ML17286A878

Exam Rept 50-397/OL-91-02 on 910402-05.Exam Results:All Eight Candidates Passed

ML17286A878
Person / Time
Site:	Columbia
Issue date:	05/16/1991
From:	Meadows T, Miller L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML17286A877	List: ML17286A876 ML17286A878
References
50-397-OL-91-02, 50-397-OL-91-2, NUDOCS 9106110211
Download: ML17286A878 (99)

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AND (3)

TO THE PDR 9106110211 91051b PDR ADOCK 05000397 V

PDR

ENCLOSURE 1, EXAMINATIONREPORT 50-397/OL-91-02

Examination Report No.:,

Facility Licensee:

50-397/OL-91-02 Washington Public Power Supply System Nuclear Reactor Facility, Unit. 2 (WNP-2)

Facility Docket No.:

Facility License No.:

50-397 NPF-21 Licensed operator initial examinations administered at WNP-2, Richland, Washington.

Chief Examiner:

T. R.

Meadows

~/r r ga Date Signed Approved by:

L. F. Miller, Jr.,

Chief, Operations Section Date Signed Summary:

Examination on April 2-5, 1991:

The NRC administered initial written examinations and operating tests to eight Reactor Operator applicants.

All eight applicants passed the examinations.

The examiners identified three findings of particular concern which were communicated with the facility management.

First, during the simulator test, the surrogate Senior Reactor Operator (SRO) misdirected three different applicant teams in a scenario involving the RPV flooding emergency procedure.

Second, none of the applicants involved in the above scenario questioned the erroneous direction that they received.

These individuals also exhibited a weak understanding of RPV level instrumentation.

Third, the examiners found that there are still no event mitigation procedures that address a slow depressurization of the RPV due to a BOP malfunction.

REPORT DETAILS

1. Personnel NRC Personnel:

• D. Lange,

NRR, OLB Headquarters L. Miller, Chief, Operations Section J. Munro,

NRR, OLB Headquarters

• T. Meadows, Chief Examiner C. Sorensen, Resident Inspector L. Vick, NRR, OLB Headquarters M. Morgan, NRC Contract Examiner

• G. Buckley, NRC Contract Examiner NNP-2 Personnel:

• J. Baker, Plant Manager

• S. McKay, Operations Manager

• B. Barmettlor,

Manager, Nuclear License Training

• D. Kobus,

Manager, Technical Training

• A. Hosier,

Manager, Licensing

• M. Baird, Control Room Supervisor

• J. Perry, Principal Training Specialist, G. Richmond, Training Specialist D. Merhar, Supervisor, Operations Procedures Group Identifies personnel present at the April 5, 1991 exit meeting.

2.

0 eratin Test The operating test consisted of two major portions:

a.

The dynamic simulator test, using prevalidated scenarios, and The walkthrough test, which included prevalidated site specific Job Performance Measures (JPMs) and associated prescripted systems questions.

The test also integrated prescripted administrative topic questions.

Both of these operating testing areas were validated with the

licensee, on site, during the period of March 18-22, 1991.

The dynamic simulator scenarios were also tested during this period.

Examination security was maintained in accordance with the Examiner Standards, NUREG 1021; Revision 6.

The operating tests were administered to eight operator candidates.

All of these candidates passed this portion of the examination.

However, the examiners identified three findings of particular concern:

First, during the simulator portion of the test, in a scenario that required the candidates to implement emergency procedure 5.1.4, "Reactor Pressure Vessel (RPU) Flooding," due to the loss of level indication, the surrogate Control Room Supervisor (CRS, a licensed Senior Reactor Operator) erroneously directed the operator candidates to terminate, or significantly restrict, all RPV injection sources before the actual core inventory recovered or stabilized.

In one instance, this CRS exited the RPV Flooding procedure with actual level still undetermined.

The operator candidates performed as directed.

However, their actions resulted in core dryout conditions that could have been avoided.

Apparently, the training that this Senior Reactor Operator (SRO) received to correctly implement emergency procedures and to diagnose actual plant conditions during such an event was inadequate.

The SRO misdirected three different candidate

teams, challenged with this scenario, in a similar manner.

Except for this deficiency, this SRO provided excellent direction for the remaining simulator tests.

Overall, his professional effort contributed significantly to the validity of the examination.

a

Second, for each of the above stated

teams, none of the candidates involved questioned their direction to significantly

reduce, or terminate, RPU injection when "RPV Flooding" was required.

After additional follow up questioning, both immediately after the simulator scenario and during the walkthrough portion of the test, it was determined that most of the candidates had some knowledge deficiency of RPV level instrumentation operation, or of when to believe indicated RPV water level was valid under emergency conditions.

These integrated system knowledge weaknesses, coup)ed with their lack of understanding of the intent of the applicable emergency procedure (to flood the RPV}, apparently resulted in the candidates complying with erroneous orders.

This type of behavior could degrade an operators ability to interact with crew members on issues, within and outside of his own control

boards, to facilitate safe plant operation.

Third, the examiners found that there were apparently no event mitigating procedures, or other forms of guidance, addressing a

slow (uncontrolled) depressurization of the RPV, under operating conditions, due to the malfunction of a Balance of Plant (BOP) component or system, before challenging the automatic isolation safety function of the reactor..

Events of this nature could involve a stuck open bypass valve, or faulted main turbine system piping.

This deficiency was also identified in a previous NRC inspection report and accompanying Notice of Violation (NRC inspection Report No. 50-397/90-09).

This violation remains open (Open Item: 50-397/90-09-01).

I The examiners communicated the following observations concerning the licensee's operating procedures to facility representatives:

a 0 Apparently, there were no coordinating support. procedures to identify the optimal systems alignment for core heat removal at very low power levels, or decay heat removal during shutdown.

For example:

The RHR system would be aligned with one or two divisions in shutdown cooling (with a given heat exchanger loading),

depending on the existing core decay heat load (time at power) and the desired time to achieve a given temperature/pressure.

These types of alignment decisions appear to be left up to the Shift Manager, on any given crew.

It was not clear to the examiners how the licensee ensures consistent optimum system alignment.

b.

Operating Procedure, PPM 2.5.7, "Synchronize the Main Turbine Generator to the Grid," has a conflicting CAUTION statement inserted before a direction step.

The CAUTION statement indicates that the generator should be loaded to 300 MWe as soon as possible (in order to avoid turbine damage at low steam loads).

However, the direction step requires the operator to "HOLD" the loading at a minimum level (40-50 MWe, depending on turbine first stage shell temperature),

to test the operation of the control system and soak the turbine, before increasing power.

This discrepancy confused all of the operator candidates that were required to perform (walkthrough) this task.

c ~

Operating Procedure, PPM 2.3.1.5.3, "Primary Containment Inerting," directs the operator to establish a 4200 cubic feet per minute (CFM) inerting flow rate, as indicated on flow instxument recorder, CN-FR-12.

However, CN-FR-12 is calibrated (and marked) in a range of 0-300 standard cubic feet per hour (SCFH).

This requires a factor of (x60) conversion, and is hard to identify on the existing scale.

All of the operator candidates, tested with this task, had difficultywith this step.

One candidate failed the JPM associated with this task.

This deficiency exits both in the simulator and in the plant control room.

The examiners identified the following generic training weaknesses:

a ~

Most of the operating candidates exhibited integrated systems knowledge weaknesses, in the area of RPV level instrumentation, as stated above.

b.

Half of the operator candidates did not understand the intent of Technical Specification 3.04, "Entry into an Operational Condition." They were asked:

What Technical Specification (TS) would prevent the shift from entering MODES 1, 2, or 3, during a plant startup (ie, fxom a lower Mode), with HPCS out. of commission.

These candidates thought entry into these higher Modes would be permitted, if HPCS could be made operational within the time allotted in the HPCS LCO action statements, applicable in Modes 1, 2,

and 3

(TS 3.1.5).

C ~

Most of the operator candidates had difficultywith, or could not perform (two failed) the task associated JPM B7, "FAZ Trip Recovery:

Z signal trip, per PPM 4.12.4.6.6.3."

They had difficulty, or could not, locate the instrumentation and/or indications necessary to verify that all of the protective system automatic actions occurred, due to a Reactor Building Ventilation, high radiation trip signal

("Z" signal).

d.

Most of the candidates did not know the significance of being issued a "Blue Radiation -Exposure Card," in that an individual would be administratively limited to 1

REM per year of

exposure, due to their high radiation exposure history.

3. Written Test The Chief Examiner conducted the Region V Office on March the licensee's staff.

These Training Specialist, and Mr.

Procedures Group (a licensed a review of the written examination in

26) 1'991'ith two representatives of individuals were Mr. J. Perry, Senior D. Merhar, Supervisor, Operations SRO).

The examination was administered on April 5, 1991 at the licensee's Plant Support Facility (PSF).

The examination's administrative arrangements were satisfactory.

At the conclusion of the examination, the author and the Chief Examiner conducted a post examination review with the same licensee staff that were involved with the (pre-)examination review.

Subsequently,'he licensee's formal letter from J.

W. Baker, WNP-2 Plant Manager, dated April 11,

1991, documenting their exam review and forwarding their comment, was received by the NRC on April 12, 1991.

This letter is included with this report (Attachment A).

The Chief Examiner accepted the licensee's comment as valid, and the answer key was modified as recommended in their letter.

All of the operator candidates passed this portion of the examination.

The examiners conducted a post examination comparison of the test results and found one additional apparent generic training weakness.

All of the candidates missed the question (question 6c) that asked them what their administrative, radiological exposure, limit was, with no documented exposure history.

The answer was 300 mRem.

All of the candidates selected 1000 mRem.

4. Exit Meetin An exit meeting was held by the NRC examiners with representatives of the licensee's staff on April 5, 1991 to discuss the NRC findings.

The licensee staff agreed to respond, in writing, to all of the above stated NRC observations within 30 days of the date of the cover letter forwarding this report.

ATTACHMENT A REACTOR OPERATOR WRITTEN POST EXAMINATIONREVIEW

WASHINGTON PUBLIC POWER SUPPLY SYSTEM F.O. Box 968

~ ~ 3000 George Washington Way

~ Richland, Washington 99352 April 11, 1991 G02"91"072 Docket No. 50-397 J.

B. Hartin Regional Administrator U. S. Nuclear Regulatory Commission Region Y

1450 Maria Lane, Suite 210 Walnut Creek, California 94596 Gentlemen:

Subject:

REACTOR OPERATOR WRITTEN EXAMINATION REVIEW Ft>

l l The Washington Public Power Supply System has completed it's review of the Reactor Operator written exam administered on April 5, 1991 and is submitting the following comment:

question 089 has two correct answers "C" as indicated by the key, and also "A".

The basis for this change is that answer "A" is also one of the Immediate Actions directed by P.P.M. 4.12.3.1, Revision 5, which is in the exam data package.

We have no other comments.

Attached you will find the final review.copy'f the exam. If you have any questions, contact John F. Perry, at 509-377-8327.

Very truly y urs, J

W. Baker WNP-2 Plant Manager JFP/JWB:jmg Attachment cc:

NRC Site Inspector (901A) - w/o attachment P. L.

Eng -

NRC Washington D.

C. - w/o attachment Document Control Desk -

NRC w/o attachment N. S. Reynolds Winston 8 Strawn - w/o attachment T. R. Meadows -

NRC Region V - w/o attachment D. L. Williams - BPA/399 w/o attachment

ENCLOSURE 4

SIMULATION FACILITY REPORT 50-397/OL-91-02

NRC Offici al Use Only 0

~Q 5 Pz~Z 4'p~~g. Wp'j~grg+C dC~g

~P~rr/y r Nuclear Regulatory Commission Operator Licensing Examination CrvreS 8pr~. ~~ yg~gg This document is removed from Official Use Only category on date of examination.

NRC Official Use Only

U. S.

NUCLEAR REGULATORY COMMISSION SITE SPECIFIC EXAMINATION REACTOR OPERATOR LICENSE REGION 5

CANDIDATE'S NAME:

FACILITY:

WPPSS 2

liialku(.Ilb'~ iv vaiiOLVAIE:

REACTOR TYPE:

BWR-GE5 DATE ADMINISTERED:

91 04 05 Use the answer sheets provided to document your answers.

