Text

Exam Rept 50-321/OL-89-01 on 890612-16.Exam Results:Nine Senior Reactor Operators & Two Reactor Operators Passed Exam.Three Senior Reactor Operators Failed Exams
	ML20247C954
Person / Time
Site:	Hatch
Issue date:	08/30/1989
From:	Casto C, Hopper G; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML20247C933	List: ML20247C928; ML20247C954; ML20247C981;
References
	50-321-OL-89-01, 50-321-OL-89-1, NUDOCS 8909140137
	Download: ML20247C954 (203)
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, ENCLOSUREi t*1 i: REPORT DETAILS y 'y, ,

, Examination Report No..50-321/0L'89 -

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Facility' Licensee:?3 Georgia Power Company -U

,,: P. 0. Box 4545

. -f, Atlanta, GA 30302

%, 'Facil.ity Docket Nos:. 50-321 and 50-366- '

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Facility-License Nos.:DPR-57 and NPF-5

Examk'a'tions'wereadministere n at 'the E. I. Hatch Nuclear Power Station in i Baxley,; Georgia.

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\ 1 Chief Examiner:

J , eorge T ppe V

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Date Signed s

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('!h IAppro'vedBy:' . @,r

\Lherr'Tes A. Casito', Chief

_. .Date F3 0Signed

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.p; Operator Licensing Section 1- .

Operations. Branch

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.", Division of Reactor Safety <

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Examinations were administeind during the pericd k ne 12 - 15, 1989.

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Written examinations and operating tests were administereri to 12 SRO and 2 R0.

' . applicants. ' Nine SRCs and two R0s passed these examinations. All othes

% -failed.

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y REPORT DETAILS

1. Examiners

'*G. T. Hopper,-NRC J..F. Hanek, EG&G-M. W. Parrish, EG&G K. M. Spencer, EG&G

Chief Examiner

2. Pre-Examination Review:

On May 30 - 31,1989, the Chief Examiner and members of the facility

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Training ~and Operation Staffs reviewed the written examination at the Region H offices in Atlanta. This effort was an initiative to improve examination validity and relevance. The pre-examination review provided und opportunity to ensure all test items were valid and concise prior to the administration of toe examination.

3. Post-Examination Review:

At the conclusion of the written examination, the examiners provided

. your staff with a copy of the written _ examination and answer key for !

review. Facility comments concerning the written examination are . '

included in this report as Enclosure 3. The NRC resolutions to comments made by the facility reviewers are inci ded i as Ench.sure. 4.

The pcst-examination review resuited in 15 comments which were cwacn to both tests and 5 comments which were unique to each exam (R0/SRO). The total of 25 post exam comments indicates that the pre-examination review was cursory and ineffective. The pre-examination review should ensure that the quettions are unambiguxs and that the answers are technically correct. Furthermore, multiple (hoice questions with tacce than one correct answer snould be id(ntifM and corrected For the r:st part, this was not accomplished. Efforts in the future will have to improve results for the pre-examination review process to continue.

4. Exit Meeting At the conclusion of the site "isit the Chief Examiner met with representatives of the plant staff to discuss the results of the examinations. The following generic weaknesses were noted during the performance of the Operating Exams:

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.a. Five SR0 candidates were observed to have skipped Emergency Operating Procedure'(E0P)~ steps, read or answered decision blocks incorrectly,

< andinotecomplete E0P. Paths prior to entering End Path Procedures. -

, ' Also,: candidates were observed to have exited one E0P Path 'and entered , another. at . incorrect grid' locations. Candidates . passed over. Notes without referring to their contentg resulting in required actions.not being performed.

b. Several candidates did not know that. resin injection into the Reactor Pressure ' Vessel (RPV)' could cause a Main Steam Line High Radiation Trip and Isolation. This was unexpected considering the number of related industry events that have occurred over the years such as:

Brunswick LER 1-79-74

. Peach' Bottom ~ LER 86022.

Hope Creek LER 36089 and 87020

c. Several candidates did not'use and/or were unaware of the MSIV High Radiation Chart Recorder 2011-R603 located on Panel 2H11-P600.

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d.- 'Several candidates were overly concerned with maintaining RPV level

.less than 60 inches and discharged from the RPV through Reactor Water Cleanup' without considering the consequences with the Main Steam Isolation Valve High Radiation condition and Group I Isolation being present.

e.- Candidates were slow to silence annunciators or made no effort to silence them at all during major casualties.

The cooperation given to the examiners and the effort to ensure an atmosphere in the control room conducive to oral examinations was noted and appraelated.

Thp . licensee did not identify as proprietary any of the material provided l to or reviewed by the examiners, a

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w U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION I

. ~.s REGION 2 !'

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'il FACILITY: E. IL Hatch 1 & 2 REACTOR TYPE: BWR-GE4 DATE ADMINISTERED: 89/06/12 CANDIDATE INSTRUCTIONS TO CANDIDATE:

Use examination paper for the answers. Write answers on one side only, beneath the question. Points for each question are indicated in parentheses after the question number. The passing grade requires at least 70% in each category and a final grade of at i lesst 80%. Examination papers will be picked up four and one-half (4.1/2) hours after the examination starts.

% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE . TOTAL SCORE VALUE CATEGORY

.L% cc 24.27

- N f.0- S6-M _ 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS (27%)

+7.oo Ge727 1

-4 & -6sree 3. PLANT SYSTEMS 938%) AND PLANT-WIDE GENERIC RESPONSIBILITIES (10%)

71. ro 46 00 % Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.

Cathidate's Signature j I

_ _ . _ _ _ . _ _ _ _ _ _ _ _ ..___m _ _________m___ _ ._. _._ _ .____ , ____

15. Uhen you complete your examination, you ahall: ,

a. Assemble your examination as follows:

(1) Exam questions on top. f I

(2) Exam aids - figures, tables, etc. 4 (3) Answer pages including figures which are part of the answer. i

b. Turn in your copy of the exe.mination and all pages used to answer the examination questions.

c. Turn in all scrap paper and the t :ance of the paper that you did not use for answering the questions. l

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Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examine.tien is still in progress, your license may be denied or revoked.

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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:

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

2. Restroom trips are to be limited and only one candidate 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.

3. Use black ink or dark pencil only to facilitate legible reproductions.

4. Print your name in the blank provided on the cover sheet of the examination.

5. Fill in the date on the cover sheet of the examination (if necessary).

6. Use only the answer sheets provided for answers.

7-. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.

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8. Separate answer sheets and place finished answer sheets face down on your desk or table.

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

10. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.

11. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.

12. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK. NOTE: Partial credit will NOT be given on multiple choice questi'ns.

13. If parts of the examination are not clear as to intent, ask questions of the examiner only.

14. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.

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0 l teIM.rYx I ' QUESTION 3rdi (1.00)

I l A reduction in recirculation pump speed has occurred while operating I et power. SELECT the statement that provides the correct actions

. based on the given conditions.

a. Three LPRMs in one quadrant are oscillating at a 15% bandwidth so rods should be inserted in that quadrant to suppress the oscillations.

b. One APRM is oscillating at a 12% bandwidth with the others q oscillating at a 7% bandwidth so recirculation flow should be increased to suppress oscillations.

c. Two LPRMs in opposite quadrants are oscillating at a 15%

bandwidth so the reactor should be scrammed.

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d. All observed LPRMS are oscillating at a 7% bandwidth so J recirculation flow should be increased to suppress the oscillations.

QUESTION 2.02 (1.00) 1 A' loss of both recirculation pumps has occurred with the plant operating at 75% power and 100% rod line. Immediate operator action i required is: i

a. Place the mode switch in shutdown.

b. Drive control rods until power is below the 80% rod line.

c. Commence a normal reactor shutdown.

d. Immediately restart a recirculation pump.

(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)

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. QUESTION 2,03' (1.00)'

.On a loss of Instrument and Service Air, a loss of condenser vacuum will occur. This.is caused by:

a. a loss of circulating water.

b. increasing hotwell level.

c. a closure of the SJAE suction valve (F004).

d. a closure of the SJAE steam supply valve (F008).

QUESTION 2.04. (1.00).

'A loss;of 120/208 V Distribution Cabinet 2A - Instrument Bus 2R25-S064 hos occurred. SELECT the correct statement concerning how to monitor plant parameters,

a. Recorders should be monitored because of the ability to compare independent' indications,

b. Meters shouldLbe used due to some meters being powered from DC sources.

c. Meters should be used because recorders will fail as is on a

' loss-of power.

d. Meters should be used because they will be easier to read during ,

the casualty.

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5,." 7 4f27%) "

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l 6C p- , QUESTION 2;05' (1'.00).

L &The. reactor;is: operating at'70% power and 58% flow. .Theloperator 1 observes a lors:of.-feedwater r heating. . Reactor ~ power:is;to:be Immediately reduced to:

n a..b'elow thei80%3 >d line by driving control: rods.

/b.?a7 thermal power of 56%:by reducing recirculation flow.

.c . a. thermal power of 50% by reducing recirculation flow. q

-d..-the minimum ~ thermal power capable without' reducing core flow lbelow 45%.

' ATTACHMENT 1 FROM'34GO'-OPS-005-2S IS PROVIDED.

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.: QUESTION 2.06 (l'.00)>

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LA'manualiscram has been insertedion'both channels and the rods have- 1 ifeiled to' insert. All blue scram 1ights are illuminated. ; IDENTIFY l "the cause for the' failure.to scram.

a.- HydraulicLlock in the scram discharge volume. l

'l g b. Failure of the. scram' valves to open.

c..BlockageLin the scram air header. '

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d. Failure of one.RPS scram trip system. 1 l

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-QUESTION 2.07 (1.00)

IDENTIFY the correct statement concerning control of systems from the remote shutdown panel (s).

a. The interlock preventing simultaneous opening of the RHR

- suction valves for shutdown cooling and from the torus are

( bypassed when operated from a remote shutdown panel.

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b. The low low set function is not operable when SRVs are transferred to a remote shutdown panel.

, c. Conte'nment isolation functions are ':perable for systems l

operated from a remote shutdown panel.

d. The only function of the Unit 2 RCIC trip throttle valve at a remote shutdown panel is the reset function.

QUESTION 2.08 (1.00)

SELECT the correct statement concerning an unrecoverable loss of RBCCW due to a leak in the system while operating in mode 1.

a. The standby RBCCW pump will not start.

b. The reactor should be scrammed within 2-3 minutes,

c. Entry into EOP flow charts is not required because they refer to the Loss of'RBCCW abnormal procedure to correct the RBCCW conditions,

d. If a scram is required then the Abnormal Procedure for Loss of RBCCW should be exited and the the EOP flowcharts entered.

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fQUESTION ~ 2.091 I(l.00); 1 w

All; service' air.. compressors have tripped. -!

Pressure:in the' entire air-

,;ovstem.isL83# - WHICH ONE of the,following' actions should have'

/ occurred.

ai Standby. Instrument Air. dryer starting, y b.-Startup flow' control valve locked,in present position.

\*; 'c.; Reactor Building Instrument: Nitrogen to Non-Interruptable

. service isolation valve opening, k _

d.' Service Air' Shutoff valve: closing.

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Q'UESTION prb,. .(1.00)-

TDhring an ATFS, severalLactions are.taken to mitigate heat addition to

/the-suppression pool. SELECT the correct statement concerning

implementation of these:aetions.

a'. Boron inject, ion istcommenced if temperature approaches 110 F to

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y ' ensure that the Technical Specifications limits are not exceeded.

.b.-Ifjpower is at 100% then immediate tripping of the recirculation 1

pumps from their present~ speed:is necessary to reduce heat added-to theJsuppression pool, c.LIf,-due=to a failure of-the SLC system, injection of boron is delayed,Lthen lowering of level should be delayed until boron is injected.

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d. Level reduction is required if power is > 3%, suppression pool-

= temperature is > 110 F, and an'SRV is'open. Level'should be

- reduced until'either level ~is at top of active fuel (TAF) or.all

_ three conditions are corrected.

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. QUESTION 2.11 (1.00)

SELECT'the correct statement concerning use of the Heat Capacity Temperature Limit and Heat Capacity Level Limit.

a. The Heat Capacity Level Limit can be complied with for all suppression pool levels by reducing reactor pressure.

b. The Heat Capacity Level Limit is less restrictive than the Heat Capacity Temperature Limit when suppression' pool level is less than 146 inches.

c. The Heat' Capacity Temperature Limit is allowed to be increased at suppression pool levels higher than 146 inches.

d. The Heat Capacity Level Limit requires that the emergency depressurization be accomplished before the downcommer vents are uncovered.

-QUESTION 2.12 (1.00)

Refer to figure 2.

SELECT the~ correct statement. '

The general caution in the Emergency Operating Procedure lists a drywell temperature and level indication for each RPV level instrument. -If drywell temperature is greater than the value stated, then level' indication is not valid below the value stated for that instrument. The level is not valid because:

a. at low reactor pressures the variable leg will flash at j temperatures greater than the stated drywell temperature. '

b. at high drywell temperatures variable leg density will decrease causing invalid readings. -j i'

c. at low indicated levels reference leg density causes on scale indications with level below the instruments monitoring range.

d. at low drywell temperatures the reference leg will cause erroneously high indicated levels, a

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. QUESTION 2.13' (1.00)

The. emergency depressurization procedure directs the operator to.the citernate.depressuriaation procedure if suppression pool level is balow 58".

IDENTIFY-the reason.for this'extion.

a. The bottom.of the downcomers is at 58".

'b. The safet'y relief valves discharge at 58'.

c. At.58' suppression pool level has insufficient heat capacity for depressurization.

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d. At 58" suppression pool level is below the suppression pool

. temperature detectors.

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QUESTION 2.14- (1.dO)

SELECT the statement that identifies the most significant contributor to reducing power when water! level is lowered during a failure to f s

, ceram event. !

a. Lowering. level-below the moisture separator removes the flowpath thereby minimising flow through the core. _j b.. Lowering. level ~ reduces the pressure in the core by reducing the head of water above the core.

c. Lowering level reduces the differential pressure between outside the shroud and inside the core.

'd. Lowering level reduces power by increasing the subcooling of the j water entering the core. )

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iQUESTION 2.15' (2.50)

For each power source in Column I below (a-f) MATCH the required

'immediate action ~ listed in Column II (1-6)-that is required to be take

.if that power source is lost. More than-one immediate action may be required for a' power source.

Column I. Column-II

c. 125/250 V DC Switchgear 2A. 1. Transfer 4160 V buses 2A and 2B.

2R22-S016 to Startup Supply.

b. 125/250 V.DC Switchgear 2B 2. Manually trip Reactor Reciro 2R22-S017 Pump B.

c .' 125 V DC Cabinet 20, 3. Secure radwaste discharge.

2R25-S003

d. 24/48V DC Cabinent 2B, 4. Manually open Main Generator 2R25-S016 output breakers.

5. None

-QUESTION 2.16 (1.00)

On.a loss of 125/250 V DC Switchgear 2A, 2R25-S001, the operator is to open the main generator output' breakers.

a. STATE the maximum time limit allowed for opening the output breakers.

b. STATE the reason that the output breakers must be opened.

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46 "T27%)

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' QUESTION 2.17 (2.50) i Given the following general EOP conditions, STATE the correct' flow path y number) to follow, which is designed to address the following conditions, in accordance with 31EO-EOP-001-25, " Emergency Operating Procedure Inside Control Room Unit 2."

a '. High radiation, loss of coolant, loss of control of primary containment integrity (under degraded conditions).

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b. High stears lir.e indiation, loss of vital power, failure of vital I equipment.

c. Malfunction of Reactivity Control System without AC power.

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l Malfunction of Reactivity Control System with AC power.

e. Reactor transients or failure of vital equipment without degraded i conditions.

QUESTION 2.18 (1.00) i The plant is in shutdown cooling with the reactor head still in place when a loss of shutdcan cooling occurs due to a loss of RHR Service J Water. Significant decay heat exists.

If reactor pressure increases to greater than (1) _ then (2) valve (s) will close. (Valve name(s) or number (s) are acceptable.)

QUESTION 2.19 (2.00) i Concerning the Drywell Spray Initiation Limit:

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a. DESCRIBE the negative effects, including the cause, of spraying the i drywell when drywell parameters are in the unsafe region of the Drywell Spray Initiation Limit. (1.0) l j

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b. STATE the parameters monitored by the Drywell Spray Initiation Limit. (1,0) 1 i

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2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS

(27%) Paga 13 QUESTION 2.20 f.2.00)

LIST FOUR pressure 1 control methods used by SOFIEFlowchart 3.

QUESTION 2'.21 (0.50)

For steam cooling the operator isl directed to open one safety relief valve when reactor level decreases to 1/3 core ~ height. This action will establish cooling to maintain peak clad temperature less than QUESTION 2.22 (2.00)

During performance of the Secondary Containment Temperature Control 125 proceduce'the operator is directed to Emergency Depressurize if a primary system is discharging into an area and area temperature / differential temperatures exceed the Maximum Safe Operating. temperature level in more than one area.

STATE TWO purposes for performing this action.

(***** END OF CATEGORY 2 *****) ;

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.p ] MSPONSIBILITIES (100 n :

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.. ,- . i 3.01o (1.00)

QUESTION

.Tochnical Specifications 1regarding Recirculation System Jet Pump OPERABILITY have some'very restrictive Limiting Conditions for

. Operation (LCO) if a jet pump is;found to be INOPERABLE. From the Kj . choices below, SELECT the concern'regarding continued plant operation ;

i with an inoperable *(or failed) jet pump. ;

a .' Invalid APRM Flow Biased SCRAM setpoints due to the change in -l flow through a failed jet pump '

b. Increased blowdown area during a' Lost of Coolant Accident (LOCA)

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c. Unbalanced. neutron. flux across the core due to' flow variations

d. Physical' core damage from a piece of a damaged jet pump

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QUESTION 3.02 -(1.00) i

The operating procedure for the High Pressure Coolant Injection (HPCI)

-system cautions against prolonged operation with turbine speed less than

'2000 RPM. SELECT the reason why this is of concern.

tr, At low turbine speeds the potential exists for exhaust check valve

. chatter and reduced oil flow / pressure to the turbine sovernor and bearings,

b. At low turbine speeds the Booster Pump may not provide adequate

. net positive suction' head to preve.it cavitation of the main pump.

c. The rate of steam flow through the HPCI turbine may not be enough to prevent it from overheating.

d.'At low turbine speeds, cooling water flow from the Booster Pump to the. lube oil system heat exchanger is inadequate.

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' RESPONSIBILITIES 710ki QUESTION 3.03 (1.00)

When the Standby Liquid Control (SBLC) system is initiated, WHICH ONE of l- the following occurs:

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a. The selected pump starts and only its associated (divisional) squib valve fires.to provide a flow path to the reactor,

b. The SBLC storage tank and pump suction line heat tracing heaters 4 cycle on to ensure all the sodium pentaborate remains in 1 solution for injectica into the reactor.

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c. Both SBLC pumps start and both squib valves fire to provide two I separate flow paths of sodium pentaborate into the reactor. ;

d.'The Reactor Water. Cleanup (RWCU) system outboard containment isolation valve (G31-F004) closes to isolate the system from the reactor.

QUESTION 3.04 (1.00)

' SELECT the set of trip channel conditions (a - b) from the Reactor Protection System shown in the diagram below that will result in a HALF SCRAM. (assume a "one-out-of-two taken twice" logic) l l TRIP SYSTEM TRIP SYSTEM A B l l l l l l l l Al A2 B1 B2 TRIP CHANNELS TRIP CHANNELS

a. Al and A2 tripped

b. A1 and B2 tripped l c. A2 and B1 and B2 tripped l

l d. A2 and B1 tripped l

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CATEGORY 3 CONTINUED ON NEXT PAGE l^

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y e n EtisPONSIBILITIES (1041

> QUESTION 3.05 (l'.00) t Ragarding the Reactor CoreLIsolation Cooling-(RCIC) system, which of the F following is NOT a function of the Suppression Pool?

g .a. Receives the discharge of the Vacuum Pump

b. Acts as a heat sink for the RCIC Turbine exhaust

c. Receives drainage from'the RCIC Turbine steam line and exhaust line drain pots

d. Provides backup source of water for the RCIC Pump

' QUESTION 3.06 (1.00)

UNIT-2 has just experienced an initiation of the Automatic Depressurisaton System (ADS). Plant conditions are as follows:

Drywell pressure: 3.2 psig !

Reactor water level: -147. inches All RHR pumps: running 13 minute timer: timed out 120 second timer: timed out i 7 ADS SRVs: open ,

Main Steam pressure: 150 psig and lowering )

WHICH of the following will cause the ADS SRVs to close? (Consider each cnswer' separately)-

a. The RHR/LPCI mode raises reactor water level to -20 inches.

b. Reactor pressure reaches 40 psig.

c. Drywell pressure decreases to 1.0 psig and the High Drywell Pressure Seal-in push button is reset by the operator,

d. 3 of the 4 RHR pumps are secured. 1

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' RESPONSIBILITIES T10W QUESTION 3.07 (1.00)

The OUTBOARD Main Steam Isolation Valve Leakage Control System (MSIV-LCS) MAY be manually initiated when WHICH of the following

.SFECIFIC sets of plant conditions exist:

a. Reactor pressure is less than 35 psig, all of the MSIVs are closed and Main Steam Line pressure between the Outboard MSIVs and the Main Turbine Stop Valves is less than 35 psig.

b. Reactor pressure has been less than 35 psig for a minimum of 10 minutes, all of the MSIVs are closed and Main Steam Line pressure between the Inboard MSIVs and the Outboard MSIVs is less than 35 psig.

c. At least-10 minutes have elapsed since the Loss of Coolant Accident (LOCA), the Outboard MSIVs are closed and pressure between the Outboard MSIVs and the Main Turbine Stop Valves and Bypass Valves is less than 35 psig,

d. At least 10 minutes have elapsed since the Loss of Coolant Accident (LOCA), all of the MSIVs are closed and pressure between the Inboard MSIVs and Outboard MSIVs has bled down to 0 psig.

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RESPONSIBILlhgS (10%)

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l QUESTION. 3.08 (1.00) 1 UNIT.2 is operating under the following conditions: 95% power with the H "B" Pressure Regulator in the Electro-Hydraulic Control (EHC) system out i of service. The "A" Pressure Regulator begins to SLOWLY increase and ,

eventually fails high. Assuming no operator action, SELECT the correct statement below. (See attached diagram)

L a, The Turbine Control Valves and Bypass valves will start to close

! Lto raise reactor pressure to meet the higher demanded pressure from the pressure regulator,

b. The Turbine Control Valves and Bypass valves will remain in the same position under control of the Maximum Combined Flow Limiter.

c. The Turbine Control Valves and Bypass valves will begin to open to reduce reactor pressure to compensate for the higher signal !

from the Pressure Regulator. The Load Limiter will halt the valve opening after a decrease of approximately 80 psig.

d. The Turbine Control Valves and Bypass valvos will open to reduce reactor pressure to compensate for the' higher signal from the Pressure Regulator. Pressure will decrease to the point of Main Steam Isolation Valve (HSIV) closure (825 psis) and subsequent Reactor SCRAM.

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GREsPONsinILITIES T10 n l

l l QUESTION 3.09 (1.00) l i !

l The Main Condenser Mechanical Vacuum Pump is used to draw the initial l vecuum up to a maximum of 5% power. SELECT the basis for this i power limit.

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a. The Mechanical Vacuum Pump does not have the capacity to draw sufficient vacuum to clear the Main Turbine and Reactor Feed i Pump low vacuum trips.

b. At' powers above 5%, the amount of steam in the Main Condenser l

-shortens the life of the seals in the Mechanical Vacuum Pump.

c. The Mechanical Vacuum Pumps are designed to draw a vacuum up to a Main Steam Pressure of 600 psig which corresponds to 5% power on a cold startup.

l l d. At powers above 5%, the amount of radioactivity being removed from the Main Condenser cannot be released to the environment and Hydrogen levels in the Condenser at this point could cause explosions from the heat in the Mechanical Vacuum Pump l

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[ETP W I N TIE U 13%)

QUESTION' 3.10 (1.00)

SELECT the FINAL plant conditions after the loss of the "A" Reactor Feed Pump from 100% power. Assume no operator action and all systems work as d3 signed.

-a. Reactor Power 50-55%

Reactor Water Level at normal level ,

Recirculation Pumps at 44% speed "B" Reactor Feed Pump speed higher than before the trip of "A"

b. Reactor Power 60-55%

Reactor Water Level at normal level Recirculation Pumps at 22% speed "B" Reactor Feed Pump speed the same as before the trip of "A"

c. Reactor Power 60-65%

Reactor Water Level below normal level Recirculation Pumps at 22% speed "B" Reactor Feed Pump speed lower than before the trip of "A" )

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d. Reactor Power 60-65% 1 Reactor Water Level at normal level i Recirculation Pumps at 44% speed I "B" Reactor Feed Pump speed higher than before the trip of "A" l

)

I j / Oci Neck f

QUESTION 3/f1 (1.00)

/

SELECT the conditions that will cause an Emergency Diesel Generator to i tie on to it's respective bus. (assume the diesel generator is already l running after a valid start signal) i l

a. Less than 90% voltage for greater than 10 seconds.

b. Less than 90% voltage for greater than 1 second.

j c. Less than 60% voltage for greater than i second.

d. Less than 60% voltage for greater than 10 seconds.

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l _ QUESTION L3.12- (0:50)

LIST the electrical busses supplying power to the "A" and "B" Control

. Rod Drive (CRD) pumps.

QUESTION' 3.13 (1.00)

May are'the Control-Rod-Drive (CRD) pumps NOT necessary to be in

-operation to support the FULL insertion of all control rods on a reactor scram with the reactor at normal pressure.

' QUESTION ~ 3.14 ( 1. 00 ')

STATE'the-purpose of the Rod Worth Minimiser (RWM). j I

QUESTION _ 3.15- -(1.50) !

l Complete _the following statements regarding the Unit 2 Rod Worth

Minimizer (RWM).

LA. The RWM may be manually bypassed under these TWO (2) conditions:

(1) and (2) .

B. Permission to manually bypass the RWM must come from (3) l C. If the RWM is manually bypassed, compliance with required rod L'

patterns and initialing of the Data Package must be done by (4) or (5) .

D. The RWM is automatically bypassed when above the (6)

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L L ' QUESTION 3.16 (2.50)

I l Answer the following questions concerning the operation of the Low L Pressure Coolant Injection (LPCI) system:

i l A. Following a Loss of Coolant Accident (LOCA) signal, what is the automatic seguence of events for the Outboard Injection Valves (FO 17A/B)? (Include in your answer any applicable time limits for both UNITS, but do NOT include pressure setpoints) (1.0) l B. What plant conditions will auto start the RHR pumps in this mode?

