Text

Exam Rept 50-413/OL-85-03 on 850909-13.Exam Results:All Nine Senior Reactor Operator Candidates Passed Oral & Written Exams.All Twelve Reactor Operator Candidates Passed Oral Exam & Eleven Passed Written Exam
	ML20133H829
Person / Time
Site:	Catawba
Issue date:	10/07/1985
From:	Rogers T, Wilson B; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML20133H816	List: ML20133H814; ML20133H829;
References
	50-413-OL-85-03, 50-413-OL-85-3, NUDOCS 8510180133
	Download: ML20133H829 (200)
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ENCLOSURE 1 EXAMINATION REPORT 413/0L-85-03 Facility Licensee: Duke Power Company 422 South Church Street Charlotte, NC 28242 Facility Name: Catawba Nuclear Station Facility Docket Nos. 50-413 and 50-414 Written and oral examinations were administered at the Catawba Nuclear Station near Clover, South Carolina.

Chief Examiner: , c, ,- /c 7 cs' Thomas Rogers ,[ Date Signed Approved by: 'f sh u,, # to /7 /s o' Brucf A. WihofSyfion Chief Date Signed Summary:

Examinations on September 9 - 13, 1985 Oral and written examinations were administered to nine SRO candidates, all of whom passed. Oral and written examinations were administered to twelve R0 candidates, all of whom passed the oral and eleven of whom passed the written.

8510180133 PDR 851010 \

G ADOCK 05000413 PDR i

4 Enclosure 1 2 i

REPORT DETAILS

1. Facility Employees Contacted:

i *H. B. Barron, Operations Superintendent j

W. H. Barron, Senior Instructor l

) *S. R. Frye, Director of Operator Training l *J. W. Hampton, Plant Manager

P. G. LeRoy, Licensing Engineer <

, *F. P. Schiffley, II, Compliance Engineer

2. Resident Inspector Contacted

P. H. Skinner, USNRC .

Attended Exit Meeting j 3. Examiners:

W. M. Dean, USNRC W. G. Douglas, USNRC P. T. Isaksen, EG&G ,

F. Jaggar, EG&G

, *T. Rogers, USNRC i

Chief Examiner

! 4. Examination Review l

i At the conclusion of the written examinationr, the exams were released to the facility staff to review the written examinations and answer keys. The following comments were made by the facility reviewers:

! a. 'R0 Exam i

, -(1) Question 1.20

! Facility Comment: Answer b or c should be considered correct because keff would be 1.0001, which is barely over the critical point.

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NRC Resolution: Barely over the critical point is supercritical.

No change to the answer key has been incorporated.

(2) Question 1.24 Facility Comment: Answ'er a or-b should be accepted because Entropy increase is so insignificant that it could be considered

{ constant. -

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Enclosure 1 3 i

NRC Resolution: This question tests the candidate's fundamental

knowledge of the second law of thermodynamics as it applies to a real pump. No change to the answer key has been incorporated, t

(3) Question 1.32 t

! Facility Comment: Candidates need access to a Mollier Diagram to answer this question.

NRC Resolution: The candidates were given a Mollier Diagram with j the copy of the steam tables distributed during the exam. No 2

change to the answer key has been incorporated.

l (4) Question 1.35(c) ,

I Facility Comment: FQZ should also be accepted as correct.

NRC Resolution: Candidates were informed prior to commencement of 2

the exam that acronyms and commonly used abbreviations are acceptable. The NRC agrees that FQZ is equivalent to the answer

' key. -

i (5) Question 2.0,1 Facility Comment: Answer b or c should be considered correct.

! Answer b would be an incorrect answer if the word " loaded" were changed to " unloaded." Refer to Attachment 2.01.

NRC Resolution: The NRC agrees with facility comment in accordance with the reference provided. The answer key has been changed to accept b or c as correct responses.

l (6) Question 2.02

, Facility Comment: Answer a or b should be considered correct.

i Secondary System pressure limit of 110*4, is protected by Steam Safeties.

NRC Resolution: Answer d is the only correct answer given by the RPS lesson text OP-CN-SPS-IC-IPE, p.5. No change has been incorporated to the answer key.

, (7) Question 2.08- ,

Facility Comment: Answer b or c should be considered correct.

For choice b the MS pumps will shutdown. Refer to 015-01.08, 4

electrical drawing, Attachment 2.08.

! NRC Resolution: The NRC agrees with facility comment in l accordance with the reference provided. The answer key has been

changed to accept b or c as a correct response.

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(8) Question 2.21 i Facility Comment: Answers a, b, or c are correct. No reference needed.

l NRC Resolution: The NRC agrees that a or c may result because of b, but b is still the only correct answer in accordance with

!' the stated reference on the answer key. No change has been j incorporated to the answer key.

(9) Qu'estion 2.25(b)

Facility Comment: This statement is not addressed in the ND I Lesson Plan as referenced on the answer key. This is not addressed in the ND System Description.

l NRC Resolution: The stated reference refers to "the Core and ;

i Reactor Coolant System." No change has been incorporated to the !

answer key.

l (10) Question 2.29(h)

] Facility Comment: The answer should be open. Refer to flow

{ diagram 1573-1.1 (D-1). Attachment 2.2.9.

NRC Resolution: The NRC agrees with the facility comment in accordance with the reference provided. The answer key has been changed to "0 PEN" as the correct response.

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(11) Question 2.32 Facility Comment:

}

These other suction sources are correct also:

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! a. Boric Acid Tank

b. Recycle Holdup Tank

c. Seal return 1

d. RMUST i e. Chemical Addition Tank NRC Resolution: The NRC agrees. The answer key has been changed j to accept the stated alternate responses. ,

(12) Question 2.33 i

Facili+y Comment: Refer to attachment 2.3.2. pH control should I be accepted for Iodine retention. Refer to Attachment 2.33. >

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Enclosure 1 5 NRC Resolution: The NRC agrees with facility comment in accordance with the reference provided. The answer key has been changed to accept "pH control" or " Iodine retention" as one of the three correct responses.

t j '(13) Question 3.09 s Facility Comment: Answer a or b should be considered correct.

i Refer to Attachment 3.09.

l NRC Resolution: Though the high steam pressure rate signal is rate sensitive, oply the low steamline pressure signal will cause both an SI and a MSLI. sNo change has beet incorporated to the j answer key. ,

(14) Question 3.10 Facility Comment: 'This question should be deleted. The l mathematical deviations indicated in the question are not a true indication of the Rod Control circuitry because the derivative function is not accounted for. Also, an2wers a or c could have been correct depending en which direction the rods were moving.

This questio,n creates a. situation of confusing (+) pluses and (-)

i minuses and does not test practical operator knowledge.

-f l NRC Resolution: The NRC agrees in part. Deperd{ng_.upor/whether a

"+" or " " signal is assumed for a rod insertion signal, answer a or c is correct. The answer key has been changed-to accept

, a or c as a correct response. ,

(15) Question 3.14 ,

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i Facility-Comment: Answer a or b is correct. No reference needed.

i NRC Resolution: The NRC-agrees. The answer key has been changad' i to accept a or b as correct answers.

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! (16) Question 3.19 k Facility Comment: Answer c is True. The RTD w'lli still function. ;

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i NRC Resolution: At low f, low rates the response time of an RTD is j . significantly increased, thereby decreasing its ability to accurately monitor temperature. No change has been incorporated to the answer key. ' '

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Enclosure,1 6

? (17) Question 3.23

, Facility Comment: Answer c should be deleted because this is not required operator knowledge. This logic test was not in the Lesson Presentation which is used to teach students. The Lesson Test is a document used by an instructor to prepare for lectures.

NRC Resolution: The " Fast Pulse" testing capability is an important feature in assuring the reliability of the RPS. The operator should be aware of this feature. No change has been incorporated to the answer key.

(18) Question 3.25(b)

Facility Comment:

"Both" or " Unit 1 only" should also be accepted as correct because the Unit 2 setpoint is 78%. No reference needed.

NRC Resolution: The question did not ask for the setpoints at which a trip would occur. It provided a condition by which both units should have tripped. No change has been incorporated to the answer key,

, (19) Question 3.26(b)

Facility Comment: Dumps will arm only because C7-B will pick up.

Refer to Attachment 3.2.6.

NRC Resolution: The NRC agrees with the comment in accordance with the reference provided. The answer key has been changed to accept " Arm Only" as the correct response.

(20) Question 3.30 Facility Comment: The additional alarms as on Attachment 3.30 should also be accepted.

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NRC Resolution: The NRC agrees with the comment. The answer key has been changed to accept the additional answers.

(21) Question 4.05 i

Facility Comment: Answer b or d is correct. Refer to Attachment 4.05.

1 NRC Resolution: The question is asking for the major evolutions during a power increase, not the conditions necessary to begin a power increase. No change has been incorporated to the answer key.

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Enclosure 1 7 (22) Question 4.13 Facility Comment: Answer a or d is correct. Choice a would be identical to what is in EP/1/A/5000/01 if the sign > was put in front of 5% pressurizer level. Refer to Attachment 4.13.

NRC Resolution: At 5% pressurizer level, termination criteria is not met. No change has been incorporated to the answer key.

(23) Question 4.34 Facility Coument:

If the SI actuated light is not lit, the operator would look at his SI parameters and determine from the conditions of the parameters what is needed. > or < should not make any difference for an answer.

NRC Resolution: The NRC agrees that there is no significance to whether a candidate answers that he would verify that an SI is required or verify that an SI is not required. No change to the answer key is necessary.

(24) Question 4.3'5 Facility Comment: The answer key does not completely agree with the procedure AP/1/A/5500/02. Refer to Attachment 4.35.

NRC Resolution: The question only asks for the operator's actions upon a failure of the turbine to trip from the control room; therefore, the ca.ididate will not be required to provide actions for a failure of the turbine to trip locally. No change has been incorporated to the answer key.

(25) Question 4.36 Facility Comment: The answer key does not address CNS but addresses only MNS. Refer to the SRO Exam Question #7.22 for the correct answer.

NRC Resolution: The NRC agrees with the comment. The answer key has been updated to the CNS requirements.

b. SRO Exam (1) Question 5.04 i

Facility Comment: Limit and reactivity on a rod ejection accident should also be an acceptable answer since this is one of the Rod

misalignment effects for the analyzed rod ejection accident.

Refer to Attachment 5.04.

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I Enclosure 1 8 NRC Resolution: If the candidate elects to itemize the analyzed accidents, partial credit will be given for each. The answer key has been changed for the alternate answer.

(2) Question 5.12 i Facility Comment: Answer c should also be considered correct because the question asked for "approximately".

NRC Resolution: The information given and the equations provided to the candidates was sufficient to determine the reactivity change to within three significant digits as indicated by the answer key. No change has been incorporated to the answer key.

(3) Question 5.36 Facility Comment: Same comment as Question 1.24 on R0 Exam.

NRC Resolution: See resolution to question 1.24.

(4) Question 6.05

Facility Comment: Answer a or b should be correct because of the confusing term " excess letdown normal seal return" would cause an operator to think that the seal return relief would divert water to the Pressurizer Relief Tank.

NRC Resolution: Even if the candidate did not understand the flowpath described, answer b would never be true since a means j to manually operate the relief valve is not available as stated by No change has been incorporated to the answer key.

the question.

1 (5) Question 6.09

! Facility Comment: The RC System should also be considered a i

correct answer. Refer to answer for R0 Exam question 2.10.

NRC Resolution: The NRC agrees. The answer key has been changed to accept the RC System.

(6) Question 6.18 Facility Comment: Same comment as question 2.01 on R0 exam.

NRC Resolution: See resolution to comment on question 2.01.

(7) Question 6.23 Facility Comment: Same comment as question 3.10 on R0 exam.

NRC Resolution: See resolution to comment on question 3.10.

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Enclosure 1 9 (8) Question 6.26 Facility Comment: Same comment as question 3.09 on RO exam.

NRC Resolution: See resolution to comment on question 3.09 (9) Question 6.39 Facility Comment: Answer c should also be considered correct.

Refer to Attachment 6.39 Resolution: The NRC agrees in accordance with the reference provided. The answer key has been changed to accept answer a or c as correct.

i (10) Question 7.09 Facility Comment: Same comment as question 4.05 on R0 Exam.

j NRC Resolution: See resolution to comment on question 4.05.

l (11) Question 7.16 Facility CoEment: During the written exam, one candidate was given a clarification of BOP operator ( ANCO) and another candidate

} was given a clarification of 0ATC (NCO), which would make answer b or c correct. Refer to attachment 7.16.

NRC Resolution: The candidate receiving the clarification of the BOP to be the NCO will be graded accordingly. All others will be graded by the answer key. The answer key has been modified to

, document this deviation.

(12) Question 7.21 Facility comment: Same comment as questions 4.35 on R0 Exam.

NRC Resolution: See resolution to comment on question 4.35.

l (13) Question 7.22 I

Facility Comment: Same comment as question 4.36 on R0 Exam.

NRC Resolution: See resolution to comment on question 4.36.

(14) Question 7.30 i Facility Comment: Same comment as question 4.34 on R0 Exam.

j NRC Resolution: See resolution to comment on question 4.34, i

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t Enclosure 1 10 (15) Question 8.11 4

i Facility Comment: This question should be deleted since there is ,

no correct answer. The answer key indicates a as the correct answer. However, venting is possible with EMF-39 declared inoperable. Refer to T.S. Table 3.3-13 action statement 45.

Refer to Attachment 8.11.

NRC Resolution: The NRC agrees. With the additional action The question statement, the venting clause makes a false too.

has been deleted from the exam.

(16) Question 8.16 Facility Comment: Answers b, c or d are correct. Refer to Attachment 8.16.

NRC Resolution: Though cold shutdown conditions allow more liberal temperature and keff conditions, changing them does not constitute a mode change from cold shutdown to refueling.

However, answer d is considered a mode change and may require some action to reduce temperature or increase the shutdown margin to be withi,n the requirements of Technical Specification. No change has been incorporated to the answer key.

(17) Question 8.24 i

Factitty Comment: Answer a or b is correct. Tech Specs also i

address these limits. Refer to Attachment 8.24.

j NRC Resolution: Tech Specs refers the operator to 10 CFR 20 for these limits which is consistent with the answer key. No change has been incorporated to the answer key.

(18) Question 8.26 Facility Comment: Answer b is the only correct answer. Refer to Attachment 8.26.

NRC Resolution: The NRC agrees. The answer key has been changed to credit b as the correct answer.

(19) Question 8.28 j Facility Comment: This question has no correct answer. Only one i person has to have a license. Refer to Attachment 8.28.

1 l NRC Resolution: The NRC agrees since the requirement does not j necessarily mean a licensed operator. The question has been deleted from the exam, i

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Enclosure 1 11 (20) Question 8.29 Facility Comment: Answer a or b is correct. Answer b is correct per 0MP-1-4 page 15 and OMP Attachment 4 page 1. Refer to Attachment 8.29.

NRC Resolution: The NRC agrees in accordance with the additional reference. The answer key has been changed to credit a or b as correct.

(21) Question 8.38 Facility Comment: " Newport" should also be considered correct for the " Maintenance Warehouse." Refer to Attachment 8.38.

NRC Resolution: The NRC agrees that " Newport" is equivalent to the answer key. No change is necessary,

c. General Comments (1) Candidates did not receive a formula sheet until after the exam started.

(2) Entropy should not be a testable item. Operator has no control over Entropy.

(3) Reactor Building integrity should not be a testable item because the only place it is used is in the definitions section of Tech Specs. It is not used in any Tech Spec LCO.

NRC Resolution: Since the preceding general comments do not request any changes to the exam answer keys and are the expressed opinion of the Catawba staff, no itemized resolution has been formulated.

5. Exit Meeting At the conclusion of the site visit the examiners met with representatives of the plant staff to discuss the results of the examination. Those individuals who clearly passed the oral examination were identified.

There were no generic weaknesses (greater than 75 percent of candidates giving incorrect answers to one examination topic) noted during the oral examination.

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

The licensee did not identify as proprietary any of the material provided to or reviewed by the examiners.

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U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: CATAWBA 1 REACTOR TYPE PWR-WEC4 DATE ADMINISTERED: 85/09/09 EXAMINER: JERRY DOUGLAS APPLICANT: ____ h ,$T_h ___________

INSTRUCTIONS TO APPLICANT:

U2o separate paper for the answers. Write answers on one side only.

Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.

% OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 40.00 25.00

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 40.00 PLANT DESIGN INCLUDING SAFETY

________ _['____5.00 ___________ ________ 2.

AND EMERGENCY SYSTEMS 40.00 25.00

________ ______ ___________ ________ 3. INSTRUMENTS AND CONTROLS 40.00 PROCEDURES - NORMAL, ADNORMAL,

________ _I'5.00

____ ___________ ________ 4.

EMERGENCY AND FADI0 LOGICAL CONTROL 160.00 100.00 TOTALS FINAL GRADE _________________%

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

PPL5C i~5~55GEhiUR5~~~~~~~~~~~~~~

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f THER 66E YC5I~UEdT TRd 5E5R d56~ELUf6~ELUU !

! QUESTION 1.01 (1.00)

) Which of the following conditions could NOT be caused by pump cavitation? ;

, s. Pump runout l b. Excess bearing wear :

c. Gas bindins ;

j d. Excess noise i

l QUESTION 1 02 (1.00) j Which of the following is NOT a reason for pressurizing the fuel rods 1 with helium?

a. Minimize clad creeping inwards toward fuel pellets. f

, b. Increase gap (pellet to clad) thermal conductivity. l 1

i c. Allow detection of clad failure by helium analysis of the ,

coolant. !

d. Maintain lower fuel centerline temperature. !

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) OUESTION 1.03 (1.00) t l An ECP is calculated for a reactor startup 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after a reactor trip 1 from 100% equilibrium conditions. Which of the following conditions i would cause the actual critical position to be lower than the ECP? :

i a. The startup is delayed until 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after the trip.

b. Actual boron concentration is 10 ppm more than the

, predicted baron concentration. !

, c. A rod fin 3er is separated from its spider assembly.

i j d. The steam dump pressure setpoint is lowered by 100 t 1

psi prior to reactor startup.

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3

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~~~~iSER5667 IMfC5~~5EdT TEd EfER d 6"EL6f6' FLUE QUESTION 1.04 (1.00)

Which of the following sets of components of the six-factor formula are most affected by a moderator temperature change?

a. Thermal Utili ation and Fast Fission

b. Resonance Escape and Reproduction

c. Reproduction and Fast Fission

d. Resonance Escape and Thermal Utilization QUESTION 1 05 (1.00)

If reactor power increases from 1000 cps to 5000 cps in 30 seconds, what is the SUR?

a. 1.0 DPH

b. 1.2 DPM

c. 1.4 DPM

d. 1.6 DPM OUESTION 1 06 (1.00)

At BOL which of the following coefficients is the major contributor to the Power Defect?

a. Fuel temperature

b. Moderator temperature

c. Void

d. Pressure

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~~~~iU5R5565 EE5C5~~55Ii~ TREE 5i E~EA6~5LUI6"iL65 GUEGTION 1.07 (1 00)

Both Pu-239 and Pu-240 concentrations increase over core life. ~

Which of the following statements concerning the effects of these increases is correct?

a. The buildup of Pu-240 increases the average delayed neutron fraction.

b. The buildup of Pu-239 decreases the core Reproduction factor.

c. The buildup of Pu-239 causes the NTC to become more negative.

d. The buildup of Pu-240 causes the FTC to become more negative. ( j 00ESTION 1 08 (1.00)

Which of the fo11owin3 parameter changes will increase the Departure from Nocleste Boiling Ratio (DNBR)? (Consider each separately.)

a. Reactor power increases.

b. RCS pressure increases.

c. RCS temperature increases.

d. RCS flow decreases.