Staple this cover sheet on top of the answer sheets.

Points for each question are indicated in parentheses after the question.

The passing grade requires a final grade of at least 80%.

Examination papers will be picked up four (4) hours after the examination starts.

TEST VALUE CANDIDATE'S SCORE 100.00 FINAL GRADE TOTALS All work done on this examination is my own.

I have neither given nor received aid.

Candidate's Signature

REACTOR OPERATOR A N S

M E

R SHEET Page 2

MultiPle Choice (Circle or X your choice)

'enre your answer, +rite your selection in the blank.

MULTIPLE CHOICE 001 a

b c

d 002 a

b c

d 003 a

b c

d 004 a

b c

d 005 a

b c

d 006 MATCHING d

MULTIPLE CHOICE 007 a

b c

d 008 a

b c

d 009 a

b c

d 010 a

b c

d Oll a

b c

d Ol'2 a

b c

d D13 a

b c

d 014 a

b c

d 013 a

b c

d D16 a

b c

d Vi7 a

b c

d 018 a

b c

d 019 a

b c

d 020 a

b c

d 021 a

b c

d 022 a

b c

d 023 a

b c

d 024 a

b c

d 025 a

b c

d 026 a

b c

d 027 a

b c

d 028 a

b c

d 029 a

b c

d 030 a

b c

d 031 a

b c

d 032 MATCHING a

b c

d MULTIPLE CHOICE 033 a

b c

d 034 a

b c

d 035 a

b c

PE<CTOR OPERATOR ANSWER SHEET Page 3

Multiple Choice (Circle or X your choice)

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

036 a

b c

d 037 a

b c

d 038 a

b c

d 039 a

b c

d

'@au a

b c

d 041 a

b c

d 04'2 a

b c

d 043 a

b c

d 044 a

b c

d ZPF~

c' c

d 046 a

b c

d 047 a

b c

d 048 a

b c

d 349 a

b c

d 050 a

b c

d 051 a

b c

d 052 a

b c

d 053 a

b c

d 054 a

b c

d 055 a

b c

d 056 a

b c

d 657 a

b c

d 058 a

b c

d 059 a

b c

060 a

b c

061 a

b c

d d

d 062 a

b c

d 063 a

b c

064 a

b c,

065 a

b c

066 a

b c

d d

d d

067 a

b c

d 068 a

b c

d 069 a

b c

070 a

b c

071 a

b c

072 a

b c

d d

d d

073 a

b c

d 0/4 a

b c

075 a

b c

076 a

b c

077 a

b c

d d

d d

078 a

b c

d 079 a

b c

d 080 a

b c

d 081 a

b c

0 REACTOR OPERATOR A N S

M E

R SHEET Page 4

Multiple Choice (Circle or X your choice)

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

082 a

b c

d 083 a

b c

d 084 a

b c

d 085 a

b c

d 086 a

b c

d 087 a

b c

d 088 a

b c

d 089 a

b c

d 090 a

b c

d 091 a

b c

d 092 a

b c

d 093 a

b c

d 094 a

b c

d D95 a

b c

d 096 a

b c

d D97 a

b c

d 098 a

b c

d

(**********

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

NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS Page 5

During the administration of this examination the following r ules apply:

1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.

2.

3.

4.

After the examination has been completed, you must sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examination.

This must be done after you complete the examination..

Restroom trips are to be limited and only one applicant at a time may leave.

You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.

Use black ink or dark pencil ONLY to facilitate legible reproductions.

Print your name in the blank provided in the upper right-hand corner of

+be examination cover sheet and each answer sheet.

6. Nark your answers on the answer sheet provided.

USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAGE.

7.

s9 10.

Before you turn in your examination, consecutively number each answer sheet, including any additional pages inserted when writing your answer s on the

'examination question page.

Use abbreviations only if they are commonly used in facility literature.

Avoid using symbols such as

< or > signs to avoid a simple transposition error resulting in an incorrect answer.

Write it out.

The point value for each question is indicated in parentheses after the question.

Show all calculations,

methods, or assumptions used to obtain an answer to

'any short answer questions.

11. Partial credit may be given except on multiple choice questions.

Therefore,

ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

12. 'Proportional grading will be applied.

Any additional wrong information that is provided may count against you.

For example, if a question is worth one point and asks for four responses, each of which is worth 0.25

points, and you give five responses, each of your responses will be worth

.0.20 points.

If one of your five responses is incorrect, 0.20 will be deducted and your total credit for that question will be 0.80 instead of 1.00 even though you got the four correct answers.

13. If the intent of a question is unclear, ask questions of the examiner only.

Page 6

14.

When turning in your examination, assemble the completed examination with examination questions, examination aids and answer sheets.

In addition, turn in all scrap paper.

15.

Ensure all information you wish to have evaluated as part of your answer is on your answer sheet.

Scrap paper will be disposed of immediately following the examination.

16.

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

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

18.

When you are done and have tur n~o in your examination, leave the examination area (EXANINER 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.

REACTOR OPERATOR Page 7

QUESTION:.001 (1.00)

The plant was operating at'00% power when a scram occurs.

While performing scram actions it is noted that RPV level reached

+11 inches and that one rod fails to fully insert.

Based on the given information, Which ONE(1) of the following procedures has privrity to be performed?

'a 4 PPH 4.1.1.4-Stuck or Inoperable Control Rod

'b.

'PPH '5.1.1-RPV Control EOP c

PPM 3.3-SCRAH

.2.4-Hain Condensate and Feed Water System QUESTION: 002 (1.00)

During'any'abnormal or emergencJ~'vent, Which ONE(1) of the following is the initial operator response for CRO1 and CR02 as prescribed in PPH 1.3.1, "Conduct of Operations" ?

a.

CROl responds to Control Room Panels

P601, P602, and then to Board A.,CR02 responds to Control Room Panel

P603, and then to Boards B

and C.

d.

CR01 responds to Control Room Panel P601 and then to Boards B and C.

CR02 responds to Control Room Panels

P602, P603, and Board A.

CROl responds to Control Room Panels

P602, P603, and Board A.

CR02 responds to P601 for initial concerns and then to Boards B and C.

C CROl responds to Control Room Panels

P602, P603, and Board A.

CR02 responds to Boards B and C for initial concerns and then to Control Room Panel P601.

REACTOR OPERATOR Page 8

QUESTION:

003 (1.00)

A clearance order is being prepared for a high temperature and pressure system.

The HININUH criteria by procedure for double valve protection is if the system pressure is greater than or equal to psig or system temperature is greater than or equal to degrees F.

100/1QG b.

200/200 c.

300/200 Lj.

1000/525 QUESTION: 004 (1.00)

A~r"!.. '-2., "Securitj Responsibilities of WNP-2 Plant Employees",

the maximum number of individuals an escort may be assigned is within the protected area and in the vital areas.

.a.

3/1 b.

5/2 c.

10/5 d.

15/10

'EAC'lOR OPERATOR Page 9

QUESTION: 005 (1.00)

A radiological survey of an area has the following results:

- 51 mrem/hr general radiation

- 125 dpm/100cm2 loose surface alpha

- 700 dpm/100cm2 loose surface beta-gamma

- airborne measured at 15! of the concentration specified in 10CFR20, Appendix B.

Based on %he survey res its, Which ONE(1) of the following correctly states the COMPLETE posting(s) required?

a.

Contamination Area, Airborne Radioactivity Area b.

Radiation Area, Contamination Area c.

Radiation Area d.

Radiation Area, Airborne Radioactivity Area

REACTOR OPERATOR Page 10 QUESTION: 006 (2.00)

For sac>> description in Column I, NATCH to it the dose limit in Column II.

Each dose limit may be used more than once or not at all.

a 0 b.

C.

d.

COLUHN I (Descrip+~an)

Federal limit per calendar quarter for skin of whole body - quarterly exposure is known.

Federal limit to whole body per calendar quarter if lifetime exposure history is known via Form NRC-4.

Administrative limit to radiation worker per calendar quarter with no documented exposure history.

Administrative limit to female radiation worker who declares herself pregnant.

Limit applies to the duration of the pregnancy.

COLUHN II (Hrem) 1.

300 2.

500 3.

1000 4.

1250 5.

1500 6.

3000 7.

5000 8.

7500 9.

12000 10.

18750 QUESTION: 007 (1.00)

Which ONE(l) of the following statements describes the authorities and responsibilities of Licens~& Operators for operation of the reactor?

a.

The Reactor Operator has the authority to shutdown the reactor'when he/she determines the safety of the reactor is in jeopardy.

b.

The Reactor Operator must obtain permission from the Shift Hanager prior to initiating any unplanned reactor scram.

c.

Only the "Operator at the Controls" has the responsibility to initiate a plant shutdown.

d.

The Reactor Operator should allow ECCS initiation only as a result of automatic actions in response to plant parameters.

REACTOR OPERATOR Page 11 QUESTION: 008 (1.00)

SELECT the set of valves which are throttled to control the reactor cooldown 1.

RMR-V-53A(B), Shutdown Cooling discharge valve.

2.

RHR-V-48A{B), Heat exchanger bypass valve.

3.

RMR-V-3A(B), Meat exchanger outlet valve.

4.

SW-V-68A(B), Meat exchanger outlet valve, Standby Service Water.

(SELECT ONE) a.

1 and 2

b.

1 and 4

c.

3 and 4

d.

2 and 3

QUESTION:

009

{1.00)

-A control rod block has occurred.

Scanning the panels results in. the fo'llowing observations:

- Source Range Honitor (SRH) channel B is reading about 95 counts per second (cps)

- All other SRH channels are reading greater than 1.90 X 10E4 cps

- Only SRH detector A is full in

- Intermediate range (IRH) channel B is on range 2

- All other

,;".,"l channels are on range 3

Which ONE(1) of the following states the cause of the rod block?

a.

SRH downscale b.

Inoperable SRH channel c.

Retract permissive interlock d.

SRH upscale

REAClOR OPERATOR Page 12 QUESTION: 010 (1.00)

The reactor is at f>ll power conditions.

Which ONE(1) of the following describes the meaning vr'n illuminated green light above each safety relief valve (SRV) remote manual switch (RHS)?

a.

The SRV RHS is in the AUTO position.

b.

The SRV solenoid open limit switch is energized.

c.

There is a normal low temperature downstream of the SRV.

d.

There is a normal low noise level downstream of the SRV.

QUESTION: 011 (1. 00)

A loss of offsite power and LOCA has occurred.

The Division 1 Diesel Generator comes up to speed and then trips on overspeed.

The Divisions 2 and 3 Diesel Generators operate as designed and power their respective buses.

Ja -.'n ,

~ pic"r: ~ondi tion~,

>ELECT from the 1'ist provided ALL the ECCS equipment that is available.

1.

'2.

3.

HPCS

'L'PCS RHR-P-2A RHR-P-2B RHR-P-2C (SELECT ONE) a

~

1,4, and 5

b.

1 and 4

C.

2,3,4, and 5

d.

1,2,4 and 5

REACTOR OPERATOR Page 13 QUESTION:

012 (1.00)

The plant is operating at 100/ power.

CR02 is reviewing instrumentation on PG03 and r aborts that wide range level instrumentation is reading approximately twenty (20) inches less than the narrow range instruments.

Which ONE(1) of the following statements correctly describes the reason for the level instrument discrepancy?

a.

Level discrepancy is due to the physical difference in height of the reference legs.

b.

The wide range level instrument is calibrated to a lower setting to

-: P~ conservative initiation of ECCS equipment.

c.

The wide range level instrument is cal.ibrated to a lower reactor pressure and drywell temperature.

d.

The wide range level instrument is calibrated to a zero jet pump flow condition.

REACTOR OPERATOR Page 14 QUESTION:

013 (1. 00)

The plant was operating at 100/o power when a

LOCA (recirc line break) occurs.

The RPS system functions properly to insert all rods.

