Text

Exam Rept 50-341/85-01 of Exam Administered on 850212.Exam Results:Two Instructors & All Reactor Operators Passed Exam
	ML20129A458
Person / Time
Site:	Fermi
Issue date:	05/17/1985
From:	Lang T, Levy I, Mcmillen J, Sly G; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:	;
Shared Package
ML20129A453	List: IR 05000341/1985001; ML20129A448;
References
	50-341-85-01, 50-341-85-1, NUDOCS 8506040613
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i U.S. NUCLEAR REGULATORY COMISSION

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REGION III

Report'No.'85001(DRS).

~ Docket'No. 50-341 License No.'CPPR-87 x

' Licensee: - The Detroit Edison Company 6400 North Dixie Highway-Newport, MI_48166 Facility Name:

Fermi Nuclear' Power Station

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. Examination Administered At:

Fermi Nucler Power Station-

-Examination Conducted:

February 12, 1985 W

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Examiners:

T. Lang J~' /4-83 Date

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hLevy,PNL

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jf.Sy,PNL

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5 //7 /75 V

Date /

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Approved By:

ll. I.-McMillen, Chief 3//7/75

)perator Licensing Section Dpe /

Examination Summary

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Examination administered on February 12, 1985 (Report No. 85001)(DRS))

Examinations were administered to one Senior Reactor Operator, seven Reactor Operators and three Instructor candidates.

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Results: Two instructors'and all of the Reactor Operators passed the examination.

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' REPORT DETAILS-

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L1 1 Exami'ers.

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' T. Lang, NRC I._ Levy,'PNL

- G.' Sly; PNL 2.

- Examination Review Meetina

- R. Bouinet

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S.' Heard

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- E.-Preston G. Overbeck m

During.the. review very few comments were made to the SRO examination.

. They were.so few that they were incorporated directly to the answer key.

. R0 examination comments are detailed in Attachment A.

13.

Exit Meeting

' The exit meeting was conducted satisfactorily and the utility was

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informed of the candidatesLwho clearly passed the oral / operating.

examination.

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ATTACHMENT A

'Coment 1.1 Why are U235 and U238 mentioned.

RESPONSE References to U-235, U-238, and Pu-239 in-the answer key were incorporated to_more fully evaluate candidate's response. Candidate was not required to mention' individual isotopes, but only that Resonance capture was a negative' reactivity effect.

Comment.1.5

' Answer:

May take it out further than this.

We would end up back in the source range until cooldown brought Rx critical again.

RESPONSE-Facility coment has no bearing on grading of question. No action taken.

Coment 1.6 Answer:

The process computer calculates Quality and-Boiling length based on Core Flow, Inlet Subcooling, Thermal Power and Pressure. ' You may get these as answers also based on " measurement"

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in question.

RESPONSE The comment is valid, Quality and Boiling length are not measurable parameters but calculated. The answer key has been modified to read:

1.

Core Flow 2.

Inlet Subcooling.

3.

Thermal Pouer 4.

Pressure.

- Any two (2) of the above will be acceptable for full credit.

Comment 2.1 Answer:

Setpoint 120 gal.

Allowable '160 gal.

for scram.

T.S. change.

RESPONSE 2.1 Since 100 inches is comparable to 120 gallons in the scram discharge volume, either answer is acceptable for the scram setpoint.

. REFERENCE: Tech. Spec.. Table 2.2.1.1 Final Draft.

Coment 2.-2 Answer:

May see reference to maximum 3" level deviation if all reference leg inside drywell flashe.

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RESPONSE 2.2-Comment was' accepted, but answer key not modified because facility conenent was; considered not the most optimum answer and. is 'self explanatory if given by a~ candidate.

Comment'2.3 Answer 4: May also see.' reference to-Iodine dissolution in water.

~ ESPONSE 2.3 R

Comment _was considered a secondary effect of the condensation process e

and was not conside. red necessary.for full credit.

Comment 2.4 Answer:. May see single coincidence or 1/18 for whole -

' system.

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RESPONSE-2.4 iSince-question' requests the trip logic for EACH Neutron Monitoring-System. a response of " single coincidence" is incomplete unless accompanied with 1Lof,18 or listed once for.each system. Answer key was modified to include single coincidence and l'of 18 or single coincidence -'SRM, IRM, APRM.-

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Comment.2.6'

Answer:

Some may specify the power supply.

RESPONSE 2.6 Comment ignored because question requests " power supplies."

If

. candidate provides bus -information it will be graded according to supplied reference materials as to correct or incorrect.

-Comment 2.9

. Answer:

CST Level Setpoint Setpoint-Allowable value-3" Ind.

may see 24" above bottom 27"'above bottom or 0" indicated RESPONSE 2.9

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Values were verified with references provided and answer key was modified to include either answer 10,000 gallons or 3" indicated or 27" above bottom for the CST level setpoint.

REFERENCE: -Tech. Spec. Table 3.3.3-2, pg. 3/4 3-27 Comment 3.1 Answer:

It means what you say, but you'may hear:

Rod not selected and odd picked up or even dropped out.

Timer timed out and even not picked up.

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RESPONSE-3.1 Rephrase of answer key was found to_be acceptable and answer key was

' modified to.use:the " trained" terminology.

Comment 3.2

- Answer:

Lower-Higher Same 4 L-3 trip back to 195" -

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-RESPONSE 3.2 Comment was considered, but should have no effect on responses because candidates were directed to respond with increase, decrease, or remain the same not higher, lower or same.

No action taken.

Comment 3.4 Answer:

Should say:

73% CV Demand We normally accept 5% if SRV opens at less than set pressure.

RESPONSE 3.4-Values (i.e., 73% and 5%)'were included to provide clarification for the grader and not required for full credit.

The answer key had, 78%

and 7% respectively and was modified to reflect the correct values even though they-were not required.

Comment 3.5 Answer:

1)

Downscale not operable unless companion-IRM upscale.

2)

I interpreted question to mean which output functions generate trips:

Answers could vary from INOP 4 - fixed / thermal /

/ Rod upsc/

upsc/

/8 locks 24 - (same for each channel)

Answer:

LPRM's not part of trip circuit.

Input to trip circuit only.

Fixed Upscale Answer:

Fixed upscale / Thermal Upscale /INOP

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- RESPONSE 3.5-a.

_The question elidits a number of responses depending upon the interpretation of the term " trip circuits. Since this is true,

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credit will be given for the following responses:

4 ; fixed (neutron). upscale

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thermal upscale.

i nop.'

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-downscale,

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. Rod-block and alarms will not be considered trip circuits but no

" ' credit will be removed if given.

b.

The answer key was modified to read fixed upscale or neutron upscale.

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The correct response'should be the same as.in part a.

The downscale trip circuit will input into the RPS system.. Rod blocks and alarms do not and credit was taken off if given.

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Comment 3.6 Answer:

-Assume below LPSP.-

RSCS group notch control.

One notch other way.

RESPONSE 3.6

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If candidate stated that.he was below the LPSP an insert response of

. one notch'is correct.

Answer key was' modified to reflect the response.

Comment 3.8 Answer:

If only one heater drain isolates,

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No Action-

.if both, your answer correct.

b RESPONSE 3.8 I

' During the examination the candidates weru told that only one heater

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drain had isolated, therefore, no action is the correct response and l-the answer key has been changed.

Comment 3.9 Answer:

1.

On LOP 1)

Load Shed f.

2)

DG Start

3)

DG BKR Closes when LS complete.

L (6 and 7 BKRs open)

4)

DG BKR closes (s and volts)

5)

Non LOCA loads sequenced on:

L DW Cooling fans L

DG Aux.

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~ Valve Power L-

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.Following" LOP;if LOCA

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.1) ' Load lShed

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'2)

-DG continue to.run

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3)

Load sequence RHR/CS etc.

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Following LOCA if. LOP-

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1)

Load Shed

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2). Start ECCS loads 4.;

No ten minute time delay TRESPONSE 3.9

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'The facility comment'is just clarification of the proper response.

a.

No action taken.

b.c The 10 minute time delay was eliminated from the answer key.

This changed.the answer key to "yes (0.25),.if you have adequate core cooling or indication of system malfunction-(0.5)."

Comment'4.5 Answer:

Or voiding causes power reduction while injecting Boron.

- Level

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RESPONSE'4.5 b.

Facility. comment was considered correct.and added to answer key.

- Both responses were necessary for full credit.

c.

No action taken " indicated zero" is throughout~ procedure and should be known.-

Comment 4.7-Question should state during changing operation.

RESPONSE 4.7 Due to the vagueness of the wording of the question, credit was given for both 1) when the nitrogen purge gas remains constant, or 2) when no'more fluid'is observed exiting the drain valve.

Comment 4.8

+ 2"

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138*F Drywell 95'F Torus 1.93 psig

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' RESPONSE 4.8-Comment reflects recent Technical Specification changes and the answer key.has_been modified.

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Additional _ facility' comments received February 26, 1985.

Comment 1.3

Answer:

This question postulates a loss of feedwater

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heating as the reason for the moderator temp reduction:of 75*F during a startup.

-Areas of confusion:

a.

A loss of feedwater heating would lead the operators to believe that we are at

. power, as such no change in moderator temperature would occur due to pressure control. We would see a change in inlet subcooling which would cause a power increase.

This would not be viewed as a heatup or a cooldown rather it would be a-power excursion.

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b.

In another view a 75' reduction in moderator temperature would result in a depressuriza -

tion and resultant closure of the MSIV's-and subsequent scram again a transient but

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no net-heatup or cooldown.

Additionally, the second part'of the q"uestion asks "which limit is more restrictive.

While we all know that cooldown stresses are more limiting, in actuality'our heatup limit is more restrictive than it needs to be since it-is equal to the cooldown limit.

Either answer with an explanation would of course be correct.

-RESPONSE 1.3 i

a.

The question'alicits a response of, heatup or cooldown not pressure or power as the facility comment suggests.

It was also clarified-during the exam that the core inlet enthalpy changed in an amount

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equal'to 75'F. -Due'to these facts no change was made to part a.

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<b.

No change was made to answer key because during the exam it was

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- stated that we were looking for the most " physically / vessel restrictive" not administrative 1y restrictive.