QUESTION 1.09 (1.00)

Tho -1/3 DPN SUR following a reactor trip is caused by which of the following?

a. The decay constant of the longest-lived group of delayed neutrons.

b. The ability of U-235 to fission with source neutrons.

c. The amount of negative reactivity added on a trip being 3reater than the Shutdown Margin.

d. The doppler effect adding positive reactivity due to the temperature decrease following a trip.

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I to PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 5

~~~~iEERbE6f d 565I~U5df~IRds5EER~d56" FLU 56~EL6U OUESTION 1.10 (1.00)

Which of the following will cause plant efficiency to increase?

a. Total S/G blowdown is changed from 30 spm to 40 spm.

b. Steam quality changes from 99.7% to 99.9%.

c. Level increase to higher than normal in a feedwater heater.

d. Absolute condenser pressure changes from 1.0 psi to 1.5 psi.

GUESTION 1.11 (1.00)

The reactor is producing 100% rated thermal power at a core delta T of 60 degrees and a mass flow rate of 100% when a blackout occurs. Natu-ral circulation is established and core delta T goes to 40 degrees.

If decay heat is 2%, what is the core mass flow rate (in %)?

a. 1.3 l b. 2.0 l c. 3.0

d. 4.0 I

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 6

--- inEEs55isisiCi- siai is3E3Eis Es5 FE0i5 FE50 QUESTION 1 12 (1.00)

During a nenon-free *eactor startup, critical data was inadvertently taken two decades below the required Intermediate Range (IR) level (1 E-10 amps). The critical data was then taken at the proper IR level (1 E-08 amps). Assuming RCS temperature and baron concentra-tion did not change, which of the following statements is correct?

a. The critical rod position taken at the proper IR level is LESS THAN the critical rod position taken two decades below the proper IR level.

b. The critical rod position taken at the proper IR level is THE SAME AS the critical rod position taken two decades below the proper IR level.

c. The critical rod position taken at the proper IR level is GREATER THAN the critical rod position taken two decades below the proper IR level.

d. There is not enough information given to determine the relationship between the critical rod position taken at the proper IR level and the critical rod position taken two decades below the proper IR level.

QUESTION 1.13 (1.00)

Which of the following statements is correct if the discharge valve from a centrifugal pump is being partially closed from the full open position?

1. Pump head decreases as head loss decreases.

b. Pump head increases as head loss increases.

c. Volume flow rate increases as head loss decreases.

d. Volume flow rate decreases as head loss decreases.

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--- iniEA557sEsiCi! AEAi iiissFEE Es5 FEUi5 FE50 GUESTION 1 14 (1.00)

Which of the following statements concerning Xenon-135 production and romoval is correct?

a. At full power, equilibrium conditions, about half of the xenon is produced by iodine decay and the other half ts produced as a direct fission product.

b. Following a reactor trip from equilibrium conditions, xenon peaks because delayed neutron precursors continue to decay to xenon while neutron absorption (burnout) has ceased.

c. Xenon production and removal increases linearly as power level increasest i.e., the value of 100% equilibrium nenon is twice that of 50% equilibrium nonon.

d. At low power levelse nonon decay is the major removal method. At high power levels, burnout is the major removal method.

QUESTION 1.15 (1.00)

Which of the following is the units of heat flux?

a. Watts / cubic centimeter

b. DTU / (hr square ft)

c. Calories / gram

d. >W/ ft

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 8

--~~iAEER557s3RICi- AEAi iEEssFEE Es5 FE015 FE50 GUESTION 1 16 (1.00)

The reactor is critical at 10,000 eps when a S/G PORV falls open.

Assuming BOL conditions, no rod motion, and no reactor tripe choose ,

the answer below that best describes the values of Tave and nuclear power for the resviting new steady state. (POAH = point of adding hoot).

a. Final Tave greater than initial Tavg, Final power above POAH.

b. Final Tave greater than initial Tav3, Final power at POAH.

c. Final Tavs less than initial Tava, Final power at POAH.

d. Final Tava less than initial Tava, Final power above POAH.

QUESTION 1.17 (1.00)

During fuel loading, which of the following will have NO effect on the shape of a 1/M plot?

a. Location of the neutron source in the core.

b. Strength of the neutron source in the core.

c. Location of the neutron detectors around the core.

d. Order of placement of fuel assemblies in the core.

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____________________________________________ 1 OUESTION 1.18 (1.00)

Which of the following statements concerning the use of water as the moderator is correct?

a. Water has a HIGH scattering cross-section, a LOW absorption cross-section, and a LARGE energy decrement per collision.

b. Water has a LOJ scattering cross-section, a HIGH absorption cross-section, and a LARGE energy decrement per collision.

c. Water has a HIGH scattering cross-section, a LOW absorption cross-section, and a SMALL energy decrement per collision.

d. Water has a LOW scattering cross-section, a HIGH absorption cross section, and a SMALL energy decrement per collision.

QUESTION 1.19 (1.00)

Which of the following statements concerning Shutdown Margin (SDM) is correct?

a. The maximum SDM requirement occurs at EOL and is based on a rod ejection accident.

b. The manimum SDM requirement occurs at EOL and is based on a steam line break accident.

c. The manimum SOM requirement occurs at DOL and is based on having a positive moderator temperature coefficient.

d. The maximum SDM requirement occurs at DOL and is based on a rod withdrawal accident while in the source range.

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

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 10

--- inEEs557s3 sics- RE3i iE3ssFEE ID5 FE0i5 FE60 OUESTION 1.20 (1.00)

With the reactor initially at a koff of 0 99, a certain reactivity change causes the count rate to double. If this same amount of reactivity is again added to the reactor, which of the following will be the status of the reactor 7

a. Subcritical

b. Critical

c. Supercritical

d. Prompt Critical GUESTION 1.21 (1 00)

During a reactor startup, the first reactivity addition caused count rate to ircrease from *0 cps to 16 cps. The second reactivity addi-tion caused count rate to increase from 16 cps to 32 eps. Which of the following statements describing the relationship between the l first and second reactivity additions is correct?

a. The first reactivity addition was larger.

b. The second reactivity addition was larger.

c. The first and second reactivity additions were equal.

d. There is not enoujh data given to determine relationship between reactivity values.

QUESTION 1.02 (1.00)

Which of the following nupress the relationship between differential rod worth (ORW) and integral rod worth (IRW)?

a. DRW is the slope of the IRW curve at that location.

b. DRW is the area under the IRW curve at that location.

c. DRW is the square root of the TRW at that location.

d. There is no relationship between ORW and IRW.

(***** CATECORY 01 CONTINUED ON NEXT PAGE *****)

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 11

~~~~iUEEE665EE55Ci! Ei37 TE3DiFis As5 Ft015 FtBE l QUESTION 1.23 (1.00) ,

Which of the following conditions would result in the highest available net positive suction head?

l s. Pump with heat source downstream and heat sink upstream with surge tank on its discharge line.

b. Pump with heat source downstream and heat sink upstream with surge tank on its svetion line.

c. Pump with heat sink downstream and heat source upstream with surge tank on its dischargo line.

d. Pump with heat sink downstream and heat source upstream with surge tank on its suction line.

QUESTION 1. 4 (1.00)

Which of the following describes the parameter chan3es that occur across a centrifujil pump?

1. Temperatura INCREASES, Enthalpy INCREASES, Entropy INCREASES

b. Temperature INCREASES, Enthalpy INCREASES, Entropy CONSTANT

c. Temperature INCREASES, Enthalpy CONSTANT, Entropy INCREASES

d. Temporatore CONSTANT, Enthalpy CONSTANT, Entropy INCREASES l, e. Temperature CONSTANT, Enthalpy INCREASES, Entropy CONSTANT 1

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

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 12

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~~~~iU5E5ddiUd55657~sEdi~IEdU5EER dU6~EE656"ELUE QUESTION 1 25 (1.00)

Which of the following statements concerning samarium reactivity effects is correct?

a. The equilibrium (at power) value of samartum is dependent upon power level. The peak value of samarium following a shutdown is dependent upon power level prior to shutdown.

b. The equilibrium (at power) valve of samarium is dependent upon power level. The peak value of samarium following a shutdown is independent of power level prior to shutdown.

c. The equilibrium (at power) value of samarium is independer t of power level. The peak value of samarium following a shutdown is dependent upon power level prior to shutdown.

d. The equilibrium (at power) valve of samarium is independent of power level. The peak value of samarium following a shutdown is independent of power level prior to shutdown.

QUESTION 1.26 (1.00)

Which of the followinj is the purpose of using soluble boron to control the excess reactivity of the reactor?

4. It does not 1ffect the flon shape.

L. It does not affect rod worth,

c. It is cheaper than adding more rods.

d. It increases reactor loading rates. e

(***** CATECORY 01 CONTIMUED ON NEXT PAGE *****)

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 13

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~~~~IU5EE655 355C5I~E55i'YEhE5EER h 6'EL0iB EtBE l

i OUESTION 1.27 (1.00) 1

! Reactivity is defined as which of the following?

a. The ratio of the number of neutrons at some point in this l generation to the number of neutrons at the same point in the previous generation.

b. The fractional change in neutron population per Generation.

c. The factor by which neutron population changes per Genera-tion.

d. The rate of change of reactor power in neutrons per second.

QUESTION 1.28 (1.00)

In which of the following conditions is the Moderator Temperature Coefficient LEAST negative (most positive)5

a. COL, low temperature .

b. 00L, high temperature

c. COL, low temperature

d. EOL, high temperature OUESTION 1.29 (2.00)

Fcr t h ., following, state whether the tensile stresses are maximum on the OUTER or INNER wall of the reactor pressure vessel.

I

i. Pressure Stress (0.5)

b. Heatup Stress (due to delta T only) (0.5)

c. Cooldown Stress (due to delta T only) (0.5)

d. Composite (Total) Stress during Cooldown (0 5) l

(***** CATECORY 01 CONTINUED ON NEXT PAGE *****)

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 14

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~~~~iE5E5665U EEC5,~Ushi~iR U5E5R h56"ELU56~iLUU GUESTION 1.30 (1.50)

Indicate whether the following statements concerning delayed neutrons ero TRUE or FALSE.

c. If the reactor is supercritical, then the fraction of delayed neutrons shifts to the shorter lived precursors and the value of the effective decay constant (lambda) decreases. (0.5)

b. Due to the significant decrease in the percentage of fast fission occuring over core life, the value of the effective delayed neutron fraction decreases over core life. (0.5)

c. Delayed neutrons are produced at some time after fission as a result of the radioactive decay of fission products. (0.5)

GUESTION 1.31 (1.50)

During a reactor startup, equal increments of reactivity are added and the count rate is allowed to reach equilibrium each time.

Choose the bracketed ( C3 ) word (s) that describe what is observed on the Source Range recorder and/or SUR meter.

a. The change in equilibrium count rate is [ larger] Cthe same]

Csmaller] each time. (0.5)

b. The time required to reach equilibrium is tionger] [the same] Cshorter] each time. (0.5)

c. The point of supercriticality can be identified by a(n)

Cincreasing] Cconstant] Cdecreasing] positive SUR several seconds after the reactivity addition is terminated. (0.5)

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

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 15

--- isEEs557sAsiEs- REAi isissFEE As5 ECUi5 FE5R QUESTION 1.32 (2.00)

If steam goes throu3h a throttling process, indicate whether the following parameters will INCREASE, DECREASE, or REMAIN THE SAME.

a. Enthalpy (0.5)

b. Pressure (0.5)

c. Entropy (0.5)

d. Temperature (0.5)

GUESTION 1.33 (2.00)

The reactor is operating at 30% power when one RCP trips. Assuming no reactor trip or turbine load change occur, indicate whether the following parameters will INCREASE, DECREASE, or REMAIN THE SAME.

a. Flow in operatin3 reactor coolant loops (0.5)

b. Core delta T (0.5)

c. Reactor vessel delta P (0.5)

d. Operatin3 100P steam generator pressure (0.5)

GUESTION 1.34 (1.50)

Hetch the heat transfer process in Column A to the equation that applies to that process in Column B.

COLUMN A COLUMN B

a. Between cold les and hot les 1. 0=mc4T of reactor (normal FC flow)

2. 0 = A m AT

b. Across S/G tubes (primary to secondary) 3. 0=UAAT

c. Across S/G (feedwater to steam) 4. 0 =mcah

5. 0=mah

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

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 16

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~~~~5H5R5U65UdbEC57~HEdT~TRd 5 FEE ~dsD FLUf6'ELEU QUESTION 1.35 (1.50)

For the following definitions, give the term that is defined.

a. The amount of heat required to change 1 lbm of water into 1 lbm of steam at a constant temperature. (0.5)

b. The ratio of the Critical Heat Flux to the actual heat flux. (0.5)

c. The ratio of the peak heat flux at core elevation : to the core average heat fivx. (0.5) l b

(***** END OF CATEGORY 01 *****)

2. PLANT DESIGN INCLUDING SAFETY AND ENERGENCY SYSTEMS PAGE 17 QUESTION 2.01 (1.00)

Which of the following statements concerning the operation of the Instrument Air (VI) system is NOT correct?

a. The BASE compressor runs continuously loading and unloading as necessary.

b. After an automatic start, the STBY 1 compressor will unload and stop after the BASE compressor is fully loaded and the 15 minute timer has timed out.

c. After an automatic start, the STBY 2 compressor must be manually stopped after the 15 minute timer has timed out.

d. At 76 psig VI header pressure, Station Air is supplied as a backup air supply.

QUESTION 2.02 (1.00)

Which of the following is NOT a purpose of the Reactor Protection System?

a. Prevent maximum DNBR from increasing above 1.3.

b. Prevent the m a ::i m u m heat flux (kW/ft) from exceeding limits.

c. Prevent NCS pressure from exceedin3 110% of design pressure.

d. Prevent secondary system pressure from exceeding 110% of design pressure.

(***** CATEGORY 02 CONTINUED ON NEXT PACE ***r*)

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 18 GUESTION 2.03 (1.00)

According to 10CFR50.46, which of the following is NOT a design criteria of the Emergency Core Coolins System subsystems.

a. The calculated peak centerline temperature shall not exceed 2000 degrees F.

b. The ma::imum cladding oxidation shall not exceed 17% of the total cladding thickness.

c. Tb. calculated total amount of hydrogen generated from the c7addins reaction with water shall not exceed 1% of the s otnt that would be generated if all cladding surrounding tie fuel reacted.

d. Caleviated changes in core geometry shall be such that the core remains amenable to cooling.

QUESTION 2.04 (1.00)

Which of the following is NOT a design feature of the Area Radiation Monitoring system?

s. Provide interlocks to automatically terminate discharge from waste systems at preset radiation levels.

b. Indicate radiation levels at various locations throughout tae station where personnel exposure is likely.

c. Sound local alarms when radiation level exceeds setpoint.

d. Indicate activity buildup in the reactor coolant filters.

QUESTION 2.05 (1.00)

Which of the following statements describing the actions assuciated with depressing the Rod Position Startup Push Button is NOT correct?

a. Resets demand position counters.

b. Recets master cycler, slave cycler, and bank overlap unit to *:ero'.

c. Resets Urgent Failure Alarm.

d. Resots P/A converter.

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

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 19 QUESTION 2.06 (1.00)

Which of the following statements concerning the operation of the Containment Air Release and Addition system is correct?

a. Initiation of air release AND initiation of air addition is AUTOMATIC.

b. Initiation of air release is AUTOMATIC and initiation of air addition is MANUAL.

c. Initiation of air release is MANUAL and initiation of air addition is AUTOMATIC.

d. Initiation of air release AND initiation of air addition is MANUAL.

QUESTION 2.07 (1.00)

Which of the following statements concerning normal pressurizer spray is correct? ,

3. The 2 normal spray lines supply water to the pressurizer through separate spray no::les.

b. A small continuous spray is provided by manual throttle valves in parallel with the spray valves.

c. The driving force for spray flow is the height difference between the S/G and the pressurizer.

d. An RTD downstream of each spray valve has a low temper-ature alarm which provides a warning of excessive spray flow.

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

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 20 QUESTION 2.08 (1.CO)

Which of the following statements describing the termination of containment spray following operation due to a LOCA is correct?

a. Containment spray will be automatically terminated by recirevlating the flow back to the FWST when containment pressure decreases below +0.25 psig.

b. Containment spray will be automatically terminated by securing the NS pumps when containment pressure decreases below +0.25 psig.

c. Containment spray will be automatically terminated by' closing the outside containment discharge isolation valve when containment pressure decreases below +0 25 psis.

d. Containment spray must be manually termie. Led by resetting the initiation signal and securing the NS pumps when ,

containment pressure decreases below +0.25 psis.

QUESTION 2.00 (1.00)

The purpcse of the flow orifice in the cold les accumulator (CLA) discharge line is given by which of the following?

s. To minimine the hydraulic resistance to flow from the CLA.

b. To entend the blowdown time of the CLA during a LOCA.

c. To provide a nethod to measure the flow from the CLA during a LOCA.

d. To provide a method to measure the flow when performing CLA discharge check valves leakage tests.

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

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 21 GUESTION 2.10 (1.00)

Which of the following gives the sources of water to the Auxiliary Feedwater (CA) system in order of perference?

a. CA CST, Hotwell, YM, RC, RN

b. CA CST, UST, Hotwell, RN, RC

c. CA CST, CST, UST, RN, RC

d. CA CST, FWST, RMWST, UST, RN OUESTION 2.11 (1.00)

Which of the following statements describin3 the design of the fuel transfer tube is correct?

a. A blank flange is used to close the transfer tube on BOTH the containment side and the spent fuel side.

b. A blank flange is used to close the transfer tube on the containment side and a valve is used on the spent fuel side.

c. A valve is used to close the transfer tube on the contain-ment side and a blank flange is used on the spent fuel side.

d. A valve is used to close the transfer tube on BOTH the con-tainment side and the spent fuel side.

QUESTION 2.12 (1.00)

Which of the following S/Gs supply steam to the auxiliary feedwater pump turbines?

a. S/Gs 'A' and 'B'

b. S/Gs 'A' and 'C'

c. S/Gs 'A' and 'D'

d. S/Gs 'B' and 'C'

e. S/Gs 'C' and 'D' (xxxxx CATEGORY 02 CONTINUED ON NEXT PAGE xxxxx)

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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 22 4

GUESTION 2.13 (1.00)

Which of the following statements concerning the automatic closure of the component cooling (KC) thermal barrier return isolation valves is correct?

a. High pressure in a return line will automatically shut its i return isolation valve.

b. High temperature in a return line will automatically shut its return isolation valve.

c. High flow in a return line will automatically shut its return isolation valve.

d. High activity in a return line will automatically shut

! its return isolation valve.