CRO]

has taken NO immediate actions associated with the SCRAM.

Present plant conditions are:

- Reactor power, 0/

- Reactor level,

+2 inches

- Reactor Pressure, 810 PSIG

- ur'ywell pressure, 2.3 PSIG No area temperature or radiation alarms Based on.the.given.plant conditions, SELECT the ONE (1) answer which correctly id'eatifias ALL of the systems that have a NS4 isolation signal present.

l.

.2.

3.

4.

5.

6.

7.

Group Gnep Group Group Group Group Group 1, Main Steam 2, Reactor Water Sample 3, Reactor Building and Containment 4, Miscellaneous 5, Residual Heat Removal (RHR) 6, RHR (Shutdown Cooling) 7, Reactor Water Clean Up (SELECT 'ONE) 1,3,4,5

& 6 b.

3,4,5

& 6 c.

1,2,'5,6

'& ~

d.

1,2,3

& 4 QUESTION:

014 (1.00)

Whi& ONE(l) of the following annunciators would alarm if a leak developed on a recirculation line weld?

a P601-A3-6.6, LEAK DET.

DRYWELL E(UIP DRAIN FLOW HIGH b.

P601-A3-6.5, LEAK DET.

DRYWELL FLOOR DRAIN FLOW HIGH c.

P602-A6-2.3, RECIRC PUMP A SEAL STAGING FLOW HIGH d.

P602-A13-3.1, REACTOR BLDG EQUIP SUMP HIGH LEVEL

REACTOR OPERATOR Page 15 QUESTION: 015 (1.00)

Valve t'."'~q is being performed on RHR System C.

RHR-V-4C, pump suction valve i "v;:.

~'n~ ~uppression

pooi, ss cio~oi with all valves/swit,ches in their normal standby position when a valid LOCA signal occurs.

Based on the provided information, Which ONE(1) of the following statements identifies the expected response of RHR System C assuming no operator action is taken?

a.

RHR pump 2C starts, trips on low suction pressure.

b.

.RHR pup "i, ~;',.;~oc attempt to start because RHR-V-4C is closed.

c.

RHR-V-4C receives an automatic open signal, pump starts when the valve is full open.

d.

RHR pump 2C starts and remains running, RHR-V-4C remains closed.

QUESTION: 016 (1.00)

WNP-2 is operating at 100! power.

The Low Pressure Core Spray system is lined up "in annal standby condition.

The injection valve, LPCS-V-5, is shut.

Which ONE(1) of the following is correct regarding the operability of LPCS-V-5?

7hz valve can be manually opened from the Control Room anytime the LPCS pump is running.

,b.

The valve will auto open if reactor pressure is less than 520 PSIG.

c.

The valve will auto open if reactor pressure is less than 470 PSIG and a

LOCA signal is present.

d.

The valve will auto open when the LPCS discharge pressure exceeds 145 PSIG.

REACTOR OPERATOR Page 16 (UESTION: 017 (1.00) iihi~h QNE(l) of the following identifies the correct sequence of components in the flow path of the Standby Gas Treatment System?

a 0 b.

C.

d.

Prefilter, HEPA filter, electric heater, moisture separator,

fans, carbon bed adsorber, HEPA filter.

Fans, electric heater, moisture separator, prefilter, HEPA filter, carbon b'd dsor ber, HEPA filter.

Moisture separator, electric heater, prefilter, HEPA filter, carbon bed adsorber, HEPA filter, fans.

Moisture separator, electric heater, HEPA filter, carbon bed

adsorber, fans, prefilter, HEPA filter'.

QUESTION: 018

{1.00)

The Average Power Range Monitoring (APRM) system provides power level

~r<<<~~m<ce ~nd core

<l~"~ ~'onals to the Rod Block Monitor (RBM) system.

Which ONE(1) of the following corresponds to the minimum APRM power level that will cause automatic initiation of RBM restraints?

a.

12/o b.

15/o z

20%%uo d.

30/o

REACTOR OPERATOR Page 17 QUESTION:

019 (1.00)

Which ONE(1) of the following is the reason for the 10 second time delay between receiving and resetting the scram signals?

a.

Allows time for the Scram Discharge Volume to drain.

b.

Allows time to ensure that the scram relays have deenergized.

c.

Allows time for all control rods to reach full in position.

d.

Allows time to ensure that the Scram Valves finish their stroke.

.(UESTJON:

020 (1.00)

Which ONE(l) of the following parameters would be the MINIMUMelectrical overspeed trip setpoint for RFW pump "B"?

a.

Turbine speed of 5300-5650 rpm b.

Turbine speed of 5700-6050 rpm c.

Turbine speed of 6100-6450 rpm d.

Turbine speed of 6500-6900 rpm QUESTION:

021 (1.00)

Drywell pressure increases to 1.68 PSIG during a core traverse by the TIP system.

Which ONE(1) of the following describes the response of the TIP system?

a.

.~',ie TIP system continues its sequence for t!:o present core position, then withdraws the probe, and the isolation valve closes.

h.

The TIP system stops its sequence and requires manual ac'tion to coni

< nue.

c.

The TIP system shifts to manual

reverse, then withdraws and the isolation valve closes.

d.

The TIP system shear valve fires and shears the probe.

REACTOR OPERATOR Page 18 QUESTION:

022 (1.00)

At the Remote Shutdown Panel the Remote Transfer Switches (Instrument power and Valve Control power) are placed in the Emergency position.

Which ONE(1) of the following effects will this action have on system operation from the main Control Room?

a.

System indication and control is still available from the main Control Room.

b.

System control is still available from the main Control Room but without system indication.

c.

System indication is still available in the main Control Room but without system control.

d.

System indication and control is no longer available in the main Control Room.

QUESTION:

023 (1.00)

WNP-2 has scrammed on loss of level.

The RPV Flowchart has been entered.

During execution of the RPV Flowchart, a High Drywell Pressure entry condition occurs.

Which ONE(1) of the following is the correct action?

x~r KC:i.r,". P'C Flowcharts.

b.

Finish execution of the RPV Flowchart, THEN enter the PC Flowchart.

c.

Return to the top of the RPV Flowchart AND enter the PC Flowchart.

d.

Exit the RPV Flowchart, THEN enter the PC Flowchart.

REACTOR OPERATOR Page 19 QUESTION: 024

{2.00)

Which ONE(l) of the following is the reason that ADS is inhibited whenever boron injection is required?

a.

To prevent a loss of boron from the vessel resulting in a reactivity increase.

b.

To prevent a rapid injection of cold, unborated water resulting in a

rapid increase in power.

c.

To present a:".

-.. essive depressurization that would cause the SLC pumps to runout.

d.

To prevent an increase in natural circulation resulting in decreased

<n.I<II-::, a;i.", a~ i~crease in power.

QUESTION:

025 (1.00)

WNP-2 has just scrammed.

The following plant conditions exist:

- Reactor Power

- Reactor Water Level

- Reactor Pressure

- 'Drywell Hydrogen Concentration

- Suppression Pool Level

- Reactor Building Delta Press.

Grgwek', F'.r=-~"::r<

3%%uo 17 inches 1047 psig

2. 3/o

+4 inches

-.05 inches 1.65 psig Which ONE(l) of the following sets of procedures must be entered?

RPV "RPV Control" PC "Primary Containment Control" SC "Secondary Containment Control" (SELECT ONE) a.

ari and

'PC b.

RPY and SC C.

PC'and SC

REACTOR OPERATOR Page 20 QUESTION: 026 (1. 00)

Which ONE(1) of the following conditions will cause a direct recirculation flow control valve runback?

ASSUME the reactor was at full power prior to the transient.

a.

Recirculation Pump "A" discharge valve drifts closed.

b.

'Reactor water level drops to 32 inches.

c.

Feedwater Level Controller runs back and reactor water level drops tn 15 '"ches.

d.

"B" Feedwater Pump trips causing reactor water level to drop to 30 inches.

QUESTION: 027 (1.00) 1 Which ONE(l) of the following components allows cooling water flow into the reactor from the control rod drive mechanism when a control rod is fully withdrawn?

a.

Buffer Orifice 1jncoupling Rod Valve Disk c.

Outer Filter d.

Velocity Limiter Valve Disk

REAClOR OPERATOR Page 21 QUESTION: 028 (1. 00) ah<un vs~i(i) of the follow'ing describes how the Scram Discharge Volume Vent and Drain Valves are closed during a reactor scram?

a ~

b.

c ~

Scram Pilot valves ENERGIZE to REMOVE control air from the operators of the vent and drain valves.

Scram Pilot valves DE-ENERGIZE to REMOVE control air from the operators of the vent and drain valves.

Scram Pilot valves ENERGIZE to SUPPLY control air to the operators of the vent and drain valves.

Scram Pilot valves DE-ENERGIZE to SUPPLY control air to the operators of the vent and drain valves.

QUESTION: 029 (1.00)

Which ONE(1) of the following statements describes the response of ECCS pumps

'to 'a 'LOCA, coincident with a loss of off-site power?

a.

b.

d.

The HPCS pump will start as soon as power is restored to its bus and the low pressure ECCS pumps will start when RPV pressure drops to 487 psig, if power is available.

The

HPCS, LPCS and RHR C pumps start as soon as power is restored to their respective buses and RHR pumps A and B start after a time delay.

All ECCS pumps start as soon as power is restored to their respective buses.

The

HPCS, RHR A and C pumps start as soon as power is restored to their respective buses and the LPCS and RHR B pumps start after a

time delay.

REACTOR OPERATOR Page 22

.AUESTJQN: 030 (1. 00)

Which ONE(1) of the following describes the effect of placing the control switch for ADS main steam relief valve, MS-RV-4D, in the "OFF" position at panel P-603.

a.

Renders all modes of operation inoperative.

b.

Renders all modes of operation inoperative except for the "Safety" mode of operation.

c.

Renders only the manual mode of operation inoperative.

d.

Render all modes of operation inoperative except for the "ADS" and "Safety" modes.

QUESTION:

031 (1.00)

While operating at 90/ power with the FWLC system in 3-element control, the selected FW level sensor "A" control signal fails low.

Assuming that the FWLC

'system responds as designed, Which ONE(1) of the following describes the effects on level in the RPV?

a.

Increases rapidly to above the high level trip setpoint

(+54.5").

b.

Decreases rapidly to below the low level trip setpoint

(+13").

c.

Decreases initially but is returned to and maintained at the desired setpoint.

d.

Remains steady at the desired setpoint.

REACTOR OPERATOR Page 23 QUESTION: 032 (2.00)

HATCH the offgas valves (Column I) with their respective automatic signals (Column II).

The signals in column II may be used

once, more than once or not at all and only one answer may occupy each answer space.

Column I (valve action) a) Close offgas isolation, OG-V-60.

b) Close offgas charcoal adsorber

bypass, OG-V-45.

c)

Open offgas adsorber inlet valves, OG-V-51 A, B.

d) Close offgas holdup

drain, OG-V-23.

Column II (signal) 1.

HSL hi/hi Radiation

2. Offgas post treatment hi radiation.

3. Offgas post treatment hi,hi, radiation.

4. Offgas post treatment hi,hi,hi radiation.

5. Offgas pre treatment hi,hi radiation.

"'~siire of offgas isolation, OG-V-60.

QUESTION:

033 (1.00)

Following a valid ADS initiation, the operator is directed to close the ADS valves.

Assuming that the initiating signals still exist, Which ONE(1) of the following operator actions will cause the ADS valves to close?

a.

Plac its contro'i switches on P-601 for the ADS valves to the OFF position.

b.

Place the ADS inhibit switches on P-601 to the NORHAL position.

c.

Secure (shutdown) all low pressure ECCS systems in either Div.

1 or Div. 2.

d.

Depress the ADS "low level/timer seal-in" reset pushbuttons on P-601.

REACTOR OPERATOR Page 24 QUESTION:

034 (1.00)

During normal power operations the operator notices RWCU delta flow indication is 40 gpm and increasing slowly.

Which ONE(1) statement below describes the TOTAL RWCU response, as indicated in the main control

room, as delta flow increases towards 75 gpm?

a.