(setpoints required) (0,5)

C. What is the pump start sequence if a concurrent Loss of Off-Site Power (LOSP) occurs with an injection signal? (1.0)

. QUESTION 3.17 (1.50)

UNIT 2 is operating with the following plant conditions:

- 100% Reactor Power

- Shifting Reactor Water Cleanup (RWCU) pumps

- Both RWCU filter /demins out of service

- The "B" RWCU pump has been out of service for 2 1/2 hours and its casing temperature is less than 130 degrees

- Preparations are being made to prewarm the "A' RWCU pump You have directed the operator in the "A" RWCU Pump Room to open the pump manual suction isolation valves. As the valves are opened, the operator in the room reports hearing noises indicating the pipes are filling with water and at the same time you have indication that an automatic RUCU isolation is occurring. The Group 5 PCIS valves (G31-F001 and G31-F004) have stroked shut. All other RWCU system parameters are normal.

A. Which RWCU system isolation has occurred?

B. What physical conditions in the RWCU system caused the isolation?

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Oco' iQUESTION 3.d81 (-1v50-)

E ' ', i JAirequired-full f' low'testlis2 currently in' progress on UNIT 2 Core Spray-ELoop B. -Just.priorLto the end'of the. test, a VALID Core Spray System-Eauto-initiation:signalDis received from High Drywell Pressure (2.2

F'ig) vindicating- 'a: Loss of Coolant Accident (LOCA) is occuring.

., a W.at'automaticiactions will-be expected to occur in the Core Spray i

. Syctem? . (valve names-.e- */or numbers may be--used) h j

~

,~ QUESTION - 3.19' 1(2.00):

'J l

From;a comparison of UNIT 1 and. UNIT 2' reactor SCRAM setpointsf. SELECT

PD:.stthel SCRAMS f rom the right-hand column Wements in the.left-hand column.

and matchcolumn (Right-hand them tochoices the appropriate may.be -t used vore,than'once) !

SCRAMS- ;

A. LSCRAM setpoints that .l. -Reactor pressure a are.different between i

-the: units.- 2. Turbine control valve fast closure.

E'q 3 .- Scram' discharge volume'high level 1B.E . SCRAM (S)'thatLare.NEVER 4. Low reactor water level bypassed.

5' . APRM High fluxE(flow biased)

6. APRM'High flux C. SCRAM (S).that are

~ AUTOMATICALLY bypassed 7. High Drywell pressure under specified plant ,

conditions. 8. Turbine Stop Valve Closure

]

9. SRM High Flux j t

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09;f # RESPONSIBILITIES Il0s F

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. QUESTION '3.20 .(1.00)

. UNIT-2 Technical Specifications requires removal of the. shorting links h: during core alterations-and'shazdown margin demonstrations. BRIEFLY describe which Reactor Protection System (RPS) TRIPS this will affect F cnd haw'it will affect them.

QUESTION 3.21 (1.20)

A Complete the following statements regarding the Intermediate Range Monitoring (IRM) System.

A. The IRM provides neutron monitoring over (1) decades of power as indicated by (2) ranges of indication.

B. IRM provides TWO (2) reactor SCRAMS, (3) and-(4) . (NO setpoints required)

C. The IRM detectors are REQUIRED to be fully withdrawn from the core

-as soon as (5)

.D. Reactor period indication is available on IRM by using the

-(.)

6 on the IRM recorders.

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QUESTION 3.22 (2.50) 4 A '. STATE FOUR (4) Reactor Core Isolation Cooling (RCIC) AUTOMATIC turbine trips including setpoints (D0 NOT include the RCIC system i isolation turbine trip) (2.0)

B. WHICH of the RCIC turbine trip (s), manual and automatic, is/are

. required to be reset locally? (0.5) 1 1

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. QUESTION- 3.23 '(1.50)

COMPLETE the following:

1. .The Automatic Depressurisation System (ADS) consists of (number) (a) of the 11 Main Steam System Safety Relief Valves (SRV).

'2. ADS is designed to protect the core during a (size) (b)

Loss.of' Coolant Accident (LOCA) where the (c) system fails to maintain reactor vessel water level.

3. The ADS SRV accumulators.are sized for at least (number)

(d) valve actuations.

QUESTION 3.24 (2.00)

' List the FIVE systems or indications that utilize the dp (flow) signal Lfrom the main steam line flow restrictors.

QUESTION 3.25 (1.00)

A. STATE the Technical Specifications setpoint for Main Steam Isolation Valve (MSIV) closure on Low Main Condenser Vacuum.

(0.25)

B. WHAT are the THREE (3) conditions that must exist to bypass the MSIV closure ~on Low Main Condenser vacuum. (0.75)

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'EE5PONSIBIL111ES (i0%.1 L QUESTION 3.26 (1.25) l -For the following plant conditions, STATE which Standby Gas Treatment Systems will be running (UNIT 1, UNIT 2, BOTH UNITS or NEITHER UNIT).

Do not consider any other system actuations that may result from these

' conditions.

A. UNIT'1 Reactor Water Level at -110 inches.

B, UNIT 2 Reactor Building exhaust ventilation radiation reading 40 mr/hr.

C.' UNIT 1 Reactor Building exhaust ventilation radiation reading 40 mr/hr.

D.-UNIT 1 Refuel' Floor exhaust' ventilation radiation reading 10 mr/hr_and UNIT 2 Refuel Floor exhaust ventilation radiation reading 45 mr/hr.

E. UNIT 2 Drywell Pressure reading 2.35 psig and UNIT 1 Reactor Water Level at -30 inches.

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~~i RESPONSIBILITIES ~ (10%)

l QUESTION 3.27 (1.80)

MATCH the following Off Gas Sy, tem components with the correct statements from the right-hand column. There may be more than one correct answer for each component and each answer from the right-hand column may be used more than once.

COMPONENTS A. Off Gas Preheater 1. Provides automatic protective l action signals to'various'Off Gas j System components B, Charcoal Adsorber 2. Provides time for Xenon and Krypton isotopes to l decay. l C. Off Gas Stack Isolation 3. Is an electric boiler on Unit 2 l Valve l

4. Reduces the levels of Hydrogen and ;

oxygen in the off-gas flow D. Off Gas Catalytic 5. Will shut on HI-HI-HI radiation j Recombiner levels from the Post Treatment Radiation Monitor j

6. Has a design 30 minute delay time E. Holdup Volume l

7. Will shut on High Hydrogen levels !

in the off-gas flow F. Post Treatment Radiation 8. Uses 250 psig steam on Unit 2 to j Monitor heat the off-gas flow ,

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9. Samples the off-gas flow going up the Main Stack G. Loop Seal Drain Valve

10. Removes Iodine isotopes from the off-gas flow. :

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L L ' QUESTION 3.28 (2.50)

L LMATCH the fire protection systems on the left with the applicable p ocatements from the right-hand column. (Statements in the right--hand l . column may be used more than once and aach system may have more r.han one

} ' correct answer)

.A . Deluge System 1. Clapper valve opened by smoke detector.

2. Uses closed sprinkler heads B. Wet Pipe System 3. Uses open sprinkler heads 4

4. Alarms locally C. Dry Pipe System 5. Alarms in the Main Control Room

6. Clapper valve opened by loss of air pressure l 7. Clapper valve opened by loss of water pressure l

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. QUESTION 3.29 (1.25)

I L MATCH the-following Reactor Pressure Vessel (RPV) Safety Limits and Dasign numbers with their appropriate values. (All pressures are steam 1 dome-prossure0 A. RPV heatup and cooldown limit (per honr) 1. 1100 psig i

B. RPV head tensioning minimum temperature 2. 90 degrees C. RPV design operating pressure 3. 1325 psig D. RPV High~ Pressure SCRAM 4. 1250 psig E. -RPV Pressure Safety Limit 5. 70 degrees

6. 100 degrees

7. 1054 psig

8. 1375 psig

9. 1120 psig i

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QUESTION 3.30 (1.00)

SELECT the statement which is NOT in conformance with 30AC-OPS-00-OS, L

Control of Equipment Clearances and Tags.

l a. When pulling fuses to deenergize power supplies, the DANGER tags L will be attached to blockout fuses if available.

b. When danges tagging an MOV as a clearance boundary, tagging the breaker and lonal operator will satisfy the tagging requirements.

c. It is permissible to repack a MOV with the valve on its backseat and a DANGER tag on its local operator, breaker, and control switch.

d. When danger tagging an MOV which requires independent verification it is permissible for the independent verifier to use the valve position indication lights to determine valve position.

QUESTION 3.31 (1.00)

SELECT the correct statement concerning symbols used on the EOP flow charts.

Refer to the symbols on figure 3.

a. The arrow within symbols A and C indicate the direction to be follo ed if the decision is answered yes,

b. Sy bol G is used to direct the operator to an End Path Manual.

c. If the answer to a decision in a symbol C changes then the operator is to return to the top of the charts.

d. Symbol D is used to identify steps that must be performed in sequence.

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QUESTION 3.32' (1.00)

IDENTIFI which one of the following is acceptable, in accordance with Control of Operator Aids (DI-OPS-05-1084N), for use as an Operator Aid.

a. Posting of.a pending change to an Emergency Operating Procedure.

Lb. Instructions while a procedure-is written and approved,

c. A caution concerning operation of a component.

'd . Instructions for emergency startup of a diesel.

. QUESTION 3.33 (1.00)

SELECT the correct statement concerning operating chart recorderc in the Main Control Room

a. One hour prior to shift change the off going . Shift operator shall mark each chart with the date, time, and his initials.

b. A chart with an incorrect scale may not be utilized on a Control Room recorder.

c. On-shift plant operators shall ensure that each completed chart is identified by the recorder number, marked with the date and time it was installed and removed, then forward it directly to Document Control for filing.

d. Used charts are to be given to the Shift Supervisor and periodically forwarded to Document control for filing.

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" 5ESiONSTBIL5 TIES T10%)

u.

QUESTION 3.34 (2.00)'

You are verifying 'a valve line-up. STATE how you' confirm position of each of the following:

a. -Closed valve,

b. Open valve.

c. Motor-opereted valve.

d. Locked throttle valve.

QUESTION- 3.35 (2.50)

'In accordance with 10CFR 20..

1D. . STATE the quarterly radiation exposure limit for each of the following:

1) Whole body- (0.5.)

2) Hands and forearms, feet and ankles (0.5)

3) Skin of the whole body (0.5)

b. The limit stated in part a for whole body dose may be exceeded if

-specific conditions are complied with. Answer the following concerning the maximum limits allowed.

1) STATE the maximum quarterly whole body limit. (0.5)

2) STATE two conditions that must be met for an individual to exceed the quarterly exposure limits stated in part a. (0.5) l l

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. c TiBESEQNSIBILITIES (10%i

- QUES': 10N - 3.36 (1.50)

Fill in the blan'ks.

n. Routine work in areas where temperatures are greater than must be done with the aid of supplemental cooling devices,

b. Work will not be done in areas where temperature is equal to or greater than ,

c. If in doubt, all heat disorders are to be treated as .

(***** END OF CATEGORY 3 *****)

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y ceiscS}

ANSWER 2<01 (1.00).

/

'c [1.0) or d

. REFERENCE E.I. Hatch, Reactor Power' Instabilities, 34AB-OPS-058-25, page 1.

1NRC Bulletin No. 88-07, supplement 1: Power Oscillations in Boiling Water Reactors.

2.5/3.3 3.8/3.7 295001K104 295001G010 ..(KA's)

ANSWER 2.02 (1.00) a-[1.0)

REFERENCE E.I. Hatch, Trip of One or Both Reactor Recirculation Pumps, 34AB-OPS-032-2S, pg 4 3.8/3.7.

295001G010 ..(KA's) t:Nchk ANSWER ,2< 06 (1.00) d [1.0)

REFERENCE E.I. Hatch, Loss of Instrument and Service Air System, 34AB-OPS-020-25, pg 2,.

2.9/2.9 .

295002K306 ..(KA's) l ANSWER 2.04 (1.00) i c [1.0] I 1

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REFERENCE-E;I. Hatch,~ Loss of Instrement Buses, 34AB-OPS-014-25, page 2.

H 4.2/4.3:

295003A202 ..(KA's)

ANSWER 2.05 (1.00) d'[1.0]

REFERENCE E.I. Hatch,' Loss of Feedwater Heating, 34AB-OPS-045-25, page 2.

4.0/3.93.6/3.8 295014G010 295014A102 ..(KA's)

ANSWER 2.06 (1.00)-

e. . [1. 0 ] '

REFERENCE E.I. Hatch,LSOFI Flowchart 1: Content and Use, LT-IH-20107-03, pg 16, EO.7.

3.8/3.9 295015K201 ..(KA's)

ANSWER 2.07' (1.00) b [1.0]

REFERENCE E.I. Hatch, Remote Shutdown Panel, LT-IH-05201-00, pg 12, 13, and 14.

EO 6.

4.1/4.1 4.0*-/4.1*

295016G006 295016K202 ..(KA's)

ANSWER 2.08 (1.00)

.b [1.0]

(

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_ .. p ,%-- gq-- - -

i p

p, l . REFERENCE p

E.I. Hatch . Loss of Reactor Building. Closed Cooling. Water,

34AB-OPS-011-25, pg 2.

'S.4/3.6 295018K202 ..(KA's) l' l

'. ANSWER 2.09- (1.00)

I c [1.0]

REFERENCE E.I. Hatch, Loss of Instrument and Service Air System, l

34AB-OPS-020-25, pg 2. Plant Air Systems, LT-IH-03501, EO-16.

3.2/3.3 3.3/3.2 3.3/3.1 295019K203 295019A104 295019A102 ..(KA's) i ANSWER- 2.10 (1.00)

a (1.0) be l *#f

d - 4 1

REFERENCE E.I. Hatch, SOFI. Flowchart 1: Content and Use, LT-IH-20107-03, pg .j 10, 20, 22, 23. EO 11, 15, 16. E.I. Hatch, Emergency Operating -

Procedure Variables and Curves, LT-IB-201113-00, pg 7, EO 1. I 3.7/4.1* . i 295026K304 ..(KA's) (

ANSWER 2.11 (1.00) .

d [1.0) .i I

REFERENCE E.I.' Hatch, Emergency Operating Procedure Variables and Curves, i LT-IH-20113-00 pg 11 and 12. EO 1 and 2. ;

3.8/4.1 3.5/3.7 3.5/3.8 3.4/3.8 i 295026K301 295026K206 295026K102 295026G007 ..(KA's)

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ANSWER- 2.12 (1.00) ci [1.0)

-REFERENCE

.E.I. Hatch, PSTG and SOFI Chart' General and Specific Cautions, LT-IH-20114-00,- Section IV.F 3.6/3.8 3.4/3.8 3.7/3.9 295028K203 295028G007 295028A203 ..(KA's)

ANSWER. 2.13 (1.00) b-[1.0]

REFERENCE ,

E.I.-Hatch, LT-IH-20114-00, pg 14. EO-2'.

3.5/3.8 295030K208 ..(KA's)

ANSWER 2.14 (1.00)

.c [1.0)

REFERENCE E.I. Hatch, SOFI Flowchart 2: Content and Use, LT-IH-20103-02, pg 20.

EO 18.

-4.1*/4.5* 4.0/4.2 4.1*/4.3*

295037K303 295037K209 295037K102 ..(KA's) 0 ANSWER 2.15 (2.561 a.f 4, f

b. 2

c. 1

f. 3

[ 2.59 [0. 5 each]

2 '. 0

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y=m _

m DREFERENCE E. I.LHatch,' Loss of DC Buses, 34 AB-OPS-013-2S.

3.2/3.4 295004G010 ..(KA's)

ANSWER 2.16- (1.00)

c. 30 seconds. [0,5].(accept 20-30 seccuds).

r b. Prevent motoring (reverse power) of the. generator. [0.5]

L REFERENCE lE.I. Hatch, Loss of DC Euses, 34AB-OPS-013-25, page 7.

-3.3/3.4 3.-2/3.4.3.4/3.6 295004K105- 295004G010 295004A103 ..(KA's)

ANSWER 2.17- (2.50) a.- Path'5

b. Path 4

c. Path:2 d.. Path.1

e. Path 3 (0.5. pts each)

-REFERENCE GPC: .31EO-EOP-001-2S, LP # LT-IH-20101-00, EO #7 1 3.8/3.4 295006G012 ..(KA's)

ANSWER '2.18 (1.00)

a. 145 psig [0.5] (accept 130 to 145)

b. RHR Suction Cooling Valves, 2E11-F008 and F009. [0.5]

' REFERENCE

'E.I. Hatch, Loss of Shutdown Cooling, 34AB-OPS-044-2S, pg 1.

RER System, LT-IH-00701, EO 7.D 3.6/3.6 295021K203 ..(KA's) h 1

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L-d ANSWER' 2.19 (2.00)

S w'Drywellmaycollapse[0.M(orotherwisefail)[A4tduetonegative .

~ pressure.[0.5] (resulting Irom rapid condensation).

b.,:Drywell temperature [0.5)

'Drywell pressure [0.5]

REFERENCE E.I. Hatch, Emergency Operating Procedure Variables and Curves,

.LT-IH-20113-00, page 10. EO 2.

,7,4.2*/4.4*

~ ~

3.9/4.0 3.6/3.9 295024A201 295024A202' 295024G007 ..(KA's)

ANSWER 2.20 (2.00)

1. Lo-Lo Set

2. Main Turbine Bypass Valves

3. SRV's

4. Alternate' Pressure Control Systems. (Accept any one alternate pressure' control system)

.[2.0) [4 at-0.5 each)

REFERENCE E.I. Hatch, SOFI Flowchart 3: Content and Use, LT-IH-20104-03, pg 14.

FO-10.

3.9*/4.5* 4.4*/4.4* 3.8/3.8 295025G012 295025A103 295025A102 ..(KA's)

ANSWER 2.21 (0.50) 2200 F. [0,5)

REFERENCE E.I. Hatch, EOP End Path Manual Content and Bases, LT-IH-20108-4, pg

47. EO 3..

4.0/4.3*

295031K304 ..(KA's)

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l3 $;- 'i27%)

bl ANSWER 2.22 -(2.00) 1.

1. Terminate energy addition to the secondary containment [1.'0]

2. Place the RPV in'a low energy state [1.0).(also accept discharge energy to suppression pool for full credit.)

' REFERENCE

-E.I. Hatch, EOP End Fath' Manual Content and Bases, LT-IH-20108-04, pg

74. EO-4.

.3.5/3.8 295032K301 ..(KA's)

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ANSWER' 3.01 (1.00) b (1.0)

REFERENCE

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E.I. HATCH LESSON PLAN,' REACTOR RECIRCULATION SYSTEM, LT-IH-00401-00,- '

JEO #25, PAGE 14

-3.5/3.7 3.4/3.9 .

202001K601 202001A201 ..(KA's)

. ANSWER 3.02 (1.00) a (1.0)

REFERENCE E.I. HATCH LESSON PLAN, HIGE PRESSURE COOLANT INJECTION, LT-IH-00501-02,

'EO #15 d.3, PAGE 45 3.8/3.7 3.3/3.3 206000A401 206000K505 ..(KA's)

ANSWER 3.03 (1.00) d ~ c c br (1.0)

REFERENCE E.I. HATCH LESSON PLAN, STANDBY LIQUID CONTROL SYSTEM, LT-IH-01101-00, EO #11 & 13, PAGES 15, 17 & 18 3.8*/3.9* 4.2*/4.2* 4.0*/4.1*

211000K407 211000K408 211000A306 ..(KA's)

(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

~ ~ ~

~ '

'RE5PO5 fEfSITTE U 15 I

(;

-ANSWER 3.04 (1.00) e (1.0)

REFERENCE E.I. HATCH-LESSON PLAN, REACTOR PROTECTIVE SYSTEM, LT-IH-01001-00, EO #6, PAGE 22 & 23 3.3/3.4 3.7/3.8 212000K502 212000K305 ..(KA's)

EANSWER 3.05 (1.00) o (1.0)

REFERENCE E.I. HATCH LESSON PLAN, REACTOR CORE ISOLATION COOLING SYSTEM, LT-IH-03901-00, EO #4, PAGE 7 3.6/3.6 217000K103 ..(KA's)

ANSWER 3.06 (1.00) b (1.0)

REFERENCE E.I. HATCH LESSON PLAN, AUTOMATIC DEPRESSURIZATION SYSTEM, LT-IH-03801-03, EO #12, PAGE 30 4.2/4.3* 4.2*/4.2*

218000A206 218000A402 ..(KA's)

L

{- (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

l I

(^ . RESPONSIBILITIES (10%)

b I

L- ANSWER 3.07 (1.00).

l-

.c. (1.0)-

/

REFERENCE:

l E.I.. HATCH LESSON PLAN, MSIVLLEAKAGE CONTROL SYSTEM, LT-IH-04901-00,

' EO #6 & 10b, PAGES 16, 21 & 22 3.1/3.3 3.1/3.1 239003K406 239003A101 ..(KA's)

ANSWER 3.08. (1.00) -

L door 'a (1.0) ,

REFERENCE E.I. HATCH LESSON PLAN, ELECTRO-HYDRAULIC CONTROL SYSTEM, LT-IH-01901-00, EO #17f, PAGE 65 3.1/3.2 3.5/3.7 3.8/3.9 245000K409 245000K602 245000A207 ..(KA's) ,

l l

t ANSWER 3.09 (1.00) d (1.0)

(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

i

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7" RESPONSIBILITIES 110%1 REFERENCE E.I. HATCH LESSON PLAN, MAIN CONDENSER, LT-IH-02501-00, EO-#15, PAGE 20

-2.8/2.8 3.'1/2.9.3.4/3.4 256000K409 256000G010 256000G007 ..(KA's)

ANSWER 3.10 (1.00) d (1.0)

. REFERENCE E.I. HATCH LESSON PLAN, CONDENSATE AND FEEDWATER, LT-IH-00201-00, EO #11, PAGES 17 & 18 3.9/3.9 3.8/3.9 3.7/3.7 3.4/3.4

.259001K301 259001K312 259001A201 259001A310 ..(KA's)

_.,- O?Abec

-ANSWER 32f_1

- (1.00) c (1.0)

REFERENCE E.I. RATCH LESSON PLAN, DIESEL GENERATORS, LT-IH-02801-00, EO #4, PAGE 17 3.8/3.7 264000K408 ..(KA's)

(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

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r - ;.ncM5MMSiBILITIES T MRC ~ ~ ~

D #8 .

' ,i-g- e

)

1 ANSWER ~ 3.12 . ( O' . 50 ) ;

, ;" A'I f CRD Pump , :- 4160 VAC. Bus 2E (Emergency. Bus 2E)

"B"lCRD> Pump .-':4160'VAC-Bus 2F (Emergency Bus 2F)

.(0.25 each)

,~

l

.j, i REFERENCE :

y E. I.:: HATCH LESSON . PLAN, CONTROL ROD DRIVE' HYDRAULICS, . LT-IH-00101-00, EO #13, PAGE 41 2.9/3.1 201001K201L ..(KA's)

' ANSWER.- 3'.13'- (1.00)

-LThe~ c-== :: t=uletsi; and-Meactor pressure functiondas the source of energy to insert control rods'on a SCRAM independent-of CRD pump

.operationc .(alternative wording acceptable) (1,0)

- REFERENCE

ESI'.cHATCH LESSON. PLAN, CONTROL ROD DRIVE HYDRAULICS, LT-IH-00101-00,

.~EO #5o,LPAGES'30 & 31-

-3.8/3.'8 3.1/3.2-L201001K405 201001K303 ..(KA's)

. ANSWER ~ 3.14 (1.00)

To limit control rod worth (0.75) so that the fuel enthalpy limit of

280 cal /gm will not-be exceeded during a rod drop accident (0.25).

(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

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

~

P RESPONSIBILITIES i10%F I.

REFERENCE E.I. HATCH LESSON PLAN, ROD WORTH MINIMIZER, LT-IH-05403-00, EO #1, PAGE 6 3.3/3.7.3.4/3.4 201006K501 201006G004 ..(KA's)

ANSWER 3.15 (1.50)

A. (1) Inoperable (2) for testing B. (3) Operations Supervisor C. (4) A second licensed operator (5) A qualified member of the Technical Staff D. (6) Low Power Setpoint (LPSP) (30% power) (0.25 each)

REFERENCE E.I. HATCH LESSON PLAN, ROD WORTH MINIMIZER, LT-IH-05403-00, EO #8, PAGE 30

.3.2/3.4 3.4/3.5 201006A401 201006K404 ..(KA's)

(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

dE*~lglsPONsIBILITIEST10%F l

p l . -

ANSWER 3.16 (2.50)

A. Valves are interlocked open (0,5)

Ten minute on UNIT 1 (0.25)

Five minutes.on UNIT 2 (0.25)

L B. High Drywell. Pressure of greater than or equal to 1.92 psig; or Reactor Water Level of less than or equal to -113 inches.