! GUESTION 2.14 (1.00)

Which of the following flowpaths corectly describes how power is normally supplied to 120 VAC Vital Panelboard 1ERPA?

a. 600 VAC from MCC 1EMXA, rectified to 125 VDC, inverted to '

120 VAC, and supplied to 1ERPA.

b. 600 VAC from MCC 1EMXA, transformed to 120 VAC, and supplied to 1ERPA.

c. 125 VDC from battery, inverted to 120 VAC, and supplied to l 1ERPA.

d. 120 VAC from Regulated Power distribution center IVRD is l supplied to 1ERPA.

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

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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 23 QUESTION 2.15 (1.00)

To protect against a loss of DC, the DC loads required for plant shutdown

! are supplied power from two sources through auctioneerins diode assemblies.

1 Which of the following are these two sources?

a. 120 VAC Vital Instrument and Control (VIRC) Power System and 125 VDC Auxiliary Control Power System.

b. 125 VDC VI&C Power System and 125 VDC Essential Diesel Auxiliary Power System.

c. 120 VAC VI&C Power System and 125 VDC VI&C Power System.

f d. 125 UDC Essential Diesel Auxiliary Power System and 125 VDC Auxiliary Control Power System.

4 QUESTION 2.16 (1.00)

The #3 NC pump seal leakoff is normally collected in which of the followins?

a. Containment Sump

b. Pressurizer Relief Tank !

c. Volume Control Tank

! d. Reactor Coolant Drain Tank

GUESTION 2.17 (1.00) i When establishin3 a cooldown rate on the ND system, the operator will MANUALLY adjust which of the followin3 valves? ;

I

a. KC outlet isolation valves from ND heat exchanger (VC-57A & 828).

b. ND heat exchanger outlet valves (ND-26 & 60). I

c. ND heat exchanger bypass valves (ND-27 & 61).

d. ND heat exchanger outlet valves AND bypass valves.

(xxxxx CATEGORY 02 CONTINUED ON NEXT PAGE xxxxx) f 4

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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 24 I

QUESTION 2.18 (1.00)

Which of the followins is the function of the flywheel on the NC pump?

a. Provide extended coastdown flow upon loss of power to pump.

i b. Resist reverse flow (counter rotation) in a secured pump. .

c. Prevent overspeed of motor in the event of a sheared shaft.

d. Provide inertia to prevent pump seizure from high bearin3 temperatures.

} GUESTION 2.19 (1.00) ,

! Listed below are the four bypass flow paths which occur within the reactor vessel.

1. Control Rod Guide Thimble and Instrumentation Guide Tube j 2. Cold Les to Hot Les '

3. Baffle Bypass

4. Upper Head Coolins j Which of the followinS choices gives the correct order of these flow

! rates stating with the highest percentage of flow and ending with the lowest.

a. 1, 4, 3, 2

b. 1, 4, 2, 3 l c. 4, 1, 3, 2 I

) d. 4, 1, 2, 3 t

1 j (***** CATEGORY 02 CONTINUED ON NEXT PAGE xxxxx) i i

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,,,_.4r_,,-.,-,---,c . , - - - _ - , . - - - , - , . , . , , , _ . - - . . , _ _ , ,

--..-----,,-,--,,,--..--y..,m,_...-.

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 25 I OUESTION 2.20 (1.00)

! When the Reactor Makeup system is in AUTOMATIC, how does i t know what strength boric acid to provide?

a. The automatic controller has an input from the boronneter.

b. The latest analysis from Chemistry is automatically inputted

. into the controller.

J

c. The controller uses the setting on the manual / automatic ;

Potentiometer. .

d. The controller uses the integrated charsing header flow j signal.

, GUESTION 2.21 (1.00)

The five safety valves for each S/G are set at different relief pressures for which of the following reasons?

4

a. Control of steam flow is smoother.

b. Prevent chattering of valves.

t c. Steam pressure oscillation is reduced.

d. Steam release in Doghouse is minimized.

QUESTION 2.22 (1.00)

The Standby Shutdown Facility contains systems designed to handle which of the following major events?

I' a. Rupture of dams impounding Lake Wylie and the service water pond.

b. A major fire disabling the Control Room controls and the Auxiliary Shutdown Panel controls.

l c. The Control Room and the Technical Support Center are unavailable.

d. A LOCA that requires evacuation of the plant site for more ,

than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

(***** CATEGORY 02 CONTINUED ON NEXT PAGE ***mm) h i

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, = . ,_ w- x, . , - - - - _. , -r. --

'r- , .H--* - w em e * - - - - - ----s -

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 26 QUESTION 2.23 (1.00)

The power range detector current comparator circuit compares which of the following?

a. Each individual power range total power signal to the average of the powe{ range total signals.

b. Each upper power range detector signal to its respective lower power range signal.

c. Each individual upper (lower) power range detector signal to the average of the upper (lower) power range detector signals.

d. The average upper power range detector signal to the average lower power range detector signal.

GUESTION 2.24 (1.50)

Indicate whether the following statements concerning diesel generator load sequencing are TRUE or FALSE.

a. If a LOCA and blackout occur simultaneously, the blackout bus will NOT be loaded on the diesel. (0.5)

b. If the diesel generator is paralleled with its essential bus a r.d t h e sequencer actuates due to a blackout, the diesel out-Pot breaker will trip. (0.5)

c. Placing the selector switch on the Auxiliary S/D Panel to ' Local' will prevent sequencer actuation due to a LOCA. (0.5)

(r**** CATEGORY 02 CONTINUED ON NEXT PAGE xxxxx) i

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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 27 QUESTION 2.25 (1.50)

Indicate whether the following statements concerning the Residual Heat Removal (ND) system are TRUE or FALSE.

a. The ND system is designed to reduce NC temperature to 140 degrees within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following a shutdown with both trains running. (0.5)

b. The design heat load handled by the ND system only includes core decay heat; it does not include RCP heat. (0.5)

c. The ND system can be used to provide p:r spray when the NC pumps are secured. (0.5) 00ESTION 2.26 (1.50)

Indicate whether the following statements concerning containment hydro-sen control are TRUE or FALSE.

a. When containment hydrogen concentration reaches 3 volume %, the H2 Purge system automatically opens an air supply line and a venting line to reduce the hydrogen concentration. (0.5)

b. The hydrogen skimmer fans start automatically 9 minutes after the receipt of a Safety Injection signal. (0.5)

c. The circonium-water reaction is the largest source of hydrogen production following a large break LOCA. (0.5)

QUESTION 2.27 (1.50)

Indicate whether the following connectionms to the Main Steam (SM) system tap off UPSTREAM or DOWNSTREAM of the main steam isolation valves.

a. Atmospheric steam dump valves (0.5)

b. Atmospheric PORVs (0.5)

c. F4PT supply (0.5)

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

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 28 GUESTION 2.28 (1.50)

Indicate whether (YES or NO) the following components are cooled by the Component Cooling (KC) system.

a. CA pump motor cooler (0.5)

b. KC pump motor cooler (0.5)

c. Fuel Pool cooling heat exchanger (0.5) l I

! QUESTION 2.29 (2.00) t t

For the following componients, indicate whether they will receive an OPEN, CLOSE, or NO signa' . upon a manual safety injection initiation. -

I

a. Control Room outside air isolation valves (0.2) l

b. Main Feedwater bypass valves (0.2) j

c. Cold Les Accumulator isolation valves (0.2) j

d. Charging header isolation valves (0.2) l I e. Main steam. isolation valves (0.2)

I l j f. FWST to centrifugal chargins pumps suction valves (0.2) i 3 NC pump seal water return isolation valve' (0.2)

! h. KC isolation valve from ND heat exchanger (0.2) ;

i. KC isolation-valve from letdown heat exchanger (0.2) ;

i ,

J. Steam supply valves to turbine-driven CA pump (0.2) l t ,

i !

[

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(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****) i P

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i 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 29 <

J l GUESTION- 2.30 (1.50) 1

Match the chemistry control in Column A to the ciaemical used for that t

, control in Column B.

COLUMN A COLUMN B

a. Control pH during startup 1. Hydrazine l b. Scavense oxygen during 2. Hydrogen i

startup from cold conditions

3. Lithium Hydroxide

c. Control oxygen during full

power operations 4. Ammonium Hydroxide QUESTION 2.31 (2.00)

,3

Match the NC system penetration in Column A to the correct location

! in Column B.

! COLUMN A COLUMN B i a. Excess Letdown 1. Loop B Hot Les

] 2. Loop D Hot Les

b. Alternate Chargins 3. Loop B Crossover Les

4. Loop C Crossover Les

! c. RHR Cooldown Suction 5. Loop C Cold Les 4

6. Loop D Cold Les j d. P:t Surse Line i

! GUESTION 2.32 (2.50)

! List the FIVE suction sources for the centrifu3al chargins pump.

l QUESTION 2.33 (1.50) ;

Other than reducing containment pressure, list the THREE functions of the Ice Condenser system that are accomplished by the sodium-tetra-

borate solution produced by meltdown of the ice.

(xxxxx END OF CATEGORY 02 xxxxx) 4 6

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3. INSTRUMENTS AND CONTROLS PAGE 30 OUESTION 3.01 (1.00)

Which of the following is NOT a function of the P-4 permissive (trip and bypass breakers open)?

a. Allows bypassing of steam dump cooldown interlock.

b. Allows operator block of SI signal.

c. Causes feedwater isolation if low Tavs is also Present,

d. Causes a turbine trip.

QUESTION 3.02 (1.00)

Which of the following reactor trip or SI signals is NOT b1 coked by permissive P-7?

a. Low P:r Pressure trip

b. Low P:t Pressure SI

c. High P:t Level trip

d. Low RCS Flow trip GUESTION 3.03 (1.00)

Which of the following does NOT automatically occur when a Feedwater Isolation signal is initiated 0

a. CF pump discharge v'Ives close

b. CF pump suction valves close

c. Tempering flow vai.>es close

d. CF bypass valves to CA no::les close (xxxxx CATEGGPY 03 CONTINUED ON NEXT PAGE **xxx)

3. INSTRUMENTS AND CONTROLS PAGE 31 GUESTION 3.04 (1.00)

Which of the following is NOT a reason for the S/G 1evel program?

a. Mu.imize cooldown of RCS following a steam break.

b. Maintain constant mass in S/G at all rower levels to i facilitate chemistry control.

c. Minimize containment pressure following a steam break.

d. Minimize carryover into the turbine blading.

QUESTION 3.05 (1.00) ;

t Which of the following reactions is used for neutron detection in the source range detector?

a. Neutron + Uranium-235 ---> 2 Fission Fragment Ions l b. Neutron + Nitrogen-16 ---> Nitrogen-17 + Gamma j c. Heutron + Baron-10 ---> Lithium-7 Ion + Helium-4 Ion

d. Neutron + F1vorine-19 ---> Nitrogen-15 Ion + Helium-4 Ion ;

I r

GUESTION 3.06 (1.00)

Which of the following statements describes the signal path from the l Source Ran3e detector to the Source Range level. meter on the MCB?

4 j

a. Detector, . Pre Amp, Discriminator, Los Integrator, Meter

b. Detector, Los Integrator, Pulse Shaper, Pulse Counter, Meter l c. Detector, Pre Amp, LoS Integrator, Discriminator, Meter

d. Detector, Los Amp, Meter i

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3. INSTRUMENTS AND CONTROLS PAGE 32 QUESTION 3.07 (1.00)

Which of the following methods is used to remove the gamma signal from the neutron signal in the intermediate range monitor?

a. The outer chamber prevents sammas from tont ing the inner chamber.

b. Inner chamber current cancels out gamma current in the outer chamber.

c. A pulse height discriminator does not allow the samma signals to be counted.

d. Squaring the combined signal makes the gamma contribu-tion insignificant.

QUESTION 3.08 (1.00)

Which of the following statements concerning the operation of the pressurizer pressure controller (PCY-455A) is correct?

a. It is possible to select all pressurizer pressure channels (I, II, III, and IV) as inputs to PCY-455A.

b. Only channels I and II may be selected as inputs to PCY-455A.

c. Only channels I and III may be selected as inputs to PCY-455A.

d. Only channels I and IV may be selected as inputs to PCY-455A.

QUESTIO.N 3.09 (1.00)

Which of the following signals cause BOTH a safety injection and a main steamline isolation?

a. Low Steamline Pressure

b. High Steam Pressure Rate

c. High Containment Pressure

d. Low Pressurizer Pressure t

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3. INSTRUMENTS AND CONTROLS PAGE 33 i

QUESTION 3.10 (1.00)

Which of the following expresses the combined error signal used by

the Reactor Control System to generate rod motion?

a. (Impulse Pressure - Nuclear Power) + (Tref - Tavs)

b. (Nuclear Power - Impulse Pressure) + (Tref - Tavs)

(Nuclear Power - Impulse Pressure) + (Tavs - Tref)

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c.

d. (Impulse Pressure - Nuclear Power) + (Tavs - Tref)

QUESTION 3.11 (1.00)

Which statement below regarding Auxiliary Feedwater System control is q correct?

a. It is not possible to auto-start the motor-driven Auxiliary

]

Feed pumps while they are in local control.

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b. The discharge valves for the motor-driven Auxiliary Feed pumps are electric motor operated and will fail open on an auto-start sequence to 55%.

c. If there is a loss of demineralized water for the Auxiliary

-l Feed pumps and 2 out of 3 low suction pressure from Train B

> activates, then suction' valves from the RC (Condenser Cire Water) and RN (Nuclear Service Water) systems will open.

d. To minimize Auxiliary Feed pump runouti the motor-driven 2

isolationvalve/onthemotor-drivenpumpsupplyline/toB(C) 4=6.S/G%g will close automatically if a high flow condition occurs.

QUESTION 3.12 (1.00)

Which of the following malfunctions would cause a pressurizer level I

indication of~0%?

a. -dP cell diaphragm rupture

b. Reference les rupture

c. Impulse line rupture

d. Equali=ing valve leakage

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3. INSTRUMENTS AND CONTROLS PAGE 34 l

QUESTION 3.13 (1.00) i With the Rod Control System in automatic, which of the following i conditions would result in a rod withdrawal?

j a. Nuclear power channel (N-44) fails high

b. Tcold (Ch. A) detector fails low

c. Impulse pressure (Ch. 1) fails high l

d. Tref fails low s

QUESTION 3.14 (1.00)

. Which of the following is the advantage of using the RTD manifold system for RCS temperature signals used for the Reactor Control and Protection System?

a. Allow the RTDs to be directly immersed into the RCS fluid.

b. Allow maintenance on the RTDs without draining the loop.

, c. Prevent having to apply ambient temperature compensation to the measured temperature.

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d. Prevent inadequate mixing of RCS fluid from causing inaccurate temperature indication.

GUESTION 3.15 (1.00) 1 With the pressurizer level control selector switch in position 1-2, a failure causes the following plant events. (Assume no operator actions taken.) ,

1. Chargins flow reduced to minimum

2. Pressurizer level decreases

3. Letdown secured and heaters off

4. Level increases until high level trip Which of the following failures occurred?

a. Level channel.I failed high

b. Level channel I failed low i

c. Le~ vel channel II failed high

d. Level channel II failed low (xxxxx CATEGORY 03 CONTINUED ON NEXT PAGE xxxxx) l 3

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3. INSTRUME"7S AND CONTROLS PAGE 35 GUESTION 3.16 (1.00)

Which of the following will cause the automatic closure of the UHI discharge isolation valves?

a. There is no automatic closure signal

b. Low nitrogen pressure in surge tank

c. Low water level in accumulator

d. Membrane leakage alarm GUESTION 3.17 (1.00)

Which of the following sets of pressurizer pressure setpoints is correct?

a. 2435 psig - High Pressure trip 2385 psig - Hi Sh Pressure alarm 2335 psis - PORV opens

b. 2210 psig - B/U heaters on 1955 psis - Low Pressure alarm 1845 psig - Low Pressure trip

c. 2385 psis - High Pressure trip 1945 psis - Low Pressure trip 1845 psis - Low Pressure SI

d. 2335 psig - PORV opens 2100 psis - PORV block 2000 psig - Low Pressure alarm (xxxxx CATEGORY 03 CONTINUED ON NEXT PAGE xxxxx)

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3. INSTRUMENTS AND CONTROLS PAGE 36 t

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GUESTION 3.18 (1.00)

Which of the following correctly describes the turbine runback asso- ;

j ciated with the loss of a MFPT? l f a. .If > 80% load, trip of either MFPT will cause a runback at I 133%/ minute until < 70% load.

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b. If > 80% load, trip of either MFPT will cause a runback at i 10%/ minute until < 70% load.

c. If > 70% load, trip of either MFPT will cause a runback at l 4 10%/ minute until < 70% load. ;

i d. If > 70% load, trip of either MFPT will cause a runback at ;

133%/ minute until < 50% load.

[

< e. If > 70% load, trip of either MFPT will cause a runback at 4 10%/ minute until < 50% load.

1 QUESTION 3.19 (1.50) .

i j Indicate whether the following statements concerning a resistance ;

j temperature detector (RTD) are TRUE or FALSE. ;

, a. An RTD is connected across one les of a bridge circuit. As l

{ temperature that is sensed by the RTD changes, a proportional

}

4 change in the output voltage (current) across the bridge occurs. (0.5) !

b. When an RTD fails open, it will indicate a downscale (low) !

]' reading on its meter. (0.5) !

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c. If an RTD is completely submerged, its ability to accurately (

j monitor temperature is unaffected by flow rate. (0.5) j

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3. INSTRUMENTS AND CONTROLS PAGE 37 I I' QUESTION 3.20 (2.00) 4 1 For the followins radiation detector types, indicate whether the

, output intensity (current or pulse height) is proportional to the l incident radiation enersyi i.e., if the incident energy increases,

, will the output intensity increase? (Answer YES or NO to each).

j a. Ion Chamber (0.5) i b. GM (0.5) !

I f i c. Proportional Counter (0.5)

J i d. Scintillation (0.5) !

4

QUESTION 3.21 (1.50) k Indicate whether the fo11owin3 statements concerning operation of the reactor trip (RT) and bypass (BY) breakers a.e TRUE or FALSE.

2 l a. If one train is placed in test while the other train's bypass breaker is closed, then both reactor trip breakers and both J bypass breakers will trip. (0.5) ;

t j b. If it is attempted to close both bypass breakers at the same ;

time, then both bypass breakers will trip but the reactor j trip breakers will remain closed. (0.5)

c. A solid state protection system (SSPS) train A reactor trip

signal will trip RTA and BYA breakers. (0.5) 4 GUESTION 3.22 (1.50) j Indicate.whether the followins statements concerning the Main Feed-j water *(CF) system are TRUE or FALSE.

i s. At 325 psis feed pump suction pressure (decreasins), a condensate

booster pump will auto start, if available. (0.5) l 1 ;

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b. 2/3 booster pumps tripping will trip the feedwater pump / turbine. (0.5) ;

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c. A feedwater pump will NOT start unless the feedwater pump recir- i culation valves are open. (0.5) !