Will immediately cause automatic closure of RMCU-V-1, 4, and 33.

After a 45 sec time delay, the RWCU pumps will trip.

b.

Will cause automatic trip of the RWCU pumps ofter

~ 45 sec time delay.

c.

Mill immediately cause automatic closure of RWCU-V-l, 4 and 33, and the RWCU pumps will trip.

d.

After a 45 sec time delay, automatic closure of valves RWCU-V-I, 4, and 33 will occur and the RWCU pumps will trip.

QUESTION:

035 (1.00)

Assume that SLC is required for injection.

Both SLC control switches are placed in the OPERATE position.

The A suction valve opened and the B suction valve failed to opem.

Which ONE(1) statement below describes the start logic associated the A and B SLC pumps?

a.

Both A and B pumps will start.

interlocks.

There are no suction valve b.

Both pumps will start.

Either suction valve open satisfies the start logic for both pumps.

c.

Pump A will start, pump B will not start.

The logic requires the respective suction valve to be open.

d.

Neither pump will start.

Both suctions must be open to satisfy the start logic for both pumps.

REACTOR OPERATOR Page 25 QUESTION: 036 (1. 00)

Which ONE(1) statement below describes the bypass of the recirculation system EOC-RPT logic?

The EOC-RPT is never automatically bypassed but can be bypassed manually at P-609 and P-611.

b.

C.

d.

The EOC-RPT is automatically bypassed when reactor power is less than 30%%u as sensed by average APRH power.

The EOC-RPT is automatically bypassed when reactor power is less than 30/

as sensed by main turbine crossover pressure.

The EOC-RPT is automatically bypassed when reactor power is less than 30! as sensed by turbine first stage pressure.

glKSTION: 037

('1.00)

Assume that RCIC is in operation following a valid initiation signal.

Suppression pool level is verified high and HPCS suction valves are in the process of swapping.

Which ONE(1) statement below describes the RCIC system response to these conditions?

a.

CST suction, RCIC-V-10, closes and THEN SP suction, RCIC-V-31, opens.

.b.

SP suction, RCIC-V-31 opens and THEN CST suction, RCIC-V-10, closes.

c.

SP suction, RCIC-V-31 opens and CST suction, RCIC-V-10, closes simultaneously.

SP suction, RGIC-v-'1, and CST suction, RCIC-V-10, are not affected and remain as is.

REACTOR OPERATOR Page 26 QUESTION: 038 (1. 00)

'lP-"." ;! tern valve operability surveillance is in progress.

Assume both CST

suction, HPCS-V-1, and SP suction, HPCS-V-15, are closed when a valid initiation signal is received.

Which ONE(1) statement below describes the proper response of the HPCS system to these conditions?

a.

CST suction, HPCS-V-1, opens AND the HPCS pump starts simultaneously.

The SP suction, HPCS-V-15, remains closed.

b.

CST suction, HPCS-V-1, opens and THEN the HPCS pump will automatically start when the valve is full open.

SP suction, HPCS-V-15. remains closed.

c.

SP suction, HPCS-V-15, opens AND the HPCS pump starts simultaneously The CST suction, HPCS-V-1, remains closed.

d.

Both CST suction, HPCS-V-l, and SP suction, HPCS-V-15, remain closed and the pump does not start.

QUESTION:

039 (1.00) 7h-

'Yr~!:

ZNS (Emergency Notification System) is manned during an emergency by the "NRC Notifier".

The "NRC Notifier" is always:

(SELECT ONE) a.

The senior equipment operator.

o.

iir~ Shifi. Technical Advisor.

c.

Assigned by the Shift Manager when an emergency is declared.

d.

Designated in the Control Room log at the beginning of each shift.

REACTOR OPERATOR Page 27 QUESTION: 040 (1.00)

An ATWS is in:progress RPV level cannot be maintained above TAF.

The Emergency Procedures have directed the crew to maintain RPV level above

-192".

Which ONE(l) of the following describes the reason for maintaining RPV level above -192"?

a.

To maintain accurate RPV level indication on the Fuel zone instrumentation when level is below TAF.

b.

io en~ui ~ chai the cov~i q~i i'ion of the oi'; generates sufficient steam to cool the uncovered portion.

c.

To control RPV power sufficiently high to maintain natural circulation cooling through the core..

d.

To prevent.core.component damage due to jet pump cavitation caused by a low 'RPV level.

QUESTION:

041 (1.00)

A loss of all high pressure injection systems has resulted in RPV level decreasing to TAF.

An Emergency RPV Depressurization has been directed.

Which ONE(1) of the following states the reason that a minimum of 5 SRVs must be opened?

a.

Ensures that the reactor will be depressurized to below ECCS shut off head before RPV level can decrease to below TAF.

b.

Ensures that at the worst case in core life, the APLHGR thermal limit will not be exceeded and inhibit adequate radiant heat transfer.

c.

Prevents exceeding I/ plastic strain on the hottest fuel pin in the core allowing fuel cladding failure to release radioactive fission products.

d.

Ensures that aufficient steam flow will exist to remove decay heat at low enough pressure for the lowest head ECCS pump to make up for steam flow.

"REAC )OR OPERATOR Page 28 QUESTION: 042 (1.00)

The plant is at 98%%u power operating just below the 100/ rod line when an electrical malfunction causes both reactor recirculation pumps to trip.

After notifying the CRS, Which ONE(l) of the following are the immediate actions which should be taken?

'a ~

b.

c ~

d.

If peak to peak APRH oscillations-are greater than 101 or periodic LPRH upscale/downscale alarms are received on the full core display, manually scram the reactor.

If entry into Region C of the WNP-2 power to flow map has occurred, exit Region C within 15 minutes by control rod insertion via the Fast Shutdown Sequence.

If entry into Region 8 of the WNP-2 power to flow map has occurred, immediately exit Region B by control rod insertion via the Fast Shutdown Sequence.

Confirm the loss of both recirc pumps and manually scram the reactor.

QUESTION: 043 (1.00)

The plant is in Hode 2 with pressure at approximately 600 PSIG.

An electrical malfunction in the startup level controller has caused a loss of reactor vessel level control.

Both startup level control valves (RFW-V-10 AEB) have failed in the open position.

Reactor vessel level is 55" and rising.

Which ONE(l) of the following represents the immediate actions which should be taken to preclude flooding the main steam lines?

a.

Stop the condensate booster pumps before reactor vessel level exceeds

+80 inches.

b.

Operate only one condensate and one condensate booster pump until reactor vessel level can be controlled within the normal operating band.

c.

Prior to reaching a reactor vessel level of +80 inches, throttle RFW-V-I]8 as necessary to maintain level below +108 inches.

d.

Stop the condensate booster pumps before reactor vessel level exceeds

+60 inches.

REACTOR OPERATOR Page 29 QUESTION: 044 (1.00)

Which ONE(l) of the following provides the signal for the Turbine Governor Valve Fast Closure scram?

a.

Position limit switches on the governor valves.

b.

Rate of position change monitors on the governor valves.

c, DEH oil pressure switch.

d.

Generator load signal.

'QUESTION: 045 (1.00)

The plant is in Node 1 at 95% power.

The plant has just experienced a partial loss of control and service air.

The service air header isolation valve has shut and the dryer bypass valve has opened.

A rod drift alarm has just annunciated.

Which ONE(l) of the following describes the immediate actions to be.taken by the operators in response to this event?

a.

Insert an immediate reactor scram.

b.

Substantially reduce reactor power within 30 seconds by control rod insertion via the fast shutdown sequence.

c.

Substantially reduce reactor power within 30 seconds AND lower core flow to 45% of rated.

d.

Substantially reduce reactor power within 30 seconds by control rod insertion via the normal shutdown sequence.

REACTOR OPERATOR Page 30 QUESTION:

046 (1.00)

A 480 VAC locally operated breaker has been racked out to the ui~cunnect position.

An equipment operator notices that the levering shutter on the breaker is in the closed position.

Which ONE(1) of the following describes the significance of this observation?

a ~

b.

The breaker is in the correct position as long as the indicator shows disconnect.

The breaker is in an intermediate position and will require repositioning.

c.

The charging springs for the breaker are still charged and should be discharged.

d.

The trip pushbutton is stuck

a> ~he

.".r >p position.

QUESTION: 047 (1.00)

The plant is in Node I at 75/ power.

Which ONE(l) of the following actions can the operator "At the Controls" (CRO 1) perform?

a.

Swap the Reactor Bldg Ventilation Fans to equalize run time.

b.

Acknowledge a Standby Gas 3'xeatment (SGT) auto start annunciator alarm on a back pa,. T.

c.

Place Standby Gas Treatment System in service for venting the primary containment.

d.

Go to the back panels to take 7.0.0 Surveillance Readings.

'EACTOR OPERATOR Page 31 QUESTION: 048 (1.00)

Haintenance has requested authorization to lift leads on RHR-V-53A motor operator to test starting contacts.

The motor is currently tagged out as part of a HWR.

Which ONE(l) of the following methods would be used to document lead lifting?

a.

10CFR50.59 Evaluation b.

Determ/Reterm Data Sheet c.

Technical Review d.

10CFR21 Evaluation QUESTION: 049 (1.00)

A reactor startup is being performed.

IRH C is on range 3 when the

-24 VDC signal fails.

Which ONE(l) of the following will be the Control Room response to this loss of signal?

a.

IRH Downscale alarm b.

IRH Upscale Hi alarm only.

c.

IRH INOP alarm d.

IRH Upscale Hi alarm and IRH Upscale Hi-Hi alarm

REACTOR OPERATOR Page 32 QUESTION: 050 (1.00)

The plant conditions are as follows:

- Hain Turbine is on turning gear.

- Turbine Throttle pressure is 100 psig.

Hain Condenser vacuum is 10 inches.

Which ONE(1) of the following describes the operation of the Bypass Valves'.

Cannot be opened.

b.

Can only be operated

-i;,.;;:.~.;i>,<L.

c.

Can be operated in BPY HANUAL and BPV AUTO.

'"""<<" -'" BPV.AUTO QUESTION: 051 (1.00)

Assumizg proper eperatic.-

..i

';i;e Reactor Hanual Control System and all interfacing systems, Which ONE(1) of the following rod manipulations could result in a "Rod Drift" alarm?

Rod tns~rtion using the Insert pushbutton.

b.

Rod insertion using the Continuous In pushbutton.

c.

Rod withdrawal using the Withdraw pushbutton only.

d.

Rod withdrawal using the Continuous Withdraw and Withdraw

REACTOR OPERATOR Page 33 OVESTION: 052 (1.00)

Which ONE(1) of the following statements correctly identifies the maximum number of main steam safety relief valves that can be operated following a Control Room evacuation AND the location for operation of these valves?

a.

7 valves from the Remote.Shutdown Panel.

D.

2 valves from the Alternate Remote Shutdown Panel.

c.

3 valves from the Remote Shutdown Panel.

d.

7 valves by opening the output breakers from the RPS MG sets.

qVESTION: 053

('1.00)

Which ONE(l) of the following systems is designed to prevent the local fuel damage that may result from a single rod withdrawal error?

a.

Reactor Manual Control System b.

Rod Worth Minimizer c

Rod.Sam~ence Control System d.

Rod Block Monitor

REACTOR OPERATOR Page 34 QUESTION:

054 (1.00)

Phil= p rforming the Rod Sequence Control System (RSCS) operability surveillance following %he withdrawal of the first in sequence rod, the operator has been directed to verify that the selected rod is not bypassed.

Which ONE(l) of the following methods should the operator use to make this ver~ ficati on?

a.

Verify no "Rod Bypass" annunciators on P603.

b.

Place the "All Rods/Free Rods" pushbutton of the RSCS display in the "Free Rods" position.

The amber light displays rods that are NOT bypassed.

c.

Observe the full core display:

bypassed rods will have a red window illuminated.

d.

Place the "Rods Full In/Bypass" pushbutton of the RSCS display in the "Bypass" position.

The red lights display rods that are bypassed.

