Text

Exam Rept 50-461/OL-86-01 on 860106-09.Exam results:8 Senior Operator & 3 Reactor Operator Candidates Passed Written Exam & 11 Senior Operator & All Reactor Operator Candidates Passed Oral Exam
	ML20205K300
Person / Time
Site:	Clinton
Issue date:	02/21/1986
From:	Lanksbury R, Mcmillen J, Morgan T, Plettner E, Sherman J; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:	;
Shared Package
ML20205K295	List: ML20205K291; ML20205K300;
References
	50-461-OL-86-01, 50-461-OL-86-1, NUDOCS 8602270279
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U.S. NUCLEAR REGULATORY COMISSION REGION III Report No. 50-461/0L-86-01 Docket No.'50-461 License No. CPPR-137 Licensee:

Illinois Power Company 500 South 27th Street Decatur, IL 62525 Facility Name:

Clinton Power Station Examination Administered At:

Clinton Power Station Examination Condu ted: 21 ary 6-9,1986

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q Operating ensing Section Dat6 Examination Summary Examination administered on January 6-9, 1986 (Report No. 50-461/0L-86-01)

Examinations were administered to 12 senior operator candidates and four reactor operator candidates.

Results:

Eight senior operator candidates and three reactor operator candidates passed the written examination and 11 senior operator candidates and all reactor operator candidates passed the oral / operating examination.

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REPORT DETAILS 1.

Examiners R. D. Lanksbury, Region III J. I. McMillen, Region III E. Plettner, Region III T. Morgan, EG&G J. Sherman, EG&G 2.

Examination Review Meeting Copies of the examination and answer keys were given to the facility personnel for review at the conclusion of the written examination.

Facility personnel gave their comments to the examiners on January 14, 1986.

These comments are enclosed as Attachment 1 to this report.

Resolution of these comments is Attachment 2 to this report.

3.

Exit Meeting On January 9, 1986, an exit meeting was held. The following personnel were present at this meeting.

D. Hall, Vice President, IP R. F. Scheller, Director-Training, IP J. I. McMillen, USNRC, Region III The facility was informed of those persons who had clearly passed the oral / operating examinations.

Attachments:

1.

Facility Comments Reactor Operator Examination 2.

Resolution of Facility Comments 2

ATTACHMENT 1 FACILITY COP # TENTS REACf0R OPERATOR EXAMINATION SECTION 1 COMMENTS 1.01 c.

The question could be interpreted to cover the period of short lived DNP decay.

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Reference:

Standard Nuclear Principles 1.02 a.

Suggest adding the following to the answer key:

Verify IRM operability.

1 Verify LPRM/APRM operability.

Neutron flux is an indication of power.

Neutron flux can affect reactor recirculation system response.

Neutron flux noise levels are required to be monitored by Technical Specifications in specified operating regions.

Neutron flux / power can be used to verify margin to LHGR.

Ensure signal to noise ratio is adequate.

Ensure a sufficient neutron population such that core response follows the kinetics modes (i.e. the neutron population is above the fedocial level).

Ensure a controllable period exists when power becomes visible.

Reference:

Standard Nuclear Principles b.

Suggest consideration of the following:

Cosmic neutrons are also an intrinsic source.

Spontaneous fission is not necessarily the major contributor at BOC.

Spontaneous fission may be the major contributor in a cold clean core or in a core with some highly exposed assemblies (greater than 12 to 13 GWDT).

Alpha - oxygen would be the major contributor at B0C prior to this.

T 4

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Photo-neutrons are the major contributor at E0C for short periods after shutdown. The photo-neutron reaction requires high energy gammas which decay away rapidly after shutdown, after this time alpha-oxygen (244) or spontaneous fission may become a major contributor (see above comment).

Reference:

G.E. chart of nuclides 1.03 c.

A restart cannot be SM free, because it is stable and can only be removed through burnup.

d.

Some candidates may answer this as false because of question wording:

50% Equilibrium Sm is exactly (not approximately) the same as 100% Equilibrium Sm.

Reference:

Standard Nuclear Principles 1.05 Point value of the question is 1.0, however answer shows three parts for a total of 1.5 points.

1.08 b.

Suggest also accepting response of motor damage due to overheating.

1.09 a.

Suggest also accepting choice 3 (CPR).

this question relies on semantics.

True, the GEXL " correlation" does what is stated, but CPR is the operating limit used by operators.

Not accepting CPR based solely on the technicality of the word " correlation" will confuse operators on the purpose of CPR.

CPR does indicate how close one is to the onset of transition boiling and is used by the operator, while GEXL is an intermediate step in CPR determination (determines Critical Power).

1.10 b.

Since a gap exists - however small - radiation, if listed, should also be accepted.

d.

Strictly speaking convection would be a secondary mode of heat transfer.

Reference:

Standard Nuclear Principles 1.12 Student response may differ if students calculated values using j~

standard temperature table vs. saturated pressure.

l 2

I

SECTION 2 COMMENTS 2.01 a.

The suppression pool maintains the' suction piping full and the discharge piping up to the discharge check valve.

b.

The question did not ask for setpoints and should not be required for credit.

Reference:

M05-1075 sh. 1 2.03 a.

This answer is incorrect:

The CCW system does not receive any automatic isolation signal to isolate the FC heat exchangers.

A loss of power will cause CCW to be lost and SX to automatically start but the FC heat exchangers will have to be manually lined up.

Reference:

E02-CC99, E02-FC99, E02-SX99 2.04 a.

Suggest accepting 250-260 psid as acceptable answers.

b.

This is true, except when driving rods d.

Suggest also accepting CRDH Sample Connection to primary sample station.

Reference:

M05-1078 RTS Lesson Plan RD Page RD 3-4 2.06 a.

This question asks for the effect on operation of the RCIC system.

High airborne contamination should not be required as a part of the answer, as this does not affect the operation of the RCIC system.

2.07 a.

Suggest adding to the answer key:

SRV above seat vents.

CGCS H2 mixing compressors.

b.

Candidates may address suppression pool makeup as providing the water supply for ECCS systems.

References:

M05-1063, M05-1002, sh. 6 2.09 a.

Some candidates may respond with SIA, this should be considered acceptable.

3

F 2.10 a.

Main Control Room Halon Fire suppression system is manually initiated,_not automatic.

b. -

There is presently a deluge system also in the DG fuel oil rooms.

Reference:

E02-FP99 l

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SECTION 3 COMMENTS 3.01 a.

Answer should be:

remains constant - this flow signal does not input to the feedwater level control system.

Reference:

E02-FW99 3.02 b.

With only 12 LPRM inputs to the APRM, an INOP signal is generated causing a Rod Block and one of four scram signal but no scram.

e.

Answer should be:

one of four scram signal but no scram or no protective action.

3.03 a.

There is insufficient information in this question to permit the examinee to determine a high containment pressure exists.

Recommend deletion.

b.

No. Auto start feature is bypassed when Remote Shutdown Control Transfer switch is in the Emergency position.

Reference:

E02-RS99 sh. 107 3.04 c.

The value for power may be up to 40% depending on reference used.

3.05 a.

1.

Data fault will also occur if one channel of the RACCS is INOP, not just RPIS.

2.

Scram valves will indicate any rods with one or more scram valves open.

However, 3304.02 does support the answer key.

4.

Also add. Mode Switch in REFUEL, and a Control Rod withdrawn.

Reference:

E02-RD99 3.06 At Clinton the terms activated and deactivated for source checking ARMS is not taught.

Students were instructed that the counting circuit is looking for at lease one count per 10 second period...

if this answer is unacceptable, recommend deletion of this question.

3.08 Answer should also include Division 1 Diesel Generator.

3.09 b.

Indication on Range 4 should be 13 times the square root of 10 = 41.10, therefore, no automatic actions.

Reference:

RTS lesson plan NR 5

SECTION 4 COMMENTS 4.03 Due to the wording of the question, the candidate may respond with a description of the area in the control room in which they are required to stay.

Reference:

1401.05s Figure 1 4.06 Also add Contamination or Radiation Mode should be acceptable.

Reference:

3402.01 4.08' The latest revision does not include:

"more than one control rod not fully inserted."

Reference:

4404.01 4.12 Answer should include Place Reactor Recirc loop controllers in manual.

Reference:

4008.01 4.13 Other possible answers:

To prevent the continuous-flow of a relatively large amount (200 gpm) of cold water to the vessel.

This will aid in level control and prevent thermal shock to the vessel.

To restore scram discharge flow path piping and equipment to normal (depressurized) conditions.

Reference:

Standard Operating Practices 6

FACILITY CO MENTS SENIOR OPERATOR EXAMINATION SECTION 5 COMMENTS 5.04 Suggest adding.the following to the answer key:

Reducing peaking.

Improves fuel utilization.

Reference:

Standard Nuclear Principles 5.05 Answer key should also include:

Spontaneous fission can occur in almost any transuranic element not just Uranium Cosmic neutrons are also'an intrinsic source

Reference:

Standard Nuclear Principles 5.07 c.

Suggest that a discussion of the short lived DNPs also be considered correct as below:

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Reference:

Standard Nuclear Principles 5.08 b.

Answer key should also include:

Fuel exposure or Control Rod exposure

Reference:

Standard Nuclear Principles 5.09 Recommend that the answer key also include:

Reference leg flashing may also occur during a LOCA as drywell temperature approaches RPV temperature.

Reference:

Standard Nuclear Principles 7

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SECTION 6 COMMENTS 6.01 SM001A and SM002A are in the same line and are both Division 1 (likewise SM001B and SM002B are Division 2).

Therefore, only one Division is required for a pool dump.

Reference:

E02-SM99 6.02 Answer should read:

The lowest value of either:

1 or the ratio of h T is applied if less than or equal to 1

Reference:

T.S. page 3/4 2-5; Standard Nuclear Principles 6.04 a.

These events will initiate an RPT only if > 30% power.

b.

Recommend the answer key also include:

At E0C the power increase will be much greater for the same transient (reactivity addition) due to the decrease in 6.05 Answer key should also accept:

'RWCU can also provide an additional means of heat removal from the reactor.

sProvide a means for draining the vessel.