(0.25 each)

C. (Upon reenergization of.the respective pump buses.) the pumps start in.the following sequence:

RHR "C" -

starts immediately RHR "A", "B" and "D" - ' start after a 10 second delay ALso ACCCAr " Tim 6 DCuY ER DrC6fL EG CrV bDDED (0.25 each)

Tb AGoog Mants/2s (T::-lZ Grd gg4 g ' - / 2 S fc REFERENCE g pg. ,'/ks *e

, +.D ,, - 2 2. Sn E.I.. HATCH LESSON PLAN, RESIDUAL HEAT REMOVAL SYSTEM, LT-IH-00701-01, EO #7a & 7b, PAGES 28, 29 & 32 3.5*/3.5*-3.7/3.9 .3.6/3.7 4.2*/4.2 203000K201 203000K407 203000K601 203000K401 ..(KA's)

ANSWER 3.17 (1.50)

A. RWCU system High Differential flow (1.0) ;

B. Opening the valves on a cool, depressurized (and partially voided) system. (0.50)

A(.so Accrer- Uoinco OAL faaratty Eccco Sysrcm i

)

i 1

i

(***** CATEGORY 3 CONTINUED ON NEXT PAGE *'****) j l

"l RESPONSIBILITIES (10'4)

REFERENCE E.I. HATCH LER 89-001-00 DATED 3/6/89, INADEQUATE PROCEDURE RESULTS IN GROUP 5 ISOLATION OF PRIMARY CONTAINMENT ISOLATION SYSTEM E.I. HATCH LESSON PLAN, REACTOR WATER CLEANUP SYSTEM, LT-IH-00301-01, EO # 7e, PAGE 17 3 5/3.6-3.6/3.6' o

204000K404 204000A303 ..(KA's)

~ ANSWER 3,18 L1..-50

- (l, CD)7

- Full' Flow'. Test valve to Torus (F015B) closes (0.50)

- Min-Flow-valve (F03134-w111 open-4when-syertem-f-low-drope--tro-400-gpm.+ - (-0225)

- Inboard Injection valve (F005B) opens when Reactor pressure is less than 500 psig. (0.50)

-Min Flow-velve (F031E) will-clese ( uhen-system-f-low - ge e c-above-450 -gpmr)- (0. 25 )-

. REFERENCE-E.I. HATCH LESSON PLAN, CORE SPRAY SYSTEM, LT-IH-00801-00, EO #11 & 12, PAGES~'31, 32 & 36 2.8/3.0 3.8/4.0 3.3/3.2 4

209001K407 209001K408 209001A108 ..(KA's) i i

k I

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(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

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_ _ - _ _ - _ ___-_-_ _ _ ___ -_ __- __ ____- - . _ __ _ _ _ _ _ . -_ __ _ __ J

~ ~ ~ ~ ~

l

BESPdNAlBILITIEB T (10W h

l l-l: s p

ANSWER 3.19 (2.00)-

JA. 3, 5,'6 w B. - 1; 4 , 7--

l C. 2, 8 (0.25 each)

REFERENCE s

E I. HATCH LESSON PLAN, REACTOR PROTECTIVE SYSTEM, LT-IH-01001-00, EO #10, PAGES 18 & 19 3.8/4.5* 3.9/4.1

>212000G005'. 212000K412 ..(KA's)

ANSWER - 3,20 (1.00)

,. :@a Nu c 1 anr_.-Inm uomi, Letrion--triHogie-changes-tro--non-coino Ldence teipc--(0-5), -any ena of 1A NT + vira uill cause a SCRA".

. (Or64 W WLCLEAR.M01Est rar>MTA-@b ( D.5) loc.tc. %<,cs To Ort; e r of

.y EM Duti (NM -LDtALCLDEN CC)(0.6)

E.I. HATCH' LESSON PLAN, REACTOR PROTECTIVE SYSTEM, LT-IH-01001-00, PAGE 28 3.3/3.5

'212000K411 ..(KA's)

(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

~ ~

4 ~ f BEEPONSIBILITIEs il6%F y

-ANSWER 3.21 -(1.20)

A. (1) 5-l (2) 10

.B. (3) IRM Inop (4) IRM Hi-Hi C. (5) --All APRMe-arun scale--(-but-before-reachi-ng408-on-Range-10 of any IRMt Apgp_ Pwew4 mocc Su>cre To " 4.uu

D. (6) Slope of.the line (will accept " time to double")

(0.2 each)

REFERENCE E.I.. HATCH LESSON PLAN, INTERMEDIATE RANGE MONITORS, LT-IH-01202-00, EO #3, 4c, 11, 13a & 13d, PAGES 11, 15, 26, 27, 30 & 32 4.0/4.0 3.3/3.3 3.0/3.1-215003K402 215003A401 215003K503 ..(KA's)

ANSWER 3.22 (2.50)

-A. High turbine exhaust pressure 40 psig (36-44 psig)

Low RCIC pump suction pressure 10" Hg vacuum (9-11" Hg)

Electrical overspeed 4950 RPM or 110%

Mechanical overspeed 5625 RPM or 125%

(0.3 for trip, 0.2 for netpoint)

B. Mechanical overspeed Local manual (0.25 each)

(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

~ ~ ~ ~

~~

% p:.. N RESPONSIBILITIES E l10 C w%:c .'

(lL: ::b? ' lQ .

- t:

, ? ^ ; L,l,[_,

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A 3

-TREFERENCE

.t s:

4E.I.: HATCH LESSONz PLAN,~ REACTOR'. CORE ISOLATION COOLING SYSTEM,

.?LT-IH-03901-00..EO.#6c & Ba,'PAGES 12, 13-O: 53.8/3.7:

<217000A202e

..(KA's)

, a:

J, t ANSWERi ' .3 ~.- 2 3 -(1; 50 ) -

o-L f a.N 7h ,. . . e,3- .

' (c><33 J B b. small breakT Ad;%1r64mcowif &C4x" . (pasT JocHigh Pressure Coolant.. Injection (HPCI) (p731 l MT

.37f each

^

Td, ;2 _

(4 37R m.m nk l .

l l

REFERENCE -'-

":E.I.. HATCH: LESSON-' PLAN,' AUTOMATIC DEPRESSURIZATION SYSTEM,

LT-IH-03801'-03,.!EO #1,-;2,-13 & 15, PAGES 9, 10 & 25

3.5/3.6 3.9*/3.9*-

k'3 2180'00K404- 218000K106- ..(KA's)-

I i

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(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

L 2_-_ _1-__ _

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T- 9 RESPONSIBILITIES (10%i' 1 ANSWER 3.24~ .(2.00)

- Individual steam line flow indication

--Steam flow' input to the Feedwater Control System

- _ Steam flow input to the Rod Worth Minimizer (RWM)

- Steam flow input to the Group I Isolation Circuit

-Proce=s computer yocess coqAer (M

pM 4' e.(b.6 EAC9)

-REFERENCE E.I. HATCH LESSON PLAN, MAIN STEAM / LOW-LOW SET, LT-IH-01401-00, EO #5c, PAGE 22 2.8/2.8-3.1/3.2 239001K505 239001K406 ..(KA's)

ANSWER. 3.25' (1.00)

A. 7" Hg vacuum (0.25)

B. - Reactor Mode Switch not in "RUN"

- Turbine Stop Valves not full open (less than 90% open)

- MSIV Low Vacuum Trip Bypass Switches in " BYPASS" (0.25 each)

(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

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t6 RESPONSIBILITIES (10%)

REFERENCE

~

-E.I. HATCH LESSON PLAN, MAIN STEAM / LOW-LOW SET, LT-IH-01401-00,

.EO'#9a & 11a, PAGES 40 &43 3.3/3.4.3.8/3.9 239001K608 239001A208 ..(KA's)

ANSWER- 3.26 (1.25)

A. BOTH B. NEITHER~

C. BOTH D. -BOTH.

E. UNIT 2 (0.25 each)

REFERENCE

.E.I. HATCH LESSON PLAN. STANDBY GAS TREATMENT SYSTEM, LT-IH-03001-00, EO #7b, PAGES.24 &25 3.7/3.8 261000K401 ..(KA's) 1:

l' 1

(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

.'~ v g g g ,

- ' ~ ~ ~ - ^ ~

.; ANSWER 3.27 (1.80) l' A. 8' B. 2, 10 C.. 'S D. ~4

- E. '2, 6 F. ;.1 ( o 7 , ->ic %dI"5 "# b w ed

, G .- 5- (0.2 each)

REFERENCE

~

. E. I . HATCH LESSON PLAN, OFF-GAS SYSTEM, LT--IH-03101-00, EO #6 & 10, PAGES 13-18, 27 & 28 3.1/3.3 3.1/3.3 3.3/3.3 3.3/3.4 271000K102 271000K408 271000A301 2710000007 ..(KA's)

ANSWER 3.28 '( 2. 50 )

A. 3, 4, 5 B. 2, 4, 5, 7 C. 2,-4, 6 (0.25 each)

(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

~

[ TRESPONji1BILITIES (10M l

' REFERENCE E.I. HATCH LESSON PLAN PLANT FIRE PROTECTION SYSTEM, LT-IH-03601-00, EO #2, PAGES 16,.19 & 20 3.8/3.9 ~3.3/3.5 l

l -286000G007 286000K402. ..(KA's)

ANSWER 3.29 (1.25)

A. 6 B. 5 C. 4 D. 7 E.. 3. (0.25 each)

' REFERENCE E.I. BATCH LESSON PLAN, REACTOR VESSEL, LT-IH-04401-00, EO #9, PAGES 20-21.

3.5/3.9 3 9/4.4*

290002K501 290002K507 ..(KA's)

ANSWER 3.30 (1.00)

b. (1.0)

REFERENCE .

E.I. Hatch 30AC-OPS-001-OS pp. 7&8 3.9/4-5 294001K102 ..(KA's)

(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

"TV RESPONSIBILITIES (100 13

.: ANSWER 3.31 (1.00) c [1.0]

' REFERENCE E.I.' Hatch, EOP Flowchart and End Path Manual Design and Use, LT-IH-20102-03, page 10 to 14. EO 4.

'4.2*/4.2*

294001A102 ..(KA's)

ANSWER. 3.32 (1.00)'

m. [1.0]

REFERENCE.

E.I.-Hatch, Control of Operator Aids, DI-OPS-05-1084N, page 3.

4.2*/4.2*

294001A102 ..(KA's)

ANSWER 3.33 (1.00)

d. (1.0)

REFERENCE E.I. Hatch Plant Operations 30AC-OPS-003-OS pp. 16 & 17 Enabling Objective 29 of LT-IB-30004-01 3.4/3.6 294001A106 ..(KA's)

(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

tBEEPONSIBILITIES (10%)

ANSWER 3.34 (2.00)

a. Turn valve in closed direction (1/4 turn max t eat)

b. Turn valve in closed direction (1/4 turn max oft backseat)

c. Verify at remote (or local) position indication

d. Confirm locking device operability.

( 0.5 each )

REFERENCE EIH: 34-GO-SUV-001-OS, EO # 3.1.3.3 3.7/3.7 294001K101 ..(KA's)

ANSWER 2. 2 (2.50)

a. i' 1.25 rem [0.5]

2) 18.75 rem [0.5]

3) 7.6 rem [0.f'

b. 1) 3 rem [0.5)

2) Exposure must not exceed 5 (N-18) [0.25]

JL f orm ? . mas t 'vu complotad [0.25]

r esa 'cu s M +o,3 It e &a n REFERENCE E. I. Hatch, Radiation Exposure Limits, 60AC-HPX-001-OS, pg 4.

10CFR20 3.3/3.8 294001K103 ..(KA's)

ANSWER 3.36 (1.50)

a. 120 F. [0.5]

b. 160 F. [0.5]

c. Heat Stroke [0.6]

(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

I g

- 's.}BESPOFSIBILITIES(lq%_1 REFERENCE G:neral Employee Handbook, pg 4.

3.1/3.4 294001K108 ..(KA's)

(***** END'OF CATEGORY 3 *****)

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

1 1

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U.'S. NUCLEAR REGULATORY' COMMISSION ~

SENIOR REACTOR-OPERATOR LICENSE EXAMINATION-REGION 2-

, FACILITY: E. I. Hatch 1 & 2 h REACTOR TYPE: BWR-GE4 L. DATE ADMINISTERED: 89/06/12 CANDIDATE

INSTRUCTIONS'TO CANDIDATE:

Use examination paper for the answers. Write answers on one side only, beneath the question. Points-for each question are indicated in. parentheses after the question number. The passing' grade requires at least 70% in each category and a-final grade of at least 80%. Examination. papers will be picked up six (6) hours a'ict the examination starte.

% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY T o, st) - 4/.J6 20.00 44.01 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS (33%)

Ath 64 43.25 ;;.00 6. PLANT SYSTEMS-(30%) AND PLANT-WIDE GENERIC RESPONSIBILITIES (13%)

T5.75~

"FT ii_, % Totals Final Grade All work done on~this' examination is my own. I have neither given nor received aid. -

Candidate's Signature l

{'

u____ _ _ _ _ _ - - - _ _ _ - - __ _ 4

m _. .. _ _ _

'I

_c NRC' RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration'of this examination the'following rules apply:

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

2. Restroom trips cre to be limited and only'one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or. possibility af. cheating.

3. Use black ink or dark pencil only to facilitate legible reproductions.

4. Print your name in the blank provided on the cover sheet of the examination.

5. Fill in the date on the cover sheet of the examination (if necessary).

6. Use only the answer sheets provided for answers.

7. ' Print your name in the upper right-hand corner of the first page of each section of the answer sheet.

8. Separate answerisheets and place finished answer sheets face down on your desk or table.

9. -Use abbreviations only ir they are commonly used in facility literature.

10. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.

11. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.

12. Partial credit may be given. Therefore. ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK. NOTE: Partial credit will NOT be given on multiple choice questions.

13. If parts of the examination are not clear as to intent, ask questions of the examiner only.

14. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.

1 l

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A :. r 16.LWhen you complete your examination, you shall:

a. Assemble your examination as follows:

(1) Exam guestions on top.

(2) Exam aids - figures, tables, etc.

( 3 )' Answer pages including figures which are part of the answer.

b. Turn in your copy of the examination and all pages used to answer the' examination questions.

{

1

.c. Turn in all scrap paper and the balance of'the paper that you did not use for answering the questions.

d. Leave the' examination area, as defined by the examiner. If after leaving, you are found lit this area while the examination is still-in progress, your license may be denied or revoked.

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u__ _ _ _ - _ _- ._ _ _

I

' QUESTION 5.01 (1.00)

A loss of both recirculation pumps has occurred with the plant operating at 75% power and 100% rod line. Immediate operator action required is:

a. Place the mode switch in shutdown.

b. Drive control rods until power is below the 80% rod line.

c. Commence a normal reactor shutdown.

d. Immediately restart a recirculation pump.

Pe)Y QUESTION p/ff (1.00)

A reduction in recirculation pump speed has occurred while operating at power. SELECT the statement that provides the correct actions based on the given conditions.

a. Three LPRMs in one quadrant are oscillating at a 15% bandwidth so rods should be inserted in that quadrant to suppress the oscillations. ,

b. One APRM is oscillating at a 12% bandwidth with the others oscillating at a 7% bandwidth so recirculation flow should be increased to suppress oscillations,

c. Two LPRMs in opposite quadrants are oscillating at a 15%

bandwidth so the reactor should be scrammed.

d. All observed LPRMS are oscillating at a 7!; bandwidth so recirculation flow should be increased to suppress the oscillations.

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

t,

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Oc.1M QUESTION Sg (1.00)

On a loss of Instrument and' Service Air, a loss of condenser vacuum will occur. This is caused by:

a. la loss of circulating water.

b. increasing hotwell level,

c. a closure of the SJAE suction valve (F004)

d. a closure of the SJAE steam supply valve (F008)

' QUESTION 5.04 (1.00)

'A? loss of 120/208 V Distribution Cabinet 2A - Instrument Bus 2R25-S064

.has occurred. SELECT the correct statement conceraing how to monitor plant parameters.

a. Recorders should be monitored because of the ability to compare independent indications.

b. Meters should be used due to some metera being powered from DC sources,

c. Meters'should be used because recorders will fail as is on a loss of power.

d. Meters should be used because they will be easier to read during the casualty.

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(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

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t$3%)'

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l QUESTION 5.05 (1.00)

SELECT the statement that identifies the most significant contributor to reducing power when water level is lowered during a failure to scram event. l

a. Lowering level below the moisture separator removes the flowpath thereby minimizing flow through the core,

b. Lowering level reduces the pressure in the core by reducing the head of water above the core.

c. Lowering level reduces the differential pressure between outside the shroud and inside the core.

d. Lowering level reduces power by increasing the subcooling of the water entering the core.

QUESTION 5.06 (1.00)

The reactor is operating at 70% power and 58% flow. The operator observes a loss of feedwater heating. Reactor power is to be immediately reduced to:

a. below the 80% rod line by driving control rods.

b. a thermal power of 56% by reducing recirculation flow,

c. a thermal power of 50% by reducing recirculation flov.

d. the minimum thermal power capable without reducing core flow below 45%.

ATTACHMENT 1 FROM 34GO-OPS-005-2S IS PROVIDED.

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

L-

? QUESTION '5'.07, (1.00)

A manual. scram has been inserted on both channels and the rods have-ifdiled to insert. All blue scram lights are illuminated. IDENTIFY the cause for the failure to scram.

a. Hydraulic lock-in the ucram discharge volume.

b) Failure of the scram valves to open.

c. Blockage in the scram air header,

d. Failure.of-one RPS scram trip system.

QUESTION 5.08 (1.00)

IDENTIFY-the correct statement concerning control of systems from the rcmote shutdown panel (s),

a. The interlock preventing simultaneous opening of the RRR suction valves for shutdown cooling and from the torus are bypassed when-operated from a remote shutdown panel,

b. The low low set function is not operable when SRVs are transferred to a remote shutdown panel.

c. . Containment isolation functions are operable for systems operated from a remote shutdown panel,

d. The only function of the Unit 2 RCIC trip throttle valve at a remote shutdown panel is the reset function.

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

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QUESTION- -5.09 (1.00) 1 l' SELECT the correct-statement concerning an unrecoverable loss of 1: RBCCW due to a leak in the system while operating in mode 1.

a. The~ standby RBCCW pump will not start. l l b. The reactor should be scrammed within 2-3 minutes.

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! c.. Entry into EOP flow charts is not required because they refer to l- .the Loss of RBCCW abnormal procedure to correct the RBCCW conditions.

d. If a scram is required then the Abnormal Procedure for Loss of RBCCW should be exited and the the EOP flowcharts entered.

QUESTION 5.10 (1.00)

All service air compressors have tripped. Pressure in the entire air system is 83#. WHICH ONE of the following actions should have occurred.

a. Standby Instrument Air dryer starting.

b. Startup flow control valve locked in present position.

c. Reactor Building Instrument Nitrogen to Non-Interruptable service isolation valve opening.

d. Service Air Shutoff valve closing.

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9edo QUESTION (1.00)

During an ATWS, several actions are taken to mitigate heat addition to the suppression pool, SELECT the correct statement concerning implementation of these actions.

a. Boron injection is commenced if temperature approaches 110 F to ensure that the Technical Specifications limits are not exceeded.

b. If power is at 100% then immediate tripping of the recirculation pumps from their present speed is necessary reduce heat added to the suppression pool.

c. If, due to a failure of the SLC system, injection of boron is delayed, then lowering of level should be delayed until boron is injected,

d. Level reduction is required if power is > 3%, suppression pool temperature is > 110 F, and an SRV is open. Level should be reduced until either level is at top of active fuel (TAF) or all three conditions are corrected.

QUESTION 5.12 (1.00)

SELECT the correct statement concerning use of the Heat Capacity Temperature Limit and Heat Capacity Level Limit.

a. The Heat Capacity Level Limit can be complied with for all suppression pool levels by reducing reactor pressure.

b. The Heat Capacity Level Limit is less restrictive than the Heat Capacity Temperature Limit when suppression pool level is less than 146 inches.

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c. The Heat Capacity Temperature Limit is allowed to be increased )

at suppression pool levels higher than 146 inches, i l

d. The Heat Capacity Level Limit requiren that the emergency i

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depressurization be accomplished before the downcommer vents are l- uncovered.

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(1.00)

' QUESTION 5.13 Refer to. figure 2.

SELECT the correct statement.

The general caution in the Emergency Operating Procedure

. lists a drywell temperature and level indication'for each RPV level instrument. If drywell temperature is greater than the value stated, then level indication is not valid below the value stated for that

-instrument. The level is not valid because:

a. at low reactor pressures the variable leg will flash at temperatures greater than the stated drywell temperature.

b. at high drywell temperatures variable leg density will decrease causing invalid readings.

c. at low indicated levels reference leg density causes on scale indications with level below the instruments monitoring range.

d. at low drywell temperatures the reference leg will cause erroneously high indicated levels.

QUESTION 5.14 (1.00)

The emergency depressurization procedure directs the operator to the alternate depressurization procedure if suppression pool level is below 58".

IDENTIFY the reason for this action.

a. The bottom of the downcomers is at 58".

b. The safety relief valves discharge at 58".

c. At 58" suppression pool level has insufficient heat capacity for depressurization,

d. At 58" level is below the suppression pool temperature detectors.

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. QUESTION 5.15 (1.00) ,

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Under'which ofuthe following conditions would 34-AB EOP-049,

'R bioactivity Release. Control,' require Emergency Depressurization of

'the reactor? t I

c. any release in excess.of 1000 mr/hr. l 1

b. any release'in excess of 1000 mr/hr outside the primary and secondary contaiment that cannot-be isolated.

c. any release in' excess of 1000 mr/hr outside the primary or secondary' containment.

d.Lany release in excess of 1000 mr/hr into the primary or secondary containment and one area exceedo max safe operating level.

QUESTION 5.16 (1.00)

On'a loss of 125/250 V DC Switchgear 2A, 2R25-S001, the operator is to open the main generator output breakers,

a. STATE the maximum' time limit allowed for opening the output breakers,

b. STATE the reason that the output breakers must be opened.

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. QUESTION 5.17 (2-50-)

For each power source in Column I below (a-f) MATCH the required

'immediate. action listed in Column II (1-6) that.is required to be take if~that-power' source is lost. :More than one immediate action may be required for a power source.

Column I Column II

a. 125/250 V DC Switchgear 2A 1. Transfer 4160 V. buses 2A anc 2B 2R22-S016 to Startup Supply.

b. 125/250 V DC Switchgear 2B 2. Manually trip Reactor Reciro 2R22-S017 Pump B.

c. 125 V DC Cabinet 2C, 3. Secure radwaste discharge.

2R25-S003

d. 24/48V DC Cabinent 2B, 4. Manually open Main Generator 2R25-S016 output breakers.

5. None QUESTION 5.18 (2.50)

Given the following general EOP conditions, STATE the correct flow pQth (by number) to follow, which is designed to address the following conditions, in accordance with 31EO-EOP-001-25, " Emergency Operating Procedure Inside Control Room Unit 2."

a. High. radiation, loss of coolant, loss of control of primary containment integrity (under degraded conditions).

b. High steam line radiation, loss of vital power, failure of vital equipment.

c. Malfunction of Reactivity Control System without AC power.

d. Malfunction of Reactivity Control System with AC power.

e. Reactor transients or failure of vital equipment eithout degraded conditions.

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QUESTION 5.15 (1.00)

A general-caution in the EOPs states " Shut MSIV's when reactor water

. :lovel exceeds 100. inches." This is to prevent the flooding of-the

't main steam' lines. . STATE TWO possible causes of damage if the main

. steam lines are flooded.

QUESTION 5.20 (2.50)

.Concerning Unit 1 Technical Specifications.

During normal power' operation, the suppression chamber water temperature shall be maintained less than or equal to (a) [ temperature). If this temperature limit is exceeded, pool cooling shall be initiated immediately.

During relief valve operation or testing of RCIC, HPCI, or other testing which adds heat to the suppression pool, the maximum water temperature shall not exceed- (b) [ temperature). In connection with such testing, the pool temperature must be reduced within (c) [ time limit) to less than or equal to (d) -[ temperature].

The reactor shall be (e) [ action] from any operating condition '

when the suppression pool temperature exceeds 110 F.

QUESTION 5.21 (1.00)

Actions to mitigate reactor power oscillations at high power / low flow conditions are taken to prevent exceeding which thermal limit?

L QUESTION 5.22 (1.00)

The plant is in shutdown cooling with the reactor head still in place when a loss of shutdown cooling occurs due to a loss of RHR Service Water. Significant decay heat exists.

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.If reactor pressure increases to greater than (1) then f L

(2) valve (s) will close. (Valve name(s) or number (s) are L

acceptable.)

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. QUESTION 5.23 (2.00)

.Concerning the Drywell Spray Initiation Limit:

a. DESCRIBE'the negative effects, including the'cause, of spraying the drywell when drywell parameters are in the unsafe region of the Drywell Spray Initiation Limit. (1.0)

b. STATE the parameters monitored by the Drywell Spray Initiation Limit.. (1.0)

QUESTION 5.24 (2.00)

LIST FOUR pressure' control methods used by SOFI Flowchart 3.

QUESTION' 5.25 (1.50)

a. For steam cooling the operator is directed to open one safety relief valve when reactor level decreases to 1/3 core height. This action will establish cooling to maintain peak clad temperature less than . (0.5)

b. STATE iWO methods of achieving adequate core cooling other than steam cooling as stated'in the definition of adequate core cooling.

(1.0)

QUESTION 5.26 (2.00)

During performance of the Secondary Containment Temperature Control 125 procedure the operator is directed to Emergency Depressurize if a primary system is discharging into an area and area temperature / differential temperatures exceed the Maximum Safe Operating temperature level in more than one area.

. STATE TWO purposes for performing this action.

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l QUESTION 6.01- (1.00)

Technical Specifications regarding Recirculation System Jet Pump OPERABILITY have some very restrictive Limiting Conditions for Operation (LCO) if-a jet pump is found to be INOPERABLE. From the choices below, SELECT the concern regarding continued plant operation

with an inoperable (or failed) jet pump.

a. Invalid APRM Flow Biased SCRAM setpoints due to the change in flow through a failed jet pump

b. Increased blowdown area during a Loss of Coolant Accident (LOCA)

c. Unbalanced neutron flux across the core due to flow variations

d. Physical core damage from a piece of a damaged jet pump QUESTION 6.02 (1.00)

The operating procedure for the High Pressure Coolant Injection (HPCI) system cautions against prolonged operation with turbine speed less than 2000 RPM. . SELECT the reason why this is of concern,

a. At low turbine speeds the potential exists for exhaust check valve chatter and reduced oil flow / pressure to the turbine governor and boarings.

b. At low turbine speeds the Booster Pump may not provide adequate net positive suction head to prevent cavitation of the main pump.

c. The rate of steam flow through the HPCI turbine may not be enough to prevent it from overheating.

d. At low turbine speeds, cooling water flow from the Booster Pump to the lube oil system heat exchanger is inadequate.

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-QUESTION 6.03 (1.00)

LUhen the Standby Liquid Control (SBLC) system is initiated, WHICH ONE of

.the.following occurs:

a. The-selected pump starts and only its associated (divisional) squib valve fires to provide a flow path to the reactor.

b..The SBLC storage ~ tank and pump suction line heat tracing heaters

' start to ensure all the sodium pentaborate remains in solution for injection into the reactor.

c. Both SBLC pumps start and both squib valves fire to provide two

. separate flow paths of sodium pentaborate into the reactor.

d. The Reactor Water Cleanup (RWCU) system outboard containment isolation valve (G31-F004) closes to isolate the system from the reactor.

QUESTION 6.04 (1.00)

SELECT the set of trip channel conditions (a - b) from the Reactor !