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

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3. INSTRUMENTS AND CONTROLS PAGE 38

.i 4 GUESTION 3.23 (1.50)

Indicate whether the followins statements about the Engineered Safe-suard Actuation System are TRUE or FALSE. ,

s. Containment High-High Pressure bistables are energized to actuate. (0.5) f

b. Two pushbuttons, one per phase, are provided on the Control Board for actuatins Containment Isolation Phase A and B. (0.5)

c. The solid state protection system uses ' fast pulse' testins l for testins logic circuits without having to bypass the reactor trip breakers. (0.5) l QUESTION 3.24 (2.00)

Indicate whether the Over Power Delta Temperature trip setpoint will f INCREASE, DECREASE, or REMAIN THE SAME for the following parameter '

chan3es. Consider each separately.

j

a. Increasing Tavs (0.5)

b. Tavs less than rated power Tavs (0.5)

c. Delta I becomins more negative (0.5) t 4

d. Pressurizer Pressure decreasins (0.5)

GUESTION 3.25 (1.50)

Indicate whether the statements below apply to UNIT 1, UNIT 2, or i BOTH UNITS.

4 i j a. Steam senerator level is programmed at a constant level of 50%

j for all power levels. (0.5)

! b. 2 out of 3 channels greater than 82% will actuate the S/G high- e

hi3h water level turbine trip. (0.5) ;

l c. The low-low water level reactor trip setpoint varies as a func-i tion of pwer level. (0.5) i

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3. INSTRUMENTS AND CONTROLS PAGE 39 GUESTION 3.26 (1.50)

For the following instrument malfunctions, indicate whether the steam dump control system would ARM ONLY, ARM & DUMP or NOT BE AFFECTED.

a. Tavs fails high with a 10% step load reduction in progress. (0.5)

b. Turbine impulse pressure (Channel I) " ails high with a 25% step load reduction in progress. (0.5)

c. Turbine impulse pressure (Channel II) fails low with a 5% step load reduction in progress. (0.5)

GUESTION 3.27 (2.50)

Match the following reactor protection and control signals in Column A to their associated logic coincidence in Column B.

COLUMN A COLUMN B ao 2 loop loss of flow trip (per loop) 1. 1/2

2. 2/2

b. P-6 (SRM turn-on on power decrease) 3. 1/3

4. 2/3

c. P-12 (Lo-Lo Tavs on temperature decrease) 5. 1/4

6. 2/4 do P:t high pressure trip (pressure increase) 7. 3/4

e. PRM high power rod stop (power increase)

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3. INSTRUMENTS AND CONTROLS PAGE 40 i

GUESTION 3.28 (2.00) ,

i Hatch the followin3 reactor conditions in Column A with the rod speed l Column B. !

! COLUMN A COLUMN B

a. Withdrawins Shutdown Bank B in Shutdown 1. 8 spm ;

e Bank B position 2. 16 spm ;

i 3. 24 spa

b. Withdrawing Bank B in Control Bank B 4. 48 spa ,

position 5. 64 spm !

6. 72 spa j c. Automatic insertion with Tavs > Tref by j 2 degrees l d. Automatic insertion with Tavs > Tref by 5.5 degrees r

GUESTION 3.29 (1 50) 1 4

Uhe following questions concern the auto start defeat function of the Auxiliary Feedwater (CA) pumps.

a. What motor-driven CA pump auto start signal (s) can be overridden i with the auto start defeat switch? (1.0) l b. What other condition (besides switch) must be met for auto start override? (0.5)

I i QUESTION 3.30 (1.50) .

j List THREE alarms that indicate the status of the Main Fire Pumps.

j (Do not include redundant alarms, i.e., Pump A High lemp and Pump L

B High Temp would only count as one alarm.) L i

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! OUESTION 3 31 (1.50) ,

! Aside from transformers 1ETXA, 1ETXC, and 1ETXE, 4160 V bus 1 ETA i supplies normal power to ten other loads. List these TEN loads.

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4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 41

~~~~RhBiBL551EAL CBNTRUL'~~~~~~~~~~~~~~~~~~~~~~~ 1 j ____________________

a QUESTION 4.01 (1.00)

If the following critical safety functions were all displayed orange, which one has priority?

I

a. Suberiticality.

b. Heat Sink.

c. Integrity. !

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d. Inventory.

QUESTION 4.02 (1.00)

While operatins Unit 1 at 40% power with all 4 loops operating, you receive ' Seal Injection Filter A Hi D/P' and 'NC Pump Seal Water lo 4

Flow' alarms. You check el seal outlet temperatures, KC flow from i NC pumps, and NC pump lower bearing temperatures and you obtain the l following data NC PUMP #1 SEAL OUT TEMP KC FLOW FROM PUMP LOWER BRG TEMP d

A 225 39 SPm 215 B 220 40 spm 227

., C 230 41 spm 220

D 215 38 SPm 210 f All temperatures are rising slowly (<1 degree / min). Which of the following actions should you take?

a. Trip NC pump C and refer to AP/1/A/5500/04, ' Loss of Reactor Coolant Pump'.

b. Trip the reactor, trip NC pumps B and C, then refer to EP/1/A/5000/01, ' Reactor Trip'.

c. Trip the reactor, trip NC pump B ,and then refer to EP/1/A/5000/01, ' Reactor Trip". .

d. Trip NC pump B and refer to AP/1/A/5500/04, ' Loss of Reactor Coolant Pump'. .

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4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 42

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RE656L66 5dL 66 TRUL QUESTION 4.03 (1.00)

While operating Unit 1 at 85% power, an inadvertent baron dilution occurs which results in a ' Control Rod Bank Lo-Lo Limit' alarm. Which of the following is an appropriate immediate operator action in accordance with AP/1/A/5500/13, ' Boron Dilution'?

a. Emergency borate to clear the ' Control Rod Bank Lo Lir.it' alarm.

b. Emergency borate to clear the ' Control Rod Bank Lo-Lo limit' alarm.

c. Stop the source of the dilution and obtain a chemistry sample to determine required boration.

d. Stop the dilution process and emergency borate to raise rods to the position they occupied prior to the inadvertent dilution.

QUESTION 4.04 (1.00)

Prior to commencing a Natural Circulation Cooldown per EP/1/A/50/1A1, it is necessary to borate the NC system to 100 ppm greater than that required to provide a Xenon free Shutdown Margin (SDM) of 1300 pcm at the minimum expected cold shutdown temperature. Which of the follow-in3 is the reason why the shutdown boron concentration should be 100 ppa greater than necessary to achieve the 1300 pcm SDM?

a. Due to improper NC system mixing with natural circulation flow,

b. Due to possible errors in NC system temperature indications with natural circulation flow.

c. Due to changes in boron worth with natural circulation flow.

d. Due to restricted reactor vessel head cooling with natural circulation flow.

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4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 43

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RA65UL55iEAt E5NTk L j

QUESTION 4.05 (1.00) 1

On a power increase from 15% to 75% power, the following actions are performed.

! 1. Start a second CF pump i 2. Verify SSRH highload valves automatically open

3. Check OAC thermal output measurement and have IaE adjust nuclear instrumentation as required.

4. Verify *P9 - Reactor Trip On Turbine Trip Permissive' light Comes one j Which of the below is the correct performance order of these actions?

a. 1, 3, 4, 2 q

j b. 1, 3, 2, 4

c. 3, 1, 4, 2 l

d. 3, 1, 2, 4

}

OUESTION 4.06 (1.00)

! On a steam generator overpressurization casualty, which of the following 2 is the correct priority (in order of highest to lowest) of the methods .

used to depressurize the affected S/G? l

a. MSIV bypass, PORV, MSIV, Turbine driven CA pump

b. PORV, MSIV bypass, MSIV, Turbine driven CA pump r

c. MSIV bypass, PORV, Turbine driven CA pump, MSIV i d. PORV, MSIV bypass, Turbine driven CA pump, MSIV l

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4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 44

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GUESTION 4.07 (1.00)

If a void exists in the reactor vessel, which approach below represents j

. the preferred actions taken to collapse the void in EP/1/A/5000/2F3,

,: " Void in Reactor Vessel *? ,

a. Decrease temperature while maintaining system pressure. l i

b. Start a SI pump to increase system pressure while keeping !

temperature constant.

I c. Increase system pressure using pressurizer heaters while

maintaining pressurizer level.

d. Fill pressuri:er solid and vent the reactor vessel head.

QUESTION 4.08 (1.00) i Followins a valid reactor trip and safety injection, which of the state-

monts below correctly describes the NC Pump Trip Criteria?

l a. SI flow indicated and NC pressure < 1500 psis.

1 4

b. SI flow indicated and p:r level < 10%.

c. SI flow indicated and NC subcooling < 0 degrees.

, d. SI flow indicated and NC temperature < 500 degrees.

I j GUESTION 4.09 (1.00) j AP/1/A/5500/04, ' Loss of Reactor Coolant Pumps', contains a CAUTION i that states that reactor power should be reduced to below a certain power level prior to restarting the affected NC pump (s). Which of the following is this power level?

a. 45%

l b. 35%

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c. 25%

d. 15% ,

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4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 45

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RE65EL 55Ehl"C ETRUL QUESTION 4.10 (1.00)

Durins a reactor coolant leak-identification process, which of the following gives the requirement for shifting centrifugal charging puap suction from the VCT to the FWST?

a. Pressurizer level decreases to 17%.

b. VCT level decreases to 20%.

4 c. Letdown isolated and pressurizer level decreasing.

d. Prior to starting second centrifugal charging pump.

l QUESTION 4.11 (1.00)

In the event of a toxic sas release in the Control Room and it is decided to evacuate the Control Room, which of the following is the i correct actions of the NCO?-

a. Take over as OATC and perform required immediate and i subsequent actions.

b. Go to ASP and initiate performance of ASP operator actions.

c. Go to AFWPTCP and initiate performance of AFWPTCP operator actions.

i d. Go to HVAC ECP's and initiate performance of HVAC ECP actions.

QUESTION 4.12 (1.00)

An immediate operator action for a loss of ND train is to verify ND suct on supply adequate. Which of the followins is the criteria for performing this verification?

a. Pzr level > 10% or NC vessel level > 10%.

b. FWST level > 50% or Containment sump level > 10%.

! c. ND suction pressure > 200 psis.

d. ND discharse pressure > 200 psis.

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GUESTION 4.13 (1.00)

In the process of determining if SI can be terminated, it is determined that the secondary heat sink is available. In which of the following

! situations could SI be t'erminated?

i i PZR LVL SUBC00 LING PRESSURL l a. 5% 60 degrees stable

b. 25% 45 degrees increasing

c. 20% 65 de3rees decreasing

d. 10% 55 degrees stable QUESTION 4.14 (1.00)

During a natural circulation cooldown following a reactor trip, which of the following criteria determine the amount of NC subcooling required?

I a. NC system cooldown rate.

i t

b. Reactor power history (decay heat rate).

c. Pressurizer level.

d. Number of CRD vent fans running.

I j GUESTION 4.15 (1.00)

) A CAUTION in EP/1/A/5000/1C, High Energy Line Break Inside Containment, states if FWST level is less than _____, then SI systems should be insediately aligned for cold les recirculation. Which of the following correctly completes this caution?

a. 37%

b. 25%

c. 15%

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d. 9%

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4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 47

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GUESTION 4.16 (1.00)

Prior to transferring rod control from MANUAL to AUTO during a reactor i

startup, Tave and Tret should be within which of the following?

a. 1 degree F.

b. 3.5 degrees F. l

c. 5 degrees F.

d. 10 degrees F.

GUESTION 4.17 (1.00) i Which of the following statements is correct concerning the status of

, the Nuclear Instrumentation Recorder prior to withdrawing control bank j rods for a reactor startup?

a. The highest reading source range channel and the highest reading intermediate range channel is selected and the NR-45 chart speed is set to "Hi' speed.

b. The highest reading source range channel and the highest reading intermediate range channel is selected and the NR-45 chart speed is set to 'Lo' speed.

c. The highest reading source ranSe channel and the lowest reading intermediate range channel is selected and the NR-45 chart speed j is set to "Hi' speed.

I d. The highest reading source range channel and the lowest reading i intermediate range channel is selected and the NR-45 chart speed ,

. -is set to "Lo* speed.

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4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 48

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R 656L 5555L'UUEiR6L QUESTION 4.18 (1.00)

During a natural circulation cooldown, the preferred method of NC system dspressurization is by which of the following?

a. openin3 the normal spray valve.

b. opening the NV auxiliary spray valve.

c. opening the pressurizer PORVs.

d. opening the reactor vessel vent valves.

GUESTION 4.19 (1.00)

The indicated AFD is considered to be outside its target band as soon as

______ operable excore channel (s) is/are indicating outside the target band.

a. one

b. two

c. three

d. four GUESTION 4.20 (1.00)

The transfer of control of the pressuri=er heaters (sub-bank of Group D) to the Safe Shutdown Facility is performed by'which of the following?

a. swapping power supplies of the motor control centers from normal to alternate in the ETA rooms.

b. swapping a plus type connector from the normal connection to the alternate connection in the SSF.

c. placing the Pressurizer Heater Selector Switch in the main control room to the ' Remote-SSF' position.

d. placing the Pressurizer Heater Selector Switch in the SSF to the ' Local' position.

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RE65ULUE UEL EU TR5L QUESTION 4.21 (1.00)

If a " Rod Control Urgent Failure' alarm occurs due to a failure in the logic cabinet, the Tave/ Tref mismatch is maintained by which of the following?

a. controlling turbine load.

b. taking manual control of individual control rod banks.

c. taking manual control of individual control rod Groups.

d. boration and dilution of the' reactor coolant system.

QUESTION 4.22 (1.00)

The trip bistables of a failed power range detector is placed in the trip condition by which of the following?

a. placing the applicable bistable test switch in the ' Test' position in the Reactor Protection Cabinet. ,

b. removing the applicable control and instrument power fuses on the power range drawers.

c. P l acing the applicable Power Mismatch Bypass switch to the failed position at the Miscellaneous Control and Indication Panel.

d. placing the applicable Comparator Channel Defeat switch to the failed channel position at the Detector Current Comparator Panel.

QUESTION 4.23 (1.00)

Which of the following is a characteristic of natural circulation?

a. SG steam pressure - STABLE or DECREASING.

1

b. NC system subcoolin3 - LESS THAN O DEGREES F.

c. NC. loop hot les temperatures - STABLE or INCREASING.

d. NC loop cold les temperatures - GREATER THAN SATURATION

, TEMPERATURE FOR SG PRESSURE.

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RE656L5555EL E5NTRUL GUESTION- 4.24 (1.00)

If a 500 mrem pocket dosimeter reads ______ or more, it should be returned to HP for re:eroing.

a. 120 mrem -

b. 200 mrem

c. 250 aren i

d. 300 mren GUESTION 4.25 (1.00)

Which of the following is a 10 CFR 20 exposure limit?

a. 5 rem / year - whole body.

b. 1 rom / quarter - whole body.

c. 18.75 rem / quarter - hands.

d. 7 rem / quarter - skin of whole body.

QUESTION 4.26 (1.00)

Which of the following radiation exposures would inflict the greatest biological damage to man? ,

a. 1 Rem of GAMMA. 1

b. 1 Rem of ALPHA.

c. 1 Rem'of NEUTRON.

d. NONE of the abovel they are all equivalent.

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(***** CATEGORY 04 CONTINUED ON NEXT PAGE xxxxx) 4 4

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l .- - -.

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 51 RADIOLOGICAL CONTROL QUESTION 4 27 (1.00)

KC. flow through the NCP seal water heat exchanger is assured by verifying which of the following?

a. proper flow indication.

b. proper pressure indication.

c. at least one KC pump is operating.

d. no seal water Hex KC Hi/Lo flow alarm ccndition exists.

QUESTION 4.28 (1.00)

When an individual has recieved ______ of an administrative exposure limit, he/she must inform his/her supervisor.

a. 75%

b. 80%

c. 85%

d. 90%

GUESTION 4.29 ( .50)

TRUE or FALSE?

It is allowable to briefly exceed the ' full steady state licensed power level' by as much as 2% for as long as 15 minutes.

QUESTION 4.30 (1.50)

List ALL the immediate operator actions for a loss of offsite power.

QUESTION 4.31 (1.00)

If KC,and NV seal cooling for any NC pump is lost, within how many ninutes AND from where must seal in,iection flow be initiated?

(***** CATEGORY 04 CONTINUED ON NEXT PAGE xxxx*)

p.

N sg 6

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 52

~~~~R E55 L55isAL C5UTR5L'~~~~~~~~~~~~~~~~~~~~~~~

GUESTION 4.32 (1.50)

List ALL immediate operator actions that are to be taken in the event of damaged spent fuel in the reactor building.

QUESTION 4.33 (2.50)

List ALL the immediate operator actions for a reactor trip.

QUESTION 4.34 (1.50)

If a reactor trip occurs and the ' Safety Injection Actuated' light is NOT lit, what conditions must the operator immediately verify?

QUESTION 4.35 (1.50)

-List ALL the immediate action steps for a turbine trip when the main control room manual trip does NOT trip the turbine.

QUESTION 4.36 (2.00)

List ALL the immediate actions required when the reactor cannot be l tripped using the control room trip switches.