RWCU flow provides an accurate bottom head drain temperature for heatup/cooldown rates and thermal shock interlocks.

RWCU,and CRD provide a means of level control during a startup.

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Reference:

StandardOperatingPrincihles 6.06 a.

Other acceptable answers should include:

High drive temperatures.

The drive any have to te w :lared IN0P for valve work.

This would require Mw -t : a and hydraulic / electrical disarming.

Reference:

Standard operating experience / practices, Technical Specifications 6.09 Should also include overcranking for Division 3 only.

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.w SECTION 7 COMMENTS 7.01 Setpoints were not required in the answer and should not be required for full'value.

7.07 Answer should also include Illinois Department of Nuclear Safety (IDNS).

Reference:

FE-06 7.08 a.

This answer should also be acceptable:

Reduce power to at least 20% below the thermal power prior to the reduction in feedwater heating by shifting reactor recirc to flux manual and reducing flow and by inserting rods in the proper sequence.

b.

Prevent exceeding local fuel limits and bulk power.

7.09 a.

Candidate may state positioning control switch for A solenoid at P601 and B solenoid on back panel to Open then to Off.

b.

Change answer key to include:

1.

SRV position indication on P601.

2.

SRV position indication on P642.

3.

SRV position indication on DCS.

4.

SRV flow monitors (acoustic monitors) on P866.

5.

SRV discharge line temperatures on SRV recorder on P614.

6.

MSL flow indication on DCS.

7.

Suppression pool temperature recorder on P601.

(Panel numbers should not be required)

Reference:

4009.01 7.10 Drywell temperature setpoint valve of 150 F should also be acceptable per latest revision.

Reference:

4402.01 9

4 SECTION 8 COMMENTS 8.01 a.

Answers should read:

Shift Supervisor or Assistant Shift Supervisor c.

The following special provisions for startup program may be mentioned by some candidates as per the latest revision Appendix A.

Tagouts may be required by authorized startup personnel.

Tagouts may be issued to any authorized startup personnel by name (i.e., responsible Lead Engineers and Startup Test Coordinators by name).

Tagouts may also be issued to Shift Test-Engineers by name only for those activities that the Shift Test Engineer is responsible.

Tagouts may be requested by authorized Baldwin Associates tagging authorities.

Tagouts may be issued to the authorized tagging authority by name.

8.02 Some candidates may respond with Line Assistant and Staff Assistant... both are Assistant Shift Supervisors and should also be accepted as correct answers.

Reference:

Operations Terminology 8.03 a.

Change employer to employee

Reference:

1402.04 8.06 The following is from the current revision of Technical Specifica-tions and should also be considered an acceptable answer:

Reactor coolant system leakage shall be limited to:

a.

No pressure boundary leakage.

b.

5 gpm unidentified leakage.

c.

25 gpm total leakage (averaged over 25 hour2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months period).

d.

0.5 gpm leakage per inch of valve size up to a maximum of 5 gpm from any reactor coolant system pressure isolation valve at rated temperature and pressure.

Reference:

Technical Specification 3.4.3.2 10

8.07 The latest revision of Technical Specifications has changed the tank levels to the numbers below, these should also be considered as an acceptable response.

Day Tank Division 1 385 gal Division 2 385 gal Division 3 240 gal Storage Tank Division 1 48,000 gal Division 2 41,500 gal Division 3 29,500 gal Additionally a requirement for each diesel generator to have a separate fuel transfer pump.

Reference:

Technical Specifications 3.8.1.1 and 3.9.1.2 8.09 a.

Only applicable when the I.F.T.S. blank flange is removed.

d.

Current Technical Specifications 3/4 9-21 states:

At least one IFTS carriage indicator is operable at each carriage position and at least one liquid level sensor is operable.

Reference:

Technical Specifications page 3/4 9-20 and 21 h

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ATTACHMENT 2 RESOLUTION OF FACILITY COMMENTS SECTION 1 1.01 c.

Disagree with comment.

Question asks for cause of immediate drop.

That short lived DNP is secondary to the immediate drop in prompt neutrons.

1.02 a.

Suggested answers were used in grading exams.

The question will be rephrased, if used in the future, to relate to all SRM's only as was originally intended.

b.

Disagree with comment. With regard to the statement about cosmic neutrons, this did not effect the grading as no candidate used this answer.

The answer key is in full agreement with the training materials provided and use of the answers provided in the comment would mean that just about any answer given by the candidate would be considered correct.

1.03 c.

The comment seems to substantiate the answer of " false." The comment is also not totally accurate.

Sm is only " stable" from an equilibrium standpoint.

As noted in the answer key Sm will increase following a reactor scram since burnup is no longer occurring but production continues from the decay of Nd to Pm to Sm.

d.

Question graded per answer key. Only one candidate answered false and his justification was that equilibrium Sm was proportional to reactor power which is incorrect.

1.05 Comment accepted.

The answer key was modified to show a total point value of 1.0 point.

1.08 b.

Comment accepted.

The answer key was changed to reflect the comment.

1.09 a.

Question graded per answer key.

Three of four candidates gave correct answers and accepting CPR would not change overall results of fourth candidate.

1.10 b.

Question graded per answer key.

None of the four candidates gave and d.

suggested answer.

Heat transfer by radiation or conviction is negligible and answer key is in agreement with facility training material.

1.12 Comment accepted.

SECTION 2 2.01 a.

Comment accepted.

b.

Comment accepted.

2.03 a.

Agree.

Answer key changed to reflect comment.

2.04 a.

Question graded per answer key.

Answer is in agreement with provided facility nparating procedure.

Three of four candidates received full credit.

Fourth candidate would not have received credit even with suggested answer.

b.

Comment accepted.

d.

Agree. Answer key changed to reflect comment.

2.06 a.

Agree.

Comment considered in grading exams.

Question will be modified, if used in the future, such that the potential for high airborne contamination will be an expected answer.

2.07 a.

Agree.

Answer key changed to reflect comment.

b.

Comment considered in grading but did not affect any grades since none of the candidates addressed suppression pool makeup.

2.09 a.

Disagree with comment.

Per the system description SA is the source of air for IA and IA is only for instruments; SA provides direct source of air for pneumatic type functions.

In addition, no reference was cited to support comment.

2.10 a.

Disagree with comment.

The Main Control Room Halon Fire suppression system is automatically initiated, however it does not service the main control room panels but just the underfloor area.

The answer key has been changed to reflect this fact.

b.

The EDG rooms do not have a deluge system, however, there is a temporary (for construction phase) wet pipe sprinkler system.

Since this is not a permanent system the answer key will not be changed.

This had no effect on grades as this was not mentioned by any of the candidates.

2 L

SECTION 3 3.01 a.

Agree.

Answer key has been changed to reflect this comment.

3.02 b.

Agree.

The question was modified during the course of the examination and e.

and so annotated on the licensee copy, however, the answer sheet was not modified until return to the region so that the new answer's could be verified against the training material.

3.03 a.

Disagree with coment.

Since the initialing event is a recirc loop leak, high containment pressure is a valid assumption and one which every candidate made.

3.03 b.

Agree.

The answer key has been changed to reflect this comment.

3.04 c.

Agree.

The answer key has been changed to 40%.

3.05 a.1. Agree.

The answer key has been changed to reflect this comment.

2. Agree.

The answer key has been changed to clarify the answer.

4. Agree. The answer key has been changed to agree with the comment.

3.06 No. reference was given therefore the question was graded per the answer key.

If the proposed answer was used no candidate would have received any credit.

3.08 Agree.

The answer key was changed to reflect the comment.

3.09 b.

Agree.

The intent of the question was to solve the problem assuming the IRM recorders vice DCS as the indication.

For the intended purpose the answer key is correct per the referenced facility training material.

However, since the question was not clear either answer was accepted.

3

_ _ =

SECTION 4 4

4.03 The response required by the question is considered to be clear.

This was apparently the case with the candidates as none responded per the comment and three of the four received full credit.

4.06 The contamination mode requested to be added by the comment already appears on the answer key and radiation mode was accepted as an alternate answer.

However, to clarify the answer, radiation mode was added as an alternate to contamination mode.

4.08 The item suggested to be deleted was given as an answer by all four candidates and the question was graded per the answer key.

4 4.12 Agree.

The answer key was changed to reflect the comment.

4.13 Disagree with comment.

The question requests a: wers in relation to i

the CRDM's and the proposed addition has nothing to do with the CROM's themselves.

In addition, the proposed answers were not provided by any of the candidates.

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4 l

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SECTION 5 5.04 Comment accepted.

5.05 Comment accepted regarding spontaneous fission.

However, with regard to the comment on cosmic neutrons Clinton training materials have defined intrinsic sources as those in the reactor - therefore cosmic neutrons do not fall into this category.

In addition, the suggested answers were not given by any of the candidates.

5.07 Disagree with comment.

Question asks for cause of immediate drop.

The short lived DNP is secondary to the immediate drop in prompt neutrons.

5.08 b.

Agree.

The answer key was changed to indicate exposure related to fuel or control rods.

5.09 Comment is not necessary.

This concept is already included in the answer key.

5

SECTION 6 6.01 Comment accepted.

Answer key corrected.

Typo in answer key also corrected.

6.02 Comment not needed.

A fraction would be acceptable.

6.04 a.

Comment accepted.

b.

Comment accepted.

6.05 Comment accepted.

6.06 Will accept.high drive temperature.

Do not accept the second comment since it would not be a problem directly related to seat leakage.

.6.09 Comment accepted.

Credit given if candidate specified Division 3.

6

SECTION 7 7.01 Comment not accepted.

In order to explain the "why" the candidate has to specify the setpoint.

J 7.07 Comment-accepted.

7.08 Comment is accepted since it is just a rewording of the answer key.

No change in answer key.

7.09 a.

Comment accepted since that is where the switches are located, b.

Comment not necessary since _it is just a clarification of the answer key.

7.10 Comment accepted.

Either value given credit.

Revision was made after material was received by examiner.

7

. Attachment 2 SECTION 8 8.01 a.

Comment accepted.

Typo corrected in answer key.

c.

Comment accepted but not for full credit.