Protection System shown in the diagram below that will result in a HALF SCRAM. (assume a "one-out-of-two taken twice" logic) ,

TRIP SYSTEM TRIP SYSTEM A B l l l l l l l l A1 A2 B1 B2 TRIP CHANNELS TRIP CHANNELS

a. Al and A2' tripped

b. Al and B2 tripped

c. A2 and B1 and B2 tripped

d. A2 and B1 tripped l

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1 ,iQUESTION! : 6.05H .

7 ( l'. 00.)l

,"[Rsgar' ding the-Resotor Core Isolation Cooling.(RCIC)' system, which of the

+' ;following-isJNOT'a functionLof.the Suppression Pool?

e

'a. Receives 3theidischarge of:the Vacuum Pump, w '

3:,-:ActsLas-a heat' sink for th'e RCIC Turbine.' exhaust =

C F

~ c. Receivesidrainage from the RCIC Turbine steam'line and exhaust-E 'line-drain pots

'd.fProvides; backup source of. water-for the RCIC Pump

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[QUESTIONi /60'6 . (1.00);

' UNIT?2 has7Just. experienced an~ initiation of the Automatic

'Dopressurizaton System-(ADS). Plant conditions are as follows:

Drywell pressure: . 3.2 psig

.ReactorLwater' level': -147. inches

.All RHR pumps: ~

running 13/ minute timer: timed out~

'120 second timer: timed out

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7 ADS SRVs: open.

Main Steam pressure: 150 psig and lowering

?WHICH of the'following will:cause'the ADS SRVs to close? -(Consider.each' Jonswerl separately)

, a. The RHR/LPCI mode raises reactor water level to -20 inches.

Lb. Reactor pressure reaches 40 psig.

.c. Drywell-pressure decreases to 1.0 psig and the High Drywell Pressure. Seal-in push button is reset byLthe operator.

d. 3 of the 4 RHR pumps are secured.

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QUESTION' 16.07 (1.00)

The. OUTBOARD Main Steam Isolation Valve Leakage Control System (MSIV-LCS) MAY be manually initiated when WHICH of the following SPECIFIC sets of plant conditions exist:

a. Reactor pressure is less than 35 psig, all of the MSIVs are closed and Main Steem Line pressure between the Outboard MSIVs and the Main Turbine Stop Valves is less than 35 psig.

b. Reactor pressure has-been less than 35 psig for a minimum of 10 minutes, all of the MSIVs are closed and Main Steam Line i pressure between the Inboard MSIVs and the Outboard MSIVs is less than'35 psig.

c. At least 10 minutes have elapsed since'the Loss of Coolant Accident.(LOCA), the Outboard MSIVs are closed and pressure between the Outboard MSIVs and the Main. Turbine Stop Valves ,

and Bypass Valves is less thar 35 psig,

d. At least'10 minutes have elapsed since the Loss of Coolant Accident (LOCA), all of the MSIVs are closed and pressure between the Inboard MSIVs and Outboard MSIVs has, bled down to O psig.

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LQUESTIONc 6.08 (1.00) i

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LThe Main ~ Condenser. Mechanical Vacuum Pump is used to draw the initial

_vrouum up to a maximum of 5% power. SELECT the basis for this power. limit./

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.a. The Mechanical Vacuum Pump does not have the capacity'to draw sufficient vacuum'to clear the Main Turbine and Reactor Feed Pump low vacuum trips.

b. At powers above 5%, the amount of steam in the Main Condenser shortens the life of the seals in the Mechanical Vacuum Pump.

c. The Mechanical Vacuum Pumps are designed to draw a vacuum up to a Main Steam Pressure of 600 psig which corresponds to 5% power on a cold startup.

d. At powers above 5%, the amount of radioactivity being removed from the Main Condenser cannot be released to the environment and Hydrogen levels in the Condenser at this point could cause explosions from'the heat in the Mechanical Vacuum Pump

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.BESEQ.HSlBILITIEST13%iE

, j-L QUESTION. 6.09 -(1.00) j SELECT.the FINAL plant conditions after the loss'of the "A" Reactor Feed

. Pump from 100% power. Assume no operator action and all systems work as d: signed..

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a. Reactor Power 50-55%

Reactor Water Level at normal level Recirculation Pumps at.44% speed "B" Reactor Feed Pump speed higher than before the trip of "A"

b. Reactor Power 50-55%

-Reactor Water Level at normal level Recirculation Pumps at 22% speed "B" Reactor Feed Pump speed the same as before the trip of "A"

c. Reactor Power 60-65%

Reactor Water Level below normal level Recirculation Pumps at 22% speed i "B" Reactor Feed Pump speed lower than before the trip of "A"

d. Reactor Pcwer 60-65%

Reactor Water Level at normal level Recirculation Pumps at 44% speed "B" Reactor Feed Pump speed higher than before the trip of "A" QUESTION 6.10 (1.00)

SELECT the statement that describes time primary objective of the Fire

. Protection program at E. I. Hatch Nuclear Plant,

a. The primary objective is to protect Safety Related equipment and areas-within the plant.

b. The primary objective is to minimise the spread of fires within the Protected Area of the plant.

c. The primary objective is to minimize both the probability and consequences of postulated fires,

d. The primary objective is to provice early detection of fires and automatic protective actions.

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l QUESTION ~ 6.11 (0.50)

LIST the electrical busses supplying power' to the "A" and "B" Control:

Rod Drive-(CRD) pumps.

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-QUESTION 6.12 (1.00)

Why.are the Control Rod Drive (CRD) pumps NOT necessary to be in

. operation to support the FULL insertion of all control rods on a reactor scram with the reactor!at normal' pressure.

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, QUESTION 6.13 (1.00)

STATE the purpose of the Rod Worth Minimizer (RWM).

o QUESTION 6.14 (1.50)

Complete the following statements regarding the Unit 2 Rod Worth Minimizer (RWM).

A. The RWM may be manually bypassed under these TWO (2) conditions:

(1) and (2) .

B. Permission to manually bypass the RWM must come from (3)

C. If'the RWM is manually bypassed, compliance with required rod patterns and initialing of the Data Package must be done by (4) or (5) .

D. The RWM is automatically bypassed when above the (6)

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RESPONSIBILITIES (13%f u

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,- QUESTION 6.15 (2.50) l Answer the followinc questions concerning the operation of the Low Pressure Coolant Injection (LPCI) system:

i A. Following a Loss of Coolant Accident (LOCA) signal, what is the automatic sequence of events for the Outboard Injection Valves (FO 17A/B)? (Include in your answer any applicable time limits for both UNITS, but do NOT include pressure setpoints) (1.0)

B. What plant conditions will auto start the RHR pumps in this mode?

(setpoints required) (0.5)

C. What is the pump start sequence if a concurrent Loss of Off-Site Power (LOSP) occurs with an injection signal? (1.0)

QUESTION 6.16 (1.00)

UNIT 2 Technical Specifications requires removal of the shorting links during core alterations and shutdown .nargin demonstrations. BRIEFLY describe which Reactor Protection System (RPS) TRIPS this will affect ,

and how-it will affect them. '

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. QUESTION '6.17 (2.00)

From a comparison of UNIT 1 and UNIT 2 reactor SCRAM setpoints, SELECT the SCRAMS from the right-hand column and match them to the appropriate

-statements in the left-hand column- . (Right-hand column choices may be used more than once)

SCRAMS A. SCRAM setpoints that 1. Reactor pressure are different-between the units. 2. Turbine control valve fast closure

3. Scram discharge volume high level

'B. SCRAM (S) that are NEVER 4. Low reactor water level bypassed.

5. APRM High flux (flow biased)

6. APRM High flux C. SCRAM (S) that are AUTOMATICALLY bypassed 7. High Drywell pressure under specified plant conditions. 8. Turbine Stop Valve Closure

9. SRM High Flux j

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1 QUESTION 6.18 (1.20) l l

l ' Complete the following statements regarding the Intermediate Range Monitoring (IRM) System.

l A. The IRM provides neutron monitoring over (1) decades I of power as indicated by (2) ranges of indication. l l

L B. IRM provides TWO (2) reactor SCRAMS, (3) and l (4) . (NO setpoints required)'

C. The IRM detectors are REQUIRED to be fully withdrawn from the core L as soon as (5)

D. Reactor period indication is available on IRM by using the l

(6) on the IRM recorders.

QUESTION 6.19 (2.50)

A. STATE FOUR (4) Reactor Core Isolation Cooling (RCIC) AUTOMATIC turbine trips including setpoints (DO NOT include the RCIC system isolation turbine trip) (2.0)

B. Which of the RCIC turbine trip (s), manual and automatic, is/are required to be reset locally? (0.5) i I

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' QUESTION 6.20 (1.50)

COMPLETE the following:

1. .The Automatic Depressurization System.(ADS) consists of (number) (a) of the 11 Main Steam System Safety Relief Valves (SRV).

2. ADS is' designed to protect the core during a (size) __(b)

. Loss of Coolant Accident (LOCA) where the (c) system fails to maintain reactor vessel water level.

3. The ADS SRV accumulators are sized for at least (number)

(d) valve actuations.

QUESTION 6.21 (0.75)

The Spent Fuel Pool Cooling and Cleanup System is designed to maintain the Spent Fuel Pool temperature below certain limits. Given the following plant. conditions, STATE the associated temperature limit.

A. During normal operations B. During refueling C. During core off-load j

l QUESTION 6.22 (1.00)

A. STATE the Technical Specifications setpoint for Main Steam Isolation Valve (MSIV) closure on Low Main Condenser Vacuum.

(0.25) )

B. What are the THREE (3) conditions that must exist to bypass the MSIV closure on Low Main Condenser vacuum. (0.75)

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QUESTION '6.23 (1.50)

LIST the' automatic start signals for the Emergency Diesel Generators.

Include in your answer the setpoints for the start signals.

(Consider each signal' separately) 1 i

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MSPONSIBILITIE3 (f3%i~

QUESTION 6.24 (1,80)

MATCH the following Off Gas System componente with the correct statements from the right-hand column. There may be more than one correct answer for each component and each answer from the right-hand' column may.be used more than once.

COMPONENTS A. Off Gas Preheater 1. Provides automatic protective action signals to various Off Gas System components D. Charcoal Adsorber 2. Provides time for Xenon and Krypton isotopes to deray.

C. Off Gas Stack Isolation 3. Is an electric boiler on Unit 2 1 Valve

4. Reduces the leve?.s of Hydrogen and oxygen in the off-gas flow D. Off Gas Catalytic 5. Will shut on HI-HI-HI radiation Recombiner levels from the Post Treatment Radiation Monitor

6. Has a design 30 minute delay time E. Holdup Volume 7 Will shut on High Hydrogen levels in the off-gas flow F. Post Treatment Radiation 8. Uses 250 psig steam on Unit 2 to Monitor heat the off-gas flow

9. Samples the off-gas flow going up the Main Stack

-G. Loop Seal Drain Valve

10. Removes Iodine isotopes from the off-gas flow.

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MSEd.LQl.BILITIEU130 QUESTION 6.25 (1.00)

SELECT the statement which is NOT in conformance with 30AC-OPS-00-OS,

' Control of Equipment Clearances and Tags.

a.-When pulling fuses i. deenergize power supplies, the DANGER tags will be attached to blockout fuses if available,

b. When danger tagging an MOV.as a clearance boundary, tagging the breaker and local operator will satisfy the tagging requirements. ;

c. It is permissible to repack a MOV with the valve on its backseat and a DANGER tag on its local operator, breaker, and control switch,

d. When danger tagging an MOV which requires independent verification it is permissible for the independent verifier to use the valve position indication lights to determine valve position.

QUESTION 6.26 (1.00)

IDENTIFY which one of the following is acceptable, in accordance with Control of Operator Aids (DI-OPS-05-1084N), for use as an Operator Aid.

a. Posting of a pending change tc an Emergency Operating Procedure.

b. Instructions while a procedure is written and approved. I

c. A caution concerning operation of a component.

l l d. Instructions for emergency startup of a diesel. j l

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-QUESTION 6.27 .(1.00) l' i l

1 SELECT.the correct statement concerning symbols used on the EOP flow I

-charts.'

. Refer to the symbols on figure 3.

l> . l L' a.'The arrow within symbols A and C indicate the direction to be L followed if the decision is answered yes,

b. Symbol G is used to direct the operator to an End Path Manual.

c..If the answer to a decision in a symbol C changes then the operator is to return to the top of the charts,

d. Symbol D is used to identify steps that must be performed in sequence.

QUESTION 6.28 (1.00)

SELECT the correct statement concerning operating chart recorders in

" =the Main Control Room

a. One hour prior to shift change the off going shift operator shall mark each chart with the date, time, and his initials.

b. A chart with'an incorrect scale may not be utilized on a Control Room recorder.

c. On-shift plant operators shall ensure that each completed chart is identified by the recorder number, marked with the date and time'it was installed and removed, then forward it directly to Document Control for filing.

d. Used charts are to be given to the Shift Supervisor and periodically forwarded to Document control for filing.

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iQUESTION 6.29 (1.00)

IDENTIFY whish'one'of the following actions the Emergency Director can de".egate, a.. Decision to authorize exceeding overtime limits.

b.EDecision'to notify offsite emergency response-agencies.

c. Decision to downgrade the emergency classification.

d. Decision to order evacuation of nonessential personnel.

QUESTION 6.30. -(2,50)

In accordance with 10CFR20.

a. STATE the quarterly radiation exposure limit for each'of the following:

'1).Whole body- (0.5)

2) Hands and forearms, feet and ankler (0.5)

3) Skin of the whole-body (0.5)

b. The-limit stated in part a for whole body dose may be exceeded if specific conditions are complied with. Answer the following concerning the maximum limits allowed.

-1) STATE'the maximum quarterly whole body limit. (0.5) 2). STATE two conditions that must be met for an individual to exceed the quarterly exposure limits stated in part a. (0.5)

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QUESTION 6.31 (2.00)

On-back shift, a fire is reported in one of the maintenance shops:

c. STATE the PRIMARY and ALTERNATE persons (position title) designated to assume Fire Brigade Chief. (1.0)

b. What is the minimum number of persons who should show up for the fire brigade? (0.5)

c. . How many of these persons should have competent knowledge of safety systems? (0.6)

QUESTION. 6.32 (2.00)

You are verifying a valve line-up. STATE how you confirm position of each of the following:

a. Closed valve,

b. 'Open valve.

c. Motor-operated valve.

d. Locked throttle valve, s,.

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' QUESTION.- 6.33 (2.00)

Match each of the. responsibilities in Column I with the person to whom

,the' responsibility is delegated to in Column II. More than one person i

in, column II may be app 31 cable.

Column I Column II

a. Maintain the broadest perspective 1. Shift Foreman of operational conditions affecting the safety of the plant. 2. Operations Supervisor on Shift

11. Responsible for compliance with requirementsLof the operating 3. Plant Operator license, Technical Specifications, and' approved plant operating 4. Shift Supervisor procedures.

c. Responsible for supervising activities of all persons, from plant' operator on down, on shift not fulfilling a shift position, d Responsible for. performing

. required surveillance.

e. Responsible for controlling conduct in the control room.

(***** END OF CATEGORY 8 *****)

(********** END OF EXAMINATION **********) '

l l

_ _ _ _ _ - - __ - - 1

y. . r -- - -

3 A

ANSWER- 5.01 (1.00)-

a [1'.0]-

' REFERENCE E.I. Hatch, Trip of One or Both Reactor Recirculation Pumps, 34AB-OPS-032-2S,-pg 4.

L 3.8/3.7 295001G010 ..(KA's)

.y t>. lea-M

- ANSWER- ~

.(1.00) o r O.

c [1.0]

REFERENCE E.I. Hatch, Reactor Power Instabilities, 34AB-OPS-058-25, page 2.,

NRC Bulletin ~No. 88-07, Supplement 1: Power Oscillations in Boiling Water-Reactors.

2.5/3.3 3.8/3.7 295001G010 295001K104 ..(KA's)

Ded ANSWER 5 (1.00)

-d [1.0)

REFERENCE E.I. Hatch, Loss of Instrument and Service Air System, o

~34AB-OPS-020-2S, pg 2.

2.9/2.9

-295002K306 ..(KA's) a

' ANSWER' 5.04 (1.00)

Ec. [1.0]

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

l i

u__---. _--

~ REFERENCE E.-I Hatch, Loss of' Instrument Buses, 34AB-OPS-014-25, page 2.

4.2/4.3 295003A202 ..(KA's)

ANSWER .5.05 (1.00) o [1.0]-

REFERENCE E.I. Hatch, SOFI Flowchart 2: Content and Use, LT-IH-20103-02, pg 20.

EO 18.

4.1*/4.5* 4.0/402 4.1*/4.3*

295037K303 295037K209 295037K102 ..(KA's)

ANSWER 5.06 (1.00) d [1.0]

REFERENCE E.I. Hatch, Loss of Feedwater. Heating, 34AB-OPS-045-2S, page.2.

4.0/3.9 3.6/3.8

'295014A102 295014G010 ..(KA's)

ANSWER 5.07 (1.00)

a. [1.0]

REFERENCE E.I. Hatch, SOFI Flowchart 1: Content and Use, LT-IH-20107-03, pg 16, EO 7.

3.8/3.9 295015K201 ..(KA's)

ANSWER 5.08 (1.00) b [1.0]

l l

l (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) i l l l'_____--.---_

L - _ _ _ . - - - _ >

' REFERENCE-

E.I. Hatch.-Remote Shutdown Panel, LT-IH-05201-00, pg 12, 13, and 14.

-E0 6.

4.1/4.1.4.0*/4.1*

295016G006 295016K202 ..(KA's)

ANSWER 5.09 (1.00)

'b [1.0)

REFERENCE. .

E.I. Hatch,. Loss of Reactor Building Closed cooling Water,

-34AB-OPS-011-25, pg 2.

3.4/3.6 295018K202 ..(KA's)

ANSWER. '5.10 (1.00) I c'[1.0]

REFERENCE E.-I. Hatch,: Loss of Instrument and Service Air System, l

'34AB-OPS-020-25, pg'2. Plant Air Systems, LT-IH-03501, EO-16. I 3.2/3.3:3.3/3.2 3.3/3.1 295019A104 295019K203 295019A102 ..(KA's)

ANSWER 5.11 '(1.00)

.e [1.0) l)e let ed REFERENCE E.I. Hatch, SOFI Flowchart 1: Content and.Use, LT-IH-20107-03, pg 18, 20, 22, 23. EO 11, 15.116. E.I. Hatch, Emergency Operating Procedure Variables and Curves, LT-IH-201113-00, pg 7, EO 1.

3.7/4.1*

295026K304 ..(KA's) j

.i l

i 1

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) ]

1 a

_ __x . ---_. _ _.

i .4

? ANSWER 5.12 (1.00)

'd- [1. 0 ] .

REFERENCE E.I. Hatch,-Emergency Operating Procedure Variables and Curves, LT-IH-20113-00 pg 11 and 12. EO 1 and 2.

'3.8/4.1.3.5/3.7.3.5/3.8 3.4/3.8 295026K102 295026K206 295026K301 295026G007 ..(KA's)'

ANSWER: 5.13 (1.00) c [1.0]

REFERENCE E.I. Hatch, PSTG and SOFI Cnart General and Specific Cautions, LT-IH-20114-00,Section IV.F 3.6/3.8 3.4/3.8 3.7/3.9 295028G007 295028K203 295028A203 ..(KA's)

ANSWER 5.14 (1.00) b [1.0]

REFERENCE E.I. Hatch, LT-IH-20114-00, pg 14. EO-2.

3.5/3.8 295030K208 ..(KA's)

. ANSWER 5.15 (1.00)

b. [1.0]

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

l REFERENCE.

E.'I. Hatch, 34AB-EOP-049, Radioactivity Release Control: Content and

'Use, LT-IH-20115-00, pg 6. E06.

3.7/4 7*'

295038K205 ..(KA's)

ANSWER 5.16 (1.00)

n. 30 seconds. [0.5] (accept 20-30 seconds) b.' Prevent motoring (reve se power) of the generator. [0.5)

REFERENCE E.I. Hatch, Loss of DC Buses, 34AB-OPS-013-2S, page 7.

3.3/3.4 3.2/3.4 3.4/3.6 {

295004G010 295004A103 295004K105 ..(KA's) 24 0 ANSWER 5.17' (3,5&)

a. 4 , Jt# '

b. 2

c. 1 cit. 3

[Br5][0.5 each]

70 REFERENCE

'E2 I. Hatch, Loss of DC Buses, 34 AB-OPS-013-2S. .

3.2/3.4 j i

'295004G010 ..(KA's)

ANSWER 5.18 (2.50) i

a. Path 5 )

b. Path 4 )

c. Path 2 {

.d. Path 1 i

o. Path 3 1 (0.5 pts each) i i

1 l

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) ]

j t___._.___._u._. _

I

l; ; ,

N '

! REFERENCE

'GPC: 31EO-EOP-001-25, LP # LT-IH-20101-00, EO:#7 N 8/3.4 295006G012 ..(KA's)

) ANSWER 5.19 (1.00)

1. weight of the water [0.5]

-2, thermal stress [0.5)

REFERENCE *

'E.I.' Hatch, PSTG and SOFI General and Specific Cautions, LT-IH-20114-00, pg 15. EO 1.

I -3.2/3;3 295008G007 ..(KA's)

. ANSWER ' 5.20 -(2.50)

a. 95 F'[0.5]

b. 105 F [0.5]

'c. 24' hours [0. 5]'

d. 95 F [0.5]

' o. scrammed [0.5]

REFERENCE

.E.I. Hatch, Technical Specifications, 3.7.A.1 3.3/4.2*

295013G003 ..(KA's)-

' ANSWER 5.21 (1.00)

MCPR .[1.0]

REFERENCE NRC Bulletin No. 88-07, Supplement 1: Power Oscillations in Boiling Water Reactors.

4.1/4.4*

295014A204 ..(KA's)

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

'm .w-~a . - - . - , = , .u a . _

k' l

L ' ANSWER' 5.22 -(1.00) l

' a. 145 psig [0.5] (accept 130 to 145)

b. RHR Suction Cooling Valves, 2E11-F008 and F009. [0.5]

tREFERENCE L -E.I. Hatch, Loss.of Shutdown Cooling, 34AB-OPS-044-23, pg 1.

L -RHR' System, LT-IH-00701, EO 7.D 3.6/3.6 295021K203 ..(KA's) l ANSWER' 5.23 (2.00) *

[ M )#due to negative

n. Drywell pressuremay [0.5]collapse (resulting'[0.43'(or from otherwise f ail) tion),

rapid condensa

b. Drywell temperature [0.5)

Drywell pressure [0.5]

REFERENCE-E.I. Hatch,. Emergency Operating Procedure Variables and Curves, LT-IH-20113-00...page 10. EO 2.

4.2*/4.4* 3.9/4.0 3.6/3.9

~295024A201 295024A202 295024G007 ..(KA's)

ANSWER 5.24- (2.00)

,1. Lo-Lo Set

2. Main Turbine Bypass Valves 3; SRV's

4. Alternate Pressure Control Systems. (Accept any one alternate pressure control system)

-[2.0] [4'at 0.5 each]

REFERENCE E.I. Hatch, SOFI Flowchart 3: Content and Use, LT-IH-201^4-03, pg 14.

EO-10.

3.9*/4.5*~4.4*/4.4* 3.8/3.8 295025G012 295025A103 295025A102 ..(KA's)

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

I

< ANSWER. 5.25 (1.50)

c. 2200 F. [0.5]

b. 1. Vessel level-above the top of_ active fuel [0.5].

2. At least one Core Spray pump injecting at rated flow [0.5].

REFERENCE E.I. Hatch, EOP End Path Manual Content and Bases, LT-IH-20108-4, pg

47. EO 3. EOP Flow chart and End Path Manual Design and Use, LT-IH-20102-03. pg.30.

4.0/4.3* 4.4*/4.7* 4.0/4.3*

295031K303 295031K302 295031K304 ..(Kh's)

ANSWER 5.26 (2.00)

1. Terminate energy addition to the secondary containment'[1.0]

2. Place the RPV in a low energy state [1.0] (also accept discharge energy to suppression pool for full credit.)

REFERENCE E.I. Hatch, EOP End Path Manual Content and Bases, LT-IH-20108-04, pg

74. EO-4.

3.5/3.8 295032K301 ..(KA's)

I l

(***** END OF CATEGORY 5 *****)

I l

i

)

L- - - - - - - - - - )

~ ~- '~

.RRESPONSIBILITIESTT13%f~

ANSWER- 6.01 (1.00)-

b (1.0)

' REFERENCE E .- I d. HATCH LESSON PLAN,. REACTOR RECIRCULATION SYSTEM, LT-IH-00401-00, EO #25, PAGE 14 3.5/3.7 3.4/3.9 202001K601 202001A201 ..(KA's)

ANSWER 6.02 (1 00) a~ (1.0)

REFERENCE DE.I. HATCH LESSONLPLAN, HIGH PRESSURE COOLANT INJECTION, LT-IH-00501-02, EO #15.d.3, PAGE 45 3.8/3.7 3.3/3.3 206000A401 206000K505 ..(KA's)

ANSWER 6~.03 (1.00) d 4" b- (1.0)

REFERENCE E.I. HATCH LESSON PLAN, STANDBY-LIQUID CONTROL SYSTEM, LT-IH-01101-00, EO #11 & 13, PAGES 15, 17 & 18 !

3.8*/3.9* 4.2*/4.2* 4.0*/4.1* !