After initializing the RWH, the RWH Full Core Display screen indicates that power is below the LPSP and LPAL, a withdraw block has been applied, and one withdraw error is present.

Which ONE(I) of the following methods should the operator use to determine the control rod responsible for the withdraw error?

a.

C.

d.

The rod position indication on the RWH Full Core Display for that control rod will be appended by a "w".

The rod position indication on the RWH Full Core Display for that control rod will be appended by a "i".

The= rod position indication on the RWH Full Core Display for that control rod will be appended by a "++".

The rod position indication on the RWH Full Core Display for that control rod will be appended by a ">".

REACTOR OPERATOR Page 35 QUESTION: 056 (1. 00)

'which ONE(1) of the following represents the response and operability of the Secondary Containment system to a loss of control air?

a.

Reactor Building heating and ventilation system isolates, the supply and exhaust fans stop.

The SGT system auto starts.

b.

Reactor Building heating and ventilation system isolates, the supply and exhaust fans must be manually stopped.

The SGT system is unavailable

~ 'c

i..'":@Lion.

c.

Reactor Building heating and ventilation system isolates, the supply and exhaust fans must be manually stopped.

The SGT system can be manually started.

d.

Reactor Building heating and ventilation system does not respond.

QUESTION: 057 (1.00)

,A LOCA has occurred.

ECCS equipment has been functioning as de~igned.

'Present plant conditions are:

RPV level, -135" and rising RPY pressure, 50 psig

- Wetwell nressure.

9 basic; The CRS has directed CR02 to place RHR loop A into the Drywell Spray mode.

CR02 simultaneously places the switches for RHR-V-16A and RHR-V-17A, the inboard and outboard drywell spray valves, into the open position.

Both valves fail to open.

Mhich ONE(l) of the following states the reason that the valves failed to open?

a.

The valves must be opened sequentially to satisfy interlocks.

b.

The valves will not open until RPV level is restored above level 1.

c.

The valves cannot be opened until a

10 minute timer has expired.

d.

Valve RKR-V-42A, the LPCI injection valve, must be fully shut to satisfy interlocks.

REACTOR OPERATOR Page 36 QUESTION: 058 (1.00)

The Divisions 1

& 2 125 VDC distribution is normally powered by the battery chargers wi".h :.h batter e." l~i.-g the backup source.

Which ONE(1) of the following is the designed capacity of these batteries during the "Design Basis Accident" ?

a.

1 Hour b.

2 Hours c.

4 Hours d.

8 Hours QUESTION: 059 (1.00)

Which ONE(l) of the following represents the HINIHUH reactor power level that would result in a reactor scram if a turbine trip occurred?

a.

20%

b.

25%

c.

30%

d.

40%

REAC)OR OPERATOR Page 37 QUESTION:

060 (1.00)

A reactor scram has occurred due to the loss of the only available Reactor Feed Pump.

All scram systems worked as designed.

HPCS and RCIC have restored level but the CRS has directed that reactor pressure be lowered to allow

'ei$=".~ox with the Condensate Booster Pumps.

Which ONE(l) of the following would be the result if CROl had NOT performed the required immediate actions following a scram?

a.

A HSIV isolation will occur due to low RPV pressure.

b.

A trip of the Recirculation pumps will occur; c.

The scram could not be

~ e~ec due to the Turbine Throttle Valve Closure signal present.

d.

The NS4 isolation signals generated at the Level 3 setpoint cannot be reset.

QUESTION:

061 (1.00)

A failure of the SM-2 bus has occurred while operating at 701 power.

The bus ha~ ii"-u.is~.;~i.,

';li~ich 0:-!'(I) of the following would be the status of the condensate system pumps following this event?

a.

Condensate pumps lA, 1B,

& 1C are running, Condensate Booster pumps are tripped.

b.

Condensate pumps lA & 1B are running, Condensate Booster pumps 2A &

2B are running, all remaining pumps tripped.

c.

Condensate pumps lA & 1C are running, Condensate Booster pumps 2A &

2C are running,all remaining pumps tripped.

d.

Condensate pumps 1C and Condensate Booster pump 2C are running, all remaining condensate system pumps are tripped.

REACTOR OPERATOR Page 38 QUESTION: 062 (1.00)

A fault has occurred in the Fire Protection Ring Header resulting in lowering header pressure.

As header pressure has been lowering, all equipment designed to auto start has been working properly to attempt to restore header pressure.

Header pressure has reduced to 115 psig and has been stabilized at that pressure for approximately ten minutes.

Based on the plant conditions, Which ONE(l) set of the following Fire Protection equipment should be presently running?

l.

2.

3.

5.

6.

FP-P-3, Jockey Pump FP-P-111, Jockey Pump FP-P-2A, Fire Protection Pump FP-P-2B, Fire Protection Pump FP-P-1, Fire Protection Diesel FP-P-110, Fire Protection Pump

'(SELECT DNE) a ~

b.

C.

1 5 2 only w(all.t 1,2,3,4, 8

5 only d.

1,2,3,4,5, E( 6 gUESTHN: C:i.;.ov')

Which ONE(l) of the following radiation monitors is powered by the Reactor Protection System (RPS) bus?

a.

Off-Gas Pre and Post Treatment Honitors b.

Hain Steam Line Rad Honitors Reactor Building Exhaust Rad Honitors d.

Refueling Floor Rad Honitor s

REACTOR OPERATOR Page 39 QUESTION:

064 (1.00)

Which ONE(l) of the following signals will NOT result in the isolation of the normal Control Room intakes and the starting of fans WMA-FN-54A 8 B?

a.

Rad Waste Building High Rad b.

High Drywell Pressure c,

Low RPV level d.

Reactor Building Ventilation High-High Rad QUESTION: 065 (1.00)

Which ONE(1) of the following would be the result of a Jet Pump instrument tap failure?

a.

Loss of Reactor Pressure indication(s) b.

Loss of Upset Level indication(s) c.

Loss of flow signal for APRM flow biased rod block and SCRAM d.

Loss of Fuel Zone Level indication(s)

QUESTION: 066 (1. 00)

'Due to 'an abnormal condition, the full reactor core must be removed for inspection purposes.

The resulting decay heat load exceeds the capacity of the Fuel Pool Cooling System.

Under these circumstances, Which ONE(l) system can.be cross-connected to the Fuel Pool Cooling System to provide additional heat removal capability?

a.

Standby Service Water b.

Residual Heat Removal c.

Reactor Closed Cooling d.

Condensate Transfer

REACTOR OPERATOR Page 40 QUESTION: 067 (1.00)

Refueling is in progress.

The Hain Hoist is loaded and the Node Switch is in REFUEL.

Which ONE(1) of the following conditions will generate a

Rod Block?

a.

A control rod is selected on P603.

b.

The refuel bridge is over the core.'.

Loss of SRH indication.

d.

Fuel Grapple control is placed in the Raise or Lower position.

QUESTION: 068 (1.00)

The.plant is operating at 100/ power when a HSIV isolation and a reactor scram occurs.

The RCIC turbine trips on overspeed.

The CRS directs that level be restored AND maintained using HPCS (suction aligned to the CST).

Which ONE(1) of the following methods would be used to control RPV level?

a.

Cycle HPCS-V-4 (injection valve).

b.

Tht ottle HPCS-V-4 (injection valve).

c.

Open HPCS-V-10 and HPCS-V-ll (return to CST valves).

d.

Cycle HPCS-V-12 (minimum flow return valve).

C.

REACTOR OPERATOR Page 41 QUESTION:

069 (1. 00)

The plant was operating at'power when a MSIV isolation resulted in a scram signal.

The reactor failed to scram.

The CRS directs the starting of the SLC system.

The SLC fails to inject boron.

Which ONE(1) of the following sets lists the systems that can be used to inject boron?

1.

HPCS

?

RRJJ

'3.

'IKIC 4.

LPCS (SELECT 'ONE) a.

1&2 b.

2&3 c.

3&4 d.

1&4 QUESTION:

070 (1.00)

C V~3 VK for buses DP-Sl-1A AND DP-Sl-2A (Division I and II 125 VDC power supplies)

'has occurred.

Which ONE(1) of the following correctly states the expected response of the Control Rods due to this power loss?

a.

Control Rods remain in their present position.

b.

Control Rods rapidly insert due to actuation of the Backup Scram Pilot Valves.

c.

Control Rods rapidly insert due to actuation of the ATWS/ARI valves.

d.

Control Rods rapidly insert due to actuation of BOTH ATWS/ARI and

>Backup Scram Pilot Valves.

REACTOR OPERATOR Page 42 QUESTION:

071 (1.00)

The following plant condit o~.. exist:

- Reactor

Power, 90%

- Feedwater RPV inlet temperature, 410 Deg.

F.

A loss of Feedwater heating occurs.

Which ONE(1) of the following represents the immediate actions required of CROI?

Proriipt;~

';"eel

- er t,o 70% ;; ing recirculation, scram the reactor if feedwater temperature reaches 370 Deg.

F; b.

d.

Shift the recirculation pumps to LFNG, scram the reactor if feedwater temperature reaches 310 Deg.. F.

Promptly reduce power by insertion of control rods, scram the reactor if feedwater temperature reaches 360 Deg.

F.

Promptly reduce power to 70% using recirculation, scram the reactor if feedwater temperature reaches 310 Deg.

F.

QUESTION: 072 (1.00)

The plant is operating in shutdown cooling.

Which ONE(1) of the following valves will close if there is shutdown cooling excess (high) flow?

a.

RHR-V-B, PAR Shutdown Cooling Suction Outboard Isolation Valve.

b.

RHR-V-49, RHR Discharge to Radwaste Valve.

c.

RHR-3-o'uA, RHR-HX-IA Sample Isolation Valve.

d.

RHR-V-123B, Testable Check Valve (RHR-V-50B) Bypass Valve.

REACTOR OPERATOR Page 43 QUESTION:

073 (1.00)

Emergency Operating Procedure PPH 5.0.0. requires that suppression pool level be maintained above the "Vortex limit".

Which ONE(1) of the following states the basis for this limit?

a.

Loss of NPSH resulting in pump cavitation.

b.

Loss of adequate core cooling via low pressure ECCS systems.

c.

Pump damage due to air entrainment.

d.

ECCS initiation will be delayed following a LOCA due to low discharge pressure.

QUESTION: 074 (1.00)

A LOCA has occurred.

Which ONE(l) of the following methods would NOT meet the criteria of providing "Adequate Core Cooling" ?

a.

Spray cooling with HPCS or LPCS when level is less than 2/3 core height or RPY level is unknown.

b.

Spray cooling with HPCS or LPCS at or above Top of Active Fuel (TAF).

c.

Steam cooling with injection of makeup water to the RPV.

d.

Core submergence.

A.

REACTOR OPERATOR Page 44 QUESTION: 075 (1.00)

An ATWS has occurred due to RPS failing to respond.

The CRS has directed that the RPS fuses be removed per EOP 5.l.2., "Failure to Scram."

Which ONE(1) of the following would be ti:; e"ult i

'he SDV vent and drain valves remained o'en?

a.

If rod insertion is successful, scram speeds would be excessive resulting in mechanism damage.

b.

Hot water and steam would be released directly into the Reactor Building sumps.

c.

Scram solenoids will not actuate to scram unless SDV vent and drain are shut.

The rods will not scram.

d.

Runout of the CRD pumps with potential trip of the pumps on overload.

QUESTION: 076 (1.00)

A HSIV isolation and subsequent reactor scram has occurred.

Given the plant conditions, Which ONE(1) of the following is the basis for the action in EOP 5..1.3..

RPV Control, which directs that reactor pressure be lowered to 930 psig with Safety Relief Valves (SRVs) IF any SRVs are cycling?

a.

Controls RPV pressure to within the capability of high pressure injection systems to inject.

b.

Controls RPV pressure below the lowest SRV lift pressure.

c.

Prevents subsequent reactor scram signals due to high pressure.

d.

Reduces pressure to a value corresponding to main turbine bypass valves being 100% open.

4 f~

REACTOR OPERATOR Page 45 QUESTION:

077 (1. 00)

A Reactor scram has occurred.