I (xxxxx END OF CATEGORY 04 xxxxx)

(xxxxxxxxxxxxx END OF EXAMINATION *****xxxxxxxxxx) i 4

/M Hb It:.K i

i

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 53

~ ~

~~~~TU5R566YOdkfC5""U5dT TRd 5E5R d 6~ELUE6~FL6U ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS ANSWER 1.01 (1.00)

D REFERENCE G neral Physics, HT & FF, pp. 319 and 320 ANSWER 1.02 (1.00) c REFERENCE G;neral Physics, HT & FF, pp. 239 and 240 ANSWER 1.03 (1.00) d REFERENCE NUS, Nuclear Energy Training - Reactor Operation VEGP, Training Text, Vol. 9, Ch. 21 DPC, Fundamentals of Nuclear Reactor Engineering 001/010-K5.13 ANSWER 1.04 (1.00) d REFERENCE NUS, Nuclear Energy Training - Reactor Operation, p. 8.3.2 W@stinghouse Reactor Physics, pp. I-5.3 and I-5.4 VESP, Training Text, Vol. 9, p. 21-60 DPC

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 54

~

~~~~T UEk 66Y IE5657~H5di~TRd EfER d 6~ELUf6~fL6U ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS ANSWER 1 05 (1.00) e REFERENCE NUS, Nuclear Energy Training - Reactor Operation, p. 6.4-2 W3stinghouse Reactor Physics, p. I-3.15 HBR, Reactor Theory, Session 43, p. 3 DPC, Fundamentals of Nuclear Reactor Engineering, p. 94 001/000-K5.49 (2 9/3.4)

ANSWER 1.06 (1 00) a REFERENCE NUS, Nuclear Energy Training - Reactor Operation, pp. 16-19 & 25 VEGP, Training Text, Vol. 9, p. 21-67 DPC, Fundamentals of Nuclear Reactor Engineerin3 ANSWER 1.07 (1.00) d REFERENCE NUS, Nuclear Energy Training - Reactor Operation, pp. 9.2-5, 11 3-2, a 11.4-3 VEGP, Training Text, Vol. 9, pp. 21-21, 33, 60, a 62 DPC, Fundamentals of Nuclear Reactor Engineering ANSWER 1.08 (1.00) b REFERENCE

.i1U S , Nuclear Ener3y Training - Plant Performance, Section 3.4 012/000-MS.01 (3.3/3.8)

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 55

~

~~~~TAEE566f dE5d5~~5EdT TRdO5EEE~ds6~FL6i6 IL64 ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS ANSWER 1.09 (1.00) e REFERENCE VEGP, Training Text, Vol. 9, p. 21-47 Westinghouse Reactor Physics, pp. I-3.17 & 19 DPC, Fundamentals of Nuclear Reactor Engineering, p. 106 001/000-K5.49 (2.9/3.4)

ANSWER 1.10 (1.00) b

. REFERENCE General Physics, Heat Transfer Thermodynamics and Fluid Flow, pp. 145 - 160.

002/000-K5.01 (3.1/3.4)

ANSWER 1.11 (1.00) c REFERENCE General Physics, HT & FF, Section 3.2 WBN, HT & FF, p. 13 002/000-K5.01 (3.1/3.4) ,

ANSWER 1.12 (1.00) b (1.0) l REFERENCE NUS, Nuclear Energy Training, Module 3, Unit 6 Westinghouse Reactor Physics, Sect. 3, Neutron Kinetics and Sect. 5, Core Physics HBR, Reactor Theory, Sessions 20 and 24 - 31 DPC, Fundamentals of Nuclear Reactor En3ineering, p. 236 0

)

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

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 56

~

~~~~T EER566Y d 5C57~s5di~TEd 5EER d 6~EL6f6~FL6E ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS 001/000-K5.13 (3.1/3.6)

ANSWER 1.13 (1.00) b REFERENCE General Physics, HT & FF, p. 328 WBN, HT & FF, p. 17 Ccaponents: Pumps - Relative relationships (2.6/2.6)

, ANSWER 1.14 (1.00) d REFERENCE Wastinghouse Reactor Physics, pp. I-5.63 - 76 f HBR, Reactor Theory, Sessions 38 and 39 DPC, Fundamentals of Nuclear Reactor Engineerins,Section VI 001/000-K5.39 (3.5/4.1)

ANSWER 1.15 (1.00) b REFERENCE General Physics, HT & FF, p. 229 l 002/000-K5.01 (3.1/3.4)

ANSWER 1 16 (1.00) d REFERENCE Wostinghouse Reactor Physics, Section I-5, MTC and Power Defect DPC, Fundamentals of Nuclear Reactor Engineerins 002/000-K5 02 (3.3/3.6)

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

,c,, .-

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 57

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~~~~_______________ ____________________________

ANSWERS -- CATAWBA 1 -85/09/09-JERRY-DOUGLAS ANSWER 1.17 (1.00) b REFERENCE Wsstinghouse Reactor Physics, pp. I-4.19 - 24 DPC, Fundamentals of Nuclear Reactor Engineering, pp. 120 - 129 ANSWER 1.18 (1.00)

C-REFERENCE Wastinghouse Reactor Physics, pp. I-2.19 - 21 HBR, Reactor Theory, Session 14, p. 3 DPC, Fundamentals of Nuclear Reactor Engineering, p. 53 001/000-K5.57 (3.0/3.2) i ANSWER 1 19 (1.00) b REFERENCE WBN, TS, p. B 3/4 1-1 HBR, TS, p. 3 10-10 CAT, TS, p. B 3/4 1-1 NAPS, TS, p. B 3/4 1-1

' ANSWER 1.20 (1 00) c REFERENCE HBR, Reactor Theory, Session 42, pp. 3& 4 DPC, Fundamentals of Nuclear Reactor Engineering i 004/000-HS.08 (2.6/3.2) i a

i l .-

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 58

~~~~is5E 667 d 165,~U5dT TRd 5F5R d 6~fLUi6 ft6E

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ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS ANSWER 1.21 (1.00)

D REFERENCE HBR, Reactor Theory, Sessions 41 and 42 DPC, Fundamentals of Nuclear Reactor Engineerin3' PP. 121 and 122 004/000-K5.08 (2.6/3.2)

ANSWER 1.22 (1.00) a REFERENCE DPC, Fundamentals of Nuclear Reactor Engineering, p. 138 001/000-K5.02 (2.9/3.4)

ANSWER 1.23 (1.00) b REFERENCE General Physics, HT & FF, p. 320 ANSWER 1.24 (1.00)

B REFERENCE General Physics, HT & FF, pp. 139, 148, 156, and 331 ANSWER 1 25 (1.00) c REFERENCE DPC, Fundamentals of Nuclear Reactor Engineering, p. 170 4

l

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 59

--- isiss557sisics- siii isisiFEE Es5 FEUi5 FE6s ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS 001/000-K5.13- (3.7/4.0)

ANSWER 1.26 (1.00) a REFERENCE DPC, Fundamentals of Nuclear Reactor Engineering, p. 170 001/000-K5.19 (3.1/3.4)

ANSWER 1.27 (1.00) b REFERENCE DPC, Fundamentals of Nuclear Reactor Engineerin3' P. 96 001/000-K5.56 (2.8/3.1)

ANSWER 1.28 (1.00) a REFERENCE VEGP, Trainin3 Text, Vol. 9, pp. 21-60 & 61 Wostinghouse Reactor Physics, Sect. I-5 HBR, Reactor Theory, Session 26, p. 2 DPC, Fundamentals of Nuclear Reactor Engineering ANSWER 1.29 (2.00)

a. INNER (0.5) i

b. OUTER (0.5)

INNER

~

c. (0 5) .

d. INNER (0.5) !

REFERENCE NUS, Nuclear Energy Training - Plant Performance, pp. 10-1.8&9 004/000-K5.09 (3.7/4.2)

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 60

~

~~~~TEER 66Y EfC5I~5Edi TRd 5E5R d 6~EL6E6'FL6U ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS ANSWER 1.30 (1.50)

o. FALSE (0.5)

b. FALSE. (0.5)

c. TRUE (0.5)

REFERENCE Wastinghouse Reactor Physics, pp. I-3.9 and I-3.4 HBR, Reactor Theory, Sessions 22 and 23 DPC, Fundamentals of Nuclear Reactor Engineerins ANSWER 1.31 (1.50)

e. LARGER (0.5)

b. LONGER (0.5)

c. CONSTANT (0.5)

REFERENCE Wostinghouse Reactor Physics, Section I-4 DPC, Fundamentals of Nuclear Reactor Engineerin3, Sect. IV 004/000-K5.08 (2.6/3.2)

ANSWER 1.32 (2.00)

a. REMAIN THE SAME (0.5)

b. DECREASE (0.5)

c. INCREASE (0.5)

d. DECREASE (0.5)

REFERENCE Steam Tables l 010/000-K5.02 (2.6/3.0) 1 I

l l

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 61

~~~~iHiss557sAsiCs, REAT TRAssFis As5 FLUi5 Ft5s ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS ANSWER 1.33 (2.00)

o. INCREASE (0.5)

6. INCREASE (0.5) )

t

c. DECREASE (0.5)

d. DECREASE (0.5) i REFERENCE General Physics, HT & FF - Fluid Flow Applications for Systems and Components 002/000-K5.01 (3.1/3.4)

ANSWER 1.34 (1.50)

a. 1 (or 5) (0.5)

b. 3 (0.5)

c. 5 (0.5)

REFERENCE General Physics, HT & FF, pp. 100 and 181 001/000-K5.47 (2.4/2.9)

ANSWER 1.35 (1.50)

a. Latent Heat (of Vapori=ation) .(<0.5)

b. DNBR (0.5)

c. Heat Flux Hot Channel Factor (0.5)

REFERENCE WBN, HT & FF, pp. 11, 22, and 24 General Physics, HT & FF, pp. 38 & 231 and HBR, TS, p. 3.10-12 002/000-K5.01 (3.1/3.4)

I

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

i

. 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 62 1 _______________________________________________________

l ANSWERS -- CATAWBA 1 -85/09/09-JERRY 00UGLAS ANSWER 2.01 (1.00)

I C OR h REFERENCE CAT, CN-PSM-SY-VI, p. 2 i 078/000-A3.01 (3.1/3.2)

. I ANSWER 2.02 (1.00) [

8

! REFERENCE !

CAT, OP-CN-SPS-IC-IPE, p. 1 012/000-PWG435 (4.1/4.3) i ANSWER 2.03 (1 00) e ,

i REFERENCE 10CFR50.46(b)

! 006/050-PWG435 (4.2/4.3)

I i

ANSWER 2.04 (1.00) 1 a

l REFERENCE i i CAT, EMF Lesson Text, p. 7 ;

} 072/000-PWG435 (3.5/3.7) 4 1

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2o PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 63 l ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS i

j ANSWER 2.05 (1.00)

! C

] REFERENCE ,

3 CAT, OP-CN-SPS-IC-IRE, p. 28 1

i 001/010-K4.02 (2.5/2.6)

-K6.04 (2.9/3.2) l

] ANSWER 2.06 (1.00) ;

I

} d REFERENCE ;

- CAT, CN-PSN-SYS-VQ, p. 1 l

ANSWER 2.07 (1.00)

I b

i REFERENCE WBN, LP - Reactor Coolant System, pp. 27 a 28 J

CAT, CNSD-0150-01, p. 5 I f 010/000-K4.01 (2.7/2 9) f F

! t

! ANSWER 2 08 (1.00) f e or b '

REFERENCE -

CAT, CN-PSH-SY-NS, p. 2 f

! 026/000-A3.01 (4 3/4 5) 2 a

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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 64 ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS ANSWER 2 09 (1.00) b REFERENCE CAT, CLA Lesson Text, p. 6 006/000-K6 02 (3.4/3.9)

ANSWER 2.10 (1.00) b REFERENCE CAT, CN-1223.42-00, pp. 7 and 8 061/000-K4.01 (3 9/4.2)

ANSWER 2.11 (1 00) ,

b REFERENCE CAT, FC Lesson Text, p. 11 034/000-K6.01 (2.1/3.0)

ANSWER 2.12 (1 00) d REFERENCE CAT, OP-CN-SPS-STM-SM, p. 5 039/000-K1.07 (3.4/3.4) 061/000-K1.03 (3.5/3.9) 4

l

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i 2o PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PACE 65 ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS ,

p ANSWER 2.13 (1.00) ,

t c

REFERENCE CAT, OP-CN-SPS-SY-KC, p. 12 i 1

008/030-A2 02 (2 5/2.8)

ANSWER 2.14 (1 00) !

a REFERENCE CAT, Figure CN-SYS-EPG-1 012/000-K4.10 (3 1/3.5)

CNSWER 2 15 (1 00) b l

REFERENCE CAT, CNS0-0010-15 (EPL), p. 3 ,

063/000-K1.03 (2.9/3.5)

ANSWER 2.16 (1.00) d REFERENCE CAT, Figure CN-CMP-NCP-5 l

003/000-K6.02 (2 7/3.1) l l

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l ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS l

I ANSWER 2.17 (1 00)

{ b i

) REFERENCE CAT, CN-PSN-SY-NO, p. 2 005/000-A1 01 (3.5/3.6) i 1

) ANSWER 2.18 (1 00) o i REFERENCE j CAT, CNSD-0150-01e p. 3 j 003/000-K4.10 (2 3/2 5) i

! ANSWER 2.19 (1.00) l j d

< t 4 REFERENCE t 1 CAT, CN-CMP-RVI, p. 1

, 002/000-K5.13 (3.3/3.6) i L

! ANSWER 2 20 (1.00) c l REFERENCE

! CAT, OP-CN-SPS-PS-NV, p. 53 j i 004/010-K6.05 (3.1/3.3) i '

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> 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 67 j ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS -

l ANSWER 2.21 (1.00)

b i REFERENCE i CAT, CN-1223.43-01, p. 10 ;

i j 035/010-K6 02 (3.1/3.5) t I

! ANSWER 2.22 (1.00) i i b

I l REFERENCE -

l

CAT, CNCS-0144-02, p. 2 l t i l ANSWER 2.23 (1.00) 1

e 1

REFERENCE i CAT, Figure CN-IC-ENB-7 j 015/020-K5.08 (2.9/3.4) .

ANSWER 2.24 (1.50) '

Li 1 c. TRUE -

b. FALSE

c. TRUE !

1

] REFERENCE .

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ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS I I

j . ANSWER 2.25 (1.50)

! a. FALSE (0.5)

b. FALSE (0.5)

c. TRUE (0.5) i REFERENCE ,

] CAT. ND Lesson Text, p(HMU(3.6/3.9) i ANSWER 2 26 (1.50) :

l' j

j c. FALSE (0.5) [

b. FALSE (0.5)

c. FALSE (0 5) ;

] REFERENCE CATe CN-PSM-SY-VXe pp. 1 & 2 and Fig.>res CN-SYS-VX-3 and CN-SYS-VY-1 i 029/000-PWG435 (3.5/3.7) j -A4.01 (4.0/4.0) l

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, ANSWER 2 27 (1.50) ;

i Ii* a. DOWNSTREAM (0.5) !

b. UPSTREAM (0.5) j c. 00WNSTREf.M (0.5) p i

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. ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS i

1 ANSWER 2.28 (1.50)

I

c. YES (0.5) l b. YES (0 5) ;

c. YES (0.5) 4 REFERENCE i CAT, OP-CN-SPS-SY-KC, p. 4

! 008/000-K1 02 (3.3/3.4) -

1 i

ANSWER 2.29 (2.00)

) (0.2 pts each) d

a. NO l

b. CLOSE

-I c. OPEN !

I d. CLOSE i o. NO j f. OPEN !

j 3 CLOSE 6

, h. Me OPsW j 1. NO 1

J. NO j REFERENCE

CATe CNSD-0151-01, p. 25 and PSM-SY-VC, p. O and PSM-SY-CF, p. 7 and :

1 PSM-PS-NV, p. 12 and PSM-SY-KCe pp. 4 a 5 and PSM-SY-CA, p.6.

013/000-A3.02 (4.1/4 2)

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, ANSWER 2.30 (1.50) j c. 3 (0.5) j

! b. 1 (0.5)

, c. 2 (0 5) l 3

REFERENCE

WBNe LP - Chemical and Volume Control System, pp. 21 & 22 7 CAT, CH-1223.04, p. 9 l

1 004/000-K4.01 (2.8/3 3)

-K4.02 (2.1/2.4) t i

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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 70 (

j ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS .

i i i 1

] ANSWER 2.31 (2.00)

a. 5 (0.5)

b. 6 (0.5) !

] c. 1 (0.5) j d. 1 (0.5) i !

l REFERENCE !

! CAT, NC Lesson Text, pp. 10 & 11

002/000-K1.06 (3.7/4.0) 3 -K1 08 (4.5/4.6) !

-K1.09 (4.1/4.1) i j ANSWER 2 32 (2.50) h 1. VCT 6 RWT (0.5)

2. Rx M/U System (blender) 7, GAT ANY FtVE (0.5)

3. FWST (0 5)

4. Emergency Boration (0.5) l 5. ND System (0.5)

~

l REFERENCE

! CAT, OP-CN-SPS-PS-NV, p. 22 004/000-K4 07 (3 0/3.3) ,

l i ANSWER 2.33 (1.50)

1. RetainIodine(teaiP*I P N (0.5) i

2. Absorbs neutrons (0 5)

! 3. Serves as cooling medium (0.5)

REFERENCE j CAT, CN-SPS-SY-NF, p. 3 025/000-PWG035 (3.4/3.6) b I

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3. INSTRUMENTS AND CONTROLS PAGE 71 i j ANSWERS -- CATAW8A 1 -85/09/09-JERRY DOUGLAS 1

l ANSWER 3.01 (1.00) k o REFERENCE

~

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012/000-K4.10 (3.3/3.5) l i

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1 b

(

l REFERENCE

Surry, Instrumentation Manuale Process Protection Instrumentation, pp.

I 1 and 2 l

/ CAT, Figure CN-IC-IPE-2 l 013/000-K4.07 (3.7/3.8) i l ANSWER 3.03 (1.00) [

a b .

, t 1 REFERENCE CAT, PSM - CFe p. 5 013/000-K1 15 (3.4/3.8) i 1

j ANSWER 3.04 (1.00) l i

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i 3. INSTRUMENTS AND CONTROLS PAGE 72

{ ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS t

j ANSWER 3 05 (1.00) c ,

REFERENCE i Nuclear Power Reactor Instrumentation System Handbook, Vol. 1, Ch. 2 l 015/000-K5.01 (2 9/3.2)

! ANSWER 3.06 (1.00) :

C REFERENCE l FNP, Excore Nuclear Instrumentation System, Fi3 7 l Surry, Instrumentation Manual, Excore Instrumentation System, p. IV-1.29 ;

j VEGP, Training Text, Volume 5e Fig. 3a-2 CAT, Fi 3ure CN-IC-ENB-4 1

) 015/000-K6.03 (2 6/3 0) l j ANSWER 3.07 (1.00) 4

. b l

i

! REFERENCE :

) Nuclear Power Reactor Instrumentation Systems Handbook, Vol. 1, Ch . 2 i 015/000-K5 02 (2.7/2.9) 4 ANSWER 3 08 (1.00) 3

e

! REFERENCE

. CAT, Figure CN-IC-ILE-1 and 8 3

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3. INSTRUMENTS AND CONTROLS PAGE 73 j ANSWERS -- CATAWBA 1 -85/09/09-JERRY 00UGLAS :

ANSWER 3.09 (1 00) l l

O REFERENCE VEGP, Training Text, Vclume 6, RPS and Question Bank, Section 6 l CAT, Figure CN-IC-ISE-5 1 013/000-K1 01 (4.2/4.4) l i

i I ANSWER 3.10 (1.00)

D or C i REFERENCE l CAT, PSM, P. 6 :

t

.I j 001/000-K4.03 (3.5/3.8) [

ANSWER 3.11 (1 00)

d I REFERENCE Cat, 50-CA, pp. 8 - 14 I 061/000-K4.04 (3.1/3.4) e i

i ANSWER 3.12 (1.00) .

i l c !

i

< REFERENCE I WBN, LP - Pressuri:er Level Control System, p. 5 j Nuclear Power Reactor Instrumentation Systems Handbook, Vol II, Ch. 15 i 000/028-K2.02 (2.6/2.7)

, 000/028-A2 01 (3.4/3.6)

+

1 !

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i :

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i 3. INSTRUMENTS AND CONTROLS PAGE 74

! ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS l ANSWER 3.13 (1.00) c :

I REFERENCE WBN, LP - Rod Control System, pp. 8 a 14 CAT, PSMe Figures CN-IC-IRE-21 & 22 001/000-K1.05 (4.5/4.4)

ANSWER 3.14 (1.00) -

b a Or b REFERENCE FNP, RCS Lesson Plane p. 23 j CAT, CN-1223.03, p. 18 1 002/000-K6.06 (2.5/2.8)

?

I ANSWER 3.15 (1.00) i a REFERENCE CAT, PSM, Figures CN-IC-ILE-11 & 22 011/000-A2.10 (3.4/3.6) l 1

ANSWER 3.16 (1.00)

)

e REFERENCE CAT, CN-PSM-SY-UHI, p. 1

) 006/000-K4.09 (3.8/4.1)

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1 i

I l

I r

i

(

,- ,,m , --,c,,,,-+-.w,--r.-~~. ,,-,-v---un-new-- - , , -

. ~--,rm-. ,-.r

- , - ~ ~ ~ , , - - - - m ee r-, , , , - ,,,w ---me,-

1

3. INSTRUMENTS AND CONTROLS PAGE 75 ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS l

ANSWER 3.17 (1.00) c REFERENCE CAT, CN-IC-ILE, p.3 l 010/000-PWG426 (3.5/3 9) i i

i ANSWER 3.18 (1.00) l b REFERENCE 4 CAT, CN-SYS-EHC, Figure CN-EHC

< 045/000-K4.12 (3.3/3.6) i ANSWER 3.19 (1.50)

O. TRUE l b. FALSE I

c. FALSE

, REFERENCE I Nuclear Power Reactor Instrumentation Systems Handbook, Vol. 1, Ch. 4 Components: Sensors / Detectors - Temperature ;

Theory (2.5/2.6) [

Failures (3.0/3 1) [

003/000-K5.03 (3.1/3.5) 1 t

i I i

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, _ . _ . . . _ . . - , . ., . - . ,-_-...m, , .. ,m. , ,-m. ..-._ . _ mm_-.m.__i.__ , . . , . _ . , . . , , , ,y. ,

3. INSTRUMENTS AND CONTR01.S PAGE 76 a ____________________________ ,

ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS I

ANSWER 3.20 (2.00)

o. YES (0.5)

' (0.5)

b. NO

c. YES (0.5)

d. YES (0.5)

REFERENCE FNP, Health Physics and Radiation Protection Lesson Plans, pp. 41-46 William J. Price, Nuclear Radiation Detection, pp. 43 - 46, 77, 138, and 196 VEGP, Training Text, Volume 9, pp. .23 42 072/00-K5.01 (2.7/3.0)

ANSWER 3.21 (1.50)

e. TRUE

b. FALSE

c. FALSE I

REFERENCE CAT, SPS-IC-IPE, p. 9

! 012/000-A3.07 (4.0/4.0)

ANSWER 3.22 (1.50)

a. TRUE

b. FALSE

c. FALSE REFERENCE CAT, CN-PSM-SY-CF, pp. 5&6

. 059/000-K4.03 (2.1/2.3)

-K4.16 (3.1/3.2)

-K4.14 (2 1/2.3)

I l

i l

f i

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

. 3. INSTRUMENTS AND CONTROLS PAGE 77 ,

+

ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS i

i ANSWER 3.23 (1 50)

I

e. TRUE (0.5)

b. FALSE (0.5)

c. TRUE (0.5)

REFERENCE CAT, ISE Lesson Text, pp. 9, 11, and 14 ANSWER 3.24 (2.00)

a. DECREASE (0.25)

b. REMAIN THE SAME (0.25)

c. REMAIN THE SAME (0.25)

d. REMAIN THE SAME ,

(0.25)

REFERENCE FNP, Tavs, Delta T, and Pimp, pp. 16 & 17 Surry, Instrumentation Manual, Sect. 9, p. IV-5.5 4

WBN, LP - Reactor Coolant Temperature Instrumentation, p. 8 CAT, OP-CN-SPS-IC-IPE, p. 23 012/000-K6.11 (2.9/2.9)

ANSWER 3.25 (1.50)

a. UNIT 2 (0.5)

b. BOTH UNITS (0.5)

c. UNIT 1 (0.5)

REFERENCE CAT, PSM-CF-IFE, p. 1 035/010-K4.01 (3.6/3.8) 4 4

]

s,

3. INSTRUMENTS AND CONTROLS PAGE 76

~

ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS ANSWER 3.26 (1.50)

a. ARM & DUMP (0.5)

b. -. vu m , u .au ARM ONLY (0.5)

c. ARM ONLY (0.5)

REFERENCE CAT, SD-IDE, pp. 19 - 21 and PSM -IDE p. 15 and CN-IC-IDE-15 041/020-K4.04 (2.1/2.3)

ANSWER 3.27 (2.50)

a. 4 (0.5)

b. 2 (0.5)

c. 6 (0.5)

d. 6 (0.5)

o. 5 (0.5)

REFERENCE CAT, PSM, CN-IC-IPE, pp. 2, 6, a7 012/000-K6.03 (3.1/3.5) 012/000-K6.10 (3.3/3.5)

ANSWER 3.28 (2.00)

a. 5 (0.5)

b. 4 (0.5)

c. 1 (0.5)

d. 6 (0.5)

REFERENCE WBN, LP - Rod Control System, pp. 5&9 CAT, PSM-IRE 001/000-K4.02 (3.8/3.8) s

3. JNSTRUMENTS AND CONTROLS PAGE 79 ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS ANSWER 3.29 (1.50)

e. 1. Low-Low Level (in 1 S/G) (0.5)

2. Both FWPTs Tripped (0.5)

b. Less than P-11 (0.5)

REFERENCE CAT, CN-PSM-SY-CA, p. 5 061/000-K4.02 (4.5/4.6)

ANSWER 3.30 (1.50) AW 3 AT O.5 NM

1. (Control Power) Trouble 4.Mator Bearidg Upper T**P (0.5)

2. Fail to Start L*Wer Te.p (0.5)

5. M o+or Bearld3

3. Runnin9 4, 5tedse Tt=P (0.5)

REFERENCE CAT, CN-PSM-SY-RF, p. 1 086/000-PWG427 (4.0/4.2)

ANSWER 3.31 (1.50)

(0.15 pts each)

1. KC pump 1A1

2. KC pump 1A2

3. NS pump 1A

4. ND pump 1A

5. NI pump 1A

6. NV pump 1A

7. CA pump 1A

8. RN pump 1A

9. KF pump 1A

10. A/C comp A (Nor)

REF ER Ei4CE Cate Plant Summary Manual -EP, Figure 3 062/000-K2.01 (3.3/3.4) 1 6

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 80