Partial credit was allowed for the suggested answer because the plant is still in the construction stage.

8.02 Comment accepted.

8.03 Comment accepted.

lypo corrected.

8.06 Comment accepted.

Technical Specifications were revised after being sent to examiner.

8.07 Comment accepted.

Technical Specifications were revised after being sent to the examiner.

8.09 a.

Comment accepted.

d.

Comment accepted.

)

8

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k-h v q._.

U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY:

CLINTON 1 REACTOR TYPE:

BWR-GE6 DATE ADMINISTERED: 86/01/06 EXAMINER:

LANKSBURY, R.

APPLICANT:

INSTRUCTIONS TO APPLICANT:

Uso separate paper for the answers.

Write answers on one side only.

Stcple question sheet on top of the answer sheets.

Points for each qu stion are indicated in parentheses after the question. The passing grede requires at least 70% in each category and a final grade of at locst 80%.

Examination papers will be picked up sin (6) hours after tho examination starts.

% OF CATECORY

% OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 25.00 25.00 1.

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

________ 2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 25.00 25.00

________ 3.

INSTRUMENTS AND CONTROLS 25 25

___!_00______!_00

________ 4.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 100.00 100.00 TOTALS FINAL GRADE _________________%

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

hPPL5 Chi 4T 5~55UUhTURE~~~~~~~~~~~~~~

I

I 1.

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 2

~~~~ 5ER566YUdU C5~~UEET TRd 5FER d 6'ELU56~EL6U

~

T QUESTION 1.01 (3.00)

A.

After notching out a rod with the reactor critical, you notice a 100 second period.

HOW MUCH reactivity was.added by the rod notch? (ASSUME BOL, SHOW ALL CALCULATIONS)

(1.0)

B. After a reactor scram from power the shortest STABLE period possible is -80 seconds.

EXPLAIN this statement.

(1.0)

C. Is the INITIAL period IMMEDIATELY following the scram shorter than -80 seconds? EXPLAIN.

(1.0)

GUESTION 1.02 (2.50) c.

Give two (2) reasons for maintaining a visible neutron level at all times in the reactor?

(0.5) b.

What are three (3) methods of producing intrinsic source neutrons?

Include which is the major contributor at BOL and EOL.

Reaction equations are NOT required.

(2.0)

QUESTION 1.03 (2.50)

Indicate whether each of the following is TRUE or FALSE.

If FALSE, oxplain WHY it is FALSE.

(2 5) c.

Xenon and Samarium concentrations increase following a scram from high power operation (within the first five hours).

b.

Samarium is not as significant an operating concern as Xenon, even though Samarium has a higher nicroscopic absorption cross section than Xenon, c.

A reactor startup several days after a scram from entended high power operation is considered to be Xenon and Samarium free.

d.

The equilibrium concentration of Samarium at 50% power is approximately the same as at 100% power.

o.

The equilibrium concentration of Xenon at 50% power is appronimately one-half the equilibrium concentration at 100% power.

(*****

CATEGORY 01 CONTINUED ON NEXT PAGE

- - - - )

1.

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3

--- isEER557sERICs-REKi isissFEE As5 FE5i5 FE5s QUESTION 1.04 (3.00)

MATCH the Failure Mechanism from column (1) AND the Limiting Condition from column (2) WITH the associated Power Distribution Limits (a-c) below.

c. Linear Heat Generation Rate (LHGR)

b. Average Planar Linear Heat Generation Rate (APLHGR)

c. Minimum Criticci Pcwer Ratio (MCPR) 1 - FAILURE MECHANISM 2 - LIMITING CONDITION

1. FUEL CLAD CRACKING DUE TO LACK 1.

1% PLASTIC STRAIN OF COOLING CAUSED BY OTB

2. FUEL CLAD CRACKING DUE TO HIGH 2.

PREVENT TRANSITION STRESS FROM PELLET EXPANSION BOILING

3. GROSS CLAD FAILURE DUE TO DECAY

3. LIMIT CLAD TEMP HEAT & STORED HEAT FOLLOWING TO 2200*F A LOCA (3.00)

GUESTION 1.05 (1.00)

Using the Steam Tables provided, CALCULATE a reactor cooldown rate assuming an initial reactor pressure of 985 psis and a reactor pressure of 385 psis one hour later.

SHOW ALL WORK.

(1.0)

GUESTION 1.06 (3.00)

A coolin3 water pump is operating at 1800 rpm.

Its capacity is 400 spm at a dischar3e head of 20 psi which requires a power of 45 kw.

Usin3 simple pump laws DETERMINE the following if pump SPEED IS INCREASED to 3600 rpm:

(Show all work) a.

Pump capacity.

(1.0) b.

Dischar3e head.

(1.0) c.

Pump power requirement.

(1.0)

(**r**

CATEGORY 01 CONTINUED ON NEXT PAGE

- - - - )

1.

PRINCIFLES OF NUCLEAR POWER PLANT OPERATION, PAGE 4

THER566Y d5EC5I~EEdT TRdUEFER dU6~FL656~FL6U

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~~~~

QUESTION 1.07 (2.00) o.

What is Critical Power?

(1.0) b.

What are the numerator and denominator of the ratio called Critical Power Ratio (CPR)?

(1.0)

GUESTION 1.08 (2.00)

Give ONE undesirable result for each of the followins (Be more specific than ' pump failure'):

o. Operating a centrifugal pump for extended periods of time with the discharse valve shut.

(1.0)

b. StartinS a motor driven centrifugal pump with the discharse valve open.

(1.0)

.g 00ESTION 1.09 (2.00) i Hotch one of the following numbered items with each of the four lettered i

otatements.

A letter-number sequence is sufficient.

(2.0) 1.

MAPRAT 5.

PCIOMR 2.

APLHGR 6.

GEXL 3.

CPR 7.

TPF 4.

FLPD 8.

LHGR a.

Correlation that can be used to predict how close actual operating conditions are to OTB.

b.

Ratio of heat flux at a location in the core to the core averase heat i

flux.

c.

Ratio of MAPLHGR to the technical specification MAPLHGR limit used to ensure that peak clad temperature is maintained less than 2200* F.

d.

Contains guidelines restricting power ramp rates above the threshold power.

(***** CATEGORY 01 CONTINUED ON NEXT PAGE

- - - - )

f a.n,

,,.-~,,,.,,,,.,,,-..._.,n.

1.

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 5

--- isEEs557sisiEs-sEEi iEsssFEE Es5 FEUi5 FE6s GUESTION 1.10 (2 00)

State the mode (s) of heat transfer for the following situations

.o.

Center of fuel pellet out to the pellet edge (0.5) b.-

Across the Helium sap in the fuel rod (0.5) c.

Clad sur. face to the center of the coolant channel (0.5) d.

Clad surface to coolant under film boiling conditions (0.5)

GUESTION 1.11 (1.00)

Which of the following statements BEST describes what happens to a fluid as it passes through a venturi?

(1.0) n.

Pressure remains constant, but the velocity increases as the diameter of the venturi decreases.

b.

Pressure increases and velocity decreases as the diameter of the venturi decreases.

c.

Pressure decreases and velocity remains constant as the diameter of the venturi increases.

d.

Pressure increases, but the velocity decreases as the diameter of the venturi increases.

QUESTION 1.12 (1.00)

A temperature instrument with an out of date calibration sticker on it is reading 400 de3rees F.

A recently calibrated pressure sage sensing in the same area indicates 350 psis.

Is the temperature instrument reading accurately (within + or - 2 de3rees F)?

If not, how close is it reading to the actual temperature?

Assume the system is under saturated conditions.

SHOW ALL WORK.

(1.0)

(***** END OF CATEGORY 01

- - - - )

2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 6

QUESTION 2.01 (3.00) o.

With RHR loop

'A' in the STANDBY service condition, what maintains the SUCTION and the DISCHARGE piping full?

(1.0) b.

What three conditions are required for automatic initiation of containment spray?

(1.5) c.

. Which LPCI loops are used for containment spray?

(0.5)

GUESTION 2.02 (3.00)

The Reactor Water Cleanup (RWCU) Blowdown Flow Control valve will automat-ically close on either two blowdown line pressure signals.

For each of the two closure signals below (Parts a.

and b.)*

1.

Where is the pressure sensed in the blowdown line (UPSTREAM or DOWNSTREAM of the blowdown FCV)?

2.

Why is the FCV auto closed if that condition exists?

a.

5 psig decreasing pressure.

(1.5) b.

140 psis increasing pressure.

(1.5)

GUESTION 2.03 (1.50) a.

Following WHAT MAJOR ACCIDENT would the Shutdown Service Water (SSW)

System be used to supply the Fuel Pool Cooling and Cleanup System heat exchangers?

Why?

(1.0) b.

Which SSW loop can be operated from the Remote Shutdown Panel?

(0 5)

(*****

CATEGORY 02 CONTINUED ON NEXT PAGE *****)

4 i

2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 7

QUESTION 2.04 (3.50) i s.

What are the normal values for CRD HYDRAULIC SYSTEM FLOW and DRIVE WATER DIFF. PRESS. as indicated in the Main Control Room?

(1.0) b.

Approximately what percentage of the flow in

'a' above is supplied to the cooling water header?

'(0.5)

I c.

Explain HOW/WHY requesting single rod insertion causes cooling header flow to vary (include by how much the flow varies).

(1.0) d.

The system flow in

'a' above is less than the normal flow output of one pump.

List two (2) taps off the CRDH system upstream of the flow sensing element.

(1.0)

GUESTION 2.05 (3 00)

Answer the following with regard to the RHR system and its various codes of operation

a.

Match the following actions, evente, or interlocks in Column A with the item in Column B that initiates that item.

(1.5)

Column A Column B 1.

Shutdown cooling isolates 125 psig 2.

Allows manual operation of the 135 psis LPCI injection valve 350 psis 3.

Input to ADS 400 psis 484 psig b.

WHICH loop of RHR is the preferred loop for Shutdown Coolin3 operation and WHY?

(1.0) c.

From where does the Shutdown Cooling Mode of RHR takes its suction?

Assume normal operation.

(0.5)

(wrxxx CATEGORY 02 CONTINUED ON NEXT PAGE

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2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 8

QUESTION 2.06 (3.00) a.

How would failure of the Gland Seal Compressor affect operation of the RCIC system?

(1.0) b.

List four (4) conditions that will auto-trip the RCIC turbine.

Setpoints are not required.

INCLUDE all trips that must be reset locally.

(2.0)

QUESTION 2.07 (2.00) a.

What are two (2) heat loads that may discharge to the suppression pool (do not include LOCA)?

(1.0) b.

Describe the flow path through the containment that will provide a continuous supply of water for the ECC Systems followins a LOCA.(1.0)

QUESTION 2.08 (2.00) a.

What are three (3) signals that will cause a diesel generator to automatically Emer3ency Start (exclude manual, setpoints not required)?

(1.0) b.

Followins an Emersency Start, what are the three (3) AUTOMATIC protective trips still in effect (setpoints not required) for the Div. I & II diesel generators?

(1 0)

GUESTION 2.09 (3.00)

With regard to the Component Coolins Water (CCW) System a.

What systen. is used to pressuri=e the CCW Expansion Tank and what purpose does this serve.

(1.5) b.

What system is used as a backup for CCW and for what loads?

(1.5)

(xxxxx CATEGORY 02 CONTINUED ON NEXT PAGE ** rr)

2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 9

QUESTION 2.10 (1.00)

With regard to the Fire Protection system a.

What method is used to extinsvish fires AUTOMATICALLY in the main control room panels?

(C.5) b.

What method is used to extinsvish fires AUTOMATICALLY in the emergency diesel generator rooms?

(0.5)

(*****

END OF CATEGORY 02 *****)

3.

INSTRUMENTS AND CONTROLS PAGE 10 QUESTION 3.01 (3.00)

Assume the FEEDWATER LEVEL CONTROL SYSTEM is being operated in 3-ELEMENT control using reactor LEVEL DETECTOR CHANNEL

'A'.

Reactor power is et 85%, STEADY STATE.

For each of the instrument or control signal failures listed below, STATE HOW REACTOR LEVEL WILL INITIALLY RESPOND (increase, decrease, or remains constant) and BRIEFLY EXPLAIN WHY in terms of WHAT is hap-pening in the Feedwater Control System IMMEDIATELY AFTER THE FAILURE.

(FOR EXAMPLE, your answers should include the following detail,

'Causes reactor level to decrease due to a steam flow / feed flow error signal, steam flow < feed flow, resulting in a signal to increase the speed of the reactor feedpump(s),' IF APPLICABLE.)

a.

B FEEDWATER inlet line FLOW signal fails HIGH (1 0) b.

Channel A REACTOR LEVEL detector signal fails LOW (1.0) c.

LOSS OF CONTROL SIGNAL to B Reactor Feed Pump Speed Controller (1.0)

QUESTION 3.02.

(2.50) ggg g go gnq For each of the following, state whether a ROD DLOCK, AFULL SCRAM, or NO PROTECTIVE ACTION is directly generated for that condition.

NOTE!

IF two or more actions are generated, i.e.

rod block and a scram, state the most severe, i.e.

scram.

a.

APRM D Downscale, Mode Switch in RUN (0.5) b.

12 LPRM inputs to APRM C, Mode Switch in STARTUP (0.5) c.

Flow Units A and D Upscale (>108% flow), Mode Switch in RUN (0.5) d.

Reactor water level 58',

Reactor power 18%, Mode Switch in RUN (0 5) e.

Main Steam Lines B ISOLATED, Mode Switch in RUN (0.5)

(** max CATEGORY 03 CONTINUED ON NEXT PAGE *****)

PAGE 11 3.

INSTRUMENTS AND CONTROLS QUESTION 3.03 (2.50)

With regard to the RHR/LPCI System

a. A break occurs in a recirculation loop.

RHR initiates in the LPCI mode.

After a period of time, the LPCI injection valve (F042A) starts to go shut.

WHAT is happening?

EXPLAIN the system functioning (INCLUDING signals and setpoints) that causes this condition.

Assume no operator action and all system: cre functioning as intended.

(2.0)

b. If RHR pump control has been transferred to the Remote Shutdown Panel and a LPCI initiation signal is received, will the transferred pumps start in the LPCI mode? (answer yes or no)

(0.5)

GUESTION 3.04 (4 50)

With regards to Recireviation & Recirculation Flow Control

a. WHAT action (s) result in the recirc system if the ' System A/B Drywell High Pressure Interlock

is actuated during operation at high power in ' Flux Auto *?

(1 0)

b. WHAT are the THREE (3) thermal shock interlocks that must be satisfied in the pump starting sequence prior to starting the pump?

(Include SETPOINTS required to satisfy this interlock.)

(1 5)

c. WHAT condition (s) actuate (s) the End of Cycle RPT trip and WHY is this trip neccessary?

(2 0)

(*****

CATEGORY 03 CONTINUED ON NEXT PAGE *****)

3.

INSTRUMENTS AND CONTROLS PAGE 12 QUESTION 3.05 (3.00)

o. BRIEFLY EXPLAIN what condition will senerate EACH of the following indications on the Operator Control Module:

1. Data Fault (0 5)

2. Scram Valves (0.5)

3. Channel Disagree (0 5) j 4.

Insert Required (0.5)

I

b. Above the High Pressure Set Point (HPSP), continuous withdrawal of a (0.5) control rod is limited to ___________

j The HPSP is determined as a function of what plant parameter?

(0 5)

QUESTION 3.06 (2.00) 1 The Area Radiation Monitors have installed check sources whiche.When octivated or deactivated will provide an indication of an ARM's operability.

BRIEFLY describe HOW operability is demonstrated with l

the check source in BOTH the activated & deactivated condition.

(2.0) l QUESTION 3.07 (3.00)

How is the integrity of the ECCS pipin3 inside the reactor vessel verified during normal operation (include sensing points, specific portion of piping verified, why it needs to be verified, and response of the instrumentation to a loss of integrity in your answer - specific sotpoints are not required)?

(3.0)

DUESTION 3 08 (1 00)

List four (4) systems that have components that can be operated or controlled from the Remote Shutdown System control panels.

(1.0)

(*****

CATEGORY 03 CONTINUED ON NEXT PAGE *****)

3.

INSTRUMENTS AND CONTROLS PAGE 13 QUESTION 3.09 (2.50)

With regard to the Nuclear Instrumentation System:

a.

During a Rx startup you start to withdraw the SRM's from the j

fully inserted position.

What two (2) IMMEDIATE indications would you have if one of the SRM's was stuck?

(0.5) 1 b.

Durin3 a Rx shutdown with the mode switch in STARTUP, the IRM's 4

^

are reading 13 on ranse 5.

You downscale all IRM's to range 4.

INDICATE the expected level AND any automatic action (s) justify your answer.

(1 0) c.

While operating at 100% power you bypass APRM Channel A.

What effect, if any, does this have on the reactor recirculation system?

(1.0)

QUESTION 3 10 (1.00)

The Recire. Flow Control system is in master manual control with flow at 5S%.

You hold the master controller slide switch in ' lower' until the controller output decreases to 20%.

What is the final recirc. flow (in %)

i cnd why?

4 1

i i

(****x END OF CATEGORY 03 *xurm) 1

.I 4

i i

.-__m,,,

4.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 14 RA656E66f6dE~66NTR6E~~~~~~~~~~~~~~~~~~~~~~~~