CR01 is verifying all rods are fully inserted.

.Whiz:h LlNE!3) of the following is the MAXIMUM number of control rods that can be NOT fully inserted and still have assurance that the reactor will shutdown under all conditions irrespective of reactor coolant temperature?

a.

0 b.

1

.C.

2 d.

6 QUESTION: 078 (1. 00)

Hain Condenser vacuum is decreasing.

Efforts to stop the loss of vacuum have

-been unsuccessful.

Which ONE(1) of the following vacuum readings corresponds to the lowest value at which the ability to maintain RPV level with the Reactor Feed Pumps will be possible?

a.

19 inches Hg Vac.

b.

10 inches Hg Vac.

c.

7 inches Hg Vac.

d.

0 inches Hg Vac.

QUESTION:

079 (1.00)

Which ONE(l) of the following corresponds to the MINIMUM reactor pressure necessary for rod insertion based solely on reactor pressure (accumulator and Control Rod Hydraulics inoperative)?

a.

200 psig b.

300 psig c.

375 psig d.

4>u psig

REACTOR OPERATOR Page 46 QUESTION: 080 (1. 00)

At time 0, a Safety Relief Valve (SRV) failed open.

Efforts to close the valve are underway.

Which ONE(l) of the following states the MINIMUM conditions which would require the operator to initiate a manual scram?

a.

When suppression pool temperature reaches 95 Deg.

F or at time 2

minutes.

b.

When Suppression pool temperature reaches 90 Deg.

F or after two unsuccessful efforts to close the SRV.

c.

When Suppression pool temperature reaches 110 Deg.

F or at time 5

minutes.

d.

When Suppression pool temperature reaches 110 Deg.

F or at time 2

minutes.

')UESTION: 081 (1.00)

If a loss of all 250 VDC power were to occur, Which ONE(1) of the following feacfi.ons.would,be affected?

a ~

Circuit breaker operation of SM-7 and SM-8 b.

RCIC operability c ~

HPCS Diesel operability d.

Remote Shutdown Panel operability

REACTOR OPERATOR Page 47 QUESTION:

082 (1.00)

Due to a

LOCA, a high wetwell and drywell pressure

exists, Which ONE(1) of

'the following statements describes why simultaneous operation of drywell and suppression pool sprays from the same RHR loop is NOT permitted?

a.

Potential runout of the RHR pump causing a

pump trip.

b.

Overriding the interlocks to accomplish simultaneous spray results in the inability to vent the primary containment via the suppression chamber.

c.

If simultaneous sprays are initiated and the RHR pump trips, the pressure suppression function of primary containment is bypassed.

d.

Simultaneous operation of sprays can result in a rapid primary

,containment depressurization that could result in a negative pressure being established.

QUESTION:

083 (1.00)

Which ONE(l) of the following is the MINIMUM reactor HIGH Steam Dome pressure v,.'hi".;f ureui~xl.result in a violation of a WNP-2 Safety Limit?

a.

1205 psig

',b.,

3~~

psig c.

1325 psig

/%7'sag

REACTOR OPERATOR Page 48 QUESTION: 084 (1.00)

The plant is operating at 100/ power when the CR01 and CR02 note a sudden decrease in reactor power and turbine load.

While investigating the cause, the CR01 notes that there are some blue SCRAM lights illuminated on the full core display.

Which ONE(1) of the following is the MINIMUM number of rods scrammed that immediately require the CROl to insert a manual scram and place the mode switch into SHUTDOWN?

a.

1 b.

2 C.

3 QUESTION: 085 (1.00)

A small steam leak LOCA has developed in the drywell.

The reactor has been

.shutdown and RPV level is being maintained automatically with RCIC.

Drywell sprays vere initiated at 13 psig in the drywell.

With drywell sprays running pressure is still rising.

The RCIC turbine will trip due to Which ONE(1) of the following if drywell pressure continues to increase due to the steam leak?

a.

High area temperature.

b.

High exhaust pressure.

c.

Overspeed due to loss of steam condensing capability.

d.

High lube oil temperatures.

REACTOR OPERATOR Page 49 QUESTION: 086 (1. 00)

Which ONE(l) of the following is the reason that, with the recirculation pumps tripped, lowering RPV water level during an ATWS reduces reactor power?

a.

Reduces reactor pressure which allows more cold water injection from low pressure systems.

b.

Prevents temperature stratification which results in localized power peaks.

c.

Reduces steam generation rate which increases the moderator temperature.

d.

Reduces static pressure head which increases core voiding.

QUESTION: 087 (1.00)

A partial loss of Reactor Closed Loop Cooling (RCC) has occurred.

RWCU has been manually isolated to reduce heat loads.

As temperatures continue to increase in the system, Which ONE(l) of the following may become a heat sink a.

Standby Service Water Fuel Pool c.

Circulating Water d.

Fire Header

[4

~

REACTOR OPERATOR Page 50

'QUESTION: '088 (1. 00)

A steam leak in the drywell has resulted in rapidly rising drywell temperature and pressure.

The CRS has directed the initiation of drywell sprays.

Which ONE(1) of the following would be the effect if drywell sprays were initiated when plant conditions were in the prohibited region of the DRYWELL SPRAY INITIATIONLIMIT (DSIL) graph?

a.

Thermal shock and potential failure of the drywell spray header piping b.

Rapid containment pressurization due to drywell spray flashing to steam.

c.

Diversion of RHR required for vessel flooding due to loss of RPV level indication.

.d.

iRapid depressurization of the primary containment causing an

excessive negative differential pressure between the drywell and the suppression chamber.

QUESTION: 089 (1.00)

Refueling is in progress.

While a spent fuel assembly is being moved, a

sudden unexpected jolt occurs.

When the operator looks down into the spent

-t'9viol'0'o afMermime the cause, it is noted that bubbles are rising to the water's surface.

Which ONE(1) of the following is the IMMEDIATE operator action?

7"ace all fuel assemblies and control rods in a safe condition.

b.

Ask the Control Room to verify Secondary Containment integrity.

'Evacuat'e 'the refuel floor.

.ci, Perform radiol ogi cal surveys.

REACTOR OPERATOR Page 51 QUESTION:

090 (1.00)

A plant accident has occurred resulting in an off-site release.

Which ONE(l) of.the following is the MINIHUH radiation level that requires entry into the

Ra&m=:. i~=

=..~..: Control Guideline (RR).

a.

0.5 mr/hr whole body at site boundary b.

50 mr/hr whole body at site boundary c.

1 R/hr whole body at site boundary d.

1.2 R/hr whole body at site boundary QUESTION:

091 (1.00)

Failed fuel has resulted in elevated Off-site release rates.

The reactor was shut down and HSIVs shut to limit the release rate.

The Hain Steam Isolation Valve Leakage Control (HSLC) system was manually started to process any leakage gast the HSIVs.

Which PNr (1) of the following is the discharge point of the HSLC system?

a.

Off Gas System b.

Standby Gas Treatment System c.

Waste Surge Tank (EDR-TK-5) d.

Waste Collector Tank (EDR-TK-2)

REACTOR OPERATOR Page 52 qUESTION:

092 (1.00)

The Reactor Building Exhaust Plenum radiation readings have been showing a

steady increase.

Whic& QWZ('.."j ';

ilowing is the MINIMUM exhaust plenum radiation level at which the Standby Gas Treatment System is designed to auto-start?

a.

2 mr. t;r b.

10 mr/hr c.

13 mr/hr d.

20 mr/hr QUESTION:

093 (1.00)

The plant was operating at 80/ reactor power when the following annunciators alarmed:

- 6.9 KV BUS 5 OVERCURRENT LOCKOUT Board C

- 6.9 KV BKR N2-5 TRIP Board C

- 6.9 KV BKR S-5 TRIP Board C

- 6.9 KV BUS 5 UNDERVOLTAGE Board C

Which ONE(1) of the following immediate actions are required of the CROl?

a.

Place the Mode Switch in SHUTDOWN.

b.

TRIP both Recirculation Pumps.

c.

Refer to PPM 4.12.4.7.,

"Unintentional Entry Into Reqion Of Potential Core Power Instability".

d.

Ensure SM-7 and SM-8 are energized.

REACTOR OPERATOR Page 53 gUESTIDN: 094 (1.00)

Hain condenser vacuum is decreasing.

Which ONE(1) of the following is the

'reason that the mechanical vacuum pump CANNOT be used at reactor powers above 5/a?

The pump wHi experience air binding.

b.

An explosion could occur inside the pump.

c.

The pump will exceed design loads.

d.

The radiation release will exceed facility design due to the Off-Gas system being bypassed.

QUESTION:

095 (1. 00)

Which ONE(l) of the following operator actions in accordance with EOP 5.3.1.,

"Secondary Containment Control", does NOT mitigate offsite doses for an accident which releases radioactivity inside secondary containment?

a.

Operating reactor building ventilation.

b.

Isolating primary systems that are discharging to the area.

c.

Emergency depressurizing.

d.

Shutting down the reactor.

REACTOR OPERATOR Page 54 DN: 096 (1.00)

A failure of the Reactor Building Ventilation system has occurred.

The CR02 has started Standby Gas Treatment.

Which ONE(1) of the following is the MINIMUM Reactor Building pressure that should be established under these circumstances to ensure the appropriate negative pressure will be maintained in secondary containment under all'conditions?

a

-.25" WG b.

-.6" WG cc.

-2.:0" WG d.

-1.2" WG QUESTION:

097 (1.00)

WNP-2 is operating at 100/ core flow, 80/ reactor power.

I L C is performing a surveillance on the EOC-RPT protective functions of the recirculation system.

Due to improper performance of the surveillance, an EOC-RPT trip is introduced.

The recirculation system responds as designed to the trip signal.

.'9;-m rJ tm'he provided information, Which ONE(l) of the following corresponds to final reactor power and core flow established assuming no operator action was taken?

(refer to attachment 1) 572 power, 40/ core flow b.

57/ power, 60/ core flow

'c.

46/ power, 41/ core flow d.

40/ power, 31/ core flow

REACTOR OPERATOR Page 55 QUESTION: 098 (1.00)

IT< ~<.~df ua<l e with PPN 1.3.1.,

't.oi<duct of Operations",

under Which ONE (1) of the following conditions can a leverage device be used to operate a valve?

a.

Closing a manual valve b.

Backseating a manual valve c.

Operating a motor operated valve d.

Opening a manual valve

(~********

END OF EXAMINATION**********)

1i0 110 100 90 80 A NATURALCIRCUL;.ION 3 LOW PUMP SPEED'ltl FCV POSITIOtl C LOW PUMP SPEEI..AAXFCV POSITION D flATEDPUMP SPE 0/MittFCV POSITIOII (At(ALYSIS)

E RATED PUMP SPE =DIMINFCV POSITION (ACTUAI.)

F ANALLOWER LIQiTOl'ASiER POWER FLOW COtlTROL 0 ANALUPPER LIL'llf OF MASTER POWER FLOW COtlTflOL N vET pulsp NozzLE chvrraTlott fl RECIRC. PUMP CAVITATIOK T PUMP CAVITATIOtIINTERLOCK U MINFLOW CONT";OLLINEFOE TRANSFER TO RATED SPEED V FCV CAVITATIO."...iERLOCK 5cG

)qO

><pi ui+

OC+

G r~ Hii 0 u.i I

60 O

60 8

A X

- 0 0'l.go pgITTE 40 30 20 27.5% PCWEB (23.5% FW -...OW 10 0

10

'20 30 40 50 60 70 80 090 100 110 COBE FLOW (%)

89 4I6 52 Lf JAN liJOO IIIC FIGURE 13.

POWER/FLOW MAP

REACTOR 'OPERATOR p~c C r~c c~yg/

6ru -w

~r~~gwg

+c J c 4 v'+

Page 56 ANSWER:

001 (1.00) f+1.0]

REFERENCE:

1.

WNP2: 1.3.1.

Pg.

42

,2..

KA 294001A102 (4. 2"/4. 2*)

294001A102

.. (KA')

ANSWER:

002 (1. 00) d.