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Comment 1.4

_Regarding the' question dealing with removal of a source.

from a level of 6000 cps:

.The correct answer per'your key was a "non-linear

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. progression."

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-Area of Concern:

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a.

The change in neutron level would describe a linear

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plot on the SRM's since they are log scales.

b. ~ The neut'ron populatidn will' stabilize at a very low

. level (probably below our indication) but it will stabilize due to intrinsic sources.

RESPONSE 1.4 The question requested the MOST correct answer which is "2" non-linear progression.

It'did not soTTcTt what the' operator saw on his script

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recorder or meter nor did it solicit an evaluation of final countrate

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with " intrinsic" sources.. Sources and subcritical multiplication are very basic fundamental area of reactor theory and the candidate should be' cognizant in this area.

No action taken.

Comment 1.12 Regarding the question dealing with Thermal Neutrons and their ability to produce embrittlement:

Forstainlessgteelsexposedtoathermalreactor

2 fluence of 10 N/cm -sec, the tensile properties show some increase in ultimate strength, an almost three fold gain in the yield strength and a drop of about 1/3 ductility.

Thought to be due to trace Boron impurities.

Ref: Basic Nuc. Rx Eng.

Foster & Wright Pg. 355 Therefore while thermal neutrons do not cause direct damage, there is some indirect damage.

Also defect production as a result of thermal neutron irradiation is largely due to the recoil of nuclei that have absorbed a thermal neutron and emitted one or more T's.

Additionally while the damage due to fast neutrons is on the order of 90 times as great as thermal, the answer is not always false.

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RESPONSE.1.12 Since the question requested an explanation of why the statement was true or false, credit would have to be given for a true response if properly justified.

The answer key was not. modified because, according to your own admission, thermal neutrons "....do not cause direct damage...." and

" damage due to fast neutrons is on the order of 90 times as great as

' thermal...."

Comment 4.9 Regarding the question dealing with what four (4) things are logged in the NSO's log when the reactor is declared critical, you may find that the examinees listed only four (4) items, i.e.,

1.

Rod Number 2.

Rod Notch 3.

Rod Sequence This was specified as one item 4.

Rod Group in your key.

5.

Period 6.

Temperature 7.

Time Four of the above, they may not have lumped all of the Rod Parameters into one group.

RESPONSE 4.9 The question states which items MUST be logged, these are:

1.

Temperature 2.

Time 3.

Period 4.

' Critical Rod Position The other items listed rod number, rod notch, rod sequence, and rod group are considered to be a subset of rod position (i.e., you could in the same token separate time into hour, minute, second).

But due to the non-explicit wording of the question partial credit will be given if the candidate broke the critical rod into its subcomponents.

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MASTER COPY QUESTIONS U. S. HUCLEAR REGULA10RY C0fflSS10N REACTOR OPERATOR LICENSE EXAlllNATION Facility:

FERMI-2 Reactor Type:

BWR

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Date Administered:

2-12-85 Examiner:

GA SLY

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Candidate:

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INSTRUCTIONS TO CANDIDATE:

Use separate paper for the answers.

Write ansvers on one side only.

Staple nuestion sheet on top of the answer sheet.

Points for cach question

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are indicated in parenthesis af ter the nuestion.

The passing grade requires at least 70% in each category end a final grade of at least 80%. Examination

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papers will be picked up s!x (6) hours after the examination starts.

.c.

Category 1 of Candide W s

% of Value Total Score Cat. Value Ca tegory

25 1. Principles of Nuclear Power

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Plant Operation, Therwodynamics, Heat Transfer and Fluid Flow

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25 2. Plant Design including Safety and Emergency Systems

25 3. Instruments and Controls

25 4. Procedures: Normal, Ahnormal,

Emergency, and Radiological Control 100 TOTALS Final Grade

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A11. work done on this examination is my own; I have neither given nor received aid.

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Cindidate's Signature

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U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAlllHATION Facility:

FERMI-2 Reactor Type:

BWR Date Administered:

2-12-85 Examiner:

GA SLY

.

Candidate:

INSTRUCTIONS TO CANDIDATE:

Use separate paper for the answers.

Write answers on one side only.

Staple ouestion sheet on top of the answer sheet.

Points for each question are indicated in parenthesis after the ouestion.

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.

Category

25 1. Princip'les of Nuclear Power Plant Operation, Therwodynamics, Heat Transfer and Fluid Flow

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25 2. Plant Design including Safety and Emergency Systems

25 3. Instruments'and Controls

25 4. Procadores: Normal, Abnormal, Emergency, and Radiological Control 100 TOTALS.

Final Grade

%

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

Candidate's Signature

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U. S. NUCLEAR REGULATORY COMMISSION

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REACTOR OPERATOR LICENSE EXAllINATION Facility:

FERMI-2 Reactor Type:

BWR Date Administered:

2-12-85 Examiner:

GA SLY Candidate:

ANSWER KEY

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INSTRUCTIONS TO CAHDIDATE:

Use separate paper for the answers.

Write answers on one side only.

Staple cuestion sheet on top of the answer sheet.

Points for each question are indicated in parenthesis after the ouestion.

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.

Category 1 of Candidate's

% of.

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Value Total Score Cat. Value Category

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25 1. Principles of Nuclear Power Plant Operation, Thermodynamics, Heat Transfer and Fluid Flow

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25 2. Plant Design Including Safety and Emergency Systems

25 3. Instruments and Controls

25 4. Procedures: Hermal, Abnormal, Emergency, and Radiological Control 100 TOTALS Final Grade

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All work done on this examination is my own; I have neither given nor received a f.d.

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Candidate's Signature

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'1.0 PRINCIPLES OF NUCLEAR POWER PLANT OPERAT10N, THERMODYNAMICS, HEAT TRANSFER AND FLUlU FLOW 1.1 ANSWER a.

Fast fission (0.25)

Resonance capture (239 Pu)

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(0.25)

b.

Fast fission has a positive reactivity effe.ct (0.5).

YvA Resonance capture:

(overal1 negative due to less neutrons)

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2350 has a negative reactivity effect 238U has a negative reactivity effect..

239Pu positive if fissions, negative if produces 240Pu (0.5)

c.

Resonance capture

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.(0,5)

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REFERENCE: 'NT/R248 (Reactor Theory) Chapter 3, pg. 9, 10 1.2 ANSWER (See Graph) Figure 1.2 (3.0)

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Grading criteria for each curve.

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0.5 curve shape (times, direction, etc.)

0.25 relative maximum 0.25 relative minimum REFERENCE:

NT/R248 (Rx Theory) Chapter 10, pg. 1-10

.1.3 ANSWER a.

Cooldown (0.25), limit not exceeded (0.75) (100*F. in one hour.)

(1.0)

h..

b.

Cooldown,- (0391 thermal stress from cooldown results.in

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h additive tensile stress.at inner vessel walls (0,5).

(1.0)

( L3)

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REFERENCE:

NT/R248 (Rx Theory) - Chapter 9 pg. 11, 21,-17, 29 Fermi Thermo - Chapter 5 pg. 8

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1.4 ANSWER

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a.

(1.0)

b.

not change (0.5), K eff is independent of source-stren h.

(1,0)

REFERENCE:

NT/R251 (Rx. Theory) Chapter 11, Chapter 12 pg. 5-8 t

- Section 1. Continued Next Page -

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1.5 ANSWER (0.5 eq., 0.4 numbers, 0.1 math)

a.

Pf = po e /g t

2000/50 = et/c t = 322 sec

$22 (1,0)

. k0 b.

1.

Assuming the reactor is initially supercritical, the doppler coefficient = adds negative reactivity before reaching the-heating range, producing a longer period and slower increase in the power rise.(0.5)

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2.

At the point of adding heat (fuel heat transfer is equal to the ambient losses) the moderator starts to heatup, adding more negative reactivity.

Period gets longer and power rise declines even more.(0.5)

3.

When the reactor becomes critical (neg.

reactivity coefficients equal pos. reactivity from fissions),. the IRM pen recorders should indicate a constant power, and the period meter indicates infinity.(0,5)

4.

Power 'will start to decrease because of the increased

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effectiveness of the moderator coefficient, causing power.

to stablize at a

lower power level than in 3.

above.(0 5)

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NOTE: (Bulk of answer for grader and clarification)

(2.0)

REFERENCE:

NT/R251 ~(Rx Theory) Chapter 11 pg. 9

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NT/R237 (Rx Theory) Chapter 13 pg. 1-3 1.6 ANSWER (>

Quality and boiling length.

(1.0)

REFERENCE:

Fermi Thermo Chapter 9 pg. 5

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- Section 1 Continued Next Page -

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1 7. ANSWER-l

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a.. increase..

(0.5)

)

.b.

increase (0.5)~

c.

increase (0.5)

d.

decrease (_ 0.5)

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. REFERENCE:

Fermi Thermo Chapter 7 pg. 9-11 Chapter'8 pg. 6, 7

'1.8 : ANSWER.

.a.

2, 3,.5,_4, 1 (0.25 for each order)

-(1.25)

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b~. - Annular, mist (0.5)-

REFERENCE:

Fermi Thermo Chapter 7 pg. 1-5

"I'.9. AN'SWER (0.5 equation,-0.4 numbers, 0.1 math)

s-a.- increase (0.25), greater head losses iJn piping (0.75)

(1.0)

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b.. decrease (0.25),- necessary - to compensate negative reactivity

.from doppler (0.75):

(1.0)

-. REFERENCE:

Fer1ni Exam Bank Questions 1.08 and 1.43

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1.10 ANSWER f

.A

' A' jet pump converts T velocity heat.into a pressure head (0.25)~.

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-Due to :the convergent section of ~ the driving flow _ nozzle,. the

' driving. flow ~ is accelerated. at a _ high velocity, _ this in turn creates L-allow-pressure.in the throat area (0.5).. Due to 'this

_ pressure differential,- - the driven flow - is accelerated and entrained 'with the ' driving flow steam.

In the diffuser. section, 3 a - further reduction in velocity. is l achieved and the resultant discharge. pressure:.is developed (0.5).

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(1.25)

- h Yes (0.25).'