~~~~

QUESTION 4.01 (2.50)

You suspect a breach of the reactor pressure vessel boundry inside the dry-wall has occured and an isolation is required but cannot be accomplished.

i What five (5) insiediate actions should you take?

(2.5)

GUESTION 4.02 (2.00)

CPS 1405.01, ' Performance of Operational Activities *, specifies five (5) 32neral occassions when a written procedure must be present and referred to directly.

LIST four (4) of these.

(2.0)

GUESTION 4.03 (1.00)

In CPS 1401 06, 'Authorites and Responsibilities of Reactor Operators for Sefe Operation and Shutdown *, what does the term ' controls

mean?

(1.0)

QUESTION 4.04 (1.50)

CPS 3001.01, ' Approach To Critical', has a precaution that during early rod withdrawal careful attention should be given to SRM's and IRM's for period indications even in the absence of significant flux increase.

Why is this?

(1.5)

QUESTION 4.05 (2 00)

The following precaution appears in various E0P's:

'Do not depressurize the RPV below 50 psig unless motor-driven l

pumps sufficient to maintain reactor water level are running l

and available for injection.'

l Why does this precaution exist?

(2.0) i l

I

(*****

CATEGORY 04 CONTINUED ON NEXT PAGE

- - - - )

i

)

I i

l l

l 1

i I

l i

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4.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 15

~~~~ I656LU656IL C6UTRUL

~

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

R QUESTION 4.06 (2.50)

List five (5) of the six (6) modes of operation available in the Control Room HVAC system.

(2.5)

QUESTION 4.07 (2.50)

As the individual discovering a fire, what five (5) initial items of information are you required to provide to the main control room?

(2.5)

QUESTION 4.08 (1.50)

What are the three (3) different conditions (symptoms), in addition to a

.*oactor scram, required for entry to CPS 4404.01, ' Reactivity Control -

Energency'?

(1.5)

QUESTION 4.09 (1.00)

What criteria shall be satisfied before you can stop LPCS injection flow following automatic initiation?

(1 0)

QUESTION 4.10 (1 00)

During reactoe heatup CPS 3002.01, 'Heatup and Pressurization', cautions yeu to limit the rate of blowdown via the RWCU system.

Why (NOTE - control of Rx water level is not the answer)?

(1.0)

(r**** CATEGORY 04 CONTINUED ON NEXT PAGE

- - - - )

v J

4.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 16

~~~~ d656[6656dE'C6NTREE'~~~~~~~~~~~~~~~~~~~~~~~

R QUESTION 4.11 (3.00)

Briefly explain why each of the following recirculation pump starting lioitations 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 ore within 50*F of each other.

(1.0) b.

When in one pump operation, the idle pump shall not be started unless the active loop flow is reduced to less than 50% rated flow.