211000K407 211000K408 211000A306 ..(KA's) {

l 1 l

l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) J l

l L: _ _ _

mx-g gggg = -- --

L

[

ANSWER 6.04 (1.00);

a (1.0)

REFERENCE E.I.' HATCH LESSON PLAN,' REACTOR PROTECTIVE SYSTEM, LT-IH-01001-00,

'EO #6,-PAGE 22 & 23 3.3/3.4 3.7/3.8 212000K502- 212000K305 ..(KA's)

' ANSWER 6.05 (1.00) c (1.0)

REFERENCE-E.I. HATCH-LESSON. PLAN, REACTOR CORE ISOLATION COOLING SYSTEM, LT-IH-03901-00, EO 114, PAGE 7 3.6/3.6 217000K103 ..(KA's)

ANSWER 6.06 (1.00) b (1.0)

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

~~ ' ~

~

T RESPONSIBILITIES 713si REFERENCE E.I. HATCH LESSON PLAN, AUTOMATIC DEPRESSURIZATION SYSTEM, LT-IH-03801-03, EO #12, PAGE 30 4.2/4.3* 4.2*/4.2*

'218000A206' 218000A402 ..(KA's)

ANSWER' 6.07. (1.00) c (1.0)

REFERENCE E.I. HATCH LESSON PLAN, MSIV LEAKAGE CONTROL SYSTEM, LT-IH-04901-00, EO #6.& 10b, PAGES 16, 21 & 22 3.1/3.3 3.1/3.1

.239003K406 239003A101 ..(KA's)

ANSWER 6.08 (1.00) d (1.0)

REFERENCE l

1 E.I. HATCH LESSON PLAN, MAIN CONDENSER, LT-IH-02501-00, EO #15, PAGE 20 2.8/2.8 3.1/2.9 3.4/3.4 i

l 256000K409 256000G010 256000G007 ..(KA's) 1 l

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

l

~ ~ ~

j- RESPONSIBILITIES T13 W L

LANSWER. 6.09 (1.00)

'd' (1.0)

" REFERENCE Y

E.I. HATCH LESSON PLAN, CONDENSATE AND FEEDWATER, LT-IH-00201-00, E0.#11. PAGES 17 & 18 3.9/3.9,3.8/3.9 3.7/3.7 3.4/3.4 259001K301 259001K312 259001A201 259001A310 ..(KA's)

' ANSWER 6.10 (1.00) e (1.0)

REFERENCE E.I. HATCH LESSON PLAN, PLANT FIRE PROTECTION SYSTEMS, LT-IH-03601-00, EO #1,.PAGE 7 3'8/3.9 286000G004 ..(KA's)

ANSWER 6.11 (0.50) i "A" CRD Pump -

4160 VAC Bus 2E (Emergency Bus 2E) I "B" CRD Pump -

4160 VAC Bus 2F (Emergency Bus 2F)

(0.25 each) i

)

1

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

~ ~

~

, " BENF_QHSIBILITIES 113'X,il h

' REFERENCE i \

E.I.' HATCH LESSON PLAN, CONTROL ROD DRIVE HYDRAULICS, LT-IH-00101-00, EO #13, PAGE 41

- 2.9/3.1 201001K201 ..(KA's)

ANSWER 6.12 (1.00)

- The scram -accumuletcre end-Maactor pressure function as the source of energy to insert control rods on a SCRAM independent of CRD pump operation. (alternative wording acceptable) (1.0)

REFERENCE E.I.. HATCH LESSON PLAN, CONTROL ROD DRIVE HYDRAULICS,'6T-IH-00101-00, EO #5o, PAGES 30'& 31 3.8/3.'8 3.1/3.2 201001K405 201001K303 ..(KA's)

ANSWER 6.13 (1.00)

To limit control rod worth (0.75) so that the fuel enthalpy limit of 280 cal /gm will not be exceeded during a rod drop accident (0.25).

l

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) I

.BESPONSIBIL E 1 5

's

?, <

w

.. REFERENCE -

'E.I. HATCH LESSON PLAN,' ROD WORTH MINIMIZER, LT-IH-05403-00, EO #1, lPAGE 6 . .

,. 3.3/3.7 3.4/3.'4

.201006K501 201006 GOO 4 ..(KA's)

ANSWER 6.14 (1.50)

A. (1) Inoperable

'(2) for testing t.

'B. (3) Operations Supervisor C. (4). A-second. licensed operator (5) A qualified member of the Technical Staff D. (6) Low Power Setpoint (LPSP) (30% power) (0.25 each)

REFERENCE E.I. HATCH LESSON PLAN, ROD WORTH MINIMIZER, LT-IH-05403-00, EO #8, PAGE 30 3.2/3.4 3.4/3.5 201006A401 201006K404 ..(KA's)

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

l' L.___.________________

~

L " RESPONSIBILITIES 113%i q i

ANSWER 6.15 (2.50) 1 l A. Valves are interlocked open (0.5)

Ten minute on UNIT 1 -(0.25)

.Five minutes on UNIT 2 (0.25)

I B. High Drywell Pressure of greater than or equal to 1.92 psig; or Reactor Water Level of less than or equal to -113 inches.

(0.25 each)

C. (Upon reenergization of the respective pump buses.) the pumps start' in the following sequence:

RHR."C" -

starts immediately RHR "A", "B" and "D" -

start after a 10 second delay

( ' **

AL60 ACcc#r 77tne DEuu/ tor _ bcerc T)6 C^t 40C0 W A60uc AluotGf45 [TlIZ) p g "[' _

(Z SCC REFERENCE s . s s

/2M ,,/1, 'd d "D - 2 2 scc E.I. HATCH LESSON PLAN, RESIDUAL HEAT REMOVAL SYSTEM, LT-IH-00701-01, EO #7a & 7b, PAGES 28, 29 & 32 3.5*/3.5* 3.7/3.9 3.6/3.7 4.2*/4.2 203000K201 203000K407 203000K601 203000K401 ..(KA's)

ANSWER 6.16 (1.00)

TheJiuc14ar- Isstrumentat-i-on-tr-i-p-4egirrtrangc5 to non-ee-iso 14ence

-tr-A pr (0.5), &ny-one--of 19 NI trips will cause a SORAM (Gr&+

~TISE N u ufAl 'IMVRJnliumDon TAtPS D.6) (pg pa m n W Oui- 6C 1 6 T w in Ona (dou- (Dractacnch ((0.5)

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

.< ~ ~ RESPONSIBILITIES'(13%) l t: ' 4 L .

REFERENCE F 'E.I. HATCH LESSON PLAN, REACTOR PROTECTIVE SYSTEM, LT-IH-01001-00,

'PAGE~28 3.3/3.5 L

212000K411- ..(KA's)

ANSWER- 6.17 (2.00_)

A. 3,.5, 6 B. 1, 4, 7 C. 2, 8 (0.25 each)

REFERENCE E.I. HATCH LESSON PLAN,. REACTOR PROTECTIVE. SYSTEM, LT-IH-01001-00, EO #16, PAGES 18 & 19 212000G005 212000K412 ..(KA's)

ANSWER 6.18 -(1.20)

.A. (1)' 5 (2) 10 B. (3) .IRM Inop

-(4) IRM.Hi-Hi C. --(5) All APRMe-er+-on--seale-4but,-before reaching-443--on-Range 10 M

-eni ITdib 4FTCA #tAcuv c, Ih obf Su)4'rcu- TD " /2un "

D. (6)- Slope of the line (will accept " time to double")

(0.2 each) ,

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

E

' RESPONSIBILITIES-(13%i o

REFERENCE' Ia

! E.I.: HATCH LESSON PLAN; INTERMEDIATE RANGE MONITORS, LT-IH-01202-00,

'EO #3, 4c,-11, 13a & 13d, PAGES 11, 15, 26, 27, 30 & 32 4 0/4.0 3.3/3.3.3.0/3.1 215003K402 215003A401 215003K503 ..(KA's)

- ANSWER, 6.19 .(2.50)

A. High turbine exhaust pressure 40 psig (36-44 psig)

Low RCIC pump suction pressure 10" Hg vacuum (9-11" Hg).

Electrical overspeed 4950 RPM or 110%

' Mechanical overspeed. 5625 RPM or 125%

(0.3 for trip, 0.2 for setpoint)

B. Mechanical overspeed 1 Local manual (0.25 each)-

' REFERENCE E I. HATCH LESSON PLAN, REACTOR CORE ISOLATION COOLING SYSTEM, LT-IH-03901-00, EO #6c & 8a, PAGES 12,~13 3.8/3.7 217000A202 ..(KA's) i

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

I u___i__. ..___:_._____

~ ' ~ ~

.BESPONSIHlklTIEs Ti3 n L

Y, .

ANSWER 6.20 (1.50)

. small break ~tWnrA/ncoiAw M 4k c; High Pressure Coolant Injection (HPCI) d.- 2 .3 tm4 (n A75eaolr)

REFERENCE E.I. HATCH LESSON PLAN, AUTOMATIC DEPRESSURIZATION SYSTEM, LT-IH-03801-03, EO #1, 2, 13 & 15, PAGES 9, 10 & 25 3.5/3.6 3.9*/3.9*

218000K404 218000K106 ..(KA's).

ANSWER 6.21 (0.75)

A. 139 degrees (accept 137 to 141)

B. 133 degrees (accept 131 to 135)

C. 150 degrees (accept 148 to 152)

(0.25 each)

REFERENCE E.I. HATCH LESSON PLAN, FUEL POOL COOLING AND CLEANUP, LT-IH-04501-01, EO.#1, PAGE 5 2.8/2.9 233000G010 ..(KA's)

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

\

T RESPOliElHILDIES ' i13%)

~ ANSWER 6.22 (1.00)

A. 7" Hg vacuum (0.25)

B. - Reactor Mode Switch not in "RUN"

- Turbine Stop Valves not full open (less than 90% open)

- MSIV Low Vacuum Trip Bypass Switches in " BYPASS" (0.25'each) ;

REFERENCE E.I. HATCH LESSON PLAN, EAIN STEAM / LOW-LOW SET, LT-IH-01401-00, EO #9a & 11a, PAGES 40 &43 3.3/3.4 3.8/3.9 239001K608 239001A208 ..(KA's)

ANSWER 6.23 (1.50)

Loss of power to respective bus

-- Less than 604 nominal voltage Er ~- i 36vNo INS W (.d

-6&%r Low reactor water level Less than or equal to -113 inches High drywell pressure Greater than or equal to 1.92 psig (0.3 for each start signal) j (0.2 for each setpoint) l l

l

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

l l

l L-- _- _ _

~ ~ ~ ~

.., RESPONSIBILITIES fi;L41} l n

' ,, 5 . ; r

? REFERENCE U<

l ..

E.I. BATCH' LESSON PLAN, DIESEL GENERATORS, LT-IB-02801-00, EO #4,

.PAGE'17 m .

K. 13'8/3.7 l;

i' p.

L264000K408 ..(KA's) b ANSWER 6.24 (1.80) t.

.A. 8 B. 2. 10 C. 5

-D. 4 E. . 2 ', 6 F. 1 _'(fjo pl$ pa/xWN e't) 4 G. 5- (0.2 each)

REFERENCE E. I . HATCH LESSON PIAN, OFF-GAS SYSTEM, LT-IH-03101-00, EO #6 & 10, PAGES 13-18, 27 & 28 3.1/3.3 3.1/3.3.3.3/3.3 3.3/3.4 271000K102 271000K408 271000A301 271000G007 ..(KA's) s ')

(***** CATEGORY 6 CONTINUED ON NEXT PAGE **c**)

l .i:

~

- = _ - _ _ _ __

~ ~~ ~ ~ '

~ ~ EESEQHSIBILIIIES li3d t

ANSWER 6.25 (1.00) b' . , (1.0) eu :.7, ,

. REFERENCE

, E.I. Hatch 30AC-OPS-001-OS pp. 7 & 8.

q 3.9/4.5-R 294001K102 ..(KA's) 4 LANSWER: 6.26 (1.00)

e. [1.03 REFERENCE E.I. Hatch, Control of' Operator Aids. DI-OPS-05-1084N, page 3.

4.2*/4.2*

294001A102 ..(KA's) m y.,

ANSWER 6.27 (1.00) c [1.0)

REFERENCE

p E.I. Hatch, EOP Flowchart and End Path Manual Design and Use,

LT-IH-20102-03, page 10 to 14. EO 4.

'4.2*/4.2*

L, 294001A102 ..(KA's)

FANSWER 6.28 (1.00) 1.i ( , 'd . ( 1. 0 )

a l<.

x

-(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

~,c

{ ,

~~ ~

7 RESPOM 1HILITIES 113'xi

' REFERENCE

'E.I'.4atch Plant Operations 30AC-OPS-003-OS pp. 16 &

J17JEnabling Objective 29 of'LT-IH-30004-01 3.4/3.6'

-294001A106 ..(KA's)

ANSWER 6.29 (1.00)

c. [1.0]

REFERENCE E.I.l Hatch... Hatch Emergency Plan, pg'13.

2.9*/4.7*

294001A116 ..(Yd's) l ANSWER '6.30 (2.50)-

... lo. 1) 1.25. rem [0.'5]

2) 18.75 rem [0.5]

3) 7.5 rem [0,5)

.b. 1.).3 rem [0.5]-

2) Exposure must not exceed 5 (N-18) [0.25]

A- Ivim- 4-muet-be-completed- [0. 25]

greesus fr)shiq Vn& n -

REFERENCE

.E. I.-' Hatch, Radiation Exposure Limits, 60AC-HPX-001-OS, pg 4, 10CFR20 3.3/3.8 294001K103 ..(KA's) 1

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

~~' RESPONSIBILITIES 113'Q ANSWER = -6.31 (2.00) a._U-2 Shift Supervisor (will report to fire as Brigade Chief if he is qualified); if not. U-1 SS will act as Chief. [1.0)

b. 5 [0.5]

o. 3 [0.5]-

' REFERENCE CPC: 40AC-FPX-001 3.5/3.8 294001K116 ..(KA's) fANSWER 6.32 (2.00)

a. Turn valve in closed direction (1/4 turn max to seat)

~b. Turn valve in closed direction (1/4 turn max off backseat)

c. Verify at remote (or lecal) position indicstio3

d. Confirm locking device operability.

( 0.5 each )

REFERENCE EIH: 34-GO-SUV-001-OS, EO # 3.1.3.3 3.7/3.7 l 294001K101 ..(KA's)

ANSWER 6.33 (2.00)

.c. 2 [0.25]

b.'1, 2, 3, 4 [1.0] i

,c. 1 [0.25] j

'd. 3 [0.25] (if 4 is also given in addition to 3 no points are to be i deducted)

e. 4 [0.25]

REFERENCE E.I. Hatch, Plant Operations, 30AC-OPS-003-OS, pg 21.

'3.3/4.3* 3.5/4.2 ,

294001A111 294001A112 ..(KA'a) l 1

(***** END OF CATEGORY 6 *****) I

(********** END OF EXAMINATION *=4*******) 4 I

E_

%t

' U ., S . NUCLEAR REGULATORY COMMISSION

^ SENIOR REACT ^R OPERATOR LICENSE EXAMINATION REGION 2 f

FACILITY: E. I._ Hatch 1 & 2

[

) . -

REACTOR TYPE: RWR-GE4 Q DATE ADMINISTERED: 69/Ofi212 ,

j$1IONSTOGANDIDAI5;_

separate paper for the answers. Write answers on one side only.

le question sheet on top of the answer sheets. Points for each

( tion are indicated in parentheses after the question. The passing requires at least 70% in each category and a final grade of at 80%. Examination papers will be pickod up six (6) hours after xamination starts.

% OF lY % OF CANDIDATE'S CATEGORY

,_ _IQIbL SCORE VALUE CAIEGORY 23.?7

_ 33-7f) 4. REACTOR PRINCIPLES (7%)

THERMODYNAMICS (7%) AND COMPONENTS (10%) (FUNDAMENTALS EXAM)

% :S d i d fr 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS (33%)

'i't.~70

.-AfL-12 6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC RESPONSIBILITIES (13%)

% TOTALS FINAL GRADE ork done on this examination is my own. I have neither given received aid.

Candidate's Signature

'a

} rn_ -

NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS

administration of thic examination the following rules apply

g,g on the examination means an automatic denial of your application

',3sld result in more severe penalties.

limited and only one candidate at a time may

- h,gsm

, Youtrips are must to beall avoid contacts with anyone outside the examination ,

j to avoid even the appearance or possibility of cheating.

Eblack ink or dark pencil only to facilitate legible reproductions.

1 at your name in the blank provided on the cover sheet of the nination.

1 in the date on the cover sheet of the examination lif necessary).

only the answer sheets provided 1or answers.

,at your name in the upper right-hand corner of the first page of each -

tion of the answer sheet.

arate answer sheets and place finished answer sheets face n on your desk or table, abbreviations only if they are commonly used in facility literature.

I point value for each question is indicated in parentheses after the stion and can be used as a guide for the depth of answer required.

a all calculations, methods, or assumptions used to obtain an answer '

. mathematical problems whether indicated in the question or not.

tial credit may be given.

Therefore, ANSWER ALL PARTS OF THE .

STION AND DO NOT LEAVE ANY ANSWER BLANK. NOTE: Partial credit <

1 NOT be given on multiple choice questions.

w parta of the examination are not clear as to intent, ask questions of examiner only, must sign the statement on the cover sheet that indicates that the n is your own and you have not received or been given assistance in pleting the examination. This must be done after the examination hau a completed.

I

3

= .

45,8 ,

b .:

l i tion, you shall:

' ,/ comp ete your exam na *

} (*,Q,.mble your examination as follows:

] p) Exam questions on top.

i f i ./2). Exam aids - figures, tables, etc.

l

'(3) Answer pages including figures which are part of the answer. ;,

\ ~

Turn in your copy of the examination and all pages used to answer the examination questions.

Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.

Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.

e 14 h

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~, . "

U. S, NUCLEAR REGULATORY' COMMISSION f* SENIOR REACTOR OPERATOR LICENSE EXAMINATION l r REGION 2 i

FACILITY: E. I. Hatch 1 & 3 J REACTOR TYPE: BWR-GE4 DATE ADMINISTERED: 89/06/12 INSTRUCTIONS 'LO CANDIMIEL Use separate paper for the answers. Write answers on one side only.

Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.

% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE_ TOTAL SCORE _VALUE CATEGQEY 72,00 21.77

-24.00 -Ber4R 4. REACTOR PRINCIPLES (?%)

THERMODYNAMICS (7%) AND COMPONENTS (10%) (FUNDAMENTALS EXAM) u c. .ro a - -G st 58- 5. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS (33%)

TI.~10

_1&uai__ -4E-42 6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC RESPONSIBILITIES (13%)

96.7,c" 404-4'r % TOTALS FINAL GRADE All work done on this examination is my own. I have neither given nor received a$d, Candidate's Signature j

NRC RULES AND GUIDELINES 70R LICENSE EXAMINATIONS During the administration of this examination the following rules apply:

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

2. Restroom trips are to be limited and only one candidate 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.

3. Use black ink or dark pencil only to facilitate legible reproductions.

4. Print your name in the blark provided on the cover sheet of the examination.

5. Fill in the date on the cever sheet of the examination lif necessary).

6. Use only the answer sheets provided for answers.

7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.

8. Separate answer sheets and place finished answer sheets face down on your desk or table.

9. Use alibreviations only if they are commonly used in facility literature.

10. The ?oint value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.

11. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.

12. Partial credit may be given. Therefore, AN3WER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER ELANK. NOTE: Partial credit will NOT be given on multiple choice questions.

13. If parts of the examination are not clear as to intent, ask questions of the examiner only.

14. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.

l l

_ - _ - _ _ i

O., O

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a.s .:

r15'. When.you complete your examination, you shall: i La. Assemble your examination as follows:

(1) Exam questions on top.

(2) Exam' aids - figures, tables, etc.

(3) . Answer pages including figures which are part of the answer.

b. Turn in your copy of the exam r nion and all pages used to answer the examination questions.

.c. Turn in all scrap paper and the balance of the paper that you did not use:for. answering the. questions,

d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.

l h

e__-_-______ -_- _ i

~- ~

7 gy AND COMPONENTS Il0%i~~TFDNDXMENTALS EXAlu h

L QUESTION- 4.01 (1.00)

[ . WHICH ONE of the following is correct concerning neutrons.

L a. The percentage of delayed neutrons produced from fission i increases as the age of the core increases.

.. b. The energy level at which delEyed neutrons are produced L categorizes them as thermal neutrons.

c. Neutrons produced from the moment of fission until 10 E-14 seconds are' considered prompt.

d. Delayed neutrons are produced ONLY as a result of thermal fission of U-235.

QUESTION 4.02 (1.00)

WHICH ONE of the following statements best describes the delayed neutron fraction, beta?

a. The delayed neutron fraction is the ratio of the number of neutrons born delayed to the number of all neutrons which are born from fission.

b. The delayed neutron fraction is the ratio of the neutrons born from delayed precursors to the number of thermal neutrons absorbed in the fuel.

c. The delayed neutron fraction is the ratio of the neutrons born from delayed precursors in the last generation to the number of neutrons born from delayed precursors in the current generation.

d. The delayed neutron fraction is the ratio of thermal neutrons absorbed in the fuel to all thermal neutrons which are absorbed.

(***** CATEGORY 4 CONTINUED ON NEXT PAGE ****4)

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

if%IMONENUd%UIFijtQldERTlisS EXAMl

~

1 QUESTION 4.03 (1.00)

WHICH ONE of the following statements is correct concerning the relative magnitudes or characteristics of the moderator coefficient, the void coefficient, and the doppler coefficient?

a. Although the moderator coefficient and void coefficient behave similarly, the void coefficient has 10 times the effect on reactivity because it is 10 times larger.

b. The-doppler coefficient reacts much faster than the moderator coefficient to add negative reactivity after reactor power transients and is approximately 10 times larger.

c. During a rapid power increase both void formation and fuel temperature increases insert negative reactivity, but the doppler coefficient has a greater effect because it is 100 times larger than the void coefficient.

d. The doppler coefficient becomes more negative with fuel temperature increase until it is 10 times smaller than the void coefficient.

QUESTION 4.04 (1.00)

WHICH ONE of the statements below describes a deep rod?

a. A control rod used to shape the thermal flux distribution of the core.

b. A control rod with limited effect on radial power distribution due to rod shadowing.

c. A control rod inserted greater than 2/3 into the core.

d. A control rod in a core with a rod density of 25% or less.

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

(7%) THERMODYNAMICS Page 6 g JENTS (10%) fFUNRAMENTALS EKAM1

%. a

(? l 7

? q. = 4.05 .

(1.00) 4?jgEo [ f the following correctly describes the formation and fj processes at equilibrium for Xe-1357

~

g (FP + I-135 burnout) - (Xe-135 decay + Xe-135 burnout) i 8.1(FP + l-135 decay) - (Xe-135 decay + Xe-136 decay)

3. (FP + Cs-135 decay) - (Xe-135 decay + Xe-135 burnout)

1. (FP + I-135 decay) - (Xe-135 decay + Xe-135 burnout)

)N 4.06 (1.00) statement explains why burnable poisons are loaded in a BWR Loading burnable poisons allows more fuel to be loaded without increasing the number of control rods required.

Loading burnable poisons provides a more controlled initial reactor startup when fission product poisons are not available to control thermal flux.

Loading burnable poisons provides a wider shutdown margin early in core life before Samarium concentrations reach equilibrium conditions.

Loading burnable poisons protects fuel pellets from excessive peak centerline temperatures in high power regions by limiting the thermal neutrons available for fission.

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

=-- _,_ _-- -.

,; , (7%) AND COMPONENTS (10%) (FUNDAMENTALS EXAM) lQUESTIONi 4.07-' (1.00)

SELECT the correct answer.

' Assume a constant change in reactivity for each rod movement during an approach to criticality. As the reactor approaches criticality a..the period will not return to infinity due to the contribution-of non-fission neutrons,

b. the time for period to stablize following each rod movement will increase, c..the increase in counts for each rod movement is constant.

d.-the change in moderator temperature will be constant for each rod movement.

QUESTION 4.08 (1.00)

WHICH ONE of the following statements is the definition of steam quality?

a.-Mass fraction of vapor present in a mixture of liquid and vapor, expressed as a percent, b.: Mass fraction of liquid present in a mixture of liquid and vapor, expressed as a percent.

c. Mass ratio of vapor present to liquid present in a mixture of liquid and vapor, expressed as a percent,

d. Mass ratio of liquid present to vapor present in a mixture of liquid and vapor, expressed as a percent.

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

l L_ _

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{ i7s FnND CDMPbNENTS (iO5FIFUNDAMENTALS EXAM)-

-QUESTION 4.05 (1.00)

WHICH.ONE of the'following correctly describes the formation and removal processes at equilibrium for Xe-1357

a. (FP + I-135 burnout) -

(Xe-135 decay + Xe-135 burnout)

b. (FP +-I-135 decay) - (Xe-135 decay + Xe-136 decay)

c. (FP + Cs-135 decay) - (Xe-135 decay + Xe-135 burnout) d.'(FP + I-135 decay) - (Xe-135 decay + Xe-135 burnout)

QUESTION 4.06 (1.00)

WHICH statement explains why burnable poisons are loaded in a BWR core?

a. Loading burnable poisons allows more fuel to be loaded without' increasing the number of control rods required.

b. Loading burnable poisons provides a more controlled initial reactor startup when fission product poisons are not available to control thermal flux,

c. Loading burnable poisons provides a wider shutdown margin early in core life before Samarium concentrations reach equilibrium conditions,

d. Loading burnable poisons protects fuel pellets from excessive peak centerline temperatures in high power regions by limiting the thermal neutrons available for fission, i

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(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

fi [], [7NANDTCOMPONUEOMM5hiiEIUS EXAML 1

g m^;

QUESTION 4'.07 .(1.00)

<:: SELECT:the correct' answer.

'EesumeJa constant change in reactivity for-each rod movementL during an. approach to. criticality. As.the~ reactor: approaches

. criticality:

- a. the period will;not return to infinity due to-the contribution:

of non-fission neutrons.

ib.'the~ time for. period to'stablize-following each rod movement will-

. increase,

c. theLincrease'in counts for each rod movement is constant.

d.:the-' change'in' moderator. temperature will be constant for each-Jrod movement.

>>?'Q ESTION- 4'.08 - ( 1. 0 0 ) -

UHICH'ONE of'the_following' statements is the definition of steam quality?

~

- a. Mass fraction of vapor present in a mixture of liquid and vapor,

'expressedLas'a' percent.

.b. Mass' fraction of' liquid;present'in a mixture of liquid'and vapor,

expressed.as a percent.

c. Mass ratio'of' vapor present to liquid present in a mixture of

liquid and vapor, expressed as a percent.

d. Mass ratio of liquid present to. vapor present in a mixture of liquid and vapor, expressed as a percent.

1^

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-(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

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____-m__. _ - . _ _ _ _ . _ _ _ . _ _ . _ _ _ _ - _ - _ - _ _ _ _ _ _ _ - - - - - - -

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L, ,.J(7%i AND COMP 6NENTS 110%i FUNDAMENTALS _EXAdl L

1 l-QUESTION 4.09 (1.00)

(

, SELECT the' correct statement concerning the operation of a jet pump.

I a. The nozzle on the jet pump converts the static head of the recirculation pumps to a low velocity jet at high static l pressure.

1 1 b. The diffuser increases the velocity of the fluid thereby I increasing its static head,

c. The constant area of the mixing section maintains the pressure constant.

d. The low pressure at the nozzle discharge draws the surrounding fluid into the jet pump throat area.

I 1

QUESTION 4.10 (1.00).