~~~~Rd656L665 L'UE TREL'~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS ANSWER 4.01 (1.00) a REFERENCE MNS EP/2/A/5000/10, p.2.

CNS EP/1/A/5000/02, p.3.

ANSWER 4.02 (1.00) d REFERENCE CAT, AP/1/A/5500/08, p. 2 000/015-A2.08 (3.4/3.5)

-PWG927 (4.2/4.5) 1 ANSWER 4.03 (1.00) b REFERENCE CAT, AP/1/A/5500/13, p. 2 000/024-K3.02 (4.2/4.4) l ANSWER 4.04 (1.00) a REFERENCE CAT, EP/1/A/5000/1A1, p. 2 PNG-424 (3.5/4.0) l l

[

9

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 81

~~~~R A5i5L55i5AL 56UTR5L'~~~~~~~~~~~~~~~~~~~~~~~

1 ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS ANSWER 4.05 (1.00) b REFERENCE CAT, OP/1/A/6100/03, p. 5 PWG-929 (3.7/4.0)

ANSWER 4.06 (1.00) b REFERENCE CAT, EP/1/A/5000/2C2, pp. 3-4 ANSWER 4.07 (1.00)

, C REFERENCE CAT, EP/1/A/5000/2F3, pp. 4 and 5 ANSWER 4.08 (1.00) c REFERENCE ,

CAT, EP/1/A/5000/01, Enclosure 1 ANSWER 4.09 (1.00) c REFERENCE CAT, AP/1/A/5500/04, p. 4 O'00/017-PWG424 (3.5/4.1) i

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 82

~

~~~~R d656L65fddL 66 TR6L"'~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS ANSWER 4.10 (1.00)

, d REFERENCE CAT, AP/1/A/5500/10, p. 2 000/009-K3.04 (4.1/4.3)

-K3.21 (4.2/4.5)

ANSWER 4.11 (1.00) b REFERENCE CAT, AP/1/A/5500/17, p. 2 and Enclosure 10 i 000/068-K3.12 (4.1/4.5) i ANSWER 4.12 (1.00) a REFERENCE ,

CAT, AP/1/A/5500/19, p. 2 i 000/025-K3.03 (3.9/4.1) ,

i

, ANSWER 4.13 (1.00) d REFERENCE CAT, EP/1/A/5000/01, p. 16 000/009-K3.24 (4.1/4.6)

J I

I

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 83

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

~~~~R d656L66E6 L"66 TR6L ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS ANSWER 4.14 (1.00) d REFERENCE CAT, EP/1/A/5000/1A1, p. 5 000/009-K3.26 (4.4/4.5)

ANSWER 4.15 (1.00) o REFERENCE CAT, EP/1/A/5000/1C, p. 1 000/040-K3.04 (4.5/4.7)

ANSWER 4.16 (1.00) l a REFERENCE MNS OP/1/A/6100/01 p. 19.

CAT OP/1/A/6100/01, Encl. 4.1, 2.95.10.

001/000-A4.03 (4.0/3.7)

ANSWER 4.17 (1.00) a REFERENCE MNS OP/1/A/6100/01, pp.14-15.

CAT OP/1/A/6100/01, Encl. 4.1, 2.70 and 2.75.

015/000-K6.01 (2.9/3.2) l i

l l

4o PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 84 i

~~~~Rd656L6656dL'66 TR6L'~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS ANSWER 4.18 (1.00) b REFERENCE MNS EP/2/A/5000/1.1, p.5.

CNS EP/1/A/5000/1A1, p.4.

ANSWER 4.19 (1.00) b REFERENCE CAT- TS, 4.2.1.1.

015/020-PWG427 (4.2/4.4)

ANSWER 4.20 (1.00) d REFERENCE MNS OP/0/A/6100/17, Encl. 4.1, p.5.

CAT, 010/000-PWG421 (3.1/3.2)

ANSWER 4.21 (1.00) a REFERENCE MNS, AP/2/A/5500/14, Case I, p.2.

CAT, AP/1/A/5500/15, Case I, p.2.

i 001/050-PWG928 (4.4/4.4) 3

)

-,n -

g- , e - - - - - - - ~ -,. + n- ,-e

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

I

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 85

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

~~~~R 656L66 6dE"_66 TREL ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS l

ANSWER 4.22 (1.00) b REFERENCE MNS, AP/2/A/5500/16, Case IV, p.9.

CAT, AP/1/A/5500/16, Case IV, p.14.

015/000-A4.03 (3.8/3.9)

ANSWER 4.23 (1.00) a REFERENCE MNS, EP/2/A/5000/01, p. Encl. 1 Foldout.

CAT, EP/2/A/5000/1A1, Encl. 1, p.15.

000/017-K1.01 (4.4/4.6)

ANSWER 4.24 (1.00) d REFERENCE MNS Orientation Manual, p.27.

CNS Directive 3.8.6, para. 2.9.

ANSWER 4.25 (1.00)

C REFERENCE 10 CFR 20.101 000/060-K1.02 (2.5/3.1)

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 86

~~~~R d65666656 L 66 TR6L'~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS ANSWER 4.26 (1.00) d REFERENCE 10CFR20 068/000-K5.04 (3 2/3.5)

ANSWER 4.27 (1.00) d REFERENCE CAT, OP/1/A/6200/01, Enci 4.1, 2.8.1.

008/000-K1.02 (3.3/3.4)

ANSWER 4.28 (1.00) 4 b REFERENCE CAT, Directive 3.8.6, p.2.

000/059-K1.02 (2.6/3.2)

ANSWER 4.29 ( .50)

TRUE REFERENCE CAT, TS-Interpretation of 1 25, 10/04/84

,,- c---

l

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 87 l

" ~~~~~~~~~~~~~~~~~~~~~~~~

~~~~E 656L 555IL EUNTRUL ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS ANSWER 4.30 (1.50)

1. Verify at least one diesel runnins (0.5)

2. Verify B/0 sequencer actuated (for runnins diesel) (0.5)

3. Ensure CA pump #1 is runnins (0.5) i REFERENCE

! CAT, AP/1/A/5500/07 000/056-K3.02 (4.4/4.7)

ANSWER 4.31 (1.00)

1. 10 minutes (0.5)

2. From the SSF (0.5)

REFERENCE CAT, AP/1/A/5500/08 000/015-PWG421 (2.7/2.8.'

l

-PWG428 (4.4/4.4e ANSWER 4.32 (1.50)

1. Ensure Containment Evacuation Alarm is sounded (if EMFs are in alarm) (0.5)

2. Terminate any waste Sas release in pro 3ress (0.5)

3. Stop Containment Purse (0.5)

REFERENCE CAT, AP/1/A/5500/25, p. 3 000/036-K3.03 (3.7/4.1) i

4 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 88

~~~~

Rd656LU6 65L'66 TRUL'~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS ANSWER 4.33 (2.50)

1. Manually exercise reactor trip train A & B switches. (0.5)

2. Verify a reactor trip has occured. (0.5)

3. Verify a turbine /senerator trip has occured. (0.5)

4. Verify ETA & ETB are enersized. (0.5)

5. Verify SI is not required. (0.5)

REFERENCE MNS, AP/2/A/5500/01, p.3.

CAT, AP/1/A/5500/01, p.3.

000/007-K3.01 (4.0/4.6)

ANSWER 4.34 (1.50) 0 0.5 points each:

Check if:

1. SGs <725 psis.

2. PZR <1845 psis.

3. Containment > 1 2 psis.

REFERENCE CAT, EP/1/A/5000/01, p.3.

000/007-K3.01 (4.0/4.6)