[+1.0]

REFERENCE:

1.

WNP2:

PPN 1.3. 1, Pg.

33 2.

KA 29400]A109 (3.3/4.2*)

294001A109

.. (KA')

ANSWER:

003 (1. 00) c.

[+1.0]

'REFERENCE':

1.

WNP2:

PPM 1.3.8, PG.

20

'2.

KA 294001K102 (3.9/4.5*)

294001K102

..(KA's)

ANSWER:

004 (1.00) c.

[+1.0]

REACTOR OPERATOR Page 57

REFERENCE:

WNP2:

PPY '. 7.2-PG.

8 2.

KA 294001K105 (3.2/3.7) 294001K105

..(KA's)

ANSWER:

005 (1.00),,

[+1.0j

REFERENCE:

1.

WNP2:

PPH 1.11.3, PGS.

11,12 2.

KA. 294001K103 (3.3/3.8) 294001K103

..(KA's)

ANSWER:

006 (2.00) a.

8 b.

6 c.

1 d.

2

[+0.5]

[+0.5]

[+0.5]

[+0.5]

REFERENCE:

'1.

WNP2:

PPH 1. 11.3, PG.

6 2.

KA 294001K103 (3.3/3.8) 294001K103

.. (KA')

ANSWER:

007 (1.00) a.

[+1.0]

REFERENCE:

1.

WNP2:

PPH 1.3.1.5.D.3 2.

TASKS: RO-0559-N-AOMIN 3.

KA 294001A103 (2.7/3.7) 294001A103

.. (KA')

R ACTOR OPERATOR Page 58 ANSWER:

OOS (1.00) d.

[+1.0]

REFERENCE:

1.

WNP2: 82-RSY-1305-T3, PG.

29

'-'-'A")Ã3~~0?

r? 8/?.9) 205000K502

..(KA's)

ANSWER:

009 (1.00)

'c.:I+1.'0]

REFERENCE:

1.

WNP2: 82-RSY-0205-T2, PG.

23 i'U~ 5%~4%401 !3.6/3.8) 215004K401

.. (KA')

ANSWER:

010 (1. 00)

/+3..%']

REFERENCE:

1.

WNP2: 82-RSY-0103-T4, PG.

10 2.

239002A407 (3.6/3.6)

Z3&02A407

..(KA s) a.

I'+1. 03

I REACTOR OPERATOR Page 59

REFERENCE:

1.

WNP2: 82-RSY-1305-T5,FIG.

1A 2.

KA 264000K301 (4.2*/4.4*)

264000K301

..(KA's)

ANSWER:

012 (1. 00) d.

[+1.0]

REFERENCE:

I.

WNP2: 82-RSY-0105-T2, PG.

13 2.

KA 216000A301 (3.4/3.4) 216000A301

.. (KA')

ANSWER:

013 (1. 00) a.

goal.0]

REFERENCE:

WNP2: 82-RSY-0905-T2)

PGS.

5-11

'2.

KA,223002A102 (3,7/3.7) 223002A102

..(KA's)

ANSWER:

014 (1. 00) b.

[+1.0]

REFERENCE:

1.

WNP2:

PPM 4.3.1.2.1.A, PG.

1 2.

KA 223001K104 (3.1/3.2) 223001K104

.. (KA')

REACTOR OPERATOR Page 60 ANSWER:

015 (1. 00) d.,[+1. Ol

REFERENCE:

1.

VNP2'2-'RSY-1305-T3, PGS.

14 E )5 2.

KA 203000A302 (4.0/3.9)

.92MDA302

...(KA's)

ANSWER:

016 (1. 00) c.

(+1.0]

REFERENCE:

1.

WNP2: 82-RSY-1204-T3, PG.

5 2.

KA 209001A403 (3.7/3.6) 209001A403

..(KA's)

ANSWER:

017 (1. 00)

[+1.0]

.KW~~E".

1.

WNP2: 82-RSY-0405-T3, FIG.

1 2

M 261000G007 (3.5/3.7) 261000G007

..(KA's)

ANSWER:

018 P,. 00$

d.

[+1. 0]

REACTOR OPERATOR Page 61

REFERENCE:

1.

WNP2: 82-RSY-0805-T1, PG.

2 2.

WNP2: 82-RSY-0506-T2, PG.

2 3.

KA 215005K103

$3.4/3.5) 213005K103 (KA')

~ANSWER:

619

'(1.00) c.

t+1.0]

REFERENCE:

1.

82-RSY-0706-T2, p.

17.

2.

KA 212000K408 (4.2/4.2) 212000K408 (KA's)

ANSWER:

020 (1.00) c.

I+> Oj

REFERENCE:

l.

82-RSY-0503-T4, pp.

22 and 23.

2.

KA 259001A310 (3.4'.4) 259002A32D

..qKA's)

ANSiit.;

921 f1..QQ) c.

[+1.0]

REFERENCE:

1.

82-RSY-0605-T2, p.

13.

2.

K/A 215001K401 (3.4/3.5) 215001K401 K'ANSWER:

822

" (.1..10)

REACTOR OPERATOR Page 62

REFERENCE:

l.

82-RSY-1602-T3, pp. 3, 6, and 7.

.2..

K/A 2950l6K201 (4.4/4.5) 295016K201

..(KA's)

ANSWER:

023 (1.00) c.

[+1.0]

REFERENCE:

'1.

MNP-2 Proceoure

~.u.'~, p. 28.

2.

K/A 294001A102 (3.8/4.4) 294001A102

.. (KA')

ANSWER:

024 (1.00) b.

[+1.0]

REFERENCE:

.1.

WNP-2 Procedure 5.0.8, p.

103

'2.

K/A 295014A203 (O.X'f4,:3) 295014A203

..(KA's)

ANSWER:

025 (1.00)

[+1.0]

REFERENCE:

1.

MNP-2, EOP Flowcharts 2.

K/A 295029G012 (3.8/4.4) 295029G012

..(KA's)

ANSWER:

026 (1.00) d.

[+1.0]

I. ~

REACTOR OPERATOR Page 63

REFERENCE:

82-RSY-1106-T1, p.

11.

2.

K/A 295009K301 (3.2/3.3) 295009K301

..(KA's)

ANSWER:

027 (1.00) b.

[+1.0]

REFERENCE:

1.

82-RSY-0304-T1.

2.

K/A 201003G007 (3.6/3.6)

.201003G007

..(KA's)

ANSWER:

028 (1.00)

,b..

[+1. 0]

REFERENCE:

l.

82-RSY-0405-Tl.

'2.

K/A 201001K406 (3.8/3.9) 201001K406

.. (KA')

ANSWER:

029 (1. 00) b.

[+1.0 pt]

REFERENCE:

82-RSY-1004-TS,

p. 25.

295003K203 (3.7/3.9) 295003K203

..(KA's)

REACTOR OPERATOR page 64 ANSWER:

030 (1.00) d e

)+I ~ V) fP.BF.ERENCE:

1.

82-RSY-0103-T4 p. 20.

2.

KA239002K408 (3.6/3.7) 239002K408

..(KA's)

ANSMER:

031 (1.00)

-d.,f+1,.0 pt]

REFERENCE:

1.

82-RSY-0603-TY.

~2.

KA259002K605 (3.5%3.5) 259002K605

..(KA's)

ANSMER:

032 (2.00)

'ck o b.

2 c.

2 (d.

6

[+0. 5]

[+0. 5]

[+0.5]

[+0.5]

REFERENCE:

l.

82-RSY-1105-T4 pp.

18, 19.

2.

KA271000K408 (3.1/3.3)

'271000K408

..(KA's)

.46ilER':

033 (1. 00) d.

[+1.0 pt.]

REACTOR OPERATOR page 65

REFERENCE:

l.

82-ASY-1105-T3 p. 9.

2.

KA218000A403 (4.2/4.2).

218000A403

..(KA's)

ANSWER:

034 (1.00)

'd.

f+1.'0:pt. ]

REFERENCE:

'cPc.-feb'i-i704-Tl p. 6.

2.

KA204000K404 (3.5/3.6) 204000K404

..(KA's)

ANSWER:

035 (1.00) b.

[+1.0 pt.]

REFMEKE:

82-RSY-0904-Tl p. 9.

2.

KA211000K408 (4.2/4.2)

.'ZMQOGK408

.. (KA')

%6WER:

036 (1. 00) d.

[+1.0 pt.]

REFERENCE:

1.

82-RSY-1006-Tl p.

15.

2.

'KA202001K128 (3. 9/4. 1) 202001K128

.. (KA')

ANSWER:

037 (1. 00) d.

[+1.0 pt.]

REACTOR OPERATOR Page 66

REFERENCE:

1.

82-RSY-100S-T3 p.

15.

2.

KA217000A217 (3.2/3.4) 217000A217

.. (KA')

ANSWER:

038 (1.00) a.

[+1.0 pt.]

REFERENCE'.

82-RSY-0904-T3 p. 5.

2.

KA209002A108 (3. 1/3.3) 209002A108

..(KA's)

ANSWER:

039 (1.00) d.

[+1.0 pt.]

REFERENCE:

l.

Emergency Planning Training 2.

KA294001A116 (2.9/4.7) 294001A116

..(KA's)

ANSWER:

040 (1.00) b.

[+1.0 pt.]

REFERENCE:

1.

PPN 5.0.8 Rev 0 (C5-3).

2.

KA295037K303 (4.1*/4.5*)

Z95mvWO3

.. (KA's)

ANSWER:

041 (1. 00) d.

I+1.0 pt.]

REACTOR OPERATOR Page 67

REFERENCE:

1.

PPM 5.0.8 Rev 0 '{C2-1.3).

2.

KA295031K305 (4.2/4,.3) 295031K305

..(KA's)

ANSWER:

042 (1. 00)

[+1.0 pt.]

REFERENCE:

1.

PPM 4.2.1.10 SEC 3

2.

KA 295001G010 (3.8/3.7)

,295001G010

..(KA's)

ANSWER:

043 (1.00) a.

[+1.0 pt.]

REFERENCE:

1.

PPM 4.2.1.2 2.

KA295008A108 {3.5/3.5) 295008A108

..(KA's)

ANSWER:

044 (1.00) c.

[+1.0 pt.]

REFERENCE:

1.

PPM 82-RSY-0706-T2 2.

KA21200A212 {3.7/3. 9) 212000A212

.. (KA')

ANSWER:

045 (1.00) c.

[+1.0 pt]

REACTOR OPERATOR Page 68

REFERENCE:

1. PPH 4.8.1.1 Sec 3.0 2.

KA295019K201 (3.8/3.9)

.295019K201

..(KA's)

ANSWER:

046 (1.00)

.a.,[+3 0 ytj

REFERENCE:

PPH 2;7.13 Rev 3, p. 5.

2.

KA262001G009 (3.6/3.6) 262001G009

..(KA's)

ANSWER:

047 (1.00) b.

[+1.0 pt.]

REFERENCE:

1.

PPH 1.3.1.

'2.

KA294001A103 (2.7/3.7) 294001A103

.. (KA')

ANSWER:

048 (1.00) b.

[+1.0 pt.]

REFERENCE:

1.

PPH 1.3.9.5.C 2.

KA294001K102 (3.9/4.5)

>~nnr~>02

..(KA's)

ANSWER:

049, (1. 00)

[+1.0]

REACTOR OPERATOR Page 69

REFERENCE:

1.

MNP2: 82-RSY-0304-T2, PG.

15 2.

KA 215003K602 (3.6/3.8)

.. (KA's)

ANSWER:

050 (1.00) b.

[+1.0]

REFERENCE.:

1.

WNP2: 82-RSY-0403-L5, L.O.

5 2.

KA 241000A406 (3.9/3.9) 241000A406

..(KA's)

ANSWER:

051 (1. 00) b.

t+1.0]

REFERENCE'RP2:

82-RSY-0506-Tl, PG.

11 2.

KA 201002A303 (3.2/3.2) 201002A303

..(KA's)

ANSWER:

052 (1.00) c,.

g+'1. 0]

REFERENCE:

1.