A feedwater heater is isolated during operation. The effect on cycle' efficiency is that:

a. cycle efficiency increases due to increased steam flow through low pressure turbine stages.

b. cycle efficiency decreases due to additional heat rejected by the condenser.

c. cycle efficiency decreases due to additional work required by feed-pumps to pump cooler feedwater,

d. cycle efficiency decreases due to increased subcooling of the condensate from the condenser.

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

e_______---_-_-___-__. . - _

,s, 1 -17%) AS.D COMPONENTS (10%) (FUNDAMENTALS EXAM 1

QUESTION 4.11 (1.00)

SELECT the. item that describes pump' head.

CHOICES:

a. The power used by.a motor driving a pump,

b. The temperature rise across a pump.

c. The pressure a pump adds to a fluid.

d. The pressure at the discharge of a pump.

QUESTION 4.12 (1.00)

WHICH ONE of the following is the best description of Departure from Nucleate Boiling (DNB).

a. The point of heat addition in which the energy from a' heated surface submerged in water must pass by radiation through a stagnant film.

'b. The point of heat addition in which bubbles formed on a heated surface submerged in water become so dense they combine and form a vapor film over the heated surface.

c..The most efficient method of heat transfer of a heated surface to its environment.

d. The heat transfer method in which energy from e heated surface is directly transferred to the fluid it is submerged in. i

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

[ (79si AND COMPONENTS (10%) (FLlblDAMENTALS EXAlil y

h L

p l QUESTION 4.13 (1.00) y' WHICH ONE'of'the following' conditions would INCREASE.the Critical Power level assuming all other variables remain unchanged?

i. ; -. NOTE: ASSUME NORMAL FULL-POWER OPERATING CONDITIONS

a. . Inlet subcooling is DECREASED

b. Reactor pressure is INCREASED

c. The axial power peak is RAISED

d. Coolant flow rate is INCREASED QUESTION 4.14 (1.00)

WHICH ONE of the following best explains the basis for the limiting condition of Linear Heat Generation Rate (LHGR).

a. Operating the reactor at power levels less than the LHGR limit will ensure 1% plastic strain on the clad is r.ot exceeded,

b. Operating the reactor at power levels less than the LHGR limit will ensure the amount of hydrogen generated during a DBA will not exceed the explosive concentration limit.

c. Operating the reactor at flux levels below the LHGR limit (s) will ensure the on set of transition boiling will not occur in greater than 1% of the core,

d. Operating.the reactor at power levels below the LHGR limit assures the core spray system can cool the core during/after a LOCA.

I

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

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[ 5,li%)~ ANDMENTS QO%) (F M LS___ EXAM) i

-QUESTION 4.15 (1.00)

-l For throttling, globe valves are preferred over gate valves because:

a. the loss coefficient is generally more linear for globe valves than for gate valves.

b. the loss coefficient for a full open globe valve is smaller than the loss coefficient for a full open gate valve.

c. valve position indication for a midpositioned valve is more reliable for glove valves than for gate valves.

d. valve motor operators are more adaptable to globe valves than to gate valves.

QUESTION 4.16 (1.00)

Thermal overload devices are used to protect circuits in nominal overload conditions. WHICH ONE of the following defines a thermal overload device?

a. A balanced circuit that compares normal current to a fixed overcurrent; exceeding the normal current trips the overcurrent relay.

b. An in-line thermal coil that, when subjected to a high current, overheats and actuates a circuit-interrupting device.

c. A temperature monitor that senses the temperature of the operating equipment and trips the circuit breaker if the temperature exceeds preset limits.

d. An induction coil that generates a secondary current to the primary current, closing the trip circuit contacts.

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

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_, l[(7%) AND COMPOBENTST(10%F (FUNDAMENTALS EXAM 1 QUESTION 4.17 (1.00)

REFER'to FIGURE 1.

-The_ purpose-of the seal-in' relay is to:

3

a. hold the valve open even if the reset pushbutton is pressed.

b. hold the valve open even if'the initiating conditions reset,

c. close the valve as soon as either initiating condition resets. ,

-d. close the valve only when both initiating conditions reset.

QUESTION 4.18 (1.00)

SELECT the statement that describes operation of centrifugal pumps during low flow conditions.

a. Operation at low flow can cause overheating due to viscous friction therefore a minimum flow valve should open during low flow' conditions.

b. Operation at low flow can cause excessive motor temperatures due to excessive current therefore motors for centrifugal pumps should be equipped with additional cooling during low flow conditions.

I

c. Operation at low ~ flow can result in excessive pressure at the discharge of the pump therefore all centrifugal putaps should be ,

equipped with a high pressure trip.

d. Operation at low flow will not result in cavitation of the pump, therefore maintaining sufficient NPSH is not required. '

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(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) I l 1

(-

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-p gi AHD COMPONENTS 710si~lFUNDAMENTILS EXAM)

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1 i

QUESTION- .19 (1.00) .

i The average starting current for an alternating current motor that 'l Lprovides; power for a centrifugal pump is approximately: Assume the i pumps discharge. valve is closed. .f 2:

.]

a. the same as its normal running current 1

b. two times its normal running' current ~ <

c. five times its normal running current

d. ten times its-normal running current

, CralM!

QUESTION -

4x20 (1.00)-

WHICH ONE of the following. conditions would result in the INCREASE in the outlet temperature of the oil-in a lube oil system.

a. Additional loads are placed on the cooling water system.

b. The bypass valve for the oil system is closed.

a

c. A flow blockage occurs in one of the tubes on the lube oil side of the heat exchanger,

d. The outlet valve for the cooling water is opened.

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(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

i L__.____ l

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jy [ ]7% FAND COMPONENTS i10siTIFUNDAMENTALS'EXAMJ>

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g. ,

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9 ..I _ .' ; : '[- . ., . . . . , . _ . m . . . . ,

.,m x ,~ (QUESTION. :4.21 (1.00){

r o &,. . . ..

.3 iWHICH(ONE:ofLthe.followingLis the purposeiof.the. ion. exchange

.y process.in theicondensatelfilt'er demineralizers, i- ,,.. . J ,

If LIA1 crease silica ~: co'ntenthof-the; condensate.

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5 b. : / Remove suspended-solids..

.. c. .-* (

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, dc. Maintain pH of the condensate. .

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.di ' Decrease.oxygenicontentlof.the condensate.

y o .

. QUESTION '4.22 . (1. 00 ): .

I t.fh [Cl ' ios ngtai. circuit breaker.between.two electrical generating systems-

thatLare:out..of phaseLwill resultiin:

~

. a. negating"theJreverse power. protection on the lagging frequency f system.

?b.ca' rapid Phase alignment which could damage generatorsJand loads iconnectedito;theLsystem.

c. a voltiage reduction in both generating systems.

d. a: reduction of the frequency of the' combined' generating system.

T

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(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

f

g "' X T7sTEAND7COMPONENTS T10 W FUNDAMENTALS EXAM 1- 1 W .; , ;;g Jil ...

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'. f QUESTION.' 4;23:: -(2.00)~

k :k '

l how:the'-fo11owing: conditions 1.will' affect ~ INDICATED reactor kA, L. STATE ivasselEwater levelf(INCREASE,-DECREASE, FAIL UPSCALE, FAIL DOWNSCALE, .

c.g;; REMAIN THE SAME)..

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}, La.'Referencei-leg-. rupture

c Lb.-: Drywe111 temperatures. increase by 80LF'

c. Detector. equalizing valve-leaks but' pressure does not equalize 3 :> 1d.1 Variable' leg: rupture: y (i

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(***** END OF CATEGORY 4 *****)

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7,43% F QUESTION- 5.01 (1.00).

Afloss of both' recirculation pumps has occurred with the plant -f operating at 75% power'and 100% rod line. Immediate operator action r; quired is:

a. Place the mode' switch in shutdown.

-b. DriveEcontrol rods until power'is below the 80% rod line,

c. Commence a normal reactor shutdown.

d. Immediately restart a recirculation pump.

DoIchoJQ QUESTION p.<d2 (1.00)

A reduction in recirculation pump speed has occurred while operating ct-power. SELECT the statement that provides the correct actions b: sed on the given conditions.

a. Three LPRMs.in one quadrant are oscillating at a 15% bandwidth so rods should be inserted in that quadrant to suppress the oscillations.

b. One APRM is oscillating at a 12% bandwidth with the others oscillating at a 7% bandwidth so recirculation flow should be increased to suppress oscillations,

c. Two LPRMs in opposite quadrants are oscillating at a 15%

bandwidth so the reactor should be scrammed.

d. All observed LPRMS are oscillating at a 7% bandwidth so recirculation flow should be increased to suppress the oscillations.

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

)

L _- - __-_

t.. ,ET33%i 1

,,. Do!# EM QUESTION (1.00) 1 On.a loss of Instrument and Service Air, a loss of condenser vacuum will occur, This is caused by:

'a.'a: loss of' circulating water.

b. increasing hotwell level.

c. a. closure of the SJAE suction valve (F004)

d. a closure of the SJAE steam supply valve (F008)

. QUESTION 5.04 (1.00)

.A loss of 120/208 V Distribution Cabinet 2A - Instrument Bus 2R25-S064 has occurred. SELECT the' correct statement concerning how to monitor

. Plant. parameters.

a. Recorders should be monitored because of the ability to compare independent indications.

b. Meters.should be used due to some meters being powered from DC sources.

c.. Meters should be used because recorders will fail as is on a loss.of. power.

d. Meters should be used because they will be easier to read during

.the casualty.

i I (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

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'.5.

EMERGENCY AND ABNORHAL PLANT EVOLUTIQEE (33%)

Pcgo 18 i

QUESTION 5.05 (1.00)

SELECT the statement that identifies the;most significant contributor )

to reducing power when water level is lowered during a failure to l scram event. !

a. Lowering level below the moisture separator removes the flowpath thereby minimizing flow through the core.

b. Lowering level reduces the pressure in the core by reducing the head of water above the core.

c. Lowering level reduces the differential pressure between outside the shroud and inside the core.

~

d. Lowering level reduces power by increasing the subcooling of the water entering the core.

QUESTION 5.06 (1.00)

The reactor 1:5 operating at 70% power and 58% flow. The operator observes a lons of feedwater heating. Reactor power is to be immediately reiduced to:

a. below the 80% rod line by driving control rods.

b. a thermal power of 56% by reducing recirculation flow.

c. a thermal power of 50% by reducing recirculation flow.

d. the minimum thermal power capable without reducing core flow below 45%.

, ATTACHMENT 1 FROM 34GO-OPS-005-25 IS PROVIDED.

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L---- - -- _ ---

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i- , QUESTION 5.07- _ (1.00).

1' A manual scram has-been' inserted on both channels and the rods have.

failed to insert. All blue scram lights are illuminated. IDENTIFY the cause for the failure to scram.

a. Hydraulic lock in the scram discharge volume.

.b.. Failure of the scram valves to open.

c. Blockage in-the scram air header.

d. Failure of one RPS scram trip system.

QUESTION 5.98 (1.00)

IDENTIFY the correct statement concerning control of systems from the remote shutdown panel (s).

a. The interlock preventing simultaneous opening of the RHR suction valves for shutdown cooling and from the torus are bypassed when operated from a remote shutdown panel.

b. The low low set function is not operable when SRVs are transferred to a remote shutdown panel..

c. Containment-isolation functions are operable for systems operated from a remote shutdown panel.

d. The only function of the Unit 2 RCIC trip throttle valve.at a remote shutdown panel is the reset function.

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QUESTION 5.09 (I.00)

, SELECT.the correct statement'concerning an unrecoverable loss of RBCCW due to*a leak in the system while operating in mode 1.

a. The standby RBCCW pump will not start, b, The reactor.should be scrammed within 2-3 minutes,

c. Entry into-EOP flow charts is not required because they refer to the Loss of RBCCW abnorma1' procedure to correct the RBCCW-conditions,

d. If a scram is required then the Abnormal Procedure for Loss of RBCCW should be exited and the the EOP flowcharts entered.

QUESTION 5.10 (1.00)

All service air compressors have tripped. Pressure in the entire' air system is 83#, WHICH ONE of the following actions should have occurred,

a. Standby Instrument Air dryer starting.

b. Startup flow control valve locked in present position.

c. Reactor Building Instrument Nitrogen to Non-Interruptable service isolation valve opening.

d. Service Air Shutoff valve closing.

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,. QUESTION 5[ (1.00)

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During an ATWS, several actions are taken to mitigate heat addition to the suppression pool. SELECT the correct statement concerning

, O .. implementation of these actions.

a. Boron injection is commenced if temperature approaches 110 F to ensure that the Technical Specifications limits are not exceeded.

b. If power is at 100% then immediate' tripping of the recirculation

. pumps from their present speed is necessary reduce heat added to the suppression pool.

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c. If, due to a failure of the SLC system, injection of boron is delayed, then lowering of level should be delayed until boron is injected,

d. Level reduction is required if power is > 3%, suppression pool temperature is > 110 F, and an SRV is open. Level should be reduced until either level is at top of active fuel (TAF) or all three conditions a e corrected.

QUESTION 5.12 (1.00)

SELECT the correct statement concerning use of the Heat Capacity Temperature L.imit and Heat Capacity Level Limit.

a. The Heat Capacity Level Limit can be complied with for all suppression pool levels by reducing reactor pressure.

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b. The Heat Capacity Level Limit is less restrictive than the Heat Capacity Temperature Limit when suppression pool level is less than 146 inches.

l c. The Heat Capacity Temperature Limit is allowed to be increased at suppression pool levels higher than 146 inches,

d. The Heat Capacity Level Limit requires that the emergency depressurization be accomplished before the downcommer vents are uncovered.

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-QUESTION. 5.13 (1.00) l Refer to figure 2.

SELECT the correct statement.

The general caution in the Emergency Operating Procedure lists a drywell temperature and level indication for each RPV level instrument. If drywell temperature is greater than the value stated, then level indication is not valid below the value stated for that instrument. The level is not valid because:

a. at. low reactor pressures the variable leg will flash at temperatures greater than the stated drywell temperature.

b. at'high drywell temperatures variable leg density will decrease causing invalid readings.

c. at low indicated levels reference leg density causes on scale indications with level below the instruments monitoring range.

d. at low drywell temperatures the reference leg will cause erroneously high indicated levels.

QUESTION 5.14- (1.00)

The emergency depressurization procedure directs the operator to the alternate depressurization procedure if suppression pool level is below 58".

IDENTIFY the reason for this action.

a. The bottom of the downcomers is at 58".

b.-The safety relief valves discharge at 58".

c. At 58" suppression pool level has insufficient heat capacity for depressurization.

d. At 58" level is below the suppression pool temperature detectors.

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41QUESTIONT . 45l151 E(1.00)-

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Unde'r,whichhSf the foilbw3ng conditions would 34-AB-EOP-049,-

sysy ' Radioactivity' Release Control, require EmergencyLDepressurization of

Ethefreactor? ,

c.Lany!rel' ease.in.exceas of 1000 mr/hr.

.b. any.' release'in-excesafof'1000 mr/hr outside the primary and a secondary contsiment1that cannot be isolated.

c.Jany/releasetin excess of11000 mr/hr outside the primary or

. secondary containment.-

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,d.4any release in excess of 1000 mr/hr into-the primary or secondary containment and one area. exceeds max safe operating level.

-- QUESTION; 5.16.' .(1.00)

'ntaiiosa O ofi12'5/250 V DC Switchgear 2A, 2R25-S001, the operator'is to-open tte main" generator. output breakers,

a. STATEithe: maximum time limit allowed for, opening the output ibreakers ,t

.b': STATE'the' reason that the output breakers must be opened.

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. QUESTION 5.17 bar50) i For each power source in Column I below (a-f) _ MATCH the required

immediate action listed in Column II (1-6) that is required to be take if that power source is lost. More than one immediate action may be rcquired for a power source.

Column'I' Column II

a. 125/250 V DC Switchgear 2A 1. Transfer 4160 V buses 2A and 2B 2R22-S036 to Startup Supply,

b. 125/250 V DC Switchgear 2B 2. Manually trip Reactor Reciro 2R22-S017 Pump B.

c. 125 V DC Cabinet 20, 3. Secure radwaste discharge.

2R25-S003

.d. 24/48V DC Cabinent 2B, 4. Manually open Main Generator 2R25-S016 output breakers.

5. None 7,ESTION 5.18 (2.50)

Given the following general EOP conditions, STATE the correct flow path (by number) to follow, which is designed to address the following

' conditions, in accordance with 31EO-EOP-001-2S, " Emergency Operating Procedure Inside Control Room Unit 2."

a. High radiation, loss of coolant, loss of control of primary containment integrity (under degraded conditions).

b. High steam line radiation, loss of vital power, failure of vital equipment.

c. Malfunction of Reactivity Control System without AC power.

d. Malfunction of Reactivity Control System with AC power.

e. Reactor transients or failure of vital equipment without degraded conditions.

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QUESTION 5.19 (1.00)

A. genera 1' caution in'the.EOPs states " Shut MSIV's when reactor-water level exceeds 100 inches." 'This is to prevent the flooding of the main steam lines. STATE TWO possible causes of damage if the main steam lines are flooded.

LQUESTION '5.20 (2.50)

Concerning Unit 1 Technical Specifications.

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'During. normal-power operation, the suppression chamber water temperature'shall be maintained less than or equal to (a) [ temperature). If this temperature limit is exceeded,

. pool cooling shall be initiated immediately.

During' relief valve operation or testing of RCIC, HPCI, or other testing which adds heat to the suppression. pool, the maximum water Temperature shall not exceed. (b) [ temperature]. In connection Lwith such testing, the pool temperature must be reduced within (c) [ time limit] to less'than or equal to (d) [ temperature].

The~ reactor shall be (e) [ action] from any operating condition when'the suppression pool temperature exceeds 110 F.

-QUESTION 5.21 (1.00) 1 1

. Actions to mitigate reactor power oscillations at high power / low flow conditions are taken.to prevent exceeding which thermal limit?

5.22 (1.00)

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-QUESTION The plant'is in shutdown cooling with the reactor head still in place when a loss of shutdown cooling occurs due to a loss of RHR Service Water. 'Significant decay heat exists.

l If reactor: pressure increases to greater than (1) then I (2) valve (s) will close. (Valve name(s) or number (s) are ecceptable.)

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QUESTION 5.23 (2.00)

Concerning the Drywell Spray Initiation Limit-

c. DESCRIBE the negative effects, including the cause, of sprayir.g . the drywell when drywell parameters are in the unsafe region of the ,

..Drywell Spray Initiation Limit.

(1.0) I

b. STATE the parameters monitored by the Drywell Spray Initiation Limit. (1.0)

QUESTION 5.24 (2.00) i LIST FOUR pressure control methods used by SOFI Flowchart 3.

-QUESTION 5.25 (1.50)

c. For steam cooling the operator is directed to open one safety relief valve when reactor level decreases to 1/3 core height. This action.will' establish cooling to maintain peak clad temperature less than . (0.5)

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b. STATE TWO methods of achieving adequate core cooling other than steam cooling as stated in the definition of adequate. core cooling.

O.0)

QUESTION 5.26 (2.00) i During performance of the Secondary Containment Temperature Control i 125 procedure the operator is directed to Emergency Depressurize if a primary system is discharging into an area and area temperature / differential temperatures exceed the Maximum Safe l Operating temperature 3evel in more than one area.

STATE TWO purposes for performing this action.

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( L._ .[RE5FOWIFdTIEU35 QUESTION 6.01 (1.00)

Technical'Specificativos regarding Recirculation System Jet Pump OPERABILITY have some very restrictive Limiting Conditions for l

Operation (LCO) if a jet-pump is found to be INOPERABLE. From.the

- choices below, SELECT the. concern regarding continued plant operation with an inoperable (or failed) jet pump.

a. Invalid APRM Flow Biased SCRAM setpoints due to.the change in flow through a failed jet pump

b. Increased blowdown area during-a Loss of Coolant Accident (LOCA)

c. Unbalanced neutron flux across the core due to flow variations

d. Physical core damage from a piece of a damaged jet pump i

QUESTION 6.02 (1.00)

The operating procedure for the High Pressure Coolant Injection (HPCI) system cautions against prolonged operatior. with turbine speed less than 2000 RPM. SELECT the reason why this is o. concern.

a. At low turbine speeds the potential exists for exhaust check valve chatter and reduced oil flow / pressure to the turbine governor and bearings.

b. At low turbine speeds the Booster Pump may not provide adequate net positive suction head to prevent cavitation of the main pump.

c. The rate of steam flow through the HPCI turbine may not be enough to prevent it from overheating.

G. At low turbine speeds, cooling water flow from the Booster Pump to the lube oil system heat exchanger is inadequate.

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QUESTION 6.03 (1.00)

When the Standby Liquid Control (SBLC) system is initiated, WHICH ONE of l the following. occurs:

a. The selected pump starts and only its associated (divisional) squib valve fires to provide a flow path to the reactor,

b. The SBLC storage tank and pump suction line heat tracing heaters start to ensure all the sodium pentaoorate remains in solution for injection into the reactor.

c. Both SBLC pumps start and both squib valves fire to provide two separate flow paths of sodium pentaborate into the reactor.

d. The Reactor Water Cleanup (RWCU) system outboard containment isolation valve (G31-F004) closes to isalste the system from the reactor.

QUESTION 6.04 (1.00)

SELECT the set of trip channel conditions (a - b) from the Reactor Protection System shown in the diagram below that will result in a HALF SCRAM. (assume a "one-out-of-two taken twice" logic)

TRIP SYSTEM TRIP SYSTEM A B l l l l l l ! l L Al A2 B1 B2 TRIP CHANNELS TRIP CHANNELS

a. Al and A2 tripped I b. A1 and B2 tripped

c. A2 and B1 and B2 tripped

d. A2 and B1 tripped 1

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L -QUESTION 6.05' (1,00)

R:garding the Reactor Core Isolation Cooling (RCIC) system, which of the following is NOT a function of the Suppression Pool?

a. Receives the discharge of the Vacuum Pump

b. Acts as a heat sink for the RCIC Turbine exhaust u

c. Receivee drainage from the RCIC Turbine steam line and exhaust line drain pots

d. Provides backup source.of water for the RCIC Pump I

QUESTION 6.06 (1.00)

' UNIT 2 has just experienced an initiation of the Automatic

.Depressurizaton System (ADS). Plant conditions are as follows:

Drywell-pressure: 3.2 psig Reactor-water level: -147 inches All RHR pumps: running 13 minute timer: timed out 120 second timer: timed out 7 ADS SRVs: open Main Steam pressure: 150 psig and lowering WHICH of the following will cause the ADS SRVs to close? (Consider each enswer separately)

a. The RHR/LPCI mode raises reactor water level to -20 inches.

b. Reactor pressure reaches 40 psig.

c. Drywell pressure decreases to 1.0 psig and the High Drywell Pressure Seal-in push button is reset by the operator,

d. 3 of the 4 RHR pumps are secured.

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l QUESTION 6.07 (1.00) l The OUTBOARD Main Steam Isolation Valve Leakage Control System (MSIV-LCS) MAY be manually initiated when WHICH of the following L SPECIFIC sets of plant conditions exist:

1 L a. Reactor pressure is less than 35 psig, all of the MSIVs are closed and Main Steam Line pressure between the Outboard MSIVs and the Main Turbine Stop Valves .4.s less than 35 psig.

u b. Reactor pressure has been less than 35 psig for a minimum of 10 minutes, all of the MSIVs are closed and Main Steam Line pressure between the Inboard MSIVs and the Outboard MSIVs is less than.35 psig.

c. At least 10 minutes have. elapsed since the Loss of Coolant .

Accident (LOCA), the Outboard MSIVs are closed and pressure I between the Outboard MSIVs and the Main Turbine Stop Valves and Bypass Valves is,less than 35 psig.

d. At least 10 minutes have elapsed since the Loss of Coolant Accident (LOCA), all of-the MSIVs are closed and pressure between the Inboard MSIVs and Outboard MSIVs has bled down to O psig.

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-QUESTION 6.08 (1.00)

The Main Condenser Mechanical Vacuum Pump is used to draw the initial vacuum up.to a maximum of 5% power. SELECT the basis for this power limit.

a. The Mechanical Vacuum Pump does not have the capacity to draw sufficient vacuum to clear the Main Turbine and Reactor Feed-Pump low vacuum trips,

b. At powers above 5%, the amount of steam in the Main Condenser shortens the life of the seals in the Mechanical Vacuum Pump.

c. The Mechanical Vacuum Pumps are designed to draw a vacuum up to a Main Steam Pressure of 600 psig which corresponds to 5% power on a cold startup,

d. At powers above 5%, the amount of radioactivity being removed from the Main Condenser cannot be released to the environment and Hydrogen levels in the Condenser at this point could cause explosions from the heat in the Mechanical Vacuum Pump l

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' QUESTION. 6.09 (1.00)

< SELECT the FINAL; plant conditions after the loss of the "A" Reactor Feed Pump from 100% power. Assume no operator action and.all systems work as designed.

a. Reactor-Power 50-55%

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Reactor Water Level at normal level Recirculation Pumps at 44% speed "B" Reactor Feed Pump speed. higher than before the trip of "A"

b. Reactor Power 50-55%

Reactor Water Level at normal level.

Recirculation Pumps at 22% speed "B" Reactor Feed Pump speed the same as before the trip of "A"

c. Peactor Power 60-65%

Wractor Water Level below normal leve)

Accirculation' Pumps at 22% speed "B" Reactor Feed Pump speed lower than before the trip of "A"

d. Reactor Power 60-65%

Reactor Water' Level at normal level Recirculation Pumps at 44% speed "B" Reactor Feed Pump speed higher than before the trip of."A" QUESTION 6.10 (1.00)

SELECT the statement that describes the primary objective of the Fire Protection program at E. I. Hatch Nuclear Plant.

a. The primary objective is to protect Safety Related equipment and areas within the plant.

b. The primary objective is to minimize the spread of fires within the Protected Area of the plant.

c. The primary objective is to minimize both the probability and consequences of postulated fires,

d. The primary objective is to provide early detection of fires and automatic protective actions.

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0 m4 N). QUESTION 6ilij l(0. 50 ) ,

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fLIST'the: electrical busses supplying power'to the;"A" and;"B"' Control"

Rod. Drive (CRD). pumps.

b. '

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, ; QUESTION, !6.12- l(1.00) J w ,

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..Why. Tare. the' ControliRod' Drive (CRD) pum:ps NOT necessary to be in 2

. Loperation..to support the'FULLLinsertion'of.all control. rods-on a reactor .]