(1.0) c.

Recirculation loop flow mistiatch shall be maintained within 10% of rated recirevlation flow with core flow less than 70% of rated core flow.

(1.0)

QUESTION 4.12 (1.50)

In the event of a single recirculation pump trip, what three (3) imn.ediate actions should you take?

(1.5)

QUESTION 4.13 (1.00)

With respect to the CRDM's why is it important to promptly reset the RpS logic after a trip?

(1.0)

QUESTION 4.14 (2.00)

What is the minimum DOSE RATE required for en area to be posted as follows a.

Radiation Area b.

High Radiation Area c.

Restricted High Radiation Area d.

Very High Radiation Area (2.0)

(*x*wn END OF CATEGORY 04

- - - - )

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

/

(y"$P;V

\\' U. /: e *i :;- w b ! w t

u s.),

t o

/..

1o PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 17

--- isEER55isAP.iEs-sEsi isissFEE As5 FCGi5 FC5E ANSWERS -- CLINTON 1

-86/01/06-LANKSDURY, R.

ANSWER I.01 (3.00)

Ao T = B-p /A p s o p = B /AT + 1 Assume E = 0.0072 (BOL)[0.13 6.545 X 10 E-4 DK/KEO.93 p = 0.0072/(100)(0.1)

+ 1 =

B.

= Ln 2/t

= 0.693/55.6 = 0.0125 see E-1 1/N T = 1/-

= 1/-0.0125 = -80 sec.

After the initial prompt drop, power cannot decrease faster than the longest lived delayed neutron appears.[1.03 (Calculation'not required for full eiedit.)

Co_Yes,EO.53 the initial drop in power will only be due to prompt neutrons.EO.53 (and could be caleviated by T = x1/p>

REFERENCE CPS Rx Theory, pg. 45-63.

ANSWER 1.02 (2.50) 1 so 1.

Verify SRM operability (0.25) 2.

Accurately predict approach to criticality (0.25) 6.

1.

Spontaneous fission (0.5) 2.

Gamma-Deuterium or Photo-Neutron (0.5) 3.

Alpha-Oxygen (0.5)

BOL - Spontaneous fission (0.25)

EOL - Photo-Neutron (0.25)

REFERENCE CPS Rx Theory, ps. 93-97.

1.

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 18

--- isEER557EARICs-sEsi TEsssFEE AR5 FCGi5 FE5s ANSWERS -- CLINTON 1

-86/01/06-LANKSBURY, R.

4 ANSWER 1.03 (2.50) o.

True (0.5) b.

False (0 25).

Xenon has a higher microscopic absorption cross section (0.25).

c.

False (0 25).

It may be Xenon free, but Samarium will increase following the scram (0.25).

d.

True (0.5).

e.

False (0.25).

50% equilibrium Xenon is > one-half the 100%

equilibrium value (0.25).

J REFERENCE CPS Rx Theory, pg. 83-93.

ANSWER 1 04 (3.00)

Failure Mechanism Limited Condition A.

LHGR 2

1 B.

APLHGR 3

3 C.

MCPR 1

2 (6 answers req. 0 0.5 each)

REFERENCE CPS Rx Theory, pg. 12-2.

ANSWER 1.05 (1.00) o Convert pressures in 'psis' to ' psia':

985 psis + 14.7 psi 1000 psia, and o.25

=

400 psia.

CA+93 385 psig + 14.7 psi

=

o Obtain corresponding temperatures from the Steam Tables:

-1000 psic = 544.6 F,

and 400 psia = 444.6 F.

CO.53 o

D e t e r n. i n e the temperature change for the hout:

o,2C 544.6 F - 444.6 F.= 100 F/hr.

ObdC

J 1.

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 19 TUEREUD5UdU5C5 IEEST TRh 5F5R 5U6~ FLU 56~ FLUE

~

~~~~

ANSWERS -- CLINTON 1

-86/01/06-LANKSBURY, R.

REFERENCE CPS Thermal Scierices' P3 3 3-4, and Steam Tables.

ANSWER 1.06 (3.00) a.

V2 RPM 2

,' RPM 2)

Where V2 & RPM 2 are new flow &

V2 = V1 (RPM 1/

speed and V1 & RPM 1 are flow &

V1 - RPM 1 or speed prior to speed change.

3600 rpm V2 = 400spm 1800 rpm V2 = 800spm (1.0) b.

Ji2, (RPM 2)*

[ RPM 2

Where H2 is the new discharge head & H1 is the head prior to H1 (RPM 1)*

or H2 = H1 (RPM 1 j the speed change.

f3600 rpm)2 I

H2 = 20 psi (1800rpmf H2 = 80 psi (1.0) c.

P2 (RPM 2)*

RPM 2 Where P2 is the new power re-I quirement and P1 is the power P1

( R P M 1 )'

or P2 = P1 (RPM 1 j prior to the speed change.

f3600 rpm 9 i

P2 = 45kw (1800rpmj P2 = 360kw (1.0)

REFERENCE CPS Thermal Sciences, ps. 17 17-26.

1 ANSWER 1.07 (2.00) a.

Critical power is the bundle power required to produce onset of trans-ition boiling somewhere in the bundle.

(1.0)

I b.

CPR = Critical Bundle Power / Actual E:undle Power (1.0)

REFERENCE CPS Thermal Sciences, Ps. 10 10-12.

1

1.

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 20

--- isEER55isERiEs-sEEi isEssFEs As5 FE5i5 FC5s ANSWERS -- CLINTON 1

-86/01/06-LANKSBURY, R.

ANSWER 1.08 (2.00)

o. The pump will eventually add a sufficient amount of heat to the fluid to cause cavitation. Also will accept overheating of the pump.

(1.0) b.

Could cause excessively long starting currents or water hammer if the downstream piping was not filleder motw bege. du to on,keat;.b ( 1. 0 )

(,ag ( tkus (. Cat hedit.),

REFERENCE NRC Exam Bank 8 CPS Thermal Sciences, Pg. 14-13.

ANSWER 1.09 (2.00) a.

6 b.

7 c.

1 d.

5 (0.5 each)

REFERENCE CPS Thermal Sciences, Pg. 10 10-21.

ANSWER 1.10 (2.00) a.

Conduction b.

Conduction c.

Conduction and convection d.

Radiation (4 G 0.5 each)

(2.0)

REFERENCE CPS Thermal Sciences, Pg. 5-31.

i ANSWER 1.11 (1.00) d (1.0)

REFERENCE CPS Thermal Sciences, Ps. 16 16-16.

1 1

1

1.

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 21

~~~~YEER566Y UEC5,"U55T iEdU5E5R 5U6"ELU56~ELEU

~

-ANSWERS -- CLINTON 1

-86/01/06-LANKSBURY, R.

ANSWER 1.12 (1.00) 350 psis + 14.7 psia = 364.7 psia (0.25)

Saturation temperature for 364 7 psia (444.6 des F - 431.73 des F)(14.7/50) + 431.73 des F = 435.5 des F (0.5)

Th2 temperature instrument is reading approximately 35.5 degrees low (0.25)

REFERENCE Steam Tables and CPS Thermal Sciences, Ps. 3 3-4.

2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 22 ANSWERS -- CLINTON 1

-86/01/06-LANKSBURY, R.

ANSWER 2.01 (3.00) a.

Suction - Static head from the suppression pool.

(0.5)

Discharse - LPCS Water Les Pump.

(0.5) b.

1.

The LPCI mode has been automatically or nianually initiated for 10

-minutes (Loop B signal delayed additional 90 seconds).

2.

High Drywell pressure (1.88 psis).

3.

High Containment pressure (9 psis).

(3 required E 0.5 each)

(1.5) c.

A & B.

(0.5)

REFERENCE RTS Lesson Plan RH2-1, RH3-4, and RH4-1.

ANSWER 2.02 (3.00) a.

Upstream CO.53.

Prevents draining isolated RWCU to the Main Condenser or Liquid Rad. Waste.

OR Prevents siphonin3 the reactor vessel to the Main Condenser or LRW E1.03.

(1 5) b.

Downstream E0.53.

Prevents over pressurization of the LRW system E1.03.

(1.5)

REFERENCE CPS 10P3303.01S, Ps. 10; RTS Lesson Plan RT3-4.

ANSWER 2.03 (1.50)

L3 m et CC4s A A c-pay 4-e =L ec-.-*A 4-a a vited 6. 55w weatt be sec.d.*Ide. to a. 444HS4 E0.5].

Component Coolins Water 3 1_

.;;-lated # ce the Fuel Pool Coolins and Cleanup systemvCO.5].

(1 0)

* '3*"*

(o,5) b.

Loop A.

REFERENCE RTS Lesson Plan FC4-1 and RS3.

l I

l

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e

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~

2.

PLANT DECIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 23 ANSWERS -- CLINTON 1

-86/01/06-LANKSBURY, R.

ANSWER 2.04

-(3.50) a.

39 spm, accept 29 - 49 spm (0.5) 250 pside accept 245 - 255 psid (0.5) b.

100% (0.5) c.

When a rod is inserted, one set of stabilizins valves, 2 valves, close (0.5) to direct 4 spm (0.5) to the CRD and away from the cooling water header.

d.- minimum flow line back to CST.

l. _ recirculation pump seal purse.

(0.5 each*,A.> c.Abe u M u w) dlen I. p 6-3 samle stab.

Jd REFERENCE CPS 3304.01, pg. 6 i RTS Lesson Plan RD3-13 & RD2-1.

ANSWER 2.05 (3.00) a.

1.

135 psis 2.

484 psis 3.

125 psis (0.5 each)

(1.5) b.

RHR loop B is preferred (0.5) because it allows the use of head spray to assist in reducing RPV pressure (0.5).

(1.0) c.

Recirculation loop B.

(0.5)

REFERENCE.

RTS Lesson Plan RH3-5, RH3-12, RH3-13, RH2-6; CPS 3312.01, pg. 25.

ANSWER 2.06 (3.00) a.

Steam leakage would occur from turbine resultin3 in airborne contamin-on in RCIC toom and possibly system isolation due to high area temperature.

(1.0) b.