WNP2: 82-RSY-1105-T3, PG.

5 2.

WNP2:

PPM 4.12.1.1.

3.

KA 218000G009 (4. 4*/4.1*)

218000G009

..(KA's)

REACTOR OPERATOR Page 70 ANSWER:

053 (1.00) d.

[+'.D]

REFERENCE:

1.

WNP2: 82-RSY-0805-T1 2.

KA 215002G004 (3.3/3.4) 2150u26004

..(KA's)

ANSWER:

054 (1. 00) d.

[+1.0]

REFERENCE:

1.

WNP2: 82-RSY-0706-T1.

PG.

7 2.

KA 201004A401 (3. 4/3. 5) 201004A401

..(KA's)

ANSWER:

055 (1.00) a.

j+1.0]

REFERENCE:

1.

WNP2: 82-RSY-0606-T1, PG.

18 2.

KA 201006K510 (3.2/3.3) 201006K510

..(KA's)

ANSWER:

056 (1.00) c.

[+1.0]

L REACTOR OPERATOR Page 71

REFERENCE:

1.

WNP2: 82-RSY-0304-T3, PG.

13 2.

KA 290001K608 (2.7/2.8) 290001K608

..(KA's)

ANSWER:

057 (l.00) d.

[+1.0]

REFERENCE:

1.

WNP2: 82-RSY-1305-73, PGS.

16820 2.

KA 226001A403 (3.5/3.4) 226001A403

..(KA's)

ANSWER:

058 (1.00) b.

[+1.0]

REFERENCE:

WNP2: 82-RSY-1102-T5 2.

KA 263000GO04 (3.4/3.6) 263000G004

..(KA's)

ANSWER:

059 (1.00) c.

[+1.0]

REFERENCE:

1.

WNP2: 82-RSY-1004-T5, PG.

21 2.

KA 245000A204 (3.7/3.8) 245000A204

..(KA's)

REACTOR OPERATOR Page 72 ANSWER:

060 (1. 00) a.

[+l.0]

REFERENCE:

1.

WNP2: 82-RSY-0103-T4, PG.22 2.

KA 239001K127 (4.0/4. 1) 239001K127

.. (KA')

ANSWER:

061 (1.00) c.

[+1.0]

REFERENCE:

1.

WNP2: 82-RSY-0205-T4, PG.33 2.

KA 256000K201 (2.7*/2.8) 256000K201

..(KA's)

ANSWER:

062 (1.00) h.

[+1.0]

REFERENCE:

1.

WNP2:

PPM 4.8.7. 1.2, PG.

2 2.

KA 286000K403 (3.3/3.4) 286000K403

..(KA's)

ANSWER:

063 (1. 00)

[+1.0]

REf'.CTOR OPERATOR Page 73

REFERENCE:

1.

WNP2: 82-RSY-0706-T2, PG.

10 2.

KA 272000K601 (3.0/3.2) 272000K601

..(KA's)

ANSWER:

064 (1.00) a.

[+1.0]

REFERENCE'.

WNP2:

PPH 4.10.3.1, PG.

2 2.

KA 290003K401 (3.1/3.2) 290003K401

..(KA's)

ANSWER:

065 (1.00)

-d.

[+1.0]

REFERENCE:

1.

WNP2: 82-RSY-0105-T2, FIGS.

4

& 6 r2 KA 290002K307 (3.1/3.1) 290002K307

..(KA's)

~VSatZR:

066 (1.00) b.

[+1.0]

REFERENCE:

3.

WNP2: 82-RSY-1406-T1 i.,

KA 233000K102 (2.9/3.0) 233000K102

..(KA's)

Ca REA CTOfl OPERATOR Page 74 ANSWER:

067 (1.00) h.

K+3. 03

REFERENCE:

1.

WNP2: 82-RSY-1504-T1, PG.

24 2.

KA 234000K402 (3.3/4.1) 2340MK402

..(KA's)

ANSWER:

068 (1.00) a.

[+1 0]

REFERENCE:

1.

WNP2: 82-RSY-0904-T3, PG.

9 2.

.KA 209002A408 (3.6/3.7) 209002A408

..(KA's)

ANSWER:

069 (1. 00) b.

[+1.0]

REFERENCE:

1.

'WNP2: 82-RSY-0904-Tl 2.

KA 211000A202 (3.6/3.9*)

211000A202

..(KA's)

ANSWER:

070 (1.00) a.

$+1.0]

REACTG}; OPERATOR Page 75

REFERENCE:

'1.

WNP2: 82-RSY-0405-T1, PG.

28 2.

KA 201001K203 (3.5*/3.6*)

203092V203

.. {KA's)

,ANSWER:

071 (1 00) d.

[+1.0]

i"Et:ERit<Ci:

1.

WNP2:

PPM 4.2.7.2.3.0 2.

KA.295014G010 (4.0~/3.9*)

295014G010

..(KA's)

ANSWER:

072 (1.00) a.

[+1.0]

REFERENCE:

%NP2". $2-RSY-1305-T3.

p.

16, L.O. 12e.

2.

KA 295021A202 (3.4/3.4) 295021A202

.. (KA')

ANSWER:

L)73 (1.00) c.

[+1.0]

REFERENCE:

W)/cV~i.". R!N 5.0.0.

2.

WNP2:

EOP L.O.

11 3.

KA 295030K102 (3.5/3.8) 295030K102

..(KA's)

REAC',OR")PERATOR Page 76

.ANSWER:

074 (1.00)

REFERENCE:

1.

WNP2:

PPM 5.0.8.,

pg.

10 2.

WNP2:

EDP L.O.

1 3.

KA 295031K101 (4.6"/4.7*)

295031K101

.. (KA')

ANSWER:

075 (1.00)

'b.

'[il.0]

REFERENCE:

1.

WNP2:

PPH 5.0.8.,

PG.

47

'2.

WNP2:

EOP L.O.

12 3.

KA 295032K303 (3.8/3.9*)

295032K303

..(KA's)

ANSWER:

076 (1.00) b.

[+1.0]

REFERENCE:

WNP2:

PPH 5.0.8, PG.

75 WMP2:

EOP L.O.

13 3.

KA 295007K304 (4.0/4. 1*)

295007K304

..(KA's)

ANSWER:

077 (1.00) b.

[+1.0]

REACTOR'PERATOR Page 77

REFERENCE:

1.

WNP2:

PPM 5.0.8.;

PG.

95'.

WNP2:

EOP L.O.

13 3.

KA 295006K102 (3.4/3.7)

'295006K102

..(KA's)

ANSWER:

078 (1.00) c.

[+1.0]

REFERENCE:

1.

WNP2:

PPM 4.6.1.1, PG.

2

.2.

KA 295002K205 (2.7/2.7) 3.

Note: based on MSIV closure resulting in loss of motive force for RFPs 295002K205

..(KA's)

AMSQER:

079 (1. 00) c.

[+1.0]

-REFSPSMCE:

1.

WNP2: 82-RSY-0304, p.

8 2.

KA 295022K101 (3.3/3.4)

.295022K101

..(KA's)

'ANSWER:

080 (1. 00) d.

[+1.0]

REFERENCE:

1.

MfP2:

PPM 4.601.A2 2.

, KA 295013G010 (4.1/4.1)

Z95013G010

.. (KA')

REACTOP.

OPERATOR Page 78 ANSMER:

081 (1. 00) b.

t+1.0]

REFERENCE:

'MNP2: '82-'RSY-1'102-T5, p.

18 2.

KA 295004K203 (3.3/3.3) 295004K203

..(KA's)

ANSMER:

082 (1.00) c.

t+1.0]

REFERENCE:

1.

MNP2:

PPH 5.0.8.,

p.

48 2.

MNP2:

EOP L.O.

12 3.

KA 295010G007 (3.6/3.8) 295010G007

..(KA's)

.ANSAL<A."

083

.,(1 AO) c.

[+1.0]

REFERENCE:

1.

MNP2: Tech Spec, Sec. 2.0 MNP2: Tech Spec L.O.

1 3.

XA 2'95025K105 (4.4*/4.7*)

295025K105

..(KA's)

REACTOR OPERATOR Page 79

REFERENCE:

1.

MNP2:

PPH 4.1.1.'7.3 2.

KA 295015G010 (4. 0*/3.9*)

'-a~UI~GQl'0

..(KA's)

ANSWER:

085

( 1. 00) b.

t+1.0]

REFERENCE:

1.

MNP2:

PPM 5.0.8.,

p.

41 2.

MNP2:

EOP L.O.

10

.3.

XA 295024G007

{3.6/3.9) 295024G007

..(KA's)

ANSWER:

086 (1.00) d.

[+1.0]

REFERENCE:

1.

MNP2: EOP.L..O.

13

'2.

%MFR': F991 5 5.$.p.

293 3.

295037K102 (4. 1/4.3)

~6fX~7Kl02

.. (KA')

ANSWER:

087

,(1.00) b.

[+1.0]

REFERENCE:

1.

MNP2:

PPM 4.8.3.2.B.3.

2.

KA 295018K307 (3. 1/3. 2)

,295M BK307

.. (KA's)

REACTOP. OPERATOR Page 80 ANSWER:

088 (1.00) d.

[+1.0]

REFERENCE:

1.

WNP2:

PPH 5.0.8.,

p.

125 2.

WNP2:

EOP L.O.

13 3.

KA 295028K303 (3.6/3.9) 295028K303

..(KA's)

ANSWER:

089 (1.00)

" c.

[+1.0]

~/gr/5/

REFERENCE:

1.

WNP2:

PPM 4.12.3.1.3.A.

KA Pq~iq>>~q]0 (3.8*/3.9*)

295023G010

..(KA's)

ANSWER:

090 (1. 00)

REFERENCE:

1.

WNP2:

PPH 5.0.8.,

p.

205 2.

WNP2:

PPH 5.4.1.

3.

WNP2:

EOP L.O.

2 4.

KA 295038GOll (4.2"/4.5*)

295038GOll

..(KA's)

ANSWER:

091 (1.00) b.

[+1.0]

RP'ACTOR OPERATOR Page 81

REFERENCE:

1.

WNP2:

PPM 82-RSY-0503-T3, Pg.

1 2..295017K211 (3. 1/3. 2) 295017K211

.. (KA')

ANSWER:

092 (1.00)

[+1.0]

REFERENCE:

1.

WNP2: 82-RSY-0304-T3, pgs.

13 8

14 2.

WNP2:

PPM 5.3.1.

3.

KA 295034K203 (4.3*/4.5*)

295034K203

..(KA's)

ANSWER:

093 (1.00) c.

[+1.0]

REFERENCE:

1.

WNP2:

PPM 4.7.1.7.3.B.

2.

KA 295003G010 (3.9"/4.1*)

295003G010

..(KA's)

ANSWER:

094 (1.00) b.

[+1.0]

REFERENCE:

1.

WNP2:

PPM 4.6.5.1.

2.

WNP2:

PPM 3.1.3.5.P.

3.

KA 295002G007 (3.2/3.2) 295002G007

..(KA's)

REACTM OPERATOR Page 82 ANSWER; 095 (1.00) a.

[+1.0]

'REFERENT:

1.

WNP2:

PPM 5.0.8.

p.

183

.2 XA 295033K103 (3.9/4.2l 295033K103

..(KA's)

'ANSWER:

096 (1.00)

REFERENCE:

1.

WNP2:

PPM 4.10.1.1.3.A.

2.

KA 295035A102 (3.8/3.8) 295035A102

..(KA's)

ANSWER:

097 (1.00)

'c.

[+1.'0]

REFERENCE:

l.

WNP2: 82-RSY-1006-Tl) p.

15 2.

WNP2: 82-RSY-1106-Tl, Fig.

13 3.

KA 202002A101 (3.2/3.2) 202002A101

..(KA's)

REACTÃ OPERATOR Page 83

REFERENCE:

1.

WNP2:

PPH 1.3.1.",

Pg.

25 2.

KA.294001K101 (3.7/3.7) 294001K101

.. (KA')

(~********

END OF EXAMINATION **********)