"5 ' scram 1withsthe.reactorJat. normal pressure. ;j .

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(QUESTION- 6.13; -(1.00)-

7' I LSTATE l the'purpo'se.offthe. Rod. Worth Minimizer (RWM).

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'QUESTIOND <6.14 :(1.50)

>? Comp 1ste the following statements regarding the Unit 2 Rod Worth

Minimiser-(RWM).-

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zA. :TheLRWM may bermanually bypassed under these TWO-(2) conditions: I

-(1) and (2) .

. B .1 Permission'to manually bypass the RWM must come from (3)

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l D. I The RWM"is automatically. bypassed when above the (6) 9 i

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. . RESPONSIBIhlIIEE_L1331

QUESTION 6.15 (2.50)-

Answer the following questions concerning the operation of the Low Pressure Coolant Injection (LPCI) system:

A. Following a Loss of Coolant Accident (LOCA) signal, what is the automatic sequence of events for the Outboard Injection Valves (FO 17A/B)? (Include in your answer.any applicable time limits for both UNITS, but do NOT include pressure setpoints) (1.0)

B. What plant' conditions will auto start the RHR pumps in this mode?

(setpoints requireu) (0.5)

C. What is the pump start sequence if a concurrent Loss of Off-Site Power (LOSP) occurs with-an injection signal? (1.0)

QUESTION 6.16 (A.00)

UNIT 2 Techr.ical Specifications requires removal of the shorting links during core alterations and shutdown margin demonstrations. BRIEFLY

' describe which Reactor Protection System (RPS) TRIPS this will affect and how it will affect them.

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' QUESTION- 6.171 (2.00)~

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From a comparison of UNIT l'and UNIT 2 reactor SCRAM setpoints, SELECT !

'the SCRAMS from the right-hand column and match them to the' appropriate

'otatements in the left-hand column. (Right-hand column choices may be Lused more than once)

SCRAMS -

A. SCRAM setpoints that 1. Reactor pressure

.are different between the units. 2. Turbine control valve fast closure

3. Scram discharge volume high level B .~ SCRAM (S) that are NEVER 4. Low reactor water level bypassed.

5. APRM High flux (flow biased)

6. APRM High flux C.- SCRAM (S) that are AUTOMATICALLY bypassed 7. High Drywell pressure under'specified plant conditions. 8. Turbine Stop Valva C) aure

9. SRM High Flux l

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'y QUESTION 6.18 (1.20)

Complete the following statements regarding the Intermediate Range Monitoring (IRM) System.

A. The IRM provides neutron monitoring over (1) decades of power as indicated by (2) ranges of indication.

B. IRM provides TWO (2) reactor SCRAMS, (3) and (4) . (NO setpoints required)

C. The IRM detectors are REQUIRED to be fully withdrawn from the core as soon as (5)

D. Reactor period indication is available on IRM by using the (6) on the IRM recorders.

QUESTION 6.19 (2.50)

A. STATE FOUR.(4) Reactor Core Isolation Cooling (RCIC) AUTOMATIC turbine trips including setpoints (DO NOT include the RCIC system isolation turbine trip) (2.0)

B. Which of the RCIC turbine trip (s), manual and automatic, is/are required to be reset locally? (0.5) 1

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L QUESTION- 6.20 (1.50) p

COMPLETE the following:

1. The Automatic Depressurization System (ADS) consists of (number) (a)- of the 11 Main Steam System Safety Relief L- Valves'(SRV).

2. ADS is designed-to. protect the core during a'(size) (b)

-Loss of Coolant' Accident (LOCA) where the (c) system fails to maintain reactor vessel water level.

3.. The-ADS SRV-accumulators are sized-for at least (number)

-(d) valve actuations.

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.. QUESTION 6.21 (0.75)

The Spent' Fuel Pool Cooling and Cleanup System is designed to maintain the Spent Fuel. Pool temperature below certain limits. Given the following plant conditions, STATE the associated temperature limit.

'A. During normal operations

'B. 'During refueling C. 'During core off-load QUESTION 6.22 (1.00)

A. STATE the Technical Specifications setpoint for Main Steam H Isolation Valve (MSIV) closure on Low Main Condenser Vacuum.

(0.25)

B. What are the THREE (3) conditions that must exist to bypass the MSIV closure on Low Main Condenser vacuum. (0.75)

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QUESTION 6'.23 (1.50).

LIST the automatic start signals for the Emergency Diesel Generators.

-Include in your answer the-setpoints for the start signals.

.(Consider each signal separately) i

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' QUESTION 6.24: :(1.80.)

MATCH the following Off Gas. System components with the correct octatements from the right-hand column. There may be more than one !

correct answer for each component and each answer from the right-hand column may be used more than once.

COMPONENTS aL Off Gas Preheater 1. Provides automatic protective action signals to various Off Gas System components ,

l B. , Charcoal Adsorber 2. Provides time for Xenon l and Krypton isotopes to i decay.

C. Off Gas Stack Isolation 3. Is an electric boiler on Unit 2 ;

Valve . !

4. Reduces the levels of Hydrogen ard oxyton in the off-gas flow D. Off Gas Catalytic 5. Will shut on HI-HI-HI radiation Recombiner- levels from the Post Treatment Radiation Monitor

6. Has a design 30 minute. delay time 3 E.. Holdup Volume

7. Will shut on High Hydrogen levels in the off-gas flow F. Post Treatment Radiation 8. Uses 250 psig steam on Unit 2 to Monitor heat the off-gas flow

9. Samples the off-gas flow going up the Main Stack G. Loop Seal Drain Valve

10. Removes Iodine isotopes from the off-gas flow.

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L h ' QUESTION 6.25 (1.00)

SELECT the statement which is NOT in conformance with 30AC-OPS-00-OS, Control of Equipment Clearances and Tags,

a. Whet. pullina fuses.to deenergize power supplies, the DANGER tags j will be attached to blockout fuses if available. i

b. When danger tagging an MOV as a clearance boundary, tagging the breaker and local operator will satisfy the tagging requirements.

c. It is permissible to repack a MOV with the valve on its backseat and a DANGER tag on its local operator, breaker, and control i switch.

d. When danger tagging an MOV which requires independent verification it is permissible for the independent verifier to !

use the valve position indication lights to determine valve position.

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QUESTION 6.26 (1.00)

IDENTIFY which one of the following is acceptable, in accordance with Control-of Operator Aids (DI-OPS-05-1C84N), for use as an Operator Aid.

a. Posting of a pending change to an Emergency Gperating Procedure.

b. Instructions while a procedure is written and approved, i

i c..A caution concerning operation of a component.

d. Instructions for emergency startup of a diesel. z i

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i QUESTION- 6.27 (1.00).

L SELECT the correct statement concerning symbols used on the EOP flow

-charts-.

' Refer-to the symbols on figure 3.

1 a.,The arrow within symbols A and C indicate the direction to be followed if the decision is answered yes.

b. Symbol G is used to direct the operator to an End Path Manual.

c. If the answer to a decision in a symbol C changes then the operator-is to return to the top of the charts,

d. Symboi D is used to identify steps that must be performed in sequence.

QUESTION- 6.28 (1.00)

SELECT the correct statement concerning operating chart recorders in the Main Control Room a.~One hour prior to shift change the off going shift operator shall-mark each chart with the date, time, and his initials.

b. A chart with an incorrect' scale may not be utilised en a Control Room. recorder.

c. On-shift plant operators shall ensure that each completed chart is identified by the recorder number, marked with the date and time it was installed and removed, then forward it directly to Document Control for filing.

d. Used charts are to be given to the Shift Supervisor and periodically forwarded to Document control for filing.

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L.1 .c EESEQNElHILIIIESJ(13%)

QUESTION. 6.29 (1.00) 1 IDENTIFY which,one of the following actions the Emergency Director can

,dalegate.

a. Decis' ion to authorize exceeding overtime limits.

b. Decision to. notify offsite emergency response agencies.

c. Decision to downgrade the emergency classification.

d. Decision to order evacuation of nonessential personnel.

QUESTION ~ 6.30 (2.50)-

.In accordance with 10CFR20.

c. STATE.the quarterly radiation exposure limit for each of the following:

1) Whole body: (0.5)

2) Hands and forearms, feet and ankles (0,5)

3) Skin of the whole body (0.5)

b. The limit' stated in part a for whole body dose may be exceeded if specific conditions are complied with. Answer the following concerning the maximum limits allowed.

1) STATE the maximum quarterly whole body limit. (0.5)

2) STATE two conditions that must be met for an individual to exceed the quarterly exposure limits stated in part a. (0.5)

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j [R5fMIUL5Ti$TTI3X[

QUESTION 6.31 (2.00) s On;back shift, a' fire is reported in one of the maintenance shops:

e. STATE the PRIMARY and ALTERNATE persons (position title) designated.

to assume Fire Brigade Chief. (1.0)

.b. .What is the minimum number of persons who should show up for the fire brigade? (0.5)

c. How'many of these persons should have competent knowledge of safety systems? (0,5)

QUESTION 6.32 (2.00)

You are verifying a valve line-up. STATE how you confirm position of'esch of.the'following

a. ' Closed valve.

b. Open valve.

c. Motor-operated valve,

d. Locked throttle valve.

l l-l l

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

1 l'

L l

p o

A

f;' l RESPONSIBILITIES fisif l ..

QUESTION 6.33 -(2.00)

Match each.of the responsibilities in Column I with the person to whom the responsibility is delegated to in Column II. More than one person in column II may be applicable.

Column I Column II

e. Maintain'the' broadest-perspective 1. Shift Foreman of operational conditions affecting the safety of the plant. 2. Operations Supervisor on Shift i

b. Responsible for compliance with requirements of the operating 3. Plant Operator

' license, Technical Specifications, and approved plant operating 4. Shift Supervisor procedures.

c. Responsible for supervising activities of all persons, from plantJoperator on down, on shift not fulfilling a shift position,

d. Responsible.for performing required surveillance,

e. Responsible for controlling I condu'ct in the control room.

l

(

l

(***** END OF CATEGORY 6 *****)

( * * * * * * * * *

END OF EXAMI NAT ION * * * * * * * * * * )

i

... m 77%) AND'COMPOHEHTS'(10%) ~(FUNDAMENTALS EXAM)'

1y

~

. ANSWER - 4.'01 (1.00) cf[1.0]-

REFERENCE lie'BNR Academics Series, Reactor Theory, Chapter 3. LO 4.1

=3.0/3.1 292001K102 ...(KA's)

ANSWER 4.02 (1.00)

a. [1.0]

LREFERENCE GE BWR Academic Series, Reactor Theory, Chapter 3. LO 4.4 292003K104 ..(KA's)

~ ANSWER. ,4.03 (1 00)

G [1.0]

REFERENCE GE BWR Academic Series, Reactor Theory, Chapter 4. LO 2.1, 2.2, 4.1 and 4.2. 3.3/3.3 292004K114 ..(KA's)

ANSWER 4.04 (1.00) c [1.0]

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

_ = _ _ _ . : _-_ _

y mp fyg-)~AND COMPONENTS B 05i (FUNDAMENTALS EXAM)-

JR5FERENCE j GE'BWR Academic Series, Reactor Theory, Chapter 5 LO 3.3 2.4*/2.5*-

292005K111 ..(KA's)-

-ANSWER 4.05 (1.00) d REFERENCE

'GE BWR Academic Series, Reactor Theory, Chap. 6. LO 2.2, 2.3.

2.9/2.9 2.9/2.9 29200SK104- 292006K103 ..(KA's)

ANSWER '4.06 (1.00) c'[1.0]

-&--4, REFERENCE GE BWR Academic Series, Reactor Theory, Chapter 6. LO 4.1.

2.9/3.1 292007K101 ..(KA's)

' ANSWER 4.07 (1.00) b [1.0]

L L.

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

l-t i

I

{ ]

p'.. ~.Ty7giEAND 1:DMPONENTS T10fi~TFONDXMfNTXLS Ex6H1~

b

-REFERENCE

~GELBWR Academics Series, Reactor Theory, Chapter 7. LO.1.7.

l , ;3.3/3.4 292008K104 ..(KAs)

- ANSWER 4.08 (1.00).

0 [1'.01 REFERENCE GE BWR Academic Series, Heat Transfer and Fluid Flow, Chapter 3. LO 3.1.n

-2.5/2.6-293003K112 ..(KA's)

ANSWER 4.09 '(1.00) d [1.0)

REFERENCE GE BWR Academics Series, Heat Transfer and Fluid Flow, Chapter 8.

LO 8.2 2.7/2.7 293004K105 ..(KA's)

. ANSWER 4.10 (1.00) b [1.0]

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

1

_mi____._...__ ._ __ k

rp ~1. T7%F AND COMPONENTS (10%) (FUNDAMENTALS EXAM 1 REFERENCE GE BWR Academics Series, Heat Transfer and Fluid Flow, Chapter 5.

LO 5.6; 2.7/2.8-203005K105. ..(KA's)

~ ANSWER- 4'.11 (1.00) c-[1.0)

, ~ REFERENCE' GE BWR Academics Series, Heat Transfer and Fluid Flow, Chapter 6.

LO 10.2

'2.5/2.6 293006K107 ..(KA's)

ANSWER 4.12 (1.00)

Lb [1.0)

REFERENCE GE BWR Academics Series, Heat Transfer and Fluid Flow, Chapter 8.

LO 2.7.a 2.9/3.1 293008K108 ..(KA's) l l

ANSWER 4.13 (1.00)

.d [J.0]

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

L-__-________-_-_

771 (7% FAND_ COMPONENTS (10%i (FUNDAMENTALS EXAM) l REF$RENCE G:neral Electric Heat Transfer and Fluid Flow, Chapter.9 LO 5.7

2.9/3.3 2.8/3.2 2.7/3.2'2.6/3.1 293009K122 293009K123 293009K124 293009K126 ..(KA's).

ANSWER 4.14 (1.00)

c. [1.0]

REFERENCE GE'BWR Academics Series, Heat Transfer and Fluid Flow, Chapter 0.

LO 3.3 2.9/3.3 2.8/3.2 2.7/3.2 2.6/3.1

'293009K107 . .. ( KA ' s )

ANSWER 4.15 (1.00) ti [1.03 REFERENCE NRC BWR Fundamentals Exam Bank.

291001K112 ..(KA's)

ANSWER 4.16 (1.00)

b. [1.0]

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

~

f ?_-

f (7% FAND COMPONENTS (10%I TFUNDKMENTAUS EXAM 1

~

l REFERENCE ~

NRC BWR Fundamentals Exam Bank

y 201008K105 ..(KA's)

' ANSWER' 4.17' (1.00) b (1.0)

' REFERENCE sNRC BWR Fundamentals Exam Bank-291003K103 ..(KA's)

ANSWERL 4.18 (1.00)

a. [1.0).

REFERENCE GELBWR Academics Series, Heat Transfer and Fluid Flow Student Text, cJuly 1985, Chapter 6. LO 10.16 3.0/3.1 291004K104 ..(KA's)

ANSWER 4.19 (1.00) c' [1.0)

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

7

.W M N Tf7XiOAND COMPONENTSE(10%F(FUNDAMENTALS' EXAM)"

M, ;&

g , ; >, %#4 w ,

,e ,

.J fE g "I@; REFERENCE' ,

4 EGE 9 BWR Academics! Series,fHeat2 Transfer and Fluid Flow Student 5 Text,.

:! July 11985,: Chapter? 6'.

9 '.(2i6/2.7; '

1 ,

29100bK10.5: . .' ( KA ' s )

m.

3 ANSWER ,7p, 2h-- . (l'. 00 )

i< g,- 4

? 1n1 - [1 0):

1., , ,

': : REFERENCE-

(GE BWR A~cademics Series, Heat Transfer-.and Fluid Flow Student Text,

, 3;- Feb. 1985, chapterL 7.

y, 1 9/3;0'2.7/2.8.;

or 29100'6K107) 291'006K108 ..(KA's)'

[ ANSWERS 4 4 211 -(1.00):

c: [1.0)

REFERENCE-

.E.I. Hatch,3 Condensate and Feedwater System, page 5.3.6.

. LOi 3. 4 f rom; Chapter' 5 'of : BWR Chemistry.

2.8/2.9-vi 291007K103 ..(KA's)

. ANSWER- :.4, 22 (1.00)

--b'.-[1'.0) 3 ._

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

c .

{

ep ,- q7g ~AND 60MPONENTS 110%f f FUNDAMENTALS EXAtil

~

it h.

REFERENCE

!r

g. 'il

p.

No facility reference identified. NRC Fundamentals Exam Bank.

3.4/3.5.

E- 291008K108 ..(KA's)-

' ANSWER 4.23- (2.00)

c. failo' upscale

, b. increase

c. increase.

d.' fails downscale REFERENCE

E. I.. l Hatch, Reactor Vessel Instrumentation, LT-IH-04404-00, EO 15, pages 21~and 22.

3.3/3.3 291002K109 ..(KA's)

(***** END OF CATEGORY 4 *****)

~ ~~ ~

77733sf [

p i ANSWER. 5.01 (1.00) l o-[1.0)_

! REFERENCE E.I.., Hatch, Trip of One"or Both Reactor Recirculation Pumps, 34AB-OPS-032-28., pg.4.

3.8/3.7 295001G010 ..(FA's)

Delded l ANSWER (1.00) dr 0.

o [1.0)

REFERENCE E.I, Hatch.. Reactor Power Instabilities. 34AB-OPS-058-25, page 1.

.NRC.BulletAn No. 88-07, Supplement 1: . Power Oscillations in Boiling Water Reactors.

2.5/3.3 3.8/3.7 295001G010 295001K104 ..(KA's) ocIck"A

' ANSWER K (1.00) d [1.0)

REFERENCE E.I. Hatch, Loss of Instrument and Service Air System, S4AB-OPS-020-2S, pg 3.

2.9/2.3 295002K306 ..(KA's)

. ANSWER 5.04 (1.00)

.c..[1.0]

(***:** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

,, 5 '. EMERGENCY AND'ABt!ORMAL PLAMI_EVOLUIlQHE Pego 54 L ...-(33%1

. REFERENCE E.I. Hatch, Loss of Instrument Buses,.34AB-OPS-014-2S, page 2.

4.2/4.3-295003A202 ..(KA's)

ANSWER 5.05 (1.00) c-[1,0]

' REFERENCE E.I. Hatch, SOFI Flowchart 2: Content and Use, LT-IH-20103-02, pg 20.

EO 18, 4.1*/4.5* 4.0/402 4.1*/4.3*

295037K303 295037K209 295037K102 ..(KA's)

~ ANSWER- 5.06 (1.00) d [1.0]

REFERENCE E.-I. Hatch, Loss of Feedwater Heating, 34AB-OPS-045-2S, page 2.

4.0/3.9 3.6/3.8 295014A102 295014G010 ..(KA's)

ANSWER 5.07 (1.00)

a. [1.0)

REFERENCE E.I. Hatch, SOFI Flowchart 1: Content and Use, LT-IH-20107-03, pg 13, i EO 7. l 3.8/3.9 295015K201 ..(KA's)

ANSWER 5.08 (1.00) l l

1 b [1.0]

l l

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

i

~

BESFONS.IBILITIES (13%)'

LREFERENCE '-

3- ,

f E.I. HATCH LESSON PLAN, REACTOR. PROTECTIVE SYSTEM, LT-IH-01001-00,

-PAGE 28-

'3.3/3.5.-

L212000K411 ..(KA's)

ANSWFR 6.17. (2.00)

A. 3, 5, 6 B. 1, 4,_7

< C. 2, A (0.25 each)

REFERENCE-E.I. HATCH LESSON PLAN, REACTOR PROTECTIVE SYSTEM, LT-IH-01001-00, EO #16, PAGES 18 & 1E'

~212000G005 212000K412 ..(KA's)

ANSWER 6.18 (1.20)

~A. (1) 5 (2) 10 B. -(3) IRM Inop (4) IRM Hi-Hi C. (5) -All-APRM&--are en sn*1 a (but-before-rearh4r>a 1nA on-Range-le of

" (4P rc2. pig.tas tuoce Su>m To " Aua "

D. (6) Slope of the line (will accept " time to double")

(0.2 each) i 1

i I

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

=_- __ .

W' 7 ~R5sPONSIBiLlllES (13%)

J

REFERENCE:

L I

-E.I._ HATCH LESSON PLAN, INTERMEDIATE RANGE MONITORS, LT-IH-01202-00, i

E0 #3, 4c, 11, 13a & 13d, PAGES 11, 15, 26, 27, 30 & 32 4.0/4.0 3.3/3.3 3.0/3.1 e:

215003K402 215003A401 215003K503 ..(KA's)

ANSWER 6-.19 (2.50)

A. High turbine exhaust pressure 40 psig (36-44 psig)-

Low RCIC pump suction pressure 10" Hg vacuum (9-11" Hg)

Electrical'overspeed 4950 RPM or 110%

Mechanical overspeed 5625 RPM or 125%

(0.3 for_ trip, 0.2 for setpoint)

B. Mechanical overspeed

' Local' manual (0.25~each)

REFERENCE E.I. HATCH ~ LESSON PLAN, REACTOR CORE ISOLATION COOLING E.3 TEM, LT-IH-03901-00, EO #6c & Ba, PAGES 12, 13 3.8/3.7 217000A202 ..(KA's)

(***** CATEGORY B CONTINUED ON NEXT PAGE *****)

= _ = -

.'- R T HESEQHgl.HILIIIES (13%iE

~~ ~~ ~ ~ ~

ANSWER 6.20 (1.50)

c. 17 e small break hbi ThrMmQlAG S4FA<

'b.

c. High Pressure Coolant Injection'(HPCI) M d.- 2 py 3 M *'d

40. 37 5e add-

~

' REFERENCE E.I. HATCH LESSON PLAN, AUTOMATIC DEPRESSURIZATION SYSTEM, LT-IH-03801-03, EO #1, 2, 13 & 15, PAGES 9. 10 & 25 3.5/3.6 3.9*/3.9*

218000K404 218000K106 ..(KA's)

ANSWER 6.21 (0.75)

A. 139 degrees (accert 137 to.141'i

-B. '133 degrees (accept 131,to 135)

C. 150 degrees. (accept 148 to '52)

(0.25 each)

REFERENCE I

i E.I. HATCH LESSON PLAN, FUEL POOL COOLING AND CLEANUP, LT-IH-04501-01, EO #1, PAGE 5 l 2.8/2.9 i

. 1 l

1 ,

k 233000G010 ..(KA's) l l

l

!r'

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) ]

l p

I

.u_m .w__._-

,; R." BKEEQNSIBILITIES ff3%i

~ -

ANSWER 6.22 (1.00)

A. 7" Hg vacuum (0.25)

B. - Reactor Mode Switch not in "RUN"

- Turbine Stop Valves not full open (less than 90% open)

MSIV Low' Vacuum Trip. Bypass Saitches~in " BYPASS" (0.25 each)

REFERENCE E.I. HATCH LESSON PLAN, MAIN STEAM / LOW-LOW SET, LT-IH-01401-00,

-EO #9a & 11a, PAGES 40 &43 3.3/3.4 3.8/3.9 239001K608 239001A208 ..(KA's)

' ANSWER 6.23 (1.50)

" Loss of power to respective bus

~Less than M nominal voltage G2 3 l SEfoALC) 4G9/o-- (u .e m 9, .f J .et nA n,g.,,3 4, 3 vy)

Low reactor water level Less than or equal to -113 inches High drywell pressure Greater than or equal to 1.52 psig (0.3 for each start signal)

(0.2 for each setpoint) l l .

l

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

y l

REFERENCE-E.I. Hatch, Remote Shutdown Panel, LT-IH-05201-00, pg 12, 13, and 14.

E0 6.

4.1/4.1 4.0*/4.1*

295016G006 29501CX202 ..(KA's)

JANSWER. 5.09 (1.00) b [1.0]

REFERENCE E.I. Hatch, Loss of Reactor Building Closed Cooling Water, 34AB-OPS-011-25, pg 2.

3.4/3.6

.295018K202- ..(KA's)

ANSWER 5.10 (1.00) c-[1.0)

REFERENCE E.I. Hatch, Loss of Instrument and Service Air System, 34AB OPS-020-25, pg 2. Plant Air Systems, LT-IH-03501, EO-16.

3.2/3.3 3.3/3.2 3.3/3.1 295019A104 295019K203 295019A102 ..(KA's)

ANSWER 5.11 (1.00) a [1.0] pal +4+d.

REFERENCE E.I. Hatch, SOFI Flowchart 1: Content and Use, LT-IH-20107-03, pg 18, 20, 22, 23. E0 11, 15, 16. E.I. Hatch, Emergency Operating Procedure Variables and Curves, LT-IH-201113-00, pg 7, E0 1.

3.7/4.1*

295026K304 ..lKA's)

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

4

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

~ ANSWER- 5.12 (1.00)

'd' [1.0)

' REFERENCE

~E.I. Hatch, Emergency Operating Procedure Variables and Curves,-

LT-IH-20113-00 pg 11 and 12. .EO 1 and 2.

3.8/4.1 3.5/3.7 3.5/3.8 3.4/3.8 295026K102 295026K206 295026K301 295026G007 ..(KA's)

LANSWER 5.13 .(1.00)

c. [1.0)

REFERENCE E.I. Hatch, PSTG and SOFI Chart General and Specific Cautions, LT-IH-20114-00,Section IV.F 3.6/3.8 3.4/3.8 3.7/3.9 295028G007 295028K203 295028A203 ..(KA's)

ANSWER 5.14' (1.00)

'bL[1.0)

REFERENCE E.I. Hatch, LT-IH-20114-00, pg 14. EO-2.

3.5/3.8-295030K208 ..(KA's) l-o

. ANSWER 5.15 (1.00) l .b. [1.0).

l l' l

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

L

F q 7-(gg5

( ,

REFERENCE.

'E..I. Hatch, 34AB-EOP-049, Radioactivity Release Control: Content and l- ;Use,- LT-IH-20115-00, pg 6. E06. .

13.7/4.7*

295038K205 ..(KA's)

/ ANSWER 5.16 (1.00)

y. c. 30 seconds. [0.5] (accept 20-30-seconds)

b. Prevent motoring (reverse power) of the generator. [0.5)

REFERENCE E.I. Hatch, Loss of DC Buses, 34AB-OPS-013-2S, page 7.