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 89

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

~~~~R 56 6L66 6 L"5 TRUL ANSWERS -- CATAWBA 1 -85/09/09-JERRY DOUGLAS ANSWER 4.35 (1.50)

MNS:

1. Stop both DEH pumps. (0.5)

2. Place the turbine in manual and close the sovernor valoes in FAST action. (0.5)

3. Locally trip the turbine. (0.5)

CNS:

1. Locally trip turbine. (0.5)

2. Unload turbine with Standby Load Set potentiometer. (0 5)

3. Depress ON and Bypass pushbuttons on Turb. CNTL. PNL. (0.5)

REFERENCE HNS, AP/2/A/5500/01, p.3.

CAT, AP/1/A/5500/02, p.2.

000/007-A1.07 (4.3/4.3)

ANSWER 4.36 (2.00)

1. Manually insert rods. . (0.5)

2. Locally ggen the tr2p breakers. + kj Pa SS b rea kc"3 (0.5)

3. Locally ope $ the MG sets cr', et b cr'er. (0.5)

4. Locally open the MG sets matas b r.e ak e r . (0.5)

REFERENCE Y"' ' #

MNS. EP/2/A/5000/11.1, p.2.

CAT, FP/8/ A /Taop /F1, p. L 000/029-K3.12 (4.4/4.7)

TEST CROSS REFERENCE PAGE 1 GUESTION VALUE REFERENCE 01.01 1.00 WGD0000530 01.02 1.00 WGD0000531 01.03 1.00 WGD0000001 01.04 1.00 WGD0000003 01.05 1.00 WGD0000012 01.06 1.00 WGD0000013 01.07 1.00 WGD0000014 01.08 1.00 WGD0000017 01.09 1.00 WGD0000098 01.10 1.00 WGD0000110 01.11 1.00 WGD0000186 01.12 1.00 WGD0000187 01 13 1.00 WGD0000188 01.14 1.00 WGD0000191 01.15 1.00 WGD0000192 01.16 1.00 WGD0000202 01.17 1.00 WGD0000263 01.18 1.00 WGD0000265 01.19 1.00 WGD0000416 01.20 1.00 WGD0000483 01.21 1.00 WGD0000486 01.22 1.00 WGD0000520 01.23 1.00 WGD0000532 01.24 1.00 WGD0000535 01.25 1.00 WGD0000536 01.26 1.00 WGD0000537 01.27 1.00 WGD0000538 01.28 1.00 WGD0000541 01.29 2.00 WGD0000020 01.30 1.50 WGD0000424 01.31 1.50 WGD0000004 01.32 2.00 WGD0000018 01.33 2.00 WGD0000022 01.34 1.50 WGD0000534 01.35 1.50 WGD0000419 40.00 02.01 1.00 WGD0000116 02.02 1.00 WGD0000605 02.03 1.00 WGD0000607 02.04 1.00 WGD0000611 02.05 1.00 WGD0000616 02.06 1.00 WGD0000118 02.07 1.00 WGD0000433 02.08 1.00 WGD0000548 02.09 1.00 WGD0000604 02.10 1.00 WGD0000606 02.11 1.00 WGD0000600 02.12 1.00 WGD0000609

A TEST CROSS REFERENCE PAGE 2 QUESTION VALUE REFERENCE 02.13 1.00 WGD0000612 02.14 1.00 WGD0000614 02.15 1 00 WGD0000615 02.16 1.00 WGD0000617 02 17 1.00 WGD0000618 02.18 1.00 WGD0000619 02.19 1.00 WGD0000620 02.20 1.00 WGD0000621 02.21 1.00 WGD0000622 1

02.22 1.00 WGD0000623 02.23 1.00 WGD0000625 02.24 1.50 WGD0000364 02.25 1.50 WGD0000547 02.26 1.50 WGD0000549 02.27 1.50 WGD0000610 02.28 1.50 WGD0000613 02.29 2.00 WGD0000542 02.30 1.50 WGD0000434 02.31 2.00 WGD0000546 02.32 2 50 ,WCD0000624 02.33 1.50 WGD0000626 40.00 03.01 1.00 WGD0000184 03.02 1.00 WGD0000283 03.03 1.00 WGD0000362 03.04 1.00 WGD0000544 03.05 1.00 WGD0000039 03.06 1.00 WGD0000048 '

03.07 1.00 WGD0000088 03.08 1.00 WGD0000104 WGD0000137 03.09 1.00 03.10 1.00 WGD0000358 03.11 1.00 WGD0000360 i 03.12 1.00 WGD0000437

03. 13 1.00 WGD0000440 -

1 03.14 1.00 WGD0000524 03.15 1.00 WGD0000543 03.16 1.00 WGD0000545 i 03.17 1.00 WGD0000602 03.18 1.00 WGD0000603 03.19 1.50 WGD0000002 03.20 2.00 WGD0000047 ,

03.21 1.50 WGD0000094 03.22 1.50 WGD0000119 I 03.23 1.50 WGD0000357 ;

03.24 2.00 WGD0000054 ;

03.25 1.50 WGD0000346 ;

03.26 1.50 WGD0000350 [

03.27 2 50 WGD0000105 6 i

A. A_ -- -- - _4,4

- >-A+-- i AJJ -4m - wa J - + +u-4 _44---d. --- .2-4. -u - a ----++- h-~ - ---Aa- + +ae3- -- . A J---

4 .

TEST CROSS REFERENCE PAGE 3

, GUESTION VALUE REFERENCE 03.28 2.00 WGD0000439 4 03.29 1.50 WGD0000114 03.30 1.50 WGD0000115 03.31 1.50 WGD0000365 40 00 04.01 1.00 WGD0000640 04.02 1 00 WGD0000372

04.03 1.00 WGD0000373 04.04 1.00 WGD0000375 04.05 1.00 WGD0000383 04.06 1.00 WGD0000386

04.07 1.00 WGD0000390 04.08 1.00 WCD0000392 l 04.09 1.00 WGD0000627 04.10 1.00 WGD0000630 04.11 1.00 WGD0000631 04.12 1.00 WGD0000632 J

04.13 1.00 WGD0000634 04.14 1.00 WGD0000635 04.15 1.00 WGD0000636

, 04.16 1.00 WGD0000637

, 04.17 1.00 WGD0000638 04.18 1.00 WGD0000639 04.19 1.00 WGD0000641 04.20 1.00 WGD0000642 04.21 1.00 WGD0000643 t 04.22 1.00 WGD0000644 04.23 1.00 WGD0000645 04.24 1.00 WGD0000646 04.25 1.00 WGD0000647 i 04.26 1.00 WGD0000648 04.27 1.00 WGD0000649 04.28 1.00 WGD0000652 04.29 .50 WGD0000655 l 04.30 1.50 WGD0000628 04.31 1.00 WGD0000629 04.32 1.50 WGD0000633 04.33 2.50 WGD0000650 04.34 1.50 WGD0000651 4

04.35 1.50 WGD0000653

, 04.36 2.00 WGD0000654

. 40.00 160.00 e

e t- .

U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: CATAWBA 1 AND 2 REACTOR TYPE: PWR-WEC4 DATE ADMINISTERED: 85/09/09 EXAMINER: TON ROGERS APPLICANT: _________________________

INSTRUCTIONS TO APPLICANT:

Uce 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%. E:: amination papers will be pteked up six (6) hours after the e:< a m i n a t i o n starts.

% OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 40.00 THEORY OF NUCLEAR POWER PLANT

________ .'5.00'____

_ ___________ ________ 5.

OPERATION, FLUIDS. AND THERMODYNAMICS 40.00 25.00 PLANT SYSTEMS DESIGN, CONTROL,

________ ______ ___________ ________ 6.

AND [NSTRUMENTATION 40.00 PROCEDURES - NORMAL, ADNORMAL,

_l5.00 ___________ ________

7.

EMERGENCY AND RADIOLOGICAL CONTROL 40 0 ADMINISTRATIVE PROCEDURES,

___I__0__ _I'_SI__ 00 ___________ ________

8.

CONDITIONS, AND LIMITATIONS 160.00 100.00 TOTALS FINAL GRADE _________________%

All work done on this e::a m i n a t i o n is my own. I have neither given nor received aid.

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

5PPL}CENiI5~55GNdiURE l

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

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 2

1 ----- -

gg--------------------------------------

r GUESTION '5 01 (1.00) ,

0.00733 delta k per k is the same as !

T

e. 0.733 pcm.

b. 7.33 pcm. .

c. 73.3 pcm. (

d. 733 pcm. .

l QUESTION 5 02 (1.00)

Which of the following conditions would cause a 1/M plot to be non-conservative during fuel loadin3? l l a. Fuel bein3 loaded closer to a source range detector than to the !

neutron source. ;

b. Loading fuel in the order of high reactivity worth to low reactivity I worth.

l

c. Loading poison rods between the source range detectors and spaces to

] be filled by fuel assemblies.

I' d. Increasing the boron concentration in the moderator.

i GUESTION 5.03 (1.00)

]

i The required Shutdown Margin is less for Mode 5 than it is for Mode 1 ,

! becLuse [

l ?

l a. negative reactivity from boron has replaced the negative reactivity J i from xenon by the time Mode 5 is reached, thus contributing to a more stable reactivity condition. !

b. the negative reactivity worth from samarium increases followin3 a plant shutdown. [

c. the reactivity transients resulting from a steam line break [

cooldown are minimal. [

d. all control bank rods are inserted.

(***** CATEGORY 05 CONTINUF.D ON NEXT PAGE xxxxx) l I

+

i s

i !

i 4 5

E l

l l

S. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3 q g GUESTION 5.04 (1.50)

List the three bases for the required rod insertion limits.

GUESTION 5.05 (1.00)

The xenon peak that occurs after a reactor trip is higher following a 100% power equilibrium xenon condition than a 25% power equilibrium condition because

c. the fission yield for xenon is higher at 100% power.

b. there is more iodine in the core at the time of a trip from 100% power.

c. there are more thermal neutrons in the core at 100% power.

d. there are more delayed neutrons in the core at 100% power.

QUESTION 5.06 (1.00)

Neutron capture as a result of resonance is more significant at EOL than BOL primarily due to

a. the reduction of fuel to clad gap distance.

b. the reduction in the moderator's boron concentration.

c. the increase of Pu-240 in the core.

d. the increase in thermal neutron flux.

QUESTION 5.07 (1.00)

Delayed neutrons allow an operator more control of the reactor because

a. there are more delayed neutrons than prompt neutrons.

b. delayed neutrons are born at a higher energy level than prompt neutrons.

c. delayed neutrons decrease the avera3e neutron generation time.

d. delayed neutrons increase the averaSe neutron generation time.

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 4 00ESTION 5.08 (1.00)

A subcritical reactor with a neutron source strength of 20 cps and a source rango count rate of 200 ces has a keff cf (assume a proportional-ity constant nf 1.0)

e. 0.85.

b. 0.90.

c. 0.95.

d. 0.99.

QUESTION 5.09 (1.00)

During a reactor startup, the first reactivity addition caused count rate to increase from 10 cps to 16 cps. The second reac-tivity addition caused count rate to increase from 16 eps to 32 eps. Which of the following statements describing the rela-tionship between the reactivity values of the first and second reactivity additions is correct?

a. The first reactivity addition was larger.

b. The second reactivity was larger.

c. Both reactivity additions were equal,

d. There is not enough data given for anyone to determine the relationship between the two reactivity additions.

QUESTION 5.10 (1.00)

As the boron concentration increases, the moderator temperature coefficient becomes

a. less neSative because there is an increase in neutron leakage.

b. more negative because there is an increase in neutron leakage.

c. less ne3ative because there is an' increase in neutron absorption.

d. more negative because there is an increase in neutron absorption.

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 5 TEERE665EAE5C5' ~' " ' "' " " ~ ''" "

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GUESTION 5.11 (1.00)

With a startup rate of 0.5 decades per minute established, reactor power will increase by a factor of 5 approximately every

a. 60 seconds.

b. 72 seconds.

i

c. 84 seconds.

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d. 96 seconds.

QUESTION 5.12 (1.00)

! The change in reactivity associated with a change in Keff from 0.920 to f

1.004 is approximately

a. 0.091.

b. 0.087.

4 c. 0.084.

d. 0.080.

l QUESTION 5.13 (1.00)

The effective delayed neutron fraction decreases over core life partly because

a. the number of delayed neutron precursor groups increase.

b. the fission yield for Pu-239 i nc r e a s'e s .

c. of a buildup of Pu-239 in the core.

d. soluble boron is removed from the core.

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 6 I QUESTION 5.14 (1.00)

The -1/3 DPM SUR followins a reactor trip is caused by which of the followins?

a. The decay constant of the longest-lived group of delayed neutrons.

b. The ability of U-235 to fission with source neutrons.

I c. The amount of negative reactivity added on a trip bein3 greater than the Shutdown Margin.

d. The doppler effect adding positive reactivity due to the temperature decrease followins a trip.

QUESTION 5.15 ( .50)

A reactor that is prompt critical is also supercritical. TRUE or FALSE?

QUESTION 5.16 (1.00)

Which of the followin3 express the relationship between differential rod worth (DRW) and integral rod worth (IRW)?

a. ORW 1s the slope of the IRW curve for a partievlar rod position.

b. DRW is the area under the IRW curve for a particular rod position.

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c. DRW is the square root of the IRW for a particular rod position.

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d. There is na relationship between DRW and IRW.

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 7 l

2 GUESTION 5.17 (1 00) .

l Which of the following statements concerning samarium reactivity i effects is correct? i

a. The equilibrium (at power) value of samarium is Dependent upon power level. The peak value of samarium following a shutdown is Dependent upon power level prior to shutdown. i

b. The equilibrium (at power) value of samarium is Dependent upon power ;

level. The peak value of samarium following a shutdown is Independent of power level prior to shutdown.

c. The equilibrium (at power) value of samarium is Independent of power i The peak value of samarium followin3 a shutdown is Dependent i

level.

upon power level prior to shutdown.

d. The equilibrium (at power) value of samarium is Independent of power level. The peak value of samarium following a shutdown is Independent of power level prior to shutdown.

QUESTION 5.18 (1.00)

A reactor shutdown from full power equilibrium conditions using control bank rods only will

a. decrease the AFD value, causing it to be less negative.

b. increase the AFD value, causing it to be more negative.

c. increase the AFD value, causing it to be more positive. l

d. have no affect on the AFD value. (

GUESTION 5.19 (1.00)

The power reactivity defect is expressed in units of

s. PCM per % full power.

. b. PCM per PPM boron.

c. PCM.

I d. % full power per PPM boron.

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5. THEORY.OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 8

--- ygggggggg gggg--------------------------------------

j QUESTION 5 20 (1.00)

The Unit 1 ARD, HZP differential baron worth in PCM/ PPM is approximately

a. 7. ,

1

! b. 14.

1

c. 24.

4.

d. 41.

QUESTION 5.21 (1.00)

Why does the xenon worth curve provided as Figure 5-21 indicate a higher positive reactivity worth for a trip following 100% power operation than one followin3 10% power operation at 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after shutdown?

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1 QUESTION 5.22 (1.00) i Which of the following will contribute to a smaller (less negative) doppler power coefficient over core life?

f a. Fuel densification.

i

b. Clad creep.

c. Crud buildup on the fuel cladding.

d. Fission gases released to the gap between the fuel and cladding.

QUESTION 5.23 (1.00) i l

Which of the following radioactive isotopes found in the reactor coolant l l system would NOT indicate a leak through the fuel cladding?

! a. I-131.

I b. Xe-133.

c. Co-60.

! d. Kr-85.

)

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 9 QUESTION 5.24 (1.00)

Which of the following is a correct statement concerning the Bases for i

the Tech Spec Pressure / Temperature Limits?

a. During cooldown, the most limiting location is at the outside of the vessel wall because of the thermal gradients produce compressive stressi.s there.

b. During cooldown, the most limitins location is at the inside of the vessel wall because the thermal gradients produce tensile stresses there.

c. During heatup, the mest limiting location is at the inside of the vessel wall because the thermal gradients produce compressive

stresses there.

d. Durin3 heatup, the most limiting location is at the outside of the vessel wall because the thermal gradients produce compressive stresses there.

QUESTION 5.25 (1.00) 1 The most serious problem with reachins the critical heat flux is caused by

a. the poor thermal conductivity of steam since steam blanketing occurs on the fuel claddins.

b. the blockage of flew through the core when steam bubble formation 7 becomes significant.

c. the displacement of boron from the core as steam bubble formaticn becomes significant.

d. the high pressure surses in the Reactor Coolant System caused by steam bubble formation.

(xxxxx CATEGORY 05 CONTINUED ON NEXT PAGE xxxxx)

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 10 QUESTION 5.26 (1.00)

The allowable heat flux at the top of the core is lower than the bottom of the core because

a. the' water at the top has the highest enthalpy.

b. the top of the core is under a lower pressure.

c. in the event of a design bases LOCA, the top of the core is uncovered longer.

d. the fuel pellets loaded at the top of the fuel rods are not designed for a high power output.

QUESTION 5.27 (1.00) i If the Reactor Coolant System is to be maintained at 350 degrees F, then the Steam Dump pressure controller should be set to the equivalent of

a. 135 psia.

b. 205 psia.

c. 295 psia.

d. 350 psia.

QUESTION 5.28 (1.00)

Which of the following meets the definition of Guadrant Power Tilt Ratio?

a. Minimum power range upper detector output divided by the average upper detector output.

b. Ma,ximum power range upper detector output divided by the average upper detector output.

c. Minimum power range upper detector output divided by the averaSe lower power range detector output.

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d. Maximum power range upper detector output divided by the average lower power range detector output.

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 11 3

QUESTION 5.29 (1.00)

The heat transfer mechanism from the fuel to the coolant becomes film boilins at what DNBR?

a. 0.0

b. 1.0

c. 1.3

d. 1.7 QUESTION 5.30 (1.00)

The reactor is at 80% power with a core delat T of 48 degrees F and a cass flow rate of 100% when a station blackout occurs. Natural circulation is established and core delta T soes to 40 degrees F. If decay heat is approximately 2% of full power, what is the mass flow rate (% of full flow)?

a. 1.9%

b. 2.1%

c. 2.4%

d. 3.0%

QUESTION 5.31 (1.00)

Which of the followins statements is correct if the discharge valve from a centrifugal pump is being partially closed from the full open position?

a. Pump head decreases as head loss decreases.

b. Pump head increases as head loss increases.

c. Volume flow rate increases as head loss decreases.

d. Volume flow rate decreases as head loss decreases.

(zuxxx CATEGORY 05 CONTINUED ON NEXT PAGE xxxxx) e

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 12 QUESTION 5.32 (1.00)

During a plant heatop with the pressurizer pressure at_1000 psis, failure of the air supply solenoids allows the PZR PORV to open slightly to a throttling position. The maximum pressure reached downstream of the valve is approximately the PRT pressure of 5 psis. What would be the condition of the fluid downstream of the valve?

a. Superheated steam

b. Dry saturated steam

c. Wet steam

d. Subcooled liquid QUESTION 5.33 (1.00)

Which of the following correctly sequences the heat transfer mechanism for an increasing fuel cladding temperature from 110F to 2000F?

a. Single phase convection, nucleate boiling, partial film boiling, film boiling.

b. Nucleate boiling, single phase convection, partial film boiling, film boiling.

c. Partial film boiling, single phase convection, nucleate boiling, film boiling.

d. Single phase convection, partial film boiling, film boilin3' nucleate boiling.

QUESTION 5.34 (1.00)

If a centrifugal pump is operating at 1800 rpm to give 400 gel / min at a discharge head of 20 psi, what would be the discharge head if the speed of the pJmp is increased in order to deliver 800 spm?

a. 40 psi.

b. 60 psi.

c. 80 psi.

d. 160 psi.

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 13 GUESTION 5.35 (1.00)

If pressure gavses are installed at the bottom of a cylindrical tank with a radius of four feet and at the bottom of a similar tank with a radius of two feet, which of the following corectly describes the relationship between the pressure gauge readings if the water level is the same in both tanks?

a. They would read the same.

b. The larger tank reading would be twice the smaller tank reading.

c. The larger tank reading would be four times the smaller tank readin3

d. The larger tank reading would be eiSht times the smaller tank readin3+

GUESTION 5.36 (1.00)

Which of the following describes the parameter changes of the fluid that occur across a centrifugal pump?

a. Temperature INCREASES, Enthalpy INCREASES, Entropy INCREASES

b. Temperature INCREASES, Enthalpy INCREASES, Entropy CONSTANT'

c. Temperature INCREASES, Enthalpy CONSTANT, Entopy INCREASES

d. Temperature CONSTANT, Enthalpy CONSTANT, Entropy INCREASES GUESTION 5.37 (2.00)

The reactor is operating at 30% power when one RCP trips. Assuming no reactor trip or turbine load change occur, indicate whether the following parameters will INCREASE, DECREASE, or REMAIN THE SAME.

a. Flow in operating reactor coolant loops (0.5)

b. Core delta T (0.5)

c. Reactor vessel delta P (0.5)

d. Operating loop steam generator pressure (0.5)

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 14 g--------------------------------------

QUESTION 5.38 (2.00)

What four conditions must be met to ensure hot channel factor limits are not exceeded during normal operation? .

(***** END OF CATEGORY 05 *****)

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 15 QUESTION 6.01 (1.00)

Wh'ich of the followin3 Reactor Coolant System temperature instruments are used by the Reactor Protection System?

a. The pressuricer steam space temperature detectors.

b. The hot-les narrow range temperature detectors.

,c. The cold-leg wide range -temperature detectors.

d. The core exit temperature detectors.

97, QUESTION 6.02 (1 00)

A pressurizer low level signal will isolate letdown and

o. trip the pressurizer heaters.

b. start a second charging pump.

c. trip the reactor.

d. initiate tafety injection.

GUESTION 6.03 (1.00)

Which of the following conditions will result in a reactor trip?

a. Low flow on 2/3 detectors in 1/4 reactor coolant loops when >P-7 and <P-8.

b. Low flow on 2/3 detectors in 1/4 reactor coolant loops when >P-8.

c. Low flow on 1/3 detectors iri 4/4 reactor coolant loops when >P-7 and <P-8.

d. Low flow on 1/3 detectors in 4/4 reactor coolant-loops when >P-8.

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 16 QUESTION 6.04 (1.00)

The Chemical and Volume Control System removes excess Lithium via the

a. cation bed demineralizer.

b. mixed bed demineralizer.

c. BTRS demineralizers.

d. volume control tank vent to the Waste Gas System.

QUESTION 6.05 (1.00)

If the excess letdown normal seal return to the VCT is isolated, then the return flow is manually diverted to 4

c. Reactor Coolant Drain Tank.

b. Pressure Relief Tank.

c. Containment Sump.

d. Reactor Coolant Sampling System.

QUESTION 6.06 (1.00)

The makeup water to the Reactor Coolant System provided from the SSF controls is supplied by

a. the NV System positive displacement pump.

b. the train A NV System centrifugal charging pump.

c. a SSF dedicated positive displacement pump.

d. a SSF dedicated centrifugal charging pump.

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6. PLANT SYSTEMS DESIGN, CONTROL. AND INSTRUMENTATION PAGE 17 GUESTION 6.07 (1.00)

Which of the following Residual Heat Removal System components can be operated from the SSF?

a. ND-2A (NC loop 'C' to ND isolation valve).

b. ND-4B (NC loop to ND pump suction cont. isolation valve).

c. Train A ND pump.

d. Train B ND pump.

QUESTION 6.08 (1.00)

The Reactor Coolant System radiation detector (EMF-48) monitors the

a. neutron radiation level.

b. samma radiation level.

c. beta radiation level.

I d. alpha radiation level.

QUESTION 6.09 (1.00)

List the four Auxiliary Feedwater Pump suction water supplies.

GUESTION 6.10 (1.00)

List the four Auxiliary Feedwater motor driven pumps automatic start signals.

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 18 GUESTION 6.11 (1.00)

The Instrument Air after coolers are cooled by the

a. Low Pressure Service Water System.

b. Nuclear Service Water System.

c. Component Cooling Water System.

d. Recirculating Water System.

QUESTION 6.12 (1.00)

Which of the followin3 is NOT a water source for the Spent Fuel Pool?

a. FWST.

I

b. RMWST.

c. Nuclear Service Water.

d. Low Pressure Service Water.

QUESTION 6.13 (1.00)

If Data A and B failures occur in the Di3 ital Rod Position Indication System for the same rod, which of the following indication will be recieved?

a. An Urgent Alarm, a Rod Bottom LiSht, a General Warning Li3ht, and a Non-Urgent Annunciator.

b. All the indication given in choice 'a' except the Rod Bottom Light.

c. All the indication given in choice 'a' except the Non-Urgent Annunciator.

d. An Ur3ent Alarm and a General Warning Light only.

4 (xxxxx CATEGORY 06 CONTINUED ON NEXT PAGE xxxxx) ,

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

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 19 GUESTION 6.14 (1.00) is added to the ice in the Containment Ice Condenser to aid in iodine removal in the event of a LOCA.

1

c. Potassium Chromate.

b. Potassium Sodium Tartrate.

c. Sodium Tetraborate.

d. Sodium Hydroxide.

QUESTION 6.15 (1.00)

If the sequencer reset button is depressed while sequencing is in progress during a blackout,

a. no futhe- load groups will be applied.

b. the sequencer will start over again delaying the loads yet to be sequenced on.

c. all loads on will ioad shed.

d. will lock in the blackout loads as first priority over LOCA loads.

QUESTION 6.16 (1.00)

If a 125 VDC vital battery charger is lost, the 120 VAC vital instrument i loads on the associated channel will be

a. lost with no operator action.

b. picked up automatically by a spare battery charger.

c. Picked up automatically by a battery.

d. picked up automatically by another operating battery charger.

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 20 QUESTION 6.17 (1.00)

A major difference between an ion chamber and a GM detector is

a. the GM detector has a photo-multiplier tube to increase it's sensitivity

b. the ion chamber is pressurized with a gas and the GM doctector operates under a vacuum.

c. the ion chamber operates at such a low voltage that a significant number of ion pairs are lost by recombination, thereby decreasing it's sensitivity.

d. the GM detector operates at a much higher volta 3e causing gas multiplication to increase the charge collected to a value independent of the type of particle initiating it.

QUESTION 6 18 (1.00)

Which of the following statements concerning the operation of the Instrument Air (VI) system is NOT correct?

a. The BASE compressor runs continuously loading and unloading as necessary.

b. After an automatic start, the STBY 1 compressor will unload and stop after the BASE compressor is fully loaded and the 15 minute timer has timed out.

c. After an automatic start, the STBY 2 compressor must be manually stopped after the 15 minute timer has timed out.

d. At 76 psi 3 VI header pressure, Station Air is supplied as a backup air supply.

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(***** CATEGORY 06 CONTINUED ON NEXT PAGE **xxx)

6. PLANT SYSTEhs DESIGN, CONTROL, AND INSTRUMENTATION PAGE 21 QUESTION 6.19 (1.00)

Which of the following statements concerning the operation of the Containment Air Release and Addition system is correct?

a. Initiation of air release AND initiation of air addition is AUTOMATIC.

b. Initiation of air release is AUTOMATIC and initiation of air addition is MANUAL.

c. Initiation of air release is MANUAL and initiation of air addition is AUTOMATIC.

d. Initiation of air release AND initiation of air addition is MANUAL.

QUESTION 6.20 (1.00)

Which of the following statements concerning normal pressurizer spray is correct?

a. The 2 normal spray lines supply water to the pressurizer through separate spray no::les.

b. A small continuous spray is provided by manual throttle valves in parallel with the spray valves.

c. The driving force for spray flow is the height difference between the S/G and the S/G and the pressurizer.

d. An RTO downstream of each spray valve has a low temperature alarm which provides a warning of e:< c e s s i v e spray flow.

QUESTION 6.21 (1.00)

Which of the following is NOT a function of the P-4 permissive (trip and bypass breakers open)?

a. Allows bypassing the steam dump cooldown interlock.

b. Allows an operator to block an SI signal.

c. Causes feedwater isolation if a low Tavs is also present.

d. Causes a turbine trip.

(***** CATEGORY 06 CONTINUED ON NEXT PAGE xxxxx)

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 22 l QUESTION 6.22 (1.00)

Which of the following reactor trip or SI signals is NOT blocked by pornissive P-7?

a. Low P:r Pressure trip.

b. Low P=r Pressure SI.

c. High P:t Level trip.

d. Low RCS Flow trip.

QUESTION 6.23 (1.00)

Which of the following expresses the combined error signal used by the Reactor Control System to generate rod motion?

a. (Impulse Pressure - Nuclear Power) + (Tref - Tavs)

b. (Nuclear Power - Impulse Pressure) + (Tref - Tavg)

c. (Nuclear Power - Impulse Pressure) + (Tavs - Tref)

d. (Impulse Pressure - Nuclear Power) + (Tavs - Tref)

GUESTION 6.24 (1.00)

Which of the following does NOT automatically occur when a Feedwater Isolation signal is initiated?

a. CF pump discharge valves close.

b. CF pump suction valves close.

c. Tempering valves close.

d. CF bypass valves to CA no::les close.

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUNENTATION PAGE 23 QUESTION 6.25 (1.00)

Which of the following methods is used to remove the samma signal from the neutron signal in the intermediate range monitor?

a. The outer chamber prevents sammas from ionizing the inner chamber.

b. The inner chamber current cancels out the gamma current from the outer chamber.

c. A pulse height discriminator does not allow the Samma signals to be counted.

d. A lead shield lines the detector well that prevents gammas to enter the detector volume while allowing neutrons to pass.

QUESTION 6.26 (1.00)

Which of the following signals cause BOTH a safety injection and a osin steamline isolation?

a. Low Steamline Pressure.

b. High Steam Pressure Rate.

c. High Containment Pressure.

d. Low Pressurizer Pressure.

(***** CATEGORY 06 CONTINUED ON NEXT PAGE ***x*)

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 24 QUESTION 6.27 (1.00)

Which statement below regarding Auxiliary Feedwater System control is correct?

a. It is not possible to auto-start the motor-driven Auxiliary Feed pumps while they are in local control.

b. The dischar3e valves for the motor-driven Auxiliary Feed pumps are electric motor operated and will fail open to 55% on an auto-start sequence.

c. If there is a loss of demineralized water for the Auxiliary Feed pumps and 2 out of 3 low suction pressure from Train B activates, then suction valves from the RC (Condenser Cire Water) and RN (Nuclear Service Water) systems will open.

d. To minimize Auxiliary Feed pump runout, the motor-driven isolation valves on the motor-driven pump supply lines to B and C SGs will close automatically if a high flow condition occurs.

QUESTION 6.28 (1.00)

With the Rod Control System in automatic, which of the following conditions would result in a rod withdrawal?

a. Nuclear power channel (N-44) fails high.

b. Teold (Ch. A) detector fails low.

c. Impulse pressure (Ch. 1) fails high.

d. Tref fails low.

QUESTION 6.29 (1.00)

Which of the following will cause the automatic closure of the UHI

discharge isolation valves?

a. There is no automatic closure signal.

b. Low nitrogen pressure in the surge tank.

c. Low water level in the accumulator,

d. Membrane leakage alarm.

(***** CATEGORY 06 CONTINUED ON NEXT PAGE xx***)

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 25 GUESTION 6.30 (1.00)

Which of the following correctly describes the turbine runback asso-ciated with the loss of a MFPT?

a. If > 80% load, trip of either NFPT will cause a runback at 133%/ minute until < 70% load.

b. If >80% load, trip of either MFPT will cause a runback at 10%/ minute until < 70% load.

c. If > 70% load, trip of either NFPT will cause a runback at 10%/ minute until < 70% load.

d. If > 70% load, trip of either NFPT will cause a runback at 133%/ minute until < 50% load.

QUESTION 6.31 (1.00)

The purpose of the flow orifice in the cold les accumulator (CLA) discharge line is given by which of the following?

a. To minimize the hydraulic resistance to flow from the CLA.

b. To extend the blowdown time of the CLA during a LOCA.

c. To provide a method to measure the flow from the CLA during a LOCA.

d. To provide a method to measure CLA check valve leakage.

QUESTION 6.32 (1.00)

The #3 NC pump seal leakoff is normally collected in which of the following?

a. Containment Sump.

b. Pr.essurizer Relief Tank.

c. Volume Control Tank.

d. Reactor Coolant Drain Tank.

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 26 QUESTION 6.33 (1.00)

The Standby Shutdown Facility contains systems designed to handle which of the following major events?

o. Rupture of dans impounding Lake Wylie and the service water pond.

b. A major fire disabling the Control Room controls and the Auxiliary Shutdown Panel controls.

c. The Control Room and the Technical Support Center are unavailable.

d. A LOCA that requires evacuation of the plant site for more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

QUESTION 6.34 (1.50)

Indicate whether the following connections to the Main Steam (SM) cystem tap off UPSTREAM or DOWNSTREAM of the main steam isolation valves.

a. Atmospheric steam dump valves. (0.5)

b. Atmospheric PORVs. (0.5)

c. Main FWPT steam supplies. (0.5) j QUESTION 6.35 (1.00)

Which of the following statements concernin3 the automatic closure of I

the component cooling (KC) thermal barrier return isolation valves is correct?

I

c. High pressure in a return line will automatically shut it's return-isolation valve.

b. High temperature in a return line will automatically shut it's return isolation valve.

c. High flow in a return line will automatically shut it's return isolation valve.

d. High activity in a return line will automatically shut it's return isolation valve.

(***** CATEGORY 06 CONTINUED ON NEXT PAGE xxxxx) i l .,.

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6. PLANT SYSTEMS DESIGNr CONTROL, AND INSTRUMENTATION PAGE 27 GUESTION 6 36 (1.00)

List the High-High SG 1evel setpoints for Feedwater Isolation, Turbine Trip (P-14 interlock) for (a) Unit 1 and (b) Unit 2.

QUESTION 6.37 (1 00)

The difference between the Unit i and Unit 2 SG programmed level setpoint in

c. Unit 1 is constant and Unit 2 is ramped with power.

b. Unit 2 is constant and Unit 1 is ramped with power.

c. Unit 2 is ramped with power from 38% to 66% and Unit 1 is ramped with power from 30% to 60%.

d. Unit 1 is camped with power from 38% to 66% and Unit 2 is ramped with power from 30% to 50%.

QUESTION 6.38 (1.00)

Which of the following is a Main Turbine Generator Stator Coolin3 Runback that is applicable to Unit 2 only?

a. Stator Coolant High Temperature,

b. Stator Coolant Low Pressure.

c. Stator Coolant Low Flow.

d. Stator Coolant High Conductivity.

GUESTION 6.39 (1.00)

Which of the following tanks is shared between both Units?

a. The Auxiliary Feedwater Condensate Storage Tank.

b. The Upper Surge Tank.

c. The Demineralized Water Storage Tank.

d, The Fueling Water Storage Tank.

(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 'u8 GUESTION 6.40 ( .50)

Both Unit 1 and Unit 2 Nuclear Service Water Pumps start on a Safety Injection Signal on either Unit. TRUE or FALSE?

(***** END OF CATEGORY 06 ramax) l 1

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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 29

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~~~~E3656E655EIL 56ET55L QUESTION 7.01 (1 00)

A reactor coolant pump should not be started unless the No. 1 seal dolta P is greater than

a. 100 psi,

b. 200 psi.

c. 300 psi.

d. 400 psi.

QUESTION 7.02 (1.00)

Prior to transferring rod control from MANUAL to AUTO during a reactor startup, Tave and Tref should be within

a. 1 degree F.

b. 3.5 de3rees F.

c. 5 degrees F.

d. 10 degrees F.

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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 30

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R 555L5555 L"55 TR5L GUESTION 7.03 (1.00)

Which of the following statements is correct concerning the status of the Nuclear Instrumentation Recorder prior to withdrawing control bcnk rods for a reactor startup?

4

c. The highest reading source range channel and the highest reading intermediate range channel is selected and the NR-45 chart speed is set to 'Hi' speed.

b. The highest reading source range channel and the highest reading l

intermediate range channel is selected and the NR-45 chart speed is set to "Lo' speed.

c. The highest reading source range channel and the lowest reading intermediate range channel is selected and the NR-45 chart speed is set to 'Hi' speed.

d. The highest reading source range channel and the lowest reading intermediate range channel is selected and the NR-45 chart speed is set to 'Lo* speed.

QUESTION 7.04 (1.00)

The transfer of control of the pressurizer heaters (sub-bank of

] Group D) to the Safe Shutdown Facility is performed by i a. swapping power supplies of the motor control centers from rior mal to alternate in the ETA rooms.

b. swapping a plus type connector from the normal connection to the alternate connection in the SSF.

c. placing the Pressurizer Heater Selector Switch in the ma,in. control room to the ' Remote-SSF' position.

d. Placing the Pressurizer Heater Selector Switch in the SSF

, to the ' Local' position.

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i 7' . PROCEDURES - NORMAL, . ABNORMAL, EMERGENCY AND PAGE 31 1 ~~~~E365EL561E L'EENTE3L"~ " " ~ " " " " ""'

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! QUESTION 7.05 (1.00)

The CRDM cooling fans should NOT be secured following a reactor shutdown d and cooldown until the reactor coolant temperature is less than

! a. 400F.

1

b. 350F.

c. 200F.

d. 160F.

QUESTION 7.06 (1.00) i KC flow through the NCP seal water heat exchanger is assured by verifying

a. proper flow indication.

b. proper pressure indication.

i

c. at least one KC pump is operating.

! d. no seal water heat exchanger KC Hi/Lo flow alarm condition exists.

QUESTION 7.07 (1.00) i How many temporary submersible neutron detectors are used during the 3

initial fuel loading?

I

a. One.

! b. Two.

j c. Three.

d. Four.

I i

i GUESTION 7.08 ( .50)

The temporary neutron detectors used during fuel loading must be equipped with scalar / timer units. TRUE or FALSE?

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~~~~k 656L65EE L EUETEUL QUESTION 7.09 (1.00)

On a power increase from 15% to 75% power. the following actions are performed.

1. Start a second CF pump

2. Verify SSRH highload valves automatically open

3. Check OAC thermal output measurement and have IRE adjust nuclear instrumentation as required.

4. Verify 'P9 - Reactor Trip On Turbine Trip Permissive' light Comes on.

Which of the below is the correct performance order of these actions?

a. 1,3,4,2

b. 1,3,2,4

c. 3,1,4,2

d. 3,1,2,4 QUESTION 7.10 (1.00)

If a ' Rod Control Urgent Failure' alarm occurs due to a failure in the I J3 i c cabinet, the Tave/ Tref mismatch is maintained by

a. controlling turbine load.

b. taking manual control of individual control rod banks.

c. taking manual control of individual control rod groups.

d. boration and dilution of the reactor coolant system.

(***** CATEGORY 07 CONTINUED ON NEXT PAGE xxxxx)

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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 33

