a normal 100% power lineup. Which ONE of the following breakers would still be closed one minute later and have to be locally opened? l a.

12 reactor coolant pump breaker b.

12 component cooling pump breaker c.

Bus 121M source breaker d.

8H16 generator output breaker ANS: b REF: C20.9 AOP2 P8186L-005 *DC Distribution' Rev 2; p. 21/Obj. 9 KA & Importance: 000.058.A2.03 [3.5/3.9) Pedigree Info: New question 81.

Given the following information: - Unit 1 is in a refueling shutdown.

- Fuel movement in progress.

- A leak has developed in the refueling cavity.

- You observe the water in the reactor cavity area decreasing.

If an irradiated fuel assembly is latched on the manipulator crane, which ONE of the following is the LEAST preferred location to lower the assembly? a.

Reactor core.

b.

RCC change fixture.

c.

Temporary storage rack.

d.

The lower cavity upender area.

ANS: d REF: D.S.2 AOP3 P8182L-003 Fuel Handling KA & Importance: 000.036.K3.03 [3.7/4.1] Pedigree Info: Pl Bank question 23G4

___. ____. _ - . _ _. _ _ _ _ _ _ _. -. _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ SENIOR REACTOR OPERATOR Pcga 63 of 75 i 82.

The following conditions exist: , - Unit 1 is operating at 100% power.

- Pressurizer Level Control Transfer switch is in Position 1-3 (Red Blue).

- Pressurizer Level transmitter LT-428 (Blue) has failed HIGH.

i Actual pressurizer level decreases from 20% to 13% LT-426 (Red) and LTA27 (White).

Which GNE of the following describes the expected condition of the Letdown isolation valves? , LVC427 LCV428 a.

OPEN OPEN b.

CLOSED CLOSED c.

OPEN CLOSED . d.

CLOSED OPEN . ANS: d REF: C51 P8170L-006 'Pzr Level Control' Rev 3; p.19/Obj. 3 , KA & Importance: 000.028.GO 11 [3.5/3.7) ! Pedigree Info: Pl Bank question dated 1994/05/09

4-m.

--- 4.,,,, g--_ _--n-_--_ --.r, ,--, --,-,. .a - , - + -, ,-,-,-,n....,,m,,,,, _._,n y v.~. ,,m n n ,-,-n-,-7,--,7-.,,,-~, ,

. . _. _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _. _.... _ _ _ _ _ _ ._ .__ SENIOR REACTOR OPERATOR Pcg]64 of 75 83.

Which ONE of the following sets of parameters provides indication of adequate natural ) circulation flow? (a) = stable (t) = increasing (4) = decreasing PZR CETC Cold leg Hot Leg SG Press , Pressure Temoerature Temocrature Temoerature Pressure

a.

2235 psig 595'F (a) 504*F (4) 590'F (a) 690 psig (4) b.

2085 psig 625'F (4) 553'F (*) 580*F (t) 1050 psig (a) c.

2285 psig 615'F (4) 560'F (t) 600'F (a) 835 psig (4) d.

2185 psig 570*F (t) 519'F (t) 568'F (t) 785 psig (4) ' ANS: a

, REF: ES-0.1 Att. A P8197L-011 *E-0 Review" Rev 2; p.13/Obj.11 , ' KA & Importance: 000.056.K1.01 [3.7/4.2) Pedigree Info: New Question 84.

While performing a surveillance procedure, the operator recognizes a step which conflicts with a precaution in an operating ('C') procedure. Which ONE of the following actions is required for this situation? ' a.

Stop the surveillance and inform the Shift Supervisor, b.

Restore the system to the inillal conditions and inform the Shift Supervisor, c.

Stop the surveillance and follow the guidance contained in the 'C' procedure, d.

Restore the system to the initial conditions and document the deviation in the Rx Log.

ANS: a REF: SWl-O-10 Operations Manual Usage P9150L-003 " Procedures" Rev 1; p.26/ Obj. 5 KA & importance: 194.001.A1.02 [4.1/3.9) < Pedigree Info: New question ! .... .- - ._ _ , ... _ . -, . - - . ..~

__ __._ . _ _ _ _ _ _ _ _ _ - _ __._________ _ _ _.___ , SENIOR REACTOR OPERATOR Page 65 of 75 i 85.

Given the following Unit 1 plant conditions: f - At 0745 a reactor trip and safety injection occurred.

At 0800 the event was classified as a Site Area Fmergency and the plant evacuation j alarm was actuated along with an announcement over the plant page.

j . . Which ONE of the following is the LATEST time by which personnel accounts.bility should be i . completed as specified h F3-107

a.