- hish turbine e::haust backpressure

- low pump suction pressure l

- high reactor water level

- system isolation

- overspeed (4 at 0.5 each, overspeed must be one of the four)

(2.0) l i

2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 24 ANSWERS -- CLINTON 1

-86/01/06-LANKSBURY, R.

REFERENCE RTS Lesson Plan RI3-2; CPS 10P3310.01S, pg. 14.

ANSWER 2.07 (2.00)

SAV Abiee 3 cat. veal.s.

m y

s.

- RCIC turbine c4c,s g2 mix;3 upnu,on.

- SRV's

- RHR HX vent (non-condensables) durins steam condensing mode

- Initial effluent flow path for Rx water condensed in the HX durins (s_ steam condensins mode.-

(2 required at 0.5 each)

(1.0) b.

Water spilling from the break fills the drywell to the top of the weit wall (0.25), where it spills over into the weit wall-suppression pool annulus (0.25), through the pressure relief vents (0.25), back into the suppression pool where it will be pumped back into the reactor (0.25).

(1.0)

REFERENCE RTS Lesson Plan RI2-li MS3-4; RH3-14.

ANSWER 2.08 (2.00) a.

- high drywell pressure

- low reactor water level

- sustained associated bus undervoltase (0.33 each) b.

- overspeed

- senerator differential current

- overcrankins (0.33 each)

REFERENCE RTS Lesson Plan DG3-47 DG4-1.

ANSWER 2.09 (3.00) a.

Service Air CO.53.

May cause pump cavitaion due to inadequate NPSH E1.0].

(1.5) b.

Shutdown Service Water E0.5].

Fuel Pool HX E0.343, Recirculation Pump 4

seal coolers E0.33], & Recire. Pump motor bearing coolers [0.333. (1.5)

REFERENCE CPS 3203.01, Ps. 4; RTS Lesson Plan CC3-2.

~ - -,,

,y

.m.-

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

_.-...m_..,,

c-

.2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 25 i

' ANSWERS -- CLINTON 1

-86/01/06-LANKSBURY, R.

i ANSWER 2.10 (1.00)

~

o.

!:r be.

b.

CO2

{

(0.5 each)

(1.0)

REFERENCE RTS Lesson Plan FP3-1.

t i

k

?

i 1

1 4

i 4

i i

!~

i i;

I 9

--gw

+..ws-=-,e.--,,,_

.-we,.,.,

w,.,

.,wr.mm...m,.,m,,,.,..,,,

3.

INSTRUMENTS AND CONTROLS PAGE 26 ANSWERS -- CLINTON 1

-86/01/06-LANKSBURY, R.

ANSWER 3.01 (3.00) n.

Causes reactor level to DECREASE EO.253 due in

+ho

' r cci Cuni. ul

~6 y s is having a STEAM FLOW /FEFn rt 0" crGUn, STEAM FLOW < FEED FLOW [0.3753 resultino i NAL to DECREASE the SPEED OF THE REACTOR FEED s

. 3753. b.t te..t -=le=as (mt t fo.as] beam

(== 6t.- i tet 1; < (t*w d.ca d

(1,o)

P m.t i-e.t t. tAe Water tev=T c -t%t splem T.c.7s].

b.

Causes reactor level to INCREASE CO.25] due to the Level Control System having a LEVEL ERROR, with NO compensating FLOW ERROR [0.3753 resulting in a SIGNAL to INCREASE the SPEED OF THE REACTOR FEED PUMPS [0.3753.

(1.0) c.

Reactor level should REMAIN CONSTANT CO.25] because the

'B' FEED PUMP Turbine Control Unit E0.375] will lock the pumps at the speed at the time of failure E0.3753.

(1.0)

REFERENCE RTS Lesson Plan FWLC.

ANSWER 3.02 (2.50) a.

tod block b.

n: ;-rath Livc uLa t /4 sme. s;pd buL n. uram c.

rod block d.

full scram e.

ne r rtert. c

t i cn g/4 scam s;pd Lt n. u%m (0.5 each)

(2.5)

REFERENCE RTS Lesson Plan RCIS and RP.

ANSWER 3.03 (2.50) a.

Containment spray is beins initiated CO.5].

Ten minutes after a l

LPCI initiation signa yig received E0.53, with 9 psig in the containment E0.53 and.,e psis in the drywell CO.53, containment l

s ray (A loop) initiates shutting F042A.

b.

[.

[0.5].

(2.5)

REFERENCE RTS Lesson Plan RH and RS.

l l

3.

INSTRUMENTS AND CONTROLS PAGE 27 ANSWERS -- CLINTON 1

-86/01/06-LANKSBURY, R.

ANSWER 3.04

-(4.50)

a. A pressure of 1.,h8psisinthedrywellresults in transfer of both loop flow controllers to manus 1EO.53 and locks the flow control valves E0.53.

(1.0)

b. Vessel Thermal Shock Interlocks' 1.' Loop / loop suction differential of 450 deg. F E0.53 2.

Steam dome / loop suction differential of 450 deg. F E0.53 3.

Steam dome / bottom head drain differential of 4100 deg. F CO.53 (1.5) 4

c. The EOC RPT trip is actuated if >20% power E0.333 and a TCV fast closure CO.333 or TSV closure occurs E0.343.

This is to insure sufficient negative reactivity is added in conjunction with the control rods to ensure thermal hydraulic limits (MCPR) are met E1.03.

(2.0)

REFERENCE RTS Lesson Plan RR.

ANSWER 3.05 (3.00)

o. tAat one chaaael o4 A At5 a

aar a.

1.

Indicates that atleast one rod has a defective position probeAEO.53.

. one or w t 09es

2. In d i c a t e s t h a ta-e++ scram valves are net in th:

o-e y w > i i s ui.

E0.53.

n 3.

Indicates that the RGDS finds disagreement between the signals received from the RACS E0.53.

4. Indicates that the selected rod must be fully inserted before any other control rod can be movedvCO.53.

(2.0) un tk m4. witA i-Aer*EL b.

2 notches [0.53; Main turbine first stage pressure E0.53.

(1.0)

REFERENCE CPS 3304.02; RTS Lesson Plan RCIS.

e-9 3.

INSTRUMENTS AND CONTROLS PAGE 28 ANSWERS -- CLINTON 1

-86/01/06-LANKSBURY, R.

ANSWER 3.06 (2.00)

When activated

- The source is positioned to irradiate the detector causing an upscale meter deflection,cr.d a ;;I Rod 21:::.

When deactivated

- There is sufficient leakage to cause a background level readins, so that a channel failure would be indicated by a downscale alarm.

(1.0 each)

(2.0)

REFERENCE NRC Exam Bank.

ANSWER 3.07 (3.00)

LPCI injection line A and LPCS, inboard of their isolation valves, feed one dp transmitter C0.53, LPCI injection lines B and C, inboard of their isola-tion valves, feed one dp transmitter EO.5], and HPCS injection line, in-board of its isolation valves, and above core plate Pressure feed one dp transmitter [0.53.

If a break occurs between the reactor vessel wall and

.the shroud CO.253, the effectiveness of that ECC system may be lost [0.253.

Under normal conditions, little or no dp should exist across the dp trans-mitters.

When a break occurs in the area of concern, the dp transmitter will theh sense the pressure drop across the shroud and actuate an out-of-service alarm and breck status light for the applicable system [1.03. (3.0)

REFERENCE RTS Lesson Plan LD.

ANSWER 3.08 (1.00)

- SRU or Nuclear Boiler or Pressure Relief

- RCIC RHR S6

-. Di v I E D F 3

(p.25 each) qg REFERENCE RTS Lesson Plan RS.

3.

INSTRUMENTS AND CONTROLS PAGE 29 ANSWERS -- CLINTON 1

-86/01/06-LANKSBURY, R.

4 ANSWER 3.09 (2.50) a.

-count rate will not change (accept no change on period meter) [0.253.

'IN' light remains energized CO.253.

(0.5) b.

130 CO.253; Rx scram (IRM's > 120/125 of scale and mode switch not in RUN) [0.753.

(1.0) c.

Rx recire. flux controller auto switchs to APRM channel C.

(1.0)

REFERENCE RTS Lesson Plan NRi' vPS 10P3306.01S.

ANSWER 3.10 (1.00) 48% (0.35);

Due to the 48% low flow limiter when the flux controller is in auto (0.65).

(1.0) j REFERENCE CPS Exam Bank.

s

+

l l

4.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 30 R 656L665656~66UTR6L'~~~~~~~~~~~~~~~~~~~~~~~

~~~~

ANSWERS -- CLINTON 1

-86/01/06-LANKSBURY, R.

ANSWER 4.01 (2.50)

-Place the Mode Switch in SHUTDOWN

-tlarify all automatic actions occur, manually perform any that did not occur

-Sound the Containment Evacuation Alarm

-Startup SGTS and es.tablish Secondary Containment

-Enter CPS 10N4401.01S, ' Level Control - Emergency'(accept name or 4)

(0 5 each)

(2.5)

REFERENCE CPS 10N4001.02S ANSWER 4.02 (2.00)

-Evolutions that are extensive or complex for the plant conditions and/or experience of the operator.

-Integrated plant startups and shutdowns.

-Tasks which are infrequently performed.

-Off-Normal procedures after immediate actions are completed to verify proper immediate action response and to perform supplementary actions.

-Test procedures.

(0.5 each - 4 required)

(2.0)

REFERENCE CPS 1405.01 ANSWER 4.03 (1.00)

Apparatus and mechanisms, the manipulation of which directly affects the reactivity or power level of the reactor.

(1.0)

REFERENCE CPS 1401.06

~

4 w

y w.--em,+

y p.,,-

w,--,w-%r,4

--.,,r.-se-----aw,=--e

-w-=

~,

--s.--m.,.-.y

es

.--.-,-wy,,9 p=-

w----,4

-.r-

4.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 31

~~-----------

~~~~ d656L55iEAL E5NTR5t R

ANSWERS -- CLINTON 1

-86/01/06-LANKSBURY, R.

ANSWER 4.04 (1.50)

With the flux peaked high in the core during early rod withdrawal, the SRM ond IRH detectors will not be in the high flux region [0.53 and criticality any occur with little observed increase in source multiplication E1.03.