~~~~

R BiBE55153E C5sTR5t QUESTION 7.11 (1.00)

The trip bistables of a failed power range detector are placed in the trip condition by

a. placing the applicable bistable test switch in the ' Test'

- position in the Reactor Protection Cabinet.

b. removing the applicable control and instrument power fuses on the power range drawers.

c. placing the applicable Power Mismatch Bypass switch to the failed position at the Miscellaneous Control and Indication Panel.

d. placing the applicable Comparator Channel Defeat switch to the failed channel position at the Detector Current Comparator Panel.

QUESTION 7 12 (1.00)

If a high activity in the reactor coolant has been identified as a crud burst, the corrective action is to

a. reduce reactor power.

b. isolate the mixed bed demineralizer from service and place the cation bed demineralizer in service.

c. isolate the mined bed demineralizer from service and open it's bypass valve.

d. increase letdown flow to 120 spm.

i (xxxxx CATEGORY 07 CONTINUED ON NEXT PAGE xxxxx)

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

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 34

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~~~~Rd656E655CIL C5NTR5t QUESTION 7.13 (1.00)

After loading of the initial eight fuel assemblies, core loading shall be suspended immediately if an unexpected increase in count rate by a factor of ------

occurs on any one of the responding nuclear monitorin3 channels.

a. two

b. five

c. ten

d. one hundred QUESTION 7.14 (1.00)

While operating Unit 1 at 85% power, an inadvertent boron dilution occurs uhich results in a ' Control Rod Bank Lo-Lo Limit' alarm. Which of the following is an appropriate immediate operator action in accordance with AP/1/A/5500/13, ' Boron Dilution'?

c. Emergency borate to cica' the ' Control Bank Lo Limit' alram.

b. Emer3ency borate to clear the Control Rod Bank Lo-Lo Limit' alarm.

c. Stop the source of the dilution and obtain a chemistry sample to determine the required boration.

d. Stop the dilution process and emergency borate to raise the rods to the position they occupied prior to the inadvertent dilution.

QUESTION 7.15 (1.00)

AP/1/A/5500/04, ' Loss of Reactor Coolant Pumps', contains a CAUTION that states that reactor power should be reduced to below a certain power level prior to restarting the affected NC pump (s). Which of the following is this power level?

a. 45%

b. 35%

c. 25%

d. 15%

(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 35

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~~~~k biUL655E L EUNTRUL QUESTION 7.16 (1.00)

In the event of a toxic gas release in the Control Room and it is dacided to evacuate the Control Room, which of the following is the

, correct actions of the NCO?

a. Take over as OATC and perform required immediate and subsequent actions.

b. Go to ASP and initiate performance of ASP operator actions.

c. Go to AFWTCP and initiate performance of AFWPTCP operator actions.

d. Go to HVAC ECP's and initiate performance of HVAC ECP actions.

QUESTION 7.17 (1.00)

If KC and NV seal cooling for any NC pump is lost, within how many cinutes AND from where must seal injection flow be initiated?

OUESTION 7.18 (1.50)

List ALL immediate operator actions that are to be taken in the event of damaged spent fuel in the reactor building.

I QUESTION 7.19 (1.00)

An immediate operator 3ction for a loss of ND train is to verify ND suction supply adequate. Which of the followin3 is the criteria for performing this verification?

a. PZR level.> 10% or NC vessel level > 10%.

b. FWST level > 50% or Containment sump level > 10%.

c. ND suction pressure > 200 psis.

d. ND discharge pressure > 200 psis.

GL'ES TION 7.20 (1.25)

List the immediate operator actions for a reactor trip.

(***** CATEGORY 07 CONTINUED ON NEXT PAGE ** mum)

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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 36

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~~~~R d656L66fCKt E6 TR6L GUESTION 7.21 ( .75) ,

List the immediate action steps for a turbine trip when the !

osin control room manual trip does not trip the turbine.

QUESTION 7.22 (1.00)

List the immediate actions required when the reactor cannot be tripped using the control room trip switches.

QUESTION 7.23 (1.00) i Which of the followin3 conditions would require all NC pumps to bo tripped during a valid safety injection?

! o. If an ND pump is running and the NC system subcooling is 10 degrees F.

b. If an NI pump is running and the NC system subcooling is 10 degrees F.

i

c. If flow is verified from an auxiliary feed water pump and NC system subcooling is less than 0 degrees F. ,

d. If flow is verified from an NV pump and NC system subcoolin3 is less than 0 degrees F. ,

0 QUESTION 7.24 (1.00) i

! Which of the following conditions satisfy the SI termination criteria of the Safety Injection procedure EP-01?

NC System Subcooling PZR Level NC System Pressure SG Total Flow

a. 25 degrees F 10% Stable 500 spa ,

l b. 60 degrees F 7% Increasing 550 apa

c. 50 degrees F 7% Stable 400 3pm

d. 50 degrees F 10% Increasing 350 spa i,

(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

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QUESTION 7.25 (1.00)

Which of the following is a characteristic of natural cireviation?

! c. SG steam pressure-STABLE or DECREASING.

b. NC system subcooling-LESS THAN O DEGREES F.

c. NC loop hot les temperatures-STABLE or INCREASING.

d. NC loop cold