0815 t ! b.

0830 c.

0845 ) d.

0900

ANS: b . REF: F310 P7410L-050 'E-plan Overview' Rev 4; p.15/Obj. 2

.

KA & Importance: - 194.001.A1.16 [3.1/4.4) Pedigree Info: New question . , ! Based on PB QUESTION: 9000338 , I a D . - ... . . . _.. - _ _. _.. _, _.. _.. -.. _ _ _. _ _. _ _.. _. _. _ _ _.. _ - _. - - _.. _ _

_.... _. .__ _ _. _. _ - _.. _. _ _ _ _ _. _.... _. _ _ _. _ _ _ l SENIOR REACTOR OPERATOR PO90 66 of 75 00.

Given the following Unit 1 plant conditions: ' A reactor trip and safety injection have occurred.

- RCS pressure is stable at 420 psig.

- Over the last hour the cold leg temperatures have decreased to 240*F as follows: 290'F 60 minutes ago - 260'F 45 minutes ago - 250*F 30 minutes ago - 245'F 15 minutos ago - 240'F Now - Which ONE of the following states the appropriate color code for the Integrity CSF7 a.

Green b.

Yellow c.

Orange d.

Red ANS: a REF: 1F-0.4 Integrity status tree (CSF Status Tree 1F-0.4 ' Integrity" is provided.)

P8197L-014 *F/FR Review' Rev 2; p. 31/Obj. 23 KA & Importance: 194.001.A1.08 [2.6/3.1) Pedigree Info: New question

- .__ _ _ _, _. _ _ - .. . _ _. _ - _ _ _ _ _ _ _. - _ _. _ _ _ - _ _ _ _ - , SENIOR REACTOR OPERATOR Page 67 of 75 . 87.

Prior to activation of the TSC and EOF, which ONE of the following personnel may authorize the INITIAL emergency classification 6NQ the protective action recommendations? i a.

Shift Manager and Shift Emergency Communicator b.

Shift Manager and Emergency Manager , c.

Affected Unit Shift Supervisor or Shift Manager d.

Unaffected Unit Shift Supervisor or Shift Manager ANS: d REF: F3 2; F3-8.1 P7405L-001 *E Plan Overview" Rev. 4 Obj. 2 ' KA & Importance: 194.001.A1.16 [3.1/4,4) Pedigree Info: New question 88.

A fuel leak has been identified on Unit 2 while at power. Which ONE of the following is a measure that will minimize the potential for an uncontrolled release of radioactivity in the plant? a.

Lower the hydrogen pressure in the VCT.

b.

Increase the sampling frequency of the RCS, , c.

Postpone maintenance on seal water return filters, , d.

Reduce the temperature of the letdown flow to the VCT.

ANS: c REF: P8182L-003 ' Fuel Handling' Rev 3; Obj. 9 SAWI 12.1.1 ' Fuel Integrity Program"; F2 section 1.7 ERTF report 94-17 ' KA & Importance: 194.001.K1.04 [3.3/3.5) I Pedigree Info: - New question , ..mr -, ,. ~, -,,.... _. -, - ,, -. - --_,# - . ~.. . - -., - - -,, _ -,. ..m.-- ,,. -,,- .., _- ,

__ . - _ - - SENIOR REACTOR OPERATOR P:03 G8 of 75 89.

One of the ALARA program's objective is to keep the annual integrated dose for all stat!on workers as low as reasonably achievable. Which of the following is a method used to minimize doses in the plant? a.

Perform periodic reactor cooiant system crud bursts during power operations.

b.

Minimize purification flow rate of the reactor coolant system during outages.

c.

Maximize dissolved oxygen levels of the reactor makeup water during shutdowns.

d.

Maintain dissolved hydrogen in the reactor coolant system during startups.

ANS: d REF: F2 page 3 P9130L-001 Radiation Protection KA & Importance: 194.001.K1.04 [3.3/3.5) Pedigree Info: Pl bank questior. 2243 90.

Which ONE of the following situations is a possible exception to the operations practice of having a procedure 'in hand"? a.

Swapping running charging pumps.

b.

Verifyir'g immediate actions in E-0.

c.

Stopping a diesei generator after inadvertent SI.

d.

Performing turbino valve test surveillance procedure.