(1.5)

REFERENCE CPS 3001.01 ANSWER 4.05 (2.00)

This precaution forewarns the operator to make sure a means to control RPV level is available before depressurizing below the low pressure isolation of RCIC.

It is a compromise between depressurination and maintaining RPV water level if RCIC is the only system available to maintain RPV water level.

(2.0)

REFERENCE CPS Exam Bank ANSWER 4.06 (2.50)

-Normal

-Chlorine or Toxic gas

-Contamination <,r Rud ht;.,

-Smoke

-Purse

-Shutdown (0.5 each - 5 required)

(2.5)

REFERENCE CPS 10P3402.01S l

4.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 32

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

~~~~6d656L66fCdL C6 TR6L

~

ANSWERS -- CLINTON 1

-86/01/06-LANKSBURY, R.

ANSWER 4.07 (2.50)

-Name and department

-Location of fire / smoke

-Type and extent of fire

-What is burning

-Number of injured personnel, if any (0.5 each)

(2.5)

REFERENCE CPS 1893.07 ANSWER 4.08 (1.50)

-Reactor Power 1 3%

-Reactor Power cannot be determined i

-Hore than one control rod not f ully inser ted3 Aelet.16 weent. p uceb recisios (0.5 each)

(1.5)

REFERENCE CPS 4404.01 ANSWER 4.09 (1.00)

-Positively verify that initiation conditions are no longer present.

-LPCS flow is not required to satisfy existing plant cooling requirements.

(0.5 each)

(1.0) t REFERENCE CPS 10P3313.01S ANSWER 4.10 (1.00)

Limit the RWCU non-regenerative heat exchanger outlet to less than 130*F l

OR to protect the RWCU resin beds from over heating.

(1.0)

REFERENCE CPS 3002.01 I

l P

I l

l

4.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 33

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

RA6fdL665UAL"C6ETRUL

~~~~

ANSWERS -- CLINTON 1

-86/01/06-LANKSBURY, R.

ANSWER 4.11 (3.00) c.

Prevents undue stress on the vessel nor:les and bottom head region.

(1.0) 1 b.

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

(1.0) c.

To ensure an adequate core flow coastdown from either recirculation pump following a LOCA.

(1.0) 4 REFERENCE CPS Exam Bank 4

ANSWER 4.12 (1.50)

-Verify appropriate actions occur and manually initiate any that have not occurred.

-Reduce power by control rod insertion below the 80% tod line usins Cram Arrays.

-Reduce the flow of the operating Pump to less than half of the rated loop l

flow of 32,500 spm (numerical value not required).

(0.5 each)

(1.5) y Ptau. Ry reck, t. p cc4. n.3 io m.oca,

p REFERENCE i

CPS 4008.01 ANSWER 4.13 (1.00)

To restore the normal flow path for cooling water to the CRDM seals.

(1.0) l REFERENCE CPS 10P3305.01S l

\\

I

- ~ -,

~.. _. _.,.

4.'