Even if the main condenser and associated. systems

'were-air l tight, gasses would collect in the main condenser in the

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. form lof Hydrogen and 0xygen (from the radiolytic decomposition of

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-cladding (0.5) and > fission-product gases (which escape - from the.

water),'

--

boundaryL through "use related" cracks),(0.5.)

'(1.25 )

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REFERENCE:

Fermi Exam Bank Q. 1.56 Fermi: Exam Bank Q. 1.14

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- Section 'l Continued Next Page -

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1.11' ANSWER a.

65%-

' (1.0 ) '

' REFERENCE:

Fermi Exam Bank Question 1.38 (rewrite)

1.12 ANSWER False (0.25), fast only(0.75)

(1.0)

REFERENCE: General.Rx Theory

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-1.13 ANSWER Increase (0.25), because more Rx - heat available.for turbine ~since less requirea 'for RWCU.(0.75) ~

(1.0)

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REFERENCE: : Fermi lhermo Text Chapter 4 pg. 22 1.14 ANSWER (0.25 pts -for each response)

First-----doppler

!?cond----moderstor Third-----void

~ (0.75)

REFERENCE:

Fermi Exam Bank Question 5.03 (rewrite)-

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- End of Section One -

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EQUATION SilEET

., ____ _____......__________.._____..______________________________ ________

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Where ml = m2 (density)1(velocity)1(area)1 = (density)2(velocity)2(area)2

_______________________________________.._____________.________________.

KE.= mv2 PE = mgh PE +KE +P V 1 1 = PE +KE +P V22 where V = specific l

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volume P = Pressure

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Q = de (Tout-Tin)

Q = UA (Tave-Tstm)

0 = m(h -hp)

p

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P = P 10sur(t)

p. p e /T SUR = _26.06 t

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______________________________.__________________. ________......._______

delta K = (Keft-1)/Keff CR ll-K rri) = CR (1-K rgg)

i e

e M = (1-Keffi)

SD!t = (1.K rg) x 100%.

e

.-

eff2I Keff

.(1-K i

decay constant 0.693 A = A e-(decay constant)x(t)

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In (2)

=

g

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t1/2 t1/2 Water Parameters 1?iscellaneous Conversions 1 gallon = 8.345 lbs-1 Curie = 3.7 x 1010 dps 1 gallon = 3.78 liters 1 kg = 2.21 lbs

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1.f t3 = 7.48 gallons I hp = 2.54 x 103 Btu /hr

Density = 62.4 lbm/ft3 i g.tw = 3.41 x 106 Btu /hr Density = 1 gm/cm3-1 inch = 2.54 centimeters Heat of Vaporization = 970 Btu /lbm Degrees F = (1.8) x (Degrees C) + 32 lleat of Fusion = 144 Btu /lbm 1 Btu = 778 f t-lbf 1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 f t-lbm/lbf-sec2

______________________..___________.._____________________________________

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FIGURE 1.2

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300 280 260

240

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160 140

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120

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100

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_

. _ _ _. _

i e

i

2.0 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 2.1 ANSWER a.

The ball check valve.

(0.5)

'

b.

Low nitrogen pressure (0.5)

liquid. leakage (0.5)

c.

3 gallons, causes an alarm in control room ~

(0.5)

f & gallons, causes rod withdrawal block (0.5)

Il 100 inches, causes scram (0.5)

.

-y.

L REFERENCE: C11 (CRDM) pg. 15, Rev. 3 PIS #C-11-50 (CRD) pg. 10, 12 2.2 ANSWER (0.25 for change, 0.75 for reason)

a.

Increase, due to reduced pressure on reference side of dP cell.

(1.0)

.

b.

Increase, due to reduced density in ference leg.

(1.0)

'

)(g(;

c.

No change, al1 piping in drywell is horizontal,. reference leg is outside drywell.

(1.0)

REFERENCE:

PIS #B21 (RPV Instrumentation) Figure 2, pg. 4,16 2.3 ANSWER (0.5 pts for each response)

.~

a.

1.

Separate the two-phase air-water mixture discharge from--

the water ring vacuum pumps.

(0.5)

2.

Allow short and.long-lived isotopes to decay to acceptable ~ level s.

(0.5)

3.

'To remove the nongaseous decay daughters and to-Ap.

attentuate a transient pressure wave, thus providing protection for the vessels downstream.

(0.5)

,

4.

Reduce the mositure content of the off-gas.

(0.5)

.

b.

4,2,3,1 (1.0)

-

_.

REFERENCE:

N62, (Off-gas System) Pg. 14, 5, 12, 11; Rev. 3_

t

- Section Two Continued Next Page -

-

., -

,

,. -. - -, - - - -.-

-

,,,,,. -,,,

--e

.

' 2.4 ANSWER

'10-1 and 106 cps (0,5)

a.

b.

decrease (0,5)

c.

14 LPRM and at least 2 from each level (0.5)

d.

One'of four, one of eight, one of six (1.5)

REFERENCE:

PIS #C51 (SRM) pg. 23, 7 PIS #C51-12, 13 (PRM) pg. 4, 19 PIS FC51-11 (IRM) pg. 13 2.5 ANSWER (0.5 'for each action and 0.5 for cause)

-

a.

Outboard isolation valve (F004) closes (0.5), tripping the-RWCU pump (0.5)

(1.0)

b.

Outboard (F004) and inboard (F001) isolation valve close, with either tripping the.RWCU pump.

(1.0)

'

1 REFERENCE: G-33-(RWCU) pg. 17, 16

.-

.

2.6 ANSWER (k1" a.

1.

480 volts AC power (M/G set) normal.

(0,5)

2.

480 volt AC via 480 VAC to 120 VAC transformer.

(0.5)

b.

energized (0.5)

c.

When either backup scram value is energized the air is vented from the lines.

This opens the scram valves and closes the scram discharge volume vent and drain values.

(1.0)

~ REFERENCE:

PIS-C71 (RPS) pg. 4, 5,-3 2.7 ANSWER By using - HPCI (0.25) or RCIC (0.25) systems and the SRVs in manual (0.5).

(1.0)

REFERENCE:

Fermi Exam Bank Question 2.09

-

_.

2

- Section Two Continued Next Page -

-

_.

2.8 ANSWER

-

a.

Two (0.5)-

b.

1.

Continuity lamps extincuish 2.

Stanaby liquid control injection valve ignition continuity loss alarm is annunciated.

3.

Pump discharge pressure equal to-Rx pressure + 40 psi.

4.

Pump run light ON 5.

Storage tank level decreasing 6.

.Rx power level decreasing 7.

Testable check valve indicating OPEN 8.

RWCU isolated (Any 'ix 0 0.33 pt. each)

(2.0)

s REFERENCE: C41 (Standby Liouid Control) pg.11,12, Rev. 3 2.9 ANSWER a.

Prevent water hammer by removing condensate. -

(0.5)

,

S K

c'

b.

1.

CST - less than 10,000 gallons.

(0.5)

Suppression pool - greater than 5" above normal (0.5)

2.

Suppression pool inboard isolation valve (F042) opens, Suppression pool outboard isolation valve (F041) opens, and CST isolation valve (F004) closes following F042 &

F041 full open.

(e.g.,

Suppression pool valves open proir to closure of CST valve)

(0,5)

. REFERENCE:.PIS E41 (HPC1) pg. 8, 14, 19, Rev. 3

'

2.10 ANSWER 1.

Joystick, bypasses one channel if INOP or for maintenance.

(0.5)

2.

<30*. power, RBM not required because RWM and RSCS " enforce" flux distributions.

(0.5)

3.

Selected. edge ' rod, can never produce a condition that

.

approaches a thermal limit.

(0.5)

REFERENCE: Fermi Exam Bank, Question 2.40 CC51-14 (RBM) pg. 12, Rev. 3

-

u

- Section Two Continued Next Page -

-

. _ -.

.

_

2.11 ANSWER a.

(1.0)

b.

Assures NPSH.

(0,5)

REFERENCE: PIS-P44 (EECW) pg. 14.

- End of Section Two -

.

v s

s W

I.

O O

.

&

+

.

..

.

-3.0 INSTRUMENTS AND CONTROLS e

3.1'. ANSWER

'

a.

The drift alarm 'will activate when the rod passes' an - odd

'

-

numbered read switch (0.5), or dual indication of two

~ different locations (0.5)

(1.0)

b.

No (0.5)

REFERENCE:

C11A-00 (RMCS) pg. 7, 4 j O'

3.2 ANSWER

,

a.

decrease (upstream flow element)

(0.5)

b.

increase (0.5)

c.

remain the same (downsteam flow element)

(0.5)

d.-

decrease (0.5)

REFERENCE:

PIS N21 (Reactor Vessel Level Control System)

pg. 12, 13, Figure 1.

'

,

.

3.3 ANSWER

-

t

.

1.

Low Vacuum Alarm 26.3" Hg. decreasing

/'r

'

2.

Low Vacuum Jurbine Trip - _

25.5" Hg. decreasing

/.? i 3.

Low Vacuum Bypass Valve Trip 16" Hg. decreasing c 7,._

4.

Low Vacuum Group 1 Isol 14" Hg. decreasing rg 5.

RFP Seal Water Return Pump Auto Start 7 psia 6.

RFP Seal Water Return Pump Auto Stop 6 psia

'

7.

RFPT High Exhaust Pressure Trip 0 psig (5 0 0.4 ea)

-- '

,

t.

(2.0)

-

,

r

,

REFERENCE: -INEL Exam Bank Question 5445 PIS #N61 (Main condenser & aux.) pg. 5, 6 N21-1 (feedwater) pg. 10,11,17 Procedure 20.125.01 pg. 2, rev. 3 210/R181/5.15 X

34 ANSWER GL'

-

a.

1.

SRV failure 7% flow. reactor pressure falls (0.5)

2.

CV start to respond to pressure decrease (voids)-

-(0.5)

_.

-

3.

CY close to 78% demand and presure returns -to normal (0,5)

.

REFERENCE: PIS #N30-12A (Gov. Pressure Control) Figure 1

'

- Section Three Contiunued Next Page -

,

.

_

..

.

c

L-

'

.

..,

..

3.5 ANSWER-

"

k (alarm, thermal, neutron, downscale(, Irpor 1

/

a.

(Cf# R b.

Neutron (LPRM)

'

"'

'

(0,5)

-

(0.5)

g'.

c.

APRM thermal power, APRM neutron power, APRM.downscale, INOP (1.0)

d.

1 (joystick)

(0.5)

REFERENCE: C51-13 (PRM) pg. 13, 19, 23, 15 3.6 ANSWER a.

Yes

.(0.5)

.j( hi

b.

It can be moved one notch out (0.5), before a withdraw block will block furu.*a-movement.tu.5). If the rod was inserted it will move as far as the operator wants (0.5), as long as it is not the third. insert e rror. ( 0.5 )

If it were the third

-

inser error, it would only go Une notch.(0.5)

(2.5)

REFERENCE:

INEL Exam Bank Parent Question 5443,

'C11-08 (RWM) pg. 20, 21

.-

.

3.7 ANSWER a.

true (K38 seal in)

(0.5)

b.

true (K9B TDC opens / resets)

(0.5)

c.

true (KGA,B & K7A,B close)

(0.5)

d.

true (KGA,B & K7A,8 - Trip)

(0.5)

REFERENCE:

(821-04) (ADS) pg. 6, 7, Figure 2.

Logic diagram (03-15-42-01-A2)

3.8~ ANSWER a.

Limiter el (30%)

.

--

(0.5)

r.

4.

^' ~

'#

j

g ec-b._ l'-4+a-73 (6GL)

(0.5)

- ' -

c.

No limiter (0.5)

d.

No limiter (0.5)

,

REFERENCE: _B-31 (FWCS) pg. 10, 13, Rev. 3

.

_.

-

A b

- Section Three Contiunued Next Page -

t

..

-

.

...

_ -

-. -.

_ -

__

3.9. ANSWER

-

jg#.

a.

1.

Loss of off-site power (LOP); ECCS~ premissives, no

' start.

(0.75)

2.

LOP followed by LOCA; LOCA signals-override LOP load squencing, ECCS start automatically.

(0.75)

3.

LOCA followed by LOP; CS permissive, auto start on LOP.

(0.75)

h ;I; k.

'1C) iiii.iuie inier kc' j,;j6-t-f !. 2 - prechde Op;r:t0r h

is a cen

~-inter

. %

A,,

p Ac, c,~

(0.75)

-

M: h J-j (0.z r}

'(0.25 No, 0.5 for reason)

.

't

Note: Yes also acceptable if reason corresponds properly REFERENCE: R30 (Emerg. Diesel Gen. and Aux. System) pg. 46, 47 3.10 ANSWER-

,

If both recire. pumps are running the-logic isAelayed for 2

.

-

second_s_ prior to selecthg-the loop for injection.

(0.5)

-

o[ both. pumps are NOT running then the logic is

.

If either

-

delayed. for 0.5-seconds (0.25), after which a' trip signal.is

'given to both pumps.(0.25)

(0.5)

'The logic then requires. that reactor pressure be below

!

.

psig (0.25), before continuing _to the 2 second time delay prior' to loop selection.(0.25)

(0.5)

!

Loop selection is made by. comparing the two riser

.

pressures.(0.5)

If. loop ' A'

riser -pressure is' greater than

'

loop 'B' riser pressure by a-preset differential -then loop 'A'

is selected.(0.25)

If loop ' A' pressure IS NOT greater than

~

loop. 'B' ' pressure by the preset differential, then loop '8' is selected for. injection.(0.25)

(1.0)

REFERENCE:

INEL' Exam Bank Question 5446 E11 (RHR) pg. 29, 30

~3.11 ANSWER =

gg 1.

As-long as the counts stay above 100 cps'

(0.5)

_.

.

2.

All the IRM'.s are on range eight or above

.(0.5)

a

!.

ri r u w a.,

a e

n-

,

REFERENCE: Fermi Exam Bank Question 3.07 (rewrite)

t

- Section Three Contiunued Next Page -

--

I

,..

..3.12 ANSWER

a.

RHR pump A-(0.5)

b.

Provides signal of pump running to ADS logic (0.5)

- c.. F007A(B) will auto open if loop flow is <2200 gpm for.10.

seconds if. either pump.is running in that loop as indicated

.

. by; closure of pump breaker.(0.5)

Will auto close on ' loop flow is ~> 2200 gpm.(0.5)

(1.0)

REFERENCE: Fermi Exam Bank Question 3.21

.

- Er.u of Section Three -

,

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_

4.0 PROCEDURES: NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL CONTROL 4.1 ANSWER-1.

Verify automatic actions 2.

Depress both manual scram pushbuttons 3.

Verify reactor power is decreasing 4.

Place mode switch in refuel and verify all control rods are in 5.

Place Rx mode switch to shutdown 6.

Maintain water level between level 3 and level 8 7.

Maintain reactor pressure between 900 psig and 1075 psig 8.

Insert SRM and.IRM detectors and range the IRMs 9.

Trip the main turbine 10. Verify recirculation pumps have run back to min. speed

11. Run program 0D7, Option ~2 12. Notify S.S. and may require classifying in the EPIP EP-101

-

'

-(Any 8 for full credit)

(2.0)

REFERENCE: AOP 20.000.21 pg. 1, 2

4.2 ANSWER

.'

1.

Avoid possibility of " fooling" LPCI loop selection logic (1.0)

2.

Back. flow through idle or reduced flow jet pumps can cause jet pump vibration and therefore unnecessary stress on jet pump components.

(1.0)

3.

Provides for some coastdown flow from the unbroken loop during a long break accident.

(1.0)

L r...

REFERENCE:

831 (RFC) pg. 9, 10 Rev. 3 T.S. 3.4.1.3 4.3 -ANSWER a.

1.

If the SRY cannot be closed within two (2) minutes (0.5)

2.

If torus temperature exceeds 95*F.

(0.5)

b.

1.

Decrease recirc. flow to minimum (0.5)

-

2.

Place mode switch to shutdown (0.5)

REFERENCE: A0P - 20.000.25, Pg. 1

_.

.

-

- Section Four Continued Next Page -

.

,

,_.

-

_

-

.

4.4 ANSWER

.

1. - Read the RWP (0.25), and resolve question (0.25)

(0.5)-

2.

Sign (0.25), "RWP. Acknowledgement Log Sheet" (0.25)

(0.5)

3.

Review copy of RWP (0.25), at work area (0.25)

(0.5)

REFERENCE:

.T,5. 4,7.1.2. pg. 379 7-S AP 12.000.13 pg. 10

.4.5 ANSWER ~

a.

1.

Reactor power cannot be determined.

(0.5)

.

2.

Reactor power > 6% (0.5) and torus water temperature reaches 110'F (6!5).

(1.0)~

err')

'Glt()

To concentrate the boron liqui (d. oveo e d ',*

<8~

(0.5)-

ggjpor b.

c.

When the SLC tank level indicates zero (0.5)

REFERENCE:

E0P'29.000.08, pg. 1-4

,.

.,

1-4.6 ANSWER

-

C MODE SWITCH AVERAGE REACTOR

'

CONDITION POSITION COOLANT TEMPERATURE

-

1. F0WER OPERATION Run Any Temperature (0.5)

2. STARTUP Startup/ Hot Standby Any Temperature (0.5)

3. HOT SHUTDOWN Shutdown

> 200*F (0.5)

4. COLD SHUTDOWN Shutdown

< 200*F (0.5)

5. REFUELING Shutdown or Refuel

][ 140*F (0.5)

(0.25 for mooe position and 0.25 for temperature)

REFERENCE: Technical Specification 4.7l ANSWER 1( h* '

a.

When the nitrogen purging gas pressure remains constant. t r (1.0)

t.a. a dn.

v. v. - W-wn b.

Gas expansion lower the temperature and you nust wait for it

-

to come to ambient equilbrium.

(1.0)

..

.

REFERENCE: S0P 23.106 pg. 9, 10

-

Section Four Continued Next Page -

-

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

.

~

.

'

4.8 ANSWER.

-

,,

1.

Torus water level > +2" (124,220 ft3)

.

2.

torus water level < -2" (121,080 ft3)

f h,-

3._ - Torus water average temperature > 1-39Y-d^~,

.

4.

Drywell atomsphere average temperature > 135*F 5.

Drywell pressure > h88 psig

.Q,

/A3 gas I.E3 wO -

(0.25 each condition, 0.25 each setpoint)

(2.5)

-REFERENCE: E0P 29.000.03, pg. 1-4.9 ANSWER

1.

Time (0.5)

2.

Critical rod position information, rod secuence, rod group, rod and rod position.

(0.5)

3.

Reactor coolant temperature (0.5)

'

4.. Reactor perico

.i (0.5)

Jga REFERENCE: GOP 2'2.000.03, pg. 6 (Startup from Cold Shutdown)

4.10 ANSWER

-

a.

(0.5)

b.

1.

Misoperation in autiomatic is confirmed (0.5)

2.

Adequate core cooling is assured.(0.5)

(1.0)

c.

Will not auto initiate (0.5)

REFERENCE:

50P 23.205, pg. 14-4.11 ANSWER a.

(0.5)

b.

(0.5)

c.

(0,5)

.-

-

d.

(0.5)

e.

(0.5)

_.

-

REFERENCE: FERMI Exam Bank Question 4.08

'

.

- Section Four Continued Next Page -

,

-

,

-

-.

.-,

,--,--

.,

1m-

-,

. _.

.

4.12 ANSWER High hydraulic. forces.

(0.5)

REFERENCE: FERMI Exam Bank Question 4.12

- End of Section Four -

,

- END OF EXAM -

c a

a e

?

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.

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$

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._

,

_

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_. _ _ _ _ _ _ _ _

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,

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U. S. NUCLEAR REGULATORY. COMMISSION SENIOR REACTOR.0PERATOR LICENSE EXAMINATION FACILITY:

FERMI 2

_________________________

' REACTOR TYPE:

BWR-GE4

_________________________

DATE ADMINISTERED: 85/02/12 l

_________________________

EXAMINER:

LANG,T.

^

___.________________L____

APPLICANT:

_________________________

INSTRUCTIONS TO APPLICANT:

__________________________

Uno separate paper for the

,swers.

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 tho examination starts.

.

-

% OF CATEGORY

% OF APPLICANT'S CATEGORY VALUE TOTAL'

SCORE VALUE CATEGORY

________ ___________________________________

________ ______

___________

________

___'88

________ 5.

THEORY OF NUCLEAR POWER PLANT 25 00

__

___________

OPERATION, FLUIDS, AND-THERMODYNAMICS

,

25.50 25.37

________ 6.

PLANT SYSTEMS DESIGN, CONTROL,

________ ______

___________

AND INSTRUMENTAlION 25.00 24*88

________ 7.

PROCEDURES - NORMAL, ABNORMAL,

________ ___ __

___________

EMERGENCY AND RADIOLOGICAL CONTROL

,

25.00 24.88 8.

ADMINISTRATIVE PROCEDURES,

________ ______

___________

_-__---

CONDITIONS, AND LIMITATIONS

'2100.50 Il00.00 TOTALS

________ ______

___________

________

FINAL GRADE _________________%

[

g All work-done on this examination is my own. I have neither

-

_

given not received aid.

APPL 5CIUT 5'5 GUITUR5-~~~~~~~~~~~~~

I

<

s i

%C7 O/3C6cCE j

.

1-

.

5.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE

y'

__--__--------

l-

$

GUESTION 5.01 (1.50)

For the following conditions, state if the FUEL TEMPERATURE COEFFICIENT

'.

-

becomes more nesitive or less nesitive.

A.

INCREASE in fuel temperature.

(0.5)

,

B. INCREASE in moderator temperature.

(0.5)

l C.

INCREASE in void fraction.

(0.5)

--

.

h QUESTION 5.02 (1.00)

Define the following terms *

A. Latent heat of vaporization.

(0.5)

B. Steam Duality.

(0.5)

3_

i

=

QUESTION 5.03 (3.00)

Explain the effects of Increasing the followins core parameters on Z

steady state critical power.

A. Core Flow.

(1.0)

e B. Inlet Subcooling.

(1.0)

C. Reactor Pressure.

(1.0)

_.-

QU.ESTION 5.04 (2.50)

1r equalibrium xenon is obtained (the reactor has been. operated at

{

constant power for many hours), and the reactor power is doubled, will the new equalibrium xenon consentration be twice as great? Explain your

}

answer.

f; 00ESTION 5.05 (3.00)

x A.After making a rod notch with the reactor critical, you notice a 100

-

second period. How much reactivity was added by the rod notch? (assume i

BOL).

(1.0)

B.After a reactor scram from power the shortest stable period possible c

is -80 seconds. Explain this statement.

(1.0)

i

>

l C.Is the initial period immediately followins the scram shorter than

(1.0)

-80' seconds? Explain your answer.

-

,

-!

-

'

.

1-

'

=

b

'=

.

-

_....

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

5.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE

GUESTION 5.06 (2 00)

A common misconception regarding rod worth is that, 'if the neutron flux increases in the vicinity of a rod, the rod worth (differential)

of that rod increases.' Explain why this statement is incomplete in oxplaining rod worth.

,

QUESTION 5.07 (2.00)

A.What are four indications inside or outside the control room, which could be used to determine if a pump is cavatating?

(1.0)

B.What is cavitation?

(1.0)

QUESTION 5.08 (1.50)

,

SBLC is designed to reduce reactor power to zero level, while allowing the neuclear system to cool to room temperature with the control rods

,

romaining withdrawn. To do this several positive reactivity effects must

'

bo overocome. What are six of these effects?

DUESTION 5.09 (3.00)

.You increase-core power by pulling. control rods around the center fuel bundle. Assuming that recirculation speed is kept constant, would the

-

flow through the center bundle increase, decrease, or remain about the ocee? Explain your answer.

'

00ESTION 5.10 (2.50)

The Reactor is on a 100 second period.

Moderator temperature is 160 des F.

With no operator action:

a.What is the associated reactivity at this period?

(1.5)

,

b.What will be the moderator temperature when the reactor is again on l

an infinite period?

(1.0)

Assume BOL'for time in Reactor Life.

State.any assumptions you make.

SHOW ALL WORK.

-

.

-

_

_ _ _ _ _ - _..

.

. _

.

5o THEORY OF. NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE

g________..____________________________

____

______________

QUESTION 5 11 (3.00)

The reactor is shutdown by 5% dk/k and the SRM's indicate

'10 cps.

If K eff of the reactor is' increased to 0.981

-

WHAT should be-the approximate count rate? (Show all work.)

(1.5)

a.

b. What is the original K eff?-

(1.5)

,

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6.

PLANT-SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE

______________________________________________________

SUESTION 6.01 (3.00)

What parameters'will be indicated on the Rod Block Monitor meter

.uith the meter switch in each of the following positions:

c.

Input b.

Count c.

Reference

d.

Block

.o.

Flow f.-

Average (3.0)

DUESTION 6 02 (2 00)

,Briefly describe the purpose of the following UPS switches:

A. Static Bypass Switch.

(1.0)

B. Static Trznsfer~ Switch.

(1.0)

.

QUESTION 6.03 (3.03)

With regard to the core spray system:

A.

State the automatic initiation signals of the system.

(1.0)

B. In the event that a Core Spray pump CMC switch is in DFF when an initiation signal is received, will placing the CMC switch-(.

in AUTO cause the pump to start? EXPLAIN.

(1.0)

C. Which valve (s) receive a CLOSE signal from the Core Spray logie upon receipt of an initiation signal?

(1.0)

D'JESTION 6.04 (2.00).

Explain how the Logic of the Reactor Protection System is altered by installing the shorting links in the Neutron Monitoring System i.e.,

links installed verses links removed.

.

-

.

-

-.

,

-

- - - - -

. _. -

_

, _ _,_

_

.

.-

.

_

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.

_

-

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6.

PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE

______________________________________________________

,

GUESTION 6.05 (2.50)

If1during.a unit start.up you receive a spuricus Group It A. What valves would you expect to close?

(1~.0)

B. What'are six_ signals you would check to determine the cause of the Isolation?

(1.5)

GUESTION

.6.06 (3.50)-

,

Which Reactor Feedpump trips provide for protection from:

A. Potential cavitation damage? (Three required for full credit).

(1.5)

'

B.; Potential' turbine blading damage? (Four required for full

'

(2 0)

credit).

- QUESTION.

6.07-(3.00)

Iar each of~the RCIC (Reactor Core Isolation Cooling) System-component

.

failures ~1isted below. STATE WHETHER.OR NOT RCIC WILL AUTO INJECT into

the reactor vessei.IF IT WILL NOT INJECT, STATE WHY, AND IF IT WILL INJECT,' provide CNE. POTENTIAL ADVERSE EFFECT OR CONSEQUENCE OF system.

_cperation with_tne failed component. Assume NO OPERATOR ACTION, and the

. component is in the failed condition at the-time RCIC receives the AUTO

.

Linitiation. signal. Consider.each item seperately.

A. The Barometric Condenser VACUUM PUMP fails.

(1.0)

B. The MINIMUM FLOW VALVE fails to AUTO open (STAYS SHUT) when system

conditions require it to be open.

_

(1.0)

C. The RCIC PUMP DISCHARGE' FLOW ELEMENT OUTPUT SIGNAL (to the RCIC flow controller) is f ailed to its-MAXIMUM output.

(1.0)

r

,

GUESTION 6.08 (1.00)

Staie four(4) EMERGENCY DIESEL GENERATOR trips that WILL NOT shutdown

_the engine with an AUTO start condition present.

QUESTIDN 6.09 (3.00)

>

What are the conditions which will cause Recirculation Pump Run Back

~

.cignals to be initiated. Be specific. Match speed limit to. condition-cnd state which runbacks must be reset manually. (Three required for.

. full credit).

.

_

i-

- -., _.. -

.

-.__._. _,- ~. _ _ _. _ _, _ _ _

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

-.-

-

.

-

.

,

6.

PLANT. SYSTEMS. DESIGN, CONTROL, AND INSTRUMENTATION PAGE

.-----------------------------------------------------

QUESTION 6 10 (2.50)

If you receive the followins alarm:

XXXXXXXXXXXXXXXXXXXXXXXXX X

APRM FLOW UNIT X

X INOP X

X X

'

XXXXXXXXXXXXXXXXXXXXXXXXX A. What AUTOMATIC action will result from the alarm?

(0.5)

B. Other than the INOP, what are two (2) signals associated with the flow units which will cause the SAME automatic action to occur as in(1.0)

part a'

above?

'

C. What are two (2) conditions which will cause an INOP trip of a flow (1.0)

unit?

.

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.

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.. _.,

. _..

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..,.

.

i i

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE-

RA6E6E6656AL C6 TR6t

~

~~~~

____________________

~

QUESTION 7.01 (3.00)

What are the entry _ conditions for E0P 29.000.08 ' Reactivity Control'?

OdESTION 7.02 (1.00)

'

FA? cording to procedure 20.129 (Loss of Station and/or Control Air Pressure) when are you required to manually scram the reactor?

"

QUESTION 7.03 (2.50)

List five (5) indications / symptoms ~of a jet pump malfunction.

.0UESTION - 7.04 (2.00)

If the Standby Liquid Control System is initiated, what are five (5)

,-

indications or parameters that should be checked or verified to insure j

.the system is' operating properly?

(0.4 each)

QUESTION 7.05 (1.50)

If plant. conditions are such that the control room must be evacuated,

'what are three (3) steps which should:be performed prior to evacuating?

OhcSTION~ 7.06 (2.00)

What must be done if MORE THAN ONE withdrawn control rod has an INOPERABLE. SCRAM ACCUMULATOR while operating the plant at power?

GUESTION 7.07 (2.00)

'With regared to EOP 29.000.08,. REACTIVITY CONTROL

~

If during a transient the Reactor fails to scram by a VALID _ scram signal,-WHAT condition (s) require the initiation of the Standby liquid

,

Control System and the isolation of the Reactor Water Cleanup System?

.

'e

L i'

.

......

..

__ -.__._,

.

- - ~

.

..

.

.-

-. -.

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. -..

7.

' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

,

~~~~~~~~~~~~~~~~~~~~~~~~

~~~~ I656E655EIE~E6UTR6L R

____________________

,

QUESTION 7.08 (1.50).

Under'what conditions-is an operator allowed to " DEFEAT THE AUTOMATIC ACTION OF AN~ECCS*?

QUESTION 7.09 (2.5';)

,

What are the entry conditions for EOP 29.000.03 ' PRIMARY CONTAINMENT CONTROL'?

QUESTION 7.10 (3.00)

A reactor water isotopic analysis for iodine and specific activity is roquired when any of.three-(3) conditions occur. What are the three (3)

' ecr.d i t ons?

QUESTION 7.11 (1.40)

How is control rod coupling verified?

QUESTION 7.12 (1.00)

Roactor level indicatrots can-be expected to___________.once drywell tcmperature reaches the. reactor vesseld saturation temperature.

(Choose one)

A. Read lower than actual vessel level.

B. Read' higher than actual vessel level.

C. fail low.

D.

fail to respond to level changes.

.

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!

W

-. -, - - - -, -..,.

-

- - - - _ -.,

.,,

,

.

f

7.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

--- EA5i5E55iEsc E6siE5E------------------------

>

____________________

QUESTION 7.13 (2.00)

Rosarding the RHR SDP 23.205 when operating in the SHUTDOWN COOLING MODE:

A.

You are cautioned to NOT use Outboard Injection Valve E11-F017A/B Jto control cooldown rate. What adverse effects could result, according to this procedure?

(1.0)

'

B. This procedure also cautions you to exercise care during one portion of a valve realignment to avoid RAPID DRAINING' of the REACTOR VESSEL. Describe THREE FLOW PATHS associated.with the RHR-system that could result in RAPID DRAINING of the REACTOR VESSEL.

(1.0)

.

?

.

O e

-

, - - -. ~, -

, - - -. - -. - - - -

-. _..

.

_

8.-

ADMINISTRATIVE PROCEDURES, CONDITkONS, AND LIMITATIONS'

PAGE

gjf, g/ 9c:xY f uc pt y,:.., g ec g QUESTION 8.01 (2.00)

.

S#i'

A. If the reactor is operating a^. equal to or greater than,3Ef RATED THERMAL POWER and under a LIMITING CONTROL ROD PATTERN, what action

.would you take if a RBM were to fail? Be specific.

(1.0)

B. What is a LIMITING CONTROL ROD PATTERN?

(1.0)

'

DUESTION 8.02 (2.00)

'A. If a temporary procedure change is required on the backshift, who~

must sign the ' Temporary Procedure Change Request *?

(1.0)

B. What limitations are placed on making temporary changes?

(1.0)

.

CUESTION

.8.03 (2.50)

lA. On shift operations personnel and security are granted UNLIMITED ACCESS to the control room. What are four (4) others (by title) who (2.0)

are also granted UNLIhITED ACCESS?

B. Who has authority to EXCLUDE non-essential personnel from the control room when their presence is hampering the operation of the (0.5)

control room?

8.04 (2.00)

}"ESTION What are the Technical Specification limits for Reactor Coolant System loakage?

C:lESTION 8.05 (2.00)

Dafine the followins ter*S*

(1.0)

A.

Core Alteration.

(1.0)

B.

Shutdown Margin.

>

.

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d

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-

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-

,

-

,.

,

-.

.

4'

8. - ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 12-

l GUESTION 8.06 (1.00)

Rogarding the TECHNICAL SPECIFICATION curves for MINIMUM TEMPERATURE FOR CRITICAL operation:

A. What could happen if critical operation was conducted LESS THAN the minimum temperature specified?

(1.0)

B. Is this temperatture expected to change over core life and if so, why?

(1.0)

GUESTION B.07 (3.00)

Briefly explain WHY each of the following-RECIRCULATION PUMP STARTING-LIMITATIONS are necessary:

A.-The pump in an-idle recirculation loop shall not be started

,

unless the temperature of the coolant within the idle and

-

operating loop are within 50 degrees F of each other.

(1.0)

.

B. When in one pump nperation, the idle pump shall not be started i.

unless the active pump speed is reduced to less than 50% rated

.

speed.

(1.0)

C. Recirculation loop flow mismatch shall be maintained within 10%

of rated recirculation flow with core flow less than 70% of rated core flow.

(1.0)

GUESTION 8.08 (2.00)

As per Technical Specification definition, what is the difference batween Hot Shutdown and Hot Standby?

.

e e

+

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-

I

. -., -

,.

-

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

.

.,

.

..

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_. _. _

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.

-

.

8.

ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE

__________________________________________________________

l QUESTION 8.09 (2 00)

The statements below define each of the four (4) emergency classifications. Match the classification (i.e.,

SITE AREA EMERGENCY, UNUSUAL EVENT, GENERAL EMERGENCY, OR ALERT) aplicable to each statement.

A. Does not require activation of the emergency centers (TSC, or (0.5)

EOF or OSC).

B. Major failures of plant functions needed for protection of the public, but the release is not expected to exceed the upper PAG's (5 rem whole body, 25 rem thyroid.)

(0.5)

C. Limited release of radioactivity in excess of Tech. Specs. limit is probable. Does not require protective actions be recommended (0.5)

to the state and county authorities.

D. A release of airborne activity which results in off-site exposures in excess of the limits specified in the USEPA

'

(0.5)

Protective Action Guides.

.

QUESTION 8 10 (2.00)

Ragarding Limiting Conditions for Operation, 3.6.2.1, for Suppression Chamber The maximum average temperature during Operational Conditior.s 1 or 2 is ____ dest ees F, except that the maximum average temperature may be permitted to increase to____ degrees F during testing which adds heat to the suppression chamber or to ____ de3rees F with thermal

.

less.than or equal to 1% of RATED THERMAL POWER or to ____

power degrees F with-the MSIV's closed following a scram.

QUESTION B.11 (1.50)

The reactor at Fermi 2 has several power distribution limits. One of these is Linear Heat Generation Rate.

(0.5)

A. What is the limit for LHGR?

(0.5)

B. When does this limit apply?

.

(0.5)

C..What must be done if this limit is exceeded?

.

sp

_

?

_

- - - _. _ -. -

. _ _ _

_

_. _

-

8.

ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE

__________________________________________________________

QUESTION 8.12 (1.50)

Answer the following questions as they pertain to Containment Systems Toch. Specs.:

A. How many Suppression Chamber-- Drywell Vacuum Breakers are required to be operable in Operational Condition 3?

(0.5)

B. How many Reactor Building -- Suppression Chamber Vacuum' Breakers are required to.be operable in Operational Condition 3?

(0.5)

C. When must the Drywell atmosphere be less than 4% 0xygen'by volume?

(0.5)

QUESTION 8.13 (1.50)

.Mosarding PLANT STARTUPS:

A.

Who must REVIEW the REACTOR STARTUP MASTER CHECKOUT 22.000.01 upon completion?

(0.5)

B. In the REACTOR STA'RTUP MASTER CHECKOUT, who determines which

. systems are required to be lined up?

(0.5)

C.

Followins a reactor SCRAM, who has responsibility for authorizing a return to power?

(0.5)

.

-

4 g

_

(

f-

.

'

.

,I

-

.

5.

. THEORY OF NUCLEAR POWER PLANT 0PERATION, FLUIDS, AND

',

PAGE

.--------------

ANSWERS -- FERMI 2-85/02/12-LANG,T.

ANSWER 5.01 (1.50)

A.less B.more

C.more

.5 each

.

. REFER'r:N CE Formi-Question Bank (5.02)

ANSWER 5.02 (1 00)

A. Energy necessary to convert a-unit of a substance from a saturated

^

' liquid to a saturated vapor.

(0.5)

B.The ratio of the mass of steam to the mass of the total fluid.

M(steam)

X=---------------------------

M(steam)+M(liquid)

(0.5)

REFERENCE Formi Question Bank (5.08)~

ANSWER 5 03 (3.00)

A.Increasiris core flow will cause critical power to increase due to

<

the increase heat removal capability.

(1.0)

D' Increasing subcooling causes an increase in critical power.-The inlet

. enthalpy will be reduced and the heat which can be removed will increase.

(1.0)

C. Increasing Reactor Pressure reduces the energy to be at critical

]

quality therefore critical power will be lower.

(1.0)

,

REFERENCE

'

Fermi Question Bank (5.09)

l

..

j

-

.

'

-

t'

i

-. -

-.

.

. _...

_.

. - - _ _.

_

-.

__ _

..

. _ _ _ _ _ - -

.

5.

THEORY DE NilCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE' 16

_-

ANSWERS -- FERMI 2-85/02/12-LANG,T.

' ANSWER 5 04 (2.50)

No. The production rate is directly proportional to power level, but removal rate is preportional to xenon concentration and it contains a ppwer dependent ter m, thermal neutron flux. Since flux is directly-proportional to power level the burnout term becomes more significant.

-This results in an equalibrium xenon value which is higher than the original value, but not twice as high.

REFERENCE Formi Question Bank (5.14)

A'NSWER 5.05

'(3.00)

A.

T=B p/xp.so p =B/xT+1-x= Lambda assume B=.0072 (BOL)

p=.0072/(100)(0.1)+1=6.545 x 10E-4 delta k/k (1.0)

B.After'the initial prompt drop, power can not decrease faster than the longest lived delayed neuton appears.

(1.0)

C.Yes The initial drop in power will only be due to the prompt neutrons.

(1.0)

.

REFERENCE G3neral Theory ANSWER 5.06 (2.00)

The worth depends on the neutron flux in its location compared to the overage neutron flux in the core. Thus, as power is increased, the flux 01so increases. This will increase the worth of the rod if the local flux is increased more than the core average flux.

REFERENCE G3nerdl Theory

-

,

.

-

k

b 5.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE

_----

CNSWERS -- FERMI 2-85/02/12-LANG,T.

ANSWER 5.07 (2.00)

A. Vibration, noise, lower or fluctuating amperage / flow /ditcharge pressure.

(1.0)

B.The formation of bubbles of steam in the suction of a pump due to operation with insuffiecent pressure. The bubbles collapse in the volute or high pressure region.

(1.0)

REFERENCE General Theroy ANSWER 5.08 (1.50)

Any six for full credit.

A. Complete decay of rate,d power xenon inventory.

.B.The elemination of steam voids.

C.The changins water density from hot to cold.

re. Reduced doppler effect.

i.The reduced neutron leakage from boiling to cold.

F. Decreased control rod worth as moderator cools.

G.Suffiecent reactivity added to insure shutdown margin met. [~MMM pas;v v6 rum rivt ' ~

O f Lc-7

aC Futt Y m ;WMO '/ -

REFERENCE Goneral Theory

' ANSWER 5.09 (3.00)

Once the control rods are pulled steam forms in the coolant channel, it forces a volume increase in the coolant, the more steam the greater.the volume. The greater the volume, the greater the frictional pressure drop

. cost be in the channel. But the pressure drop across the core can not be changed by any one bundle. The pressure drop across-a bundle is locked botween the pressure in the lower plenum and the pressure inside the chroud. This means that as power and consequently steam production are increased in the bundle, the pressure drop can not change. If the pressure drop can not change then flow must decrease. Core orificing was installed to minimize the effects of two phase flow.HAs the pressure drop due to two phase flow increases bundle flow decreases. However, as bundle flow d: creases the pressure drop across the orifice decreases. The total effect io that bundle flow will remain about the same.

_-

NOTE Shorter e>:plainations would be excepted.

.

- _ _ _ _ _ _ -

- _ _ _ _ _ _ _ _ _ _ _ _

!

5.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE

_---__

ANSWERS -- FERMI 2-85/02/12-LANG,T.

ANSWER 5.10 (2.50)

Assumptions:

Lambda = 0.1 see -1 (given)

0.0070 (0.25)

Beta Bar

=

Alpha T = -1 X 10 E-4 (0.25)

Period = 100 see (given)

(0.5)

T= (B-p)/(.p):

solve for p p = B/1+T.:

p = 0.0070 / 1 + 100 X 0.1:

p = 6.3 X 10 E-4 delta K/k (1 0)

NOTE: The second part will be graded independent 1v of the first part.

dalta T mod = p / alpha T

= 6.3 X 10 E-4 delta K/K / 1 X 10 E-4 delta K/K des F

= 6.3 des F,

Moderator Temp = 160 des F + 6.3 des F = 166.3 des F (1.0)

REFERENCE Roactor Theory ANSWER 5.11 (3.00)

formulas!

CR2(1-Keff2)

p =-(Keff - 1)/Koff CR1(1-Keff1)

=

.05 = (Keff-1)/Keff 10(1.952) e CR2(1.98)

.05 = 1 - 1/Keff 10(1.952)/1.98 = CR2-1.05 = -1/Keff

.48/.02 = CR2-1.05/Keff = -1 24 = CR2 (1 5)

Koff = -1/-1 05 Keff =.952 (1.5)

(The two calculations will be graded seperately)

REFERENCE Reactor Theory

_

-

.

s

_

6.

PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE

. ANSWERS -- FERMI 2-85/02/12-LANG,T.

ANSWER 6.01 (3.00)

o.

Input - Any of the LPRM inputs b.

Count: -The number of LPRM inputs which are operable c.

Reference:

The reference APRM input d.

Block - The trip level reference

.o.

Flow:

The flow input to the slope and bias circuit

't.

. Average:.The RBM channel output (0.5.each)

(3.0)

REFERENCE RBM Lesson Plan f

ANSWER'

6.02 (2.00)

A. Transfers load to an alternate AC supply should the inverter voltage dr op. below a ' pr e-set ' level. Mv+4-bewanuall" ee et4eet-oce-nomi power-supply.

Q m.

>> o eg. y FL-DC /IW' N,mr New A 's " U L B. Automatically transfers the load from the inverter to the voltage r e g u l a t o r i-f-e n y --e f--the--f ol-low i ng-e x i st4 w/:. t i nen r 1.Undervoltage a) slow

@bWN CU Am M W@

W

{

-b) fast WHM co w C> t T i p o L.

2. Abnormal Frequency 3 0verload

'EFERENCE Formi Question Bank (6.36)

ANSWER 6.03 (3 00)

A. Rx low low low level-level 1 UwF CC uA e

(v:

.. J (1.0)

B.~No. The CMC will have to be placed in RUN to cause the CS pump to sta'et. The ' Manual Override Seal-In' indicator above the pump CMC will be illuminated.

(1.0)

C.-Full flow test valves, F015A/B.

(1.0)

.

e REFERENCE u-Formi Questin Bank (6.37)

t

_

.

>

~6.

PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE

..-----------------------------------------------------

. ANSWERS ---FERMI 2-85/02/12-LANG,T.

,

.CNSWER.

.6.04 (2.00)

With shorting links installed:

A.SRM Hi-Hi. scram is bypassed

.

(0.5)

B. Neutron Monitoring System for RPS is one out of two-twice'los,ic. (0.5)

"

With shorting links removed:

Any one of.the 4 SRMse ori any one of the 8 IRMs, ori any one of the

.

6LAPRMs is capable of imposing scrams if their trip points are exceeded i.e. non-coincident logic.

(1.0)

REFERENCE-

- Formi Guestion Bar (6.28)

ANSWER 6.05 (2.50)

a MSIV's

-

(0.5)

'.. :MSL Drain' Isolation valves.

(0.5)

'B.

1. RPV low-low level Any six for full credit.

2. Main steamline high radiation-3. High steamline flow

,

4. High steamline tunnel temperature 5. Low-steamline. pressure 6.-Condenser high pressure 7. Turbine building high area temperature (tunnel)

(0.25 each)

,

REFERENCE Formi Question Bank (6.19)

ANSWER 6.06 (3.50)

.A.

Main condenser hotwell level low.

All heater feedpumps off.

.Feedpump suction pressure Iow.

(0.5 each)

B. Exhaust hood temperature high.

Exhaust hood pressure high.

Overspeed.

(0.5 each)

Res~ctor vessel water level high.

-

REFERENCE

-

Fermi Question Bank (6.20)

,

~-,r,e

- - -, -

,v,

,,- -

-.v.+n-

---n-----c.

,--.-.-c-w

--_- - -

v,-,

,.~,--r,.,,~w,m,-

--,--,c,-

--n-~

r-n-

-, -

e---.:

-.- -.

.

_

E6.

PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE

CNSWERS -- FERMI 2-85/02/12-LANG,T.

ANSWER 6.07 (3.00)

A. Will inject. Turbine seal leakage resulting in potential airborne activity in the RCIC room.

(1.0)

B. Will inject. Pump overheating and seal damage result during low or

~

'

non-flow' conditions.

(1.0)

C. Will not inject. Maximum signal from the flow element will result in the flow controller keeping the turbine speed at minimum.

(1.0)

REFERENCE Formi Question Bank (6.16)

ANSWER 6.08 (1.00)

i. Etr% few f%

W A

or for full credit.

(.25 each)

C #ch 1. Lo ibe oil pressure.

og v+tica.

2. Overspe 3, y,,,f,f.o#t m.A 3. High crankca ressure.

q d u M.pd f urd b.

4.

Generator differe

' a'1.

../%

g g, g (;,

5. Overcrank (time delay

.

,

b " ~"' N " '

W 'E

WM. blU *t REFERENCE

'

s Formi Question Bank (6.17)

7, !/wl Oml. jr.wa.6-M

.

H,io k I'

" * "r.," " " P ' ' "s,

e s

ANSWER 6.09 (3 00)

A.

Feedwater flow decrease below 20% or discharge valve not full open, runback to minimum speed, 30% -- signal--euto-reset. vwi met (1 0)

s D. Reactor level decrease below level 4 and RFP_. trip,, runback to,45%

speed,.si nal manually reset.

j,:-[.

g' pg (1.0)

,

C. Loss of heater drains, runback to 80% speed, signal manually reset.(1.0)

REFERENCE Formi Question Bank (6.15)

.

e

.;,'

e

W (

__

i 6.-

PLANT SYSTEMS DESIGN, CONTROL', AND INSTRUMENTATION PAGE

ANSWERS ~--- FERMI 2-85/02/12-LANG,T.

ANSWER 6.10 (2.50)

A..An alarm and a Rod Block will bu generated.

". > 10% difference between two flow units. 4 - '

h t ce<,,Ye Nua t:. ' >f.

..

High flow (100%).

om i a r._ feu

,

C. Module unpluged.

Mode switch out of operate.

REFERENCE Formi Question Bank (6'.08)

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.

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.

- -,

.-.

.

..

.. _ _,

--.

-

. - -,

,

..

.

7.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

~

~~~~~~~~~~~~~~~~~~~~~~~~

R565UE665C5L 66NTR L

~~~~

____________________

CNSWERS -- FERMI 2-85/02/12-LANG,T.

CNSWER 7.01 (3 00)

A. Receipt of a full scram signal and sustained APRM indication-is greater than or equal to 6%.

(1.0)

B. Receipt of.a full scram signal and more than.one rod does not insert to position 04 or less.

(1.0)

C.-Recept of a full scram signal and reactor power cannot be determined.

(1.0)

REFERENCE Formi Question Bank (7.44)

ANSWER 7.02-(1.00)

When the scram air heade'r low pressure alarms annunciates and more than ene control rod drift alarm is received.

REFERENCE Formi Question Bank (7.42)

ANSWER 7.03 (2.50)

Any five for full credit.

1. Unplanned change in core flow.

2. Unplanned change in recirculation loop flow.

3. Unplanned decrease in core differential pressure.

4.A~ Jet pump percent differential pressure deviates excessively from the average of the remaining pumps.

5. Unplanned decreased in reactor power.

6. Unplanned decrease in MWE output.

REFERENCE

.Formi Question Bank (7.39)

~

ANSWER :

7.04 (2.00)

l-Loss of squib valve continuity /A p ^ b

A.

B. In]ection check valves indicate open.

-

C. Pump run indication (run light and amps).

~'

D. Tank level decreasing.

-

E. Power decreasins y,G"' k

.

(0.4 each)

l p,

74,,

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'e n t,, M w e $r*

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);, 0 25 IS'l EU

,,,

pcf (, ? :M. *A ! =

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. - - - -.

-

.

7.-

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

~

~~~~~~~~~~~~~~~~~~~~~~~~

R5656L66565L 66UTR6L

~~~~~

____________________

ANSWERS -- FERMI 2-85/02/12-LANG,T.

REFERENCE Formi Question Bank (7.36)

ANSWER 7.05~

(1.50)

,

A.

Place the mode switch in SHUTDOWN.

B. Depress both manual scram pushbuttons.

C. Arm and depress the main turbine trip pushbuttons.

REFERENCE Formi Question Bank (7.34)

ANSWER 7.06 (2.00)

A.~ Declare the rods inop and immediatelyi (0.5)

B. Verify one CRD pump i.s operating by inserting a withdrawn control rod at least one notch, (0.5)

OR Place the Mode Switch in the SHUTDOWN position.

(0.5)

l C.. Insert inop rods and disarm electrically and manually.

(0.5)

REFERENCE Formi Question Bank (7.20)

ANSWER 7.07 (2 00)

A. If Rx power cannot be determined (0.5)

OR B. If MSIV's closed--or--condenser not available (0.5).

AND Rx power greater than or equal to 6%

(0.5)

AND Torus temperature reaches 110 degrees F (0.5)

l REFERENCE Forni;Guestion Bank (7.16)

.

ANSWERu 7.08 (1 50)

.

Ccnfirm by at least two independent indications that!

(0.5)

1. Misoperation in auto initiated; or 2.. Adequate core cooling is assured.

(0.5)

,

,, -.

,. - -,.,,,,,., - - -

.

_

..

-

..

.

.j.

7.

' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

RA656L G5 CAL"66NTR6t

~~~~

____________________

-ANSWERS - < FERMI 2-85/02/12-LANG,T.

REFERENCE Forni Question Bank (7.14)

,

ANSWER-7.09 (2.50)

.

Five for full; credit.

A. Torus water level above

+2, (434t40^ T t 0 )

B. Torus water. level below-2, (4204 800-f44 )

'C. Torus water average temperature above 95 degrees F D. Drywell atmosphere average temperature above.130' degrees F-E.

Drywell presure above 1 S3'psig.

(S N (0.5 each)

M*

"T' "

/ REFERENCE Formi Question Bank (7.10)

ANSWER

.7.10 (3.0V)

LCO exceeded and:

A. Thermal power changed by more than 15% of rated thermal power in one hour, or B. The off-sas level, at the delay pipe, increased by more than 10,000 microcuries per second in one hour during steady state

~ operation at release rates less than 75,000 microcuries per

second, or C. The off sas level, at the delay pipe, increased by more than 15%

in any one hour during steady state' operation at release rates grater-than 75,000 microcuries per second.

REFERENCE Formi Question Bank (7.05)

ANSWER 7.11 (1 00)

l Ar--Verif y ' FULL-0UT*"indicatur--ts iTIUnsirruted 1snd position indicator indica te s-8-482..

M M vC.,

.(ASP]

Jr. Attempt to withdraw each control rod to the OVERTRAVEL position., c/-J2 Verify control rod is coupled by:

,

~ g\\, {[,

~1.

Observing ' stall flow'

/, 2 g' )

,

'-

.

2. Rod settling back to position 48.

(

,

-,,

"

t c

3.

'0VERTRAVEL" light extinguished.

(.3 5)

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L

_

l

~

.

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7.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

~~~~~~~~~~~~~~~~~~~~~~~~

~~~~ I656L665C5E~66ATR6L R

ANSWERS -- FERMI 2-85/02/12-LANG,T.

REFERENCE Forni Question. Bank (7.02)

ANSWER 7.12 (1 00)

,

B. Read higher than actual vesses level.

' REFERENCE-Formi Question Bank (7.50)

ANSWER 7.13 (2.00)

,A.

Thermal stratification, improper mixing, loop temperatures-appear normal, could result in exceedin3 200 degrees F and possibly pressurize the vessel.

(1.0)

p. Through.the shutdown' cooling suction and out through the suppre.tsion pool suction. Through the shutdown cooling suction and out through the

.

RHR pump minimum flow line. Through the shutdown suction and out through the full fimw test valve.

(1 0)

REFERENCE Formi Question Bank (7.12)

.

O w

L*

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-8

r

.

~

,

8.

ADMINISTRATIVE ~ PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE

__________________________________________________________

ANSWERS -- FERMI 2-85/02/12-LANGrT.

ANSWER 8.01 (2.00)

A. For failure of a single RBM channel 1. Verify not operating on a Limiting Control Rod Pattern (0.5)

AND 2. Restore the inop RBM channel to operable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(0.5)

'

OR 1.

Place the'inop RBM in a tripped condition within the next hour.

(1.0)

B. A rod pattern which exists that should a control rod be moved out of the core, the core would be placed on a thermal limit.

REFERENCE Formi Question Bank (8.03)

ANSWER 8.02 (2.00)

A. Two cognizant members' of plant staff-one with an SRO license.

(1.0)

B.

1.

Must not change the intent of the procedure.

(0.33)

2. Must be approved by OSRO within 14 days.

(0.33)

3.

<=5 changes per procedure.

(0.33)

REFERENCE Formi Question Bank (8.05)

ANSWER 8.03 (2.50)

A. Plant Superintendent.

Assistant Plant Superintendent.

Operatin3 Engineer.

Assistant Operating Engineer.

(0.5 each)

( d * C)

B. NSS may exclude personnel.

WSO -may -r equest NSS _to-exc4vde-pereennel.

(0.25 each)

DN

' ~ ' ' ' -

"'E M

REFERENCY

~

Formi 0uestion Bank (8.07)

'

,

.

6*

S

J

i

_..

_

(.'

.

a 8.

ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE

-ANSWERS -- FERMI 2-85/02/12-LANG,T.

- ANSWER 8.04 (2.00)

A. No. pressure boundary leakage.

g B. 5 spa unidentified leakaSee L. 25 Spa total leakage average over any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period.

, D.. 1 spa leakage at an RCS pressure through valves.

.

E'

2 spm increase in Unidentified Leakage.

(0.4 each)

' REFERENCE'

Formi Question Bank (8.10)

ANSWER 8.05 (2 00)

A. The addition, removal, relocction, or movement of fuel, sources, incore instruments or reactivity' controls within the reactor pressure vessel with the head Temoved and fuel in the vessel. (Suspension of-Core Alteration shall not preclude completion of the movement to a safe conservative position.

(1.0)

B. SDM shall be the amount of reactivity by which the reactor is or would be subcritical assuming all. control rods are fully inserted except for the single control rod of the highest reactivity worth which is assumed to be fully withdrawn and the reactor is in the shutdown condition,

~

cold, xenon free.

(1.0)

.

REFERENCE Formi Question Bank (8.11)

ANSWER-8.06 (1.00)

A. Brittle fracture of reactor pressure vessel could occur.

(0.5)

B. Yes. Due to neutron embrittlement.

(0.5)

REFERENCE Forni, Question Bank (0.13)

.

e a'

.

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J I

-

__

_ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _

_-

.

8.

ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE

ANSWERS -- FERMI ~2-85/02/12-LANG,T.

ANSWER 8.07 (3.00)

A. Prt: vents undue stress on the vessel no: les and bottom head region.(1.0)

B. Prevents creation of abnormal conditions in the idle loop jet pumps (vibration)'

(1.0)

C. To insure an adequate core flow coastdown from either recirculation I

pump followins a LOCA -- prevent "foolins' LPCI Loop Selection Logic.

(1.0)

l REFERENCE

~

,

FGrmi Question Bank (8.10)

ANSWER 8.08 (2.00)

HOT S/D --MSS in S/D and >200 destees F.

(1.0)

HOT S/B---MSS in S/U -, HOT S/B and any temperature.

(1.0)

REFERENCE Formi Question Bank'(8 22)

l

. ANSWER 8.09 (2 00)

1. Unusual Event.

2. Atert.

%nt

' * ' '

t*

' ' ' ' '

.

3. Site -Emergency.

F ut t 4

'

4.

General Emergency.

(0.5 each)

l l

REFERENCE Fermi Question Bank (8.27)

I ANSWER 8.10 (2.00)

.

...

95...

105...

110....

120.

r REFERENCE Formi-Question Bank (8.35)

.

1*

,

Y

.____-_

___

'

.

.____5____

_ __

$___

_

_ ___ __

I

__

_ _____ _ __ __

CNSWERS -- FERMI 2-85/02/12-LANG,T.

ANSWER 8.11 (1.50)

A.

13.4 kw/ft B. Operational Condition 1>= 25% of RTP C. Take actions within 15 min. to restore LHGR to within limits (within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or-reduce power to < 25% of RTP).

(0.5 each)

REFERENCE t

Formi Question Bank (8.40)

l ANSWER 8.12 (1.50)

A.

10'

'

l

B. 2'

,

.

following a S/U C. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after attaining 15% reactor power

'

to within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to decreasing power below 15% on a scheduled Rx. shutdown.

(0.5 each)-

'

REFERENCE Fermi Question Bank (8.36)

.

ANSWER 8.13 (1.50)

NSS (0.5)

B. OE (0.5)

(

...

(0.25)

C. C e v e s ' r.sw r,

=-NSS (.,. '

'# ^' #'

....,..

'

" ' ' " ' "

I

'Cause unknown---Plant Superintendent..

~~

(0.25)

!

I REFERENCE I

Formi. Question Bank (8.12)

L

.

e

% *

L

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

v t e Inspection Report - Fermi - 1985001
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \| 25 \|
1985 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 7(0) 11(0) 23(0) 24(0) 25(0) 26(0) 27(0) 28(0) 29(0) 30(0) 31(0) 32(0) 33(0) 34(0) 35(0) 36(0) 37(0) 38(0) 39(0) 40(0) 41(0) 42(0) 43(0) 44(0) 45(0) 46(0) 47(0) 48(0) 49(0) 50(0) 51(0) 52(0) Security Operator Exams Emergency Planning Other

IR 05000341/1985001

Text

REGION III

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