ANS: a REF: SWl-O-10 P9150L-003 " Procedures' Rev 1; p.18/Obj. 4 KA & Importance: 194.001.A1.01 [3.3/3.4] Pedigree Info: New question

_ _ _ -. _. _ . _ _ _ _ _ _ _ _ - _ _ _. - _ _.. _ _ _ _ -. _ - i . SENIOR REACTOR OPERATOR Pcge 69 of 75 , 91, Which ONE of the following procedures should be entered first upon a loss of RHR cooling with pressurizer level at 30%7 s.

E-4, " Core Cooling Following Loss of RHR Flow" b.

FR l.2, * Response to Low System Inventory" c.

E-0, " Reactor Trip or Safety injection" d.

C15 AOP1,'RHR Flow Restoration' , ANS: d REF: E-4, C15 AOP1 P8180L-003 'RHR' Rev 3; p. 46/Obj. 8 KA & Importance: 194.001.A1.02 [4.1/3.9) Pedigree Info: New quertion; Based on PB QUESTION: 9000279 92.

. You are a Shift Supervisor assigned to the Operations Support Pool. Your work history for the plant includes: - Two (2) 12 hour shifts as Unit 1 SS . - One (1) 12 hour shift as Unit 2 SS . - One (1) 12 hour shift as Work Control Centar (WCC) SS Which ONE of the following is the MINIMUM required to maintain your license active? a.

One (1) 12 hour shift as Unit 1 SS, Unit 2 SS, or WCC SS b.

Two (2) 12 hour shifts as Unit 1 SS or Unit 2 SS c.

Three (3) 12 hour shift as Unit 1 SS , ' d.

Four (4) 12 hour shifts as Unit 2 SS or WCC SS ANS: b REF: SWI O-43; 10CFRSS P9150L-025 'NRC Operator Licenses' Rev 0; p.11/Obj.1 KA & importance: 194.001.A1.03 [2.5/3.4] , Pedigree Info: New question ' --. - .. -. .. - - - - - - - -- .--.-.-

. _. _ _.

-_ __ . _. SENIOR REACTOR OPERATOR Paga 70 of 75 93.

You are the Unit i Shift Suporvisor making a nonemergency call for one individual to work on Unit 1. The individuals are called per the calllist. Their answers to the timo of their last drink of alcohol and whether they feel impaired or not is provided. It is 1730 and you have exhausted the call list. Who should be asked to report to work? NAME LAST DRINK FEEL IMPAlRED7 a.

Joe 1300 YES b.

Dirk ';400 NO c.

Steve 1500 YES d.

Tim 1600 NO ANS: b REF: P9150L-017 "Secunty" Rev. 2; obj.1 ' SAWI 3.151 KA & Importanco: 194.001.A1.03 [2.5/3.4) Podigreo info: Significantly altered form of Pl question 2849 PB QUESTION: 9000281 94.

You rocolved a phone call reporting a medical omergency. Which ONE of the following lists of questions should be asked in order to carry out emergency medical response? i a.

Name of victim, extent of injuries, first aid givon.

b.

Namo of caller, name of victim, first aid given.

c.

Name of caller, extent of injuries, casualty location.

d.

Name of victim, first aid given, casualty location.

ANS: c REF: SS OJT Manual-D3;-lTA 3440150303 , F4; PINGP 788 Rev. 22 KA & Importance: 194.001.A1.05 [3.6/3.8) Podigroo Info: New question ... .--

_ _.. _ _ _ _ _. _ _ _.. _. _ _ _.. _. _ _ _ _ _ _ _ _. _ _ _ _ _ . _. _. _ _ , SENIOR REACTOR OPERATOR Pogs 71 of 75 l 95.

While performing a surveillance on Containment Spray, an instrumentation problem was identified that will take 90 days to repair or replace. Which ONE of the following should be issued to provide additional guidance in performance of the procedure while awaiting repairs? a.

Temporary instruction b.

Temporary Memo ~ c.

Special Order d.

Procedure Change

ANS: b . - REF: P9150L-003 ' Procedures' Rev.1; obj.10 SWI O 19 KA & Importance: 194.001.A1.03 [2.5/3.4] Pedigree Info: New question . I 96.

A fire has occurred in the control room requiring evacuation. The evacuation procedure directs depressurizing the instrument air header. From the list of names below, who is responsible to perform this task? e a.

Unit 1 or 2 Shift Supervisor , b.

Unit 1 or 2 Reactor Operator , c.

Unit 1 Lead Reactor Operator d.

Unit 2 Lead Reactor Operator ANS: a REF: P8179L-009 *FS App B/C Review" Rev 3; pp.11 12/Obj. 4

KA & Importance: 194.001.A1.10 [2.9/3.9) Pedigree Info: New question -.. . .. . . ,. .-.. - -... . . - ... -.