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 34 RA656L66E6AL'66NTR5E------------------------

~~~~

ANSWERS -- CLINTON 1

-86/01/06-LANKSBURY, R.

ANSWER 4.14 (2.00) o.

5 mr/hr or 100 mr/five consecutive days b.

100 mr/hr c.

1000 mr/hr d.

500 mr/hr (0.5 each)

(2.0)

~ REFERENCE CPS 1905.20

sq To' f

.l'...,2 _(2 %). I [.'

K <(J*j f;

S

.. m 2.

~

.w U.S. NUCLEAR REGULATORY COMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY:

Clinton REACTOR TYPE:

BWR DATE ADMINISTERED: January 6, 1985 EXAMINER:

J. McMillen APPLICANT:

INSTRUCTIONS TO APPLICANT:

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%.

% 0f j

Reference:

T.S. definitions.

5.2 When power is changing, the relative concentration of the precursors change because short-lived large precursor concentrations change faster while approaching their final, steady-state values, than do the long-lived precursors.

(2.0)

Reference:

Reactor Theory, pages 53 and 54.

will become more negative as void fraction increases.

(.5) 5.3 an AY voids increase, the slowing down length and slowing down time are very long.

Since the neutrons spend a longer period of time in the resonance energy spectrum, more neutrons will be resonantly captured as a will be D

(1.5) more negative.

3

Reference:

Reactor Theory, page 70.

5.4 They increase neutron moderation near the center of the fuel assembly.

This tends to flatten the flux distribution across the entire assembly.fueluo. g al q (1.5)

/~~fu /M. a4.24 3A-

Reference:

Fuel lesson plan.

5.5 a.

Spontaneous fission - Uranium.

(.5) b.

Photo neutrons - T induced dissociation of deuterium (2,H).

(.5)

c. Alpha neutrons (a,n) reaction of 0-18.

(.5)

Reference:

Reactor Theory lesson plan.

5.6 a.

The central rods are more likely to exceed the 2200*F limit.

(.5)

In the event of a LOCA, the fuel would dry out rather quickly and the primary heat transfer mechanism prior to rewetting would be thermal radiation.

The edge rods can radiate heat away from the bundle while the central rods radiate much of their heat to other central rods.

(1.5)

Clinton - Answer Key b.

No.

The edge rods, and the corner rods in particular, have higher local peaking factors.

This is due to the water gaps.

(1.5)

Reference:

Standard Nuclear Theory.

T= gp so

~A.= Lambda 5.7 a.

7p p, 3 +1 Assume p =.0072 (BOL) and '7 =.1

~4 d'Ita k p =.0072/(100)(0.1) + 1 = 6.545 x 10 k

(1.0) b.

After the initial prompt drop, power cannot decrease faster than the longest lived delayed neutron appears.

(1.0) c.

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

(1.0) 5.8 a.

The worth depends on the neutron flux in its location compared to the average neutron flux in the core.

Thus, as power is increased, the flux also increases.

This will increase the worth of the rod if the local flux is increased more than the core average flux.

(1.5) b.

1.

Voids 2.

Fuel loading 3.

Control rod pattern 4.

Moderator temperature 5.

Exposure (F-url./ccMkl/

)

l 6.

Peaking factors l

Any 4 at.5 each.

Reference:

Reactor Theory lesson plan.

5.9 The level sensors are calibrated for a fixed primary

)

system pressure and fluid density. A height of water l

within the pressure vessel is compared to a height in a fixed reference leg across a differential pressure cell to give level indication. During a LOCA, as density in l

the vessel decreases, a given height of fluid would exert I

less pressure and indicated level would be below actual l

level. Variations in density during the LOCA transient would result in level fluctuation.

Indicated level will l

also vary with changes in density of the reference leg caused by temperature changes in the drywell and containment.

(2.0) 2 l

Clinton - Answer Key 5.10 a.

Critical power increases. A greater power input is needed to raise the coolant enthalpy to saturation and change water to steam.

(1.0) b.

Critical power increases. The enthalpy rise required to boil water is higher at the lower pressure.

(1.0) c.

Critical power decreases. This highest quality would be expected to occur around the hottest pin.

As local peaking factor increases it is expected that the hottest pin is closer to OTB.

(1.0) d.

Critical power decreases.

In a bundle with a bottom peak, the most rapid enthalpy rise occurs in areas of low quality, thus providing a large margin to OTB.

In a top peaked bundle, the highest enthalpy rise occurs in a region which is closer to OTB.

(1.0) 5.11 Circulating water is maintained subcooled while the steam undergoes a change in phase. The heat removal required to condense the steam (i.e., latent heat of condensation) 4 accounts for the large difference in flow rates.

(1.0)

Reference:

Standard Thermodynamics.

3

- -.. =.

Clinton - Answer Key 6.

PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 0+

6.1 1.

Division 1 and 2 keylock switches positioned to " Enable".

(1.0) 2.

Low Reactor Vessel water level (level 1) or High Drywell Pressure (2#); and (1.0) 3.

Suppression Pool Low-Low Water level (17'5") or the 30 minute,. gime Delay Clock is ty4ed out.

(1.0)

TM

Reference:

CPS 4402.01, page 3.

6.2 Lowest value of the ratio of FRACTION OF RATED THERMAL POWER divided by the MAXIMUM FRACTION OF LIMITING POWER DENSITY. T is applied only if less than or equal to 1.0.

(1.5)

Reference:

T.S. page 3/4 2-5 6.3 a.

The individual TIP probes are all scanned at the common location (Channel 10) and the computer adjusts the channel outputs so that they are all equal. Then all TIP traces can be compared to the same base.

(2.0) b.

Suppression pool or in the reactor vessel below the core.

(1.0)

Reference:

Standard system operation and 10P 3322.015

(.5) 6.4 a.

1.

Closure of the turbine stop valves.

j g go 2.

Fast closure of the turbine control valves.

p

(.5) b.

The void reactivity feedback due to a pressurization l

transient can add positive reactivity to the reactor system at a faster rate than the control rods add negative reactivity on a. scram. E t C F%ew 9mA (2.0)

Am of.Awuo.a ~ M

Reference:

T.S. B 3/4 3-4.

6.5 1.

Reduce concentrations of fission products in the reactor coolant.

2.

Reduce corrosion products and impurities in the reactor coolant to minimize fouling of heat transfer surfaces anditeduce the generation of secondary sources of beta and gamma radiation.

3.

Provide a means of water removal from the reactor.

4.

Establish and maintain reactor water clarity prior to and during refueling.

( cJd g o( oAdctt e <J c w w pt.

etig 4c~. 4

c Clinton - Answer Key 5.

Maintain circulation in the Reactor Pressure Vessel to prevent thermal stratification during low / loss of coolant flow conditions.

4 of above at (1.0) ea.

i

Reference:

RT lesson plan.

6.6 a.

A seat leak on the scram outlet valve could cause depressurization of the overpiston area. Cooling water pressure on the underpiston area can cause the drive to drift in. Mt C bydua.

(1.5) b.

With a 50V rupture, reactor water at high temperature continuously leaks past the CR0 seals to the SDV.

This will cause high CR0 temperatures.

(1.0)

Reference:

Generic system operation and RD lesson plan.

6.7 With shorting links installed:

I a.

SRM HI-HI is bypassed b.

Neutron monitoring system for RPS is "two-out-of four" l

logic.

(1.0)

With shorting links removed:

i l

Any one of the-four SRM's or Any one of the eight IRM's or Any one of the four APRM's is capable of imposing scram if their trip points are exceeded, i.e., non-coincident i

logic.

(1.0) i

Reference:

RP and NR lesson plans.

6.8 a.

Manual Pressure Relief Mode from Control Room Panels P601, P642, or Remote Shutdown Panel.

(1.5)

ADS Mode from Control Room Panel P601 only.

(1.0) b.

The area of the venturi is designed such that at 170%

flow the steam will reach sonic velocity. The turbulence at this velocity is such that this velocity cannot be exceeded, therefore restricting the flow. At sonic velocities, any downstream pressure affect cannot be sensed upstream.

Therefore, flow is independent of

+

venture dp at sonic velocities.

(1.0)

Reference:

MS lesson plan and standard fluid flow theory.

l l

2

6.9 1.

Low Lube Oil Pressure 2.

High Crankcase Pressure (Div I and II only) 3.

Generator Overcurrent 4.

Generator Reverse Power 5.

Loss of Excitation 6.

High Coolant Temperature 7.

Generator Ground Fault (Div I and II only) 9 oMW 4 ( D,4 3) any 5 at.5 each

Reference:

D.G. lesson plan.

l l

1 i

I i

3 l

l

Clinton - Answer Key 7.

PROCEDURES - NORMAL, ABNORMAL EhERGENCY, AND RADIOLOGICAL CONTROL 7.1 a.

The mode switch should be left in run to ensure the MSIV's shut at 850 psi steam line pressure.

(1.0) b.

Open the RPS breakers for the A and B Scram solenoids.

(1.0)

Reference:

CPS 4003.01, page 2 7.2 1.

HPCS water leg pump 2.

LPCS/RHR a water leg pump 3.

RHR D/C water leg pump 4.

SLC test tank 5.

FPCC 6.

RHR service water injection 7.

Fire Protection-system 8.

SLC storage tank

.25 pt each

Reference:

CPS 4401.01, page 12 7.3 1.

All control rods are inserted to position 02 or beyond.

(1.0) 2.

No more than eight control rods are not fully inserted and the rods out are at least two cells apart.

(1.5) 3.

A Qualified Nuclear Engineer has determined that the existing rod positions cannot result in reactor criticality in cold shutdown condition.

(1.0)

Reference:

CPS 4401.01, page 53 7.4 1.

Main Turbine bypass valve 2.

RCIC 3.

RRH steam condensing 4.

SRV's 5.

Main Steam line drains 6.

RWCU

.25 each

Clinton - Answer Key CPS 4403.01, page 4 7.5 a.

Place mode switch in Shutdown, Perform Reactor Scram procedure and Proceed to cold shutdown by performing C00LDOWN - EMERGENCY procedure.

(2.0) b.

Emergency depressurization is required.

(1.0)

Reference:

CPS No. 10N4406.01S page 7 and 9 7.6 a.

1.

The supervisor in charge of the fuel movement shall direct an immediate evacuation of the fuel floor.

(1.0) 2.

The MCR shall direct all personnel to evacuate the Fuel Building.

(1.0) b.

Secure Fuel Building ventilation and route the fuel Building exhaust through the SGTS.

(1.0)

Reference:

00NY96=79.09N page 1 and 2 7.7 The Illinois Emergency Services and Disaster Agency (ESDA) Headquarters in Springfield and the DeWitt County ESDA office in Clinton. (I D N 51 (1.0)

Reference:

EC-07 page 3 7.8 a.

Reduce recirculation flow to reduce power 20% and l

insert control rods to decrease power to the 100% rod pattern line.

(1.0) b.

Recirculation flow is reduced to lower bulk power to prevent exceeding any local fuel limits.

Rods are inserted to prevent scramming due to being above the 100% rod line and possible getting the APRM HI-HI's.

(1.0)

Reference:

CPS No. 4005.01 l

7.9 a.

Take the control switch for the appropriate safety relief valve to the OPEN position and (1.0) then to the 0FF position.

gj b.

1.

SRV position indication on 1H13-P642_on Main Steam DCS display.

MW 2.

SRV flow monitors on 1H13-P866 2

l

Clinton - Answer Key 3.

SRV discharge line temperatures on the ADS safety valves recorder on 1H13-P611 y.

u n yt-at' ls 9 a u t u a s~r f4 eI an 2 at.5 each.

c.

If the SRV cannot be closed within two (2) minutes, or if suppression pool temperature reaches 110'F.

(1.0)

Reference:

CPS No. 4009.01 and T.S. 3.6.3.1 7.10 1.

Suppression Pool Temperature above 95'F 2.

Drywell temperature above 135*F es /5c'#

3.

Suppression Pool level below 18'11" 4.

Suppression Pool level above 19'5" 5.

Containment temperature above 122 F 6.

Drywell pressure above 2 psig

.5 each

Reference:

CPS No. 4402.01 7.11 This level promotes natural circulation in the reactor vessel.

(1.0)

Reference:

CPS No. 3312.01 i

l l

Clinton - Answer Key 8.

ADMINISTRATIVE PROCEDURES, CONDITIONS AND LIMITATIONS 8.1 a.

Either - The person (s) responsible for administering the provisions of the safety tagging procedure - or The shift supervisor an the Assistant Shift Supervisor.

(1.0) 5 b.

The red tags are Danger Tags that prohibit operation or status change of a component.

The yellow tags are Caution Tags that provide special instructions to indicate that unusual' caution must be exercised to operate or change status of component.

(1.0) c.

Only maintenance personnel, the equivalent of a repairman second class (or above), permanent plant exempt personnel, or personnel as approved by the Supervisor Plant Operation. Gu.wdrAwcMM"") (1.0)

L ap.apJ4 W q f. s.J.w.w,-, M ~~4 b ptnu )

d.

Yes.

The Shift / assistant Shift Supervisor may release a tagout after obtaining approval of a supervisor in the same department as the person to whom the tagout was issued.

(1.0)

Reference:

CPS 1014.01 8.2 The Shift Supervisor, Assistant Shift Supervisor, and Control Room Operator. A Wa~( et C4 */ add (1.0) 7

Reference:

CPS OAP 1401.05S and 1401.0 8.3 1.

Theemployekshouldbeup-to-dateonall Qualification Requirements.

(1.0) 2.

The employee should stand 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months as Shift /

Assistant and Shift Supervision within the previous 28 days.

(1.0)

Reference:

1402.04 8.4 Remote (motor) operation can be demonstrated by cycling the valve twice.

(1.0)

Reference:

1405.01 8.5 a.

A red T should be used at the time a piece of equipment is removed from service for maintenance, which shall require retest per T.S. before being considered operable.

(.5)

Clinton - Answer Key b.

A red I should be used for equipment that becomes inoperable per T.S. for reasons other than removal from service for maintenance.

(.5) c.

A red arrow should be used for significant abnormalities that, in the opinion of the logbook writer, require special attention by those using the logbook.

(.5)

Reference:

CPS 1417.01 8.6 a.

.No pressure boundary leakage.

(.5) b.

5 gpm unidentified leakage.

(.5) sf c.

25 gpm total leakage (averaged over any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period).

(.5) d.

1 gpm leakage at a reactor coolant system pressure of 1000+ 10 psig from any reactor coolant system pressure isolation valve specified in the table.

(.5) e.

2 gpm increase in unidentified leakage within any four hour period.

(.5) J

c. Q P ~fuet ys-otve %

l'^< ~

y p 4 w'o h onix (4 A f(, 9. p t.- my g

wmJ

Reference:

T.S. page 3/4 4-8 $

8.7 a.

Two physically independent circuits between the offsite transmission network and the onsite Class 1E distribution system and Three separate Q o;7a A pr(e p r) S

1. 5 and independent diesel generators, each with:

I *,

3 3N b.

1.

A separate day fuel tank containing a minimum SM vfI I"4 on.

4' h'v o of 330 gallons of fuel for diesel generators

>-163##

1A and 18 and 194 gallons of fuel for diesel generator 1C.

(1. 0)3 - > f, f #

2.

A separate fuel storage system containing a minimum of 46,520 gallons of fuel for diesel generators 1A and 1B and 19,582 gallons of fuel for diesel generator 1C.

(1.0)

Reference:

T.S. page 3/4 8-1 8.8 Noncompliance with a Specification exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within i

the specified time intervals.

(1.5)

Reference:

T.S. page 3/4 0-1 2

Clinton - Answer Key A O '#"I 8.9 a.

The access doors of all rooms through which the t

r transfer system penetrates are closed and locked.

jd0 #

b.

All access door interlocks are OPERABLE.

c.

The blocking valve located in the fuel building IFTS hydraulic power unit is OPERABLE.

Al kN'"

A4 A Y"I& J tad d.

All IFTS p + ry r ::: r.t ry carriage position >

J w b and liquid level indicator / ere OPERABLE.

o x

g e.

Any keylock switch that provides IFTS access control-transfer system lockout is OPERABLE.

f.

All warning lights outside of access doors are OPERABLE.

Reference:

T.S. page 3/4 9-20 8.10 a failed jet pump would:

a.

Increase the cross-sectional flow area for blowdown following a Recirculation line break (factor not considered in DBA).

(1.0) b.

Reduce the capability to reflood the core to the 2/3 height level following a recirculation line break.

(1.0)

Reference:

T.S. B 3/4 4-1 8.11 1.

Thermal power changed by more than 15% of rated thermal power in one hour.

2.

The off gas level, at the SJAE, 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.

3.

The off gas level, at the SJAE, increased by more than 15% in or,e hour during steady-state operation at release rates greater than 75,000 microcuries per second.

4.

Primary coolant specific activity is greater than 0.2 microcuries/ gram dose equivalent II31.

5.

Primary coolant specific activity is greater than 100/E.

Any 3 at 1.0 ea.

Reference:

T.S. 3.4.5 B+C i

3

ML20205K300

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