3.3/3.4 3.2/3.4 3.4/3.6

'295004G010- 295004A103 295004K105 ..(KA's) 2.0 ANSWER 5.17 (er50)

n. 4 , J*

Lb. 2

c. 1

f. 3

[3r&][0.5 each]

2.0 REFERENCE E. I. Hatch, Loss of DC Buses, 34 AB-OPS-013-2S.

3.2/3.4 295004G010 ..(KA's)

ANSWER 5.18 (2.50)

a. Path 5

b. Path 4

c. .4ath 2

d. Path 1

e. Path 3.

(0.5 pts each)

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

s

~ REFERENCE GPC: 31EO-EOP-001-25, LP # LT-IH-20101-00, EO #7

3. 8/3.'4 :

'295006G012 ..(KA'c)

ANSWER 5.19 (1.00) 1.-weight of the water [0.5]'

-2. thermal stress [0.5]

-REFERENCE

E.I. Hatch, PSTG and SOFI General and Specific Cautions, LT-IH-20114-00, pg 15. E0 1.

3.2/3'.3 29500aG007 ..(KA's)

ANSWER 5.20 (2,50)

a. 95 F [0.5]-

b. 105'F [0.5]

c. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> [0.5]

d. 95 F [0.5]

e. scrammed [0~5] .

REFERENCE E.I. Hatch, Technical Specifications, 3.7.A.1

.3.3/4.2*

295013G003- ..(KA's)

ANSWER 5.21 (1.00)

MCPR [1.0]

REFERENCE NRC Bulletin No. 88-07, .Supplemen+ 1: Power Oscillations in Boiling Water Reactors.

4.1/4.4*

295014A204 ..(KA's) l l

l l

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) I 1

I i

_L___-__-_____ -_ - _

v q y

" ANSWER 5.22 (1.00)

Tea 145 psig.[0.5] (accept 130 to 145)

-:. .b. RHR1 Suction-Cooling Valves,-2E11-F008 and F009. [0.5] i

REFERENCE:

-E.I. Hatch,~ Loss of Shutdown Cooling, 34AB-OPS-044-25, pg 1.

ERER System-, LT-IH-00701, EO'7.D 1 13.6/3.6 .

4

, 295021K203- ..(KA's)

-ANSWER ~ 5.23 (2.00) 5~

[-OvF& due to negative a.Drywellmaycollapse[0.Ah(orotherwisefail) pressure [0.5] (resulting from rapid condensa tion).

ib. Drywell temperature [0,5)

Drywell pressure [0.5] i

-REFERENCE

~E.I. Hatch, Emergency Operating Procedure Variables and Curves, LT-IH-20113-00, page 10. EO.2.

'4.2*/4.4* 3.9/4.0 3.6/3.9 295024A201 295024A202 295024G007 ..(KA's)

ANSWER 5.24 (2.00)

1. Lo-Lc1 Set

12. Main Turbine Bypass Valves 3'. SRV's' 4.1 Alternate Pressure Control Systems. (Accept any one alternate pressure control system)

~[2.0] [4.at.0.5 each]

REFERENCE E.1. Hatch, SOFI Flowchart 3: Content and Use, LT-IH-20104-03, pg 14.

EO-10.

3.9*/4.5* 4.4*/4.4* 3.8/3.8

'295025G012' 295025A103 295025A102 ..(KA's)

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

s

~_- _x :____--_- _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _

p., f12(33%i 7 JdNSWER 5.25 (1.50)

.. .c. 2200 F.-[0'.5]

L- b. li Vessel-level above the top of active fuel [0.5].

" 2. At least one Core Spray pump injecting at rated flow [0.5].

REFERENCE E.I.. Hatch, EOP End Path Manual Content and Bases, LT-IH-20108-4, pg 47 EO 3. EOP Flow' chart and End Path Manual Design and Use, LT-IH-20102-03 pg 30, 4.0/4.3* 4.4*/4.7* 4.0/4.3* i

~295031K303 295031K302 295031K304 ..(KA's)

ANSWER 5.26 (2.00)

1. Terminate energy addition to the secondary containment [1.0]

2. Place the RPV in a low energy state [1.0) (also accept discharge energy to suppression' pool for full credit.)

REFERENCE E.I. Hatch, EG.D End Path Manual Content and Bases, LT-IH-20108-04, pg

74. EO-4, 3.5/3.8 295032K301 ..(KA's).

(***** END OF CATEGORY 5 *****)

~ ~

5 ,

'RESPONSIB GITIES (13 E

!o l:

LANSWER- _ 6.01~ (1.00) c b: (1.0)

REFERENCE

'E.I. HATCH LESSON PLAN. REACTOR RECIRCULATION SYSTEM, LT-IH-00401-00, EO #25, PAGE 14~

3;5/3.7 3.4/3.9 202001K601 202001A201 ..(KA's)

ANSWER '6.02 .(1.00) a (1.0)

REFERENCE JE.I. HATCH LESSON PLAN, HIGH PRESSURE COOLANT INJECTION, LT-IH-00501-02, EO #15.d.3, PAGE 45 3.8/3.7 3.3/3.3

, 206000A401 206000K505 ..(KA's)

-i ANSWER 6.03 (1.00) i d <?R- k- (1.0)

REFERENCE E.I. HATCH LESSON PLAN, STANDBY LIQUID CONTROL SYSTEM, LT-IH-01101-00, EO #11 & 13, PAGES 15, 17 & 18 3.8*/3.9* 4.2*/4.2* 4.0*/4.1*

211000K407 211000K408 211000A306 ..(KA's) l

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

1

___._a__._- . . _ _ . l

~ ~ ~ ~ ~ ~ ' ~

Q. )~BEEPONSIBILITIEs (13%i c

i l ANSWER- ~6.04 (1.00) e 1

h. c.. (1.0)

YREFERENCE E.I. HATCH LESSON PLAN, REACTOR PROTECTIVE SYSTEM, LT-IH-01001-00, EO #6, PAGE 22 & 23-3.3/3.413.7/3.8 !

-212000K502L 212000K305 ..(KA's)

ANSWER- 6.05 (1.00) ;

I I

c- (1.0) ']

4

-REFERENCE i E I. HATCH LESSON PLAN, REACTOR CORE ISOLATION COOLING SYSTEM, LT-IH-03901-00, EO #4, PAGE 7 3.6/3.6.

217000K103 ..(KA's)

ANSWER 6.06' (1.00) b (3.0)

I .~

l 1'

l (**N** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

l-J , k

_ _ - - _ _ _ l

~ ~~ ~ ~

,EC .(BESfQMSIBILIIJES 11'3%f

' REFERENCE TE.'I. HATCH LESSON PLAN, AUTOMATIC DEPRESSURIZATION SYSTEM,.

LT-IH-03801-03, EO #12, PAGE 30 4r2/4.3* 4.2*/4.2*

218000A206. '218000A402 ..(KA's)

ANSWER 6.07 (1.00) 4 c- (1.0)

REFERENCE-6 E.I. HATCH LESSON' PLAN, MSIV LEAKAGE CONTROL SYSTEM, LT-IH-Oe901-00,

- EO-#6:& 10b, PAGES' 16 ,21 & 22 3.1/3.3 3.1/3.1 239003K406 239003A101 ..(KA's)

. ANSWER 6.08 (1.00)'

d (1.0)

REFERENCE E.I. HATCH LESSON PLAN, MAIN CONDENSER, LT-IH-02501-00, EO #15, PAGE 20 2.8/2.8 3.1/2.9 3.4/3.4 l

256000K409 256000G010 256000G007 ..(KA's) I i

l l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

L 1 l

L I i

l i

l I.

0 7 R T RESPONSIBILITIES (13%)'

(

' ANSWER 6.09 (1.00) ,

d (1.0)

REFERENCE E.I.-HATCH LESSON PLAN, CONDENSATE AND FEEDWATER, LT-IH-00201-00, EO #11, PAGES 17 & 18 3,9/3.9 3.8/3.9 3.7/3.7 3.4/3.4 259001K301 259001K312 259001A201 259001A310 ..(KA's)

' ANSWER 6.10- (1.00)

.o (1.0)

REFERENCE E.I. HATCH LESSON PLAN, PLANT FIRE PROTECTION SYSTEMS, LT-IH-03601-00, EO #1, PAGE 7 .

3.8/3.9 l 286000G004 ..(KA's)

ANSWER ~ 6.11 (0.50) ,

"A" CRD Pump -

4360 VAC Bus 2E (Emergency Bus 2E)

"B" CRD Pump -

4160 VAC Bus 2F (Emergency Bus 2F)

(0.25 each) 1 i

1

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

l 1

____ _ _ _ _ _ a

' ~

N m RESPONSIBILITIES (13 h '. 4

, . .?

, 9. . .. ,

,w (1 <? REFERENCEC ,

~

j ;E.11.4HATCHiLESSON. PLAN,-CONTROL ROD' DRIVE HYDRAULICS,:LT-IH-00101-00,,

iEO:#13i.:PAGE.41 p y 72.9/341'--

-f' L',,

J201001K2011 x...(KA's) o

(.(1,00)'

. ANSWER 7 6.12~.

1 4f _4 g G2__..a,igeactorpressurefunction'asthesource.of-i

<energyLtolinsert control rods on a' SCRAM independent.of CRD pump j' .~ operation. (alternative wording acceptable) (1.0)

REFERENCE E.I.> HATCH. LESSON. PLAN', CONTROL ROD DRIVE HYDRAULICS, LT-IH-00101-00,-

-EO #So', PAGES 30 & 31 3.8/3;8 3.1/3.2- .

i' \ .

201001K405 '201001K303 ..(KA's) i LANSWER: 6.13 . ( l'. 0 0 )

s.

> ITo : limit: control: rod worth (0.75) so that the fuel enthalpy limit of 280 cal /gm will.not be exceeded during a rod drop accident.(0.25).

(***** CATEGORY 6 CONTINUED Oh !4 EXT PAGE ***** )

w -_= _ _ - - - _

k T 7 5R2sP6NSIBIL,ITIES Ti13%Y p

u <

h', REFERENCE lEil. HATCH-LESSON PLAN, ROD WORTH MINIMIZER, LT-IH-05403-00, EO #1,-

LPAGE 6-

3.3/3.7 3.4/3.4-

[; ' 201006K501 .01006G004 ..(KA's)

ANSWER 6.14 (1.50)

IA. -(1) Inoperable

.(2) for testing-B. (3) Operations Supervisor C. (4) A second licensed operator (5) A qualified member of the Technical Staff

'D. (6) Low Power Setpoint (LPSP)- (30% power) (0.25 each)

REFERENCE E.I. HATCH LESSON PLAN, ROD WORTH MINIMIZER, LT-IH-05403-00, EO #8, PAGE 30

3.2/3.4 3.4/3.5 201006A401 201006K404 ..(KA's)

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

u__m__ _

i

I@ YdHf6E5}BILITIES (13%)-

ls

! ANSWER' 6.15~ (2.50)'

l :;

-A. Valves are interlocked open (0.5)

Ten minute on UNIT 1 (0.25)

Five. minutes on UNIT'2 (0.25)

B.- High Drywell' Pressure of greater than or equal to 1.92 psig; or Reactor Water: Level of less than or equal to -113 inches.

(0.25 each)

C. (Upon reenergization of the respective pump buses ) the pumps start in the following sequence:

RHR "C" -

starts immediately RHR "A, "B" and "D" - start after a 10 second delay A.So ACLCpr nmc bELA1 ~-va r bicsCL Tic ON ADDCb REFERENCE SM ~

IDA R (42 ~ A 'g W $6' ~ 7_ Z. S Gc E.I. HATCH LESSON PLAN, RESIDUAL HEAT REMOVAL SYSTEM, LT-IH-00701-01, EO #7a & 7b, PAGES 28, 29 & 32 3.5*/3.5* 3.7/3.9 3.6/3.7 4.2*/4.2 203000K201 203000K407 203000K601 203000K401 ..(KA's) 1 ANSWER 6.16 (1.00) !

i The Pucle e Tnatr" mentation trip logh-changes-to-non-coi-noidence ;

tr4ps-44&}Ay ^na ^f 18 NI-tr4ps-wiL1-oause-e-SGRAM. (0.54 ;

T42 Nacce~M hl&TearsupmLw TRtPS lD& lc>M Ct4%IMS TO DAti 02 G=

l % Yx'CA D^tL2 {Non - CCCdWGLib(0 b i

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

l I

l 1

i

. RESPONSIBILITIES (13%)

IRIFERENCE.

E.

I E.I. HATCHTLESSON PLAN, DIESEL GENERATORS, LT-IH-02801-00, EO #4, PAGE 17 3.3/3.7 F 264000K408 ..(KA's)

ANSWER 6'.24 (1.80)'

A. 8

'B.- 2, 10 C. 5 D. 4 -

E. 2, 6 F. '1-(tb e 4J c 'e<A$ cA"3 Dor *I)

G. 5 (0.2 each)

. REFERENCE E.I. HATCH LESSON PLAN, OFF-GAS SYSTEM, LT-IH-03101-00, EO #6 & 10, PAGES 13-18, 27 & 28 3.1/3.3 3.1/3.3 3.3/3.3 3.3/3.4 271000K102 271000K408 271000A301 271000G007 ..(KA's)

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

i 4

P.spf J'BEgfDHglBILITIESl13%f l fi H>

y(! l r'

lANSWER- '6.25 (1.00)

'b.. (1.0)

? REFERENCE n E.I; Hatch 30AC-OPS-001-OS pp.-7 & 8.

{ 13.9/4.5 294001K102- ..(KA's)

' ANSWER 6.26 (1.00) u, G. [1.0) c REFERENCE E.I.. Hatch, ControlLof Operator Aids, DI-OPS-05-1084N, page 3.

.4.2*/4.2*

294001A102 ..(KA's)

ANSWER 6.27 (1.00) ,

c (1.0)

'n .

REFERENCE E.I. Hatch, EOP Flowchart and End Path Manual Design and Use, LT-IH-20102-03, page 10 to 14. EO 4.

'4.2*/4~.2*

294001A102 ..(KA's) 1 ANSWER- 6.28 (1.00) 'I d.(1.0)

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

~~ ' ~ ~ - -

RESPONSIBILITIES 713sF g 4Rs s @ W; n

' I.

sesREFERENCEi 3 .m . , ,

'(

6 , . . , ,

Q', Q E!I. Hatch Plant.09erations;30AC-OPS-003-OS pp. 16l&1 !

Z17/ Enabling ~ Objective:29Jof.LT-IH-30004-01"

, ; 3'. 4 /3. 6 x cr :;-294001A106:' n. .:.(KA's) i S I v

,sifANSWER.' (6.291 L(l'. 00 ) ; '!

O f.g;l.ci[120]

L . ..

$,[ REFERENCE.- >

g,. .

/E.I.LHatch,(Hatch Emergency Plan,- pg 13.

J2.9*/4.7*.. ..

( ,

294001A116: -..(KA's)

! r. .

-ANSWER 6.30 -(2.50) a.41) ' 1. 25 rem'- [0. 5]

2) 18'.75' rem [0.5]:

7

3). 7.5. rem [0.5]

lb.!1):32. rem 3[0.5)-

[0.25]

2) Exposure myst not' exceed 5~(N-18) i#^- ' rset bc coa +1ci,ed [0.25)

, ' (r er Iws X, s 4&r .

) caw.

LREFERENCE

!E.1I. Hatch,: Radiation' Exposure Limits,-60AC-HPX-001-OS',.pg 4.

- 110CFR20-

H3s3/3'.'8-i-

1' .

294001K103- .-(KA's) l ,

i

-(*****LCATEGORY 6 CONTINUED ON NEXT PAGE *****)

1 L

L e .

mjQisRESPONSIBILITIES'(13%}n:

w

^

+

Q![ i i n f SANSWER'.'- I6231L (2.00)'

T' g,)g 2fShiftLSupervisor.(will report to1 fire as Brigade. Chief if-he:is

, .# ' qualified);;if.not,.U-1 SS'will act.as~ Chief. [1.0)-

~ Th ti. 5 [0. 5] L .

c.f3;[0.5] .

.m .

p RE N NCEi-TGPCi 40AC-FPX-001 9

.13.5/3.8'

,y. [294001K116 ,.(KA's) b IANSWER- 6.32 .

(2.00)'

'a. Turn valve ~inLclosed direction (1/4 turn max to seat)-

b.

Turn-valve in" closed direction (1/4 turn max off backseat).

-oh ? Verify at; remote (or-local) position indication

!s 'd..

,Confirmclocking-device operability.

l(? O . 51 each . ) -

_ : REFERENCE ~

'E.EIHel 34-GO-SUV-001-05, EO # 341.3.3'

,,13.7/3;7~

J294001K101- ..(KA's).

+

-ANSWER; 6433- (2.00)-

-a. 2 [0 253'.

b.11,.2, 3,c4;[1.0) l

'o;711[0.25]. j d: 3;[0.25]"(if 4 is also given in addition to 3 no points are to be j JN ; L .

deducted) j y io. 4 [0.25] .

H REFERENCE 1 I

1 E.I.. Hatch, Plant. Operations, 30AC-OPS-003-OS, pg 21.

"3.3/4.3* 3.5/4.2 j

,r 294001A111' '294001A112 ..(KA's) j I

l

(***** END OF CATEGORY 6 *****) l D (********** END OF EXAMINATION **********)

i ,

1 i * , -I Q1 . . . . .

i

ENCLOSURE 4 NRC RESOLUTION OF FACILITY COMMENTS (QuestionR0/SRO)

(1) Question 2.01 / 5.02 Comment partially accepted. The question was deleted because there was more than one correct response.

(2) Question 2.03 / 5.03 Commer.t partially accepted. The question was deleted because there was more than one correct response.

(3) Question 2.10 / 5.11 Comment partially accepted. The question was deleted because there was no correct response. Answer (a) is incorrect for reasons stated by the facility. Answer (d) is incorrect because level reduction is stopped when any one of the three criteria listed (not all) are no longer met.

(4)_ Question 2.14 / 5.05 Comment not accepted. The K/A 295037EK303 specifically requires that the candidates have " Knowledge of the reasons for the following response as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Lowering reactor water level." This KA has an importance rating of 4.1 / 4.5. This is sufficient to justify its inclusion in the examination. The facility's Enabling Objective 13 of LT-IH-20107-03 also states, " State how natural circulation is reduced during an ATWS."

The Boiling Water Reactor Owners Group EPGs indicates that lowering level reduces the natural circulation driving head. The Plant Specific Technical' Guidelines (PSTG) statement is an effect of the loss of driving head from the lower RPV level. The required answer is obtainable from the facility lesson plan LT-IH-20107-03 page 20 which states: " Natural circulation is a function of the heights of the water columns inside and outside the shroud. Lowering the heights of these water columns reduces the natural circulation driving head."

l l

), _ . .,

ll ~

'2

. Enclosure 4

. (5) Question' J2.15 / 5.17 Comment' partially accepted. Number 1 was deleted from the part (a) answer. ~ No changes were n'ade to part (c) of the question for the following reasons:

(a) . Immediate ,0perator Actions r.ecessitate rapid implementation of required steps. The referenced K/A specifically requires that,immediate actions be performed without reference..

(b)' The correct answer could easily be determined based upon systems knowledge and the effects that a' loss of DC Control Power has on automatic bus transfers. The following list of K/A's are applicable:

262001K4.03 3.1/3.4 262001K6.01 3.1/3.4 262001A3.02 3.2/3.3 263000K3.02 3.5/3.8 263000K4.01 3.1/3.4 (6).. Question 3.03 / 6.03 Comment not accepted. No change to the answer key was made. The j terminology differences mentioned by the facility are not significant i enough to warrant choice (b) as an acceptable answer.

(7) Question 3.08 / ***

Comment accepted. It is NRC policy to delete questions with more than one correct response. However, since one candidate was told by a proctor that the pressure setpoint was increasing, either answer (a) or (d) will be accepted.

(8) Question 3.11 / ***

Comment not accepted. The question was deleted because of the possibility of there being more than one correct response. The facility reference material coes not provide adequate justification foranswer(b). Furthermore, subsequent conversations with facility

' personnel revealed that the actual voltage conditions that will tie the Diesel Generators to their respective busses differ from those taught by the training department or published in the training material .

(9) Question 3.13 / 6.12 Comment accepted. The answer key was changed to accept reactor pressure for full credit.

1

' Enclosure 4 3 (10) Question. 3.16 / 6.15' Comment accepted. The answer key was changed to accept either time notation for full credit since they are equivalent.

(11) Question 3.17 / ***

Comment accepted. The answer key was changed to accept " voided or partially filled system" for full credit.

.(12) Question 3.18 / ***

Comment accepted. The answer key was modified so that the minimum flow valve cycling was not required for full credit. The question point-value was reduced to 1.0 points.

(13) Question. 3.19 / 6.17 (a) Comment not accepted. The only difference in the Turbine Control Valve fast closure .between the units is in actuation times, not setpoints. No change to the answer key was made.

(b) Comment not accepted. The question specifically refers to scram setpoints and not circuitry. The APRM High Flux (flow biased) scram is bypassed when the mode switch is not in Run. The APRM High Flux Scram (118%) is also bypassed when the mode switch is not in Run. No change to the answer key was made.

(14) Question 3.20 / 6.16 Comment partially accepted. The answer key was modified to accept "The Nuclear Instrumentation Trip logic changes to one out of 18 taken once (non-coincidence)" in accordance with the supplied reference.

(15) Question 3.21 / 6.18 Comment accepted. The answer key was modified to accept "after placing the Mode Switch to Run" as the correct answer. The answer was changed because the reference provided with the facility's

. comment, disagrees with the originel reference material provided to develop the written examination. The original reference material .

(LT-IH-01202-00) should be corrected. ,

1 (16) Question 3.22 / 6.19 l

a. Comment not accepted. The question specifically asked for RCIC !

Turbine Trips. High reactor water level results in closure of I the F045 valve (Steam Supply Valve) and not a turbine trip.

1

_ _ _ _ _ . _ i

y -

t na.

g ,

Enclosure 4- 4-

,u b'. Comment.not accepted. The answer was not changed because the i ,' i facility's ' reference inaterial - (3450-E51-001-2S Rev 7) 'is in -

error.1 The leal manual- trip of the .RCIC Turbine is listed as a RCIC Turbine Trip in this reference. -The note at the bottom of ~

f the page.. states that all tripi except the mechanical overspeed 1

. trip'can.be reset from'the Control. Room.: This is incorrect since

the . local manual: trip is a manual actuation of the mechanical

? .overspeed trip linkage'and thus-cannot be reset from the Control Room.

. (17) Question 13.23 / 6.20

. Comment accepted. The answer key was modified to accept small or intermediate for full credit. The answer was modified because the.

-reference provided by the facility expands the information in the lesson planLLT-IH-03801-03 which was used to develop the exam.

L

,This. lesson plan should be revised to include this information.

(18)QuestionJ3.24/***

Comment accepted. 'The.' answer key was changed to require only four Ecorrect responses for full credit. The Process Computer is a valid answer toithis question. It is recommended that the lesson plan

.LT-IH-01401-00 be' revised to include the Process Computer in the list of systems / indications which. utilize the dp flow signal from the Main Steam Line Flow Restrictors.

.(19) Question 3.27./ 6.24

- Comment accepted. N3 points were deducted if #9 was included for

< the answer to part F of-the question.

(20) Question '3.35 / 6.30 Comment accepted. The answer key was modified to accept " previous history known" for full credit.

(21) Question *** / 4.06 Comment not accepted. While burnable poisons are used locally to flatten flux peaks, this is to even out fuel depletion and not to prevent excessive peak centerline temperatures.

. (22)-Question:***/4.20 I. Comment accepted. The question was deleted because there was more than one' correct response.

i

..-___-----~-:.--__._-- - -- -

i .r.. -

' Enclosure 4 5 (23) Question *** / 5.19 Comment partially accepted. The referenced K/A requires the candidate to " Explain .. . . all system limits and precautions."

Understanding the effects of failure to comply with the precaution is an integral part of this explanation. Alternative answers which are viable causes' of damage will be accepted based upon the examiner's expertise, due to the failure of the facility to provide additional answers and support them with references.

(24) Question *** / 6.21 Comment not accepted. The supplied reference material does not support the facility's comment.

(25) Question *** / 6.23 Comment partially accepted. The answer was graded without regards to percent value of undervoltage. . Subsequent conversations with facility personnel revealed that the actual voltage conditions that will start the Diesel Generators differ from those taught by the' training department or published in the training material.

.,,< . 1.-i U ~.-

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hk .g % m k' '

G e h- z

s. ,

lk i ENCLOSURE 5

, SIMULATION FACILITV' REPORT

Facility Licensee: Georgia' Power Company.

[ Facility Docket Nos.: 50'-321'and 50-366 I

LOperating1 Tests.AoministeredOnil June 12 - 16, 1989

' This; form.. is t'o be .used only to report -observations. The'se observations do

" not. constitute, audit Tor . inspection findings and; are not, without' further verification- and review,. indicative of noncompliance with 10 CFR 55.45(b).

These : observations do not Laffect: NRC' certification' or. approval 'of the

simulation 1 facility other. than to provide information which may be' used in future : evaluations. . No licensee action is- required in respunse to these

' observations.:

. During, the- conduct of; the : simulator portion of the operating tests, the Lfo11owing items were observed:

Item . DescH ption MALF'171. 600V. Bus2C Fagit:

The ' description of the malfunction in the Simulator Instructor's Handbook 'is incorrect due to vital power supply being lined up to 600V Gus 2C with no alternate supply' avamble. This electrical bus. lireup is a mimic of the- prese.it in-plant conditions. The loss of this bus will result in a reactor scram and is not addressed in the malfunction description.

"MALF 50 Recirculation Pump Master Flow Controller Failure-(LOW):

The description of the malfunction in the Simulator Instructor's Handbook is incorrect. The malfunction does not run Recire Pumps'back to 22 percent as stated.

MALF 161. Loss of Off-Site Power:

The running of this malfunction resulted .in unrealistic plant conditions in the drywell. Drywell temperature increased to over 400 degrees F almost instantaneously a'

with drywell pressure increasing to over 5.2 lbs. No LOCA or steam leak was' simulated. This item was identified by the facility as'a software problem.

t Enclosure 5 3

.Itgm Description LPRMs LPRM malfunctions are limited in number and scope and are only marginally useful.

MSL RAD No m61 functions exist'for the Main Steam Line Radiation Monitois Monitors.

MALF 211 Scram Discharge Volume' Hydraulic Block:

During the performance- of this malfunction at 20 percent severity, reactor power was observed to have stabilized at 50 percent. This power ' level appears to be higher than would be expected since at 50 percent severity the rods are in a modified black / white pattern.

- - _ _ _ _ _ _ _ _ _ _ _

ML20247C954

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