-.._ _ _._ _ _ _ _ _ _ _ -- __ _ _ _ _ _. _. _. _ _. _ _ _. _. _ _.. _ _ _ _ , SENIOR REACTOR OPERATOR Pcg)72 of 75 , 97.

The following conditions exist: - Operator "A" is working in the Control Room on 6/9/97.

' - He realizes he made an error in his log entry the previous day 6/8/97.

Which ONE of the following methods should Operator "A" use to correct this error? a.

Cross out the incorrect information on 6/0/97. Insert the correct information in the ! space above the incorrect information, b.

Leave the incorrect entry as-is. Enter the correct information in the log for 6/9/97 referencing the page and line number of the incorrect entry.

c.

Draw one line and inillal through the incorrect information 6/8/97. Enter the correct information at the bottom of the same page.

, , . d.

Draw one line and initial through the incorrect information on 6/8/97. Enter the correct information in the log for 6/9/97 referencirig the page and line number of the incorrect entry.

ANS: d REF: P9150L-012 " Records / Logs", Objective 2, page 12 SWI O 25

KA & Importance: 194.001.A1.06 [3.4/3/4] Pedigree Info: Pl Bank question dated 1994/05/09

i d . - . . -. _- _ . . - - - - ._ .- .

._ _ _ SENIOR REACTOR OPERATOR PcDo 73 of 76 98.

An oncoming Shift Supervisor has worked the folicwing schedulo:

DAY: 1

3

5

7

9 10 11 12 13 14 Hrs : 12 12 12 12 12 8 16 8 12 12 12 8 14 7 Which ONE of the following is the MAXIMUM number of hours the individual may work on Day 14 without obtaining special authorization assuming the operator has a minimum of 8 hours off between each shift? o.

O hours b.

10 hours c.

12 hours d.

14 hours ANS: b REF: 5 AWI 3.15 Rev 4 Section G.4.2; P9150L-004 ' Plant Operation" Roy 2; p.10/Obj.1 KA & Importanco: 194.001.A1.03 [2.5/3.4) Podigroo info: Significantly altered form of NRC Exam 95 question 88 ' . - -.

~.. - _ - SENIOR REACTOR OPERATOR Pago 74 of 75 99.

An independent verification of the closo position of SI Suction from BAST 'C' valvo is nooded. The local stem position indicator is in question and is not able to provide absolute evidonce of the valvo position. Which ONE of the following is required por procedure SAWI 3.10.0, ' Control of Plant Equipment' in determinlag the valvo position? (Assume the valve broe.kor is in OFF position.)

NOT a.

Vorify the SI Ready light is * LIT'. b.

Attempt to close the valvo using the local handwheel.

c.

Energize the valvo and attempt to closs the valve remotely, d.

Turn the breaker on and check the red and green indicating lights.

ANS: d REF: P9150L-024 ' Control & Operation of Plant Equipment' Rev. 0; obj.1 SAWI 3.10.0 KA & Importance: 194.001.K1.01 [3.6/3.7) Podigroo info: New question 100.

Safoguard hold tags which havo boon removed, shall be returned to the Shift Supervisor.

Which ONE of the following is an exception ~? The tag is... a.

being kept with its work package, b.

to be replaced prior to the end of the shift.

c.

locked to the valve or breaker in its alternato position.

d.

removed during plant conditions when the component is not required to be operable.

ANS: b REF: SWI O 3 Section 5.0 P9150L-005 * Work Control' Rev 3; pp, 25-26/Obj. 2 KA & Importance: 194.001.K1.02 [3.7/4.1) Podigroo info: New question; based on PB QUESTION: 9000288

. __ SENIOR REACTOR OPERATOR Pago 75 of 75 """"""END OF EXAMINATION *"**"*"*

_.. _. _ _ _ _ _____._ _ ____-_.

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v t e Letter Sequence Request
TAC:M98944, (Open) TAC:M98945, (Open)
Initiation Request, Request, Request, Request Acceptance... Supplement Administration Questions 13 August 1997 Answers Results Approval... Other: ML20217R052
MONTHYEAR1997-10-28 [Table View]

v t e Inspection Report - Prairie Island - 1997306 (Operator Exams)
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \| 25 \|
1997 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 7(0) 8(0) 9(0) 10(0) 11(0) 12(0) 13(0) 14(0) 15(0) 16(0) 17(0) 18(0) 19(0) 20(0) 21(0) 22(0) 23(0) Security Operator Exams 306(0) Emergency Planning Other

IR 05000282/1997306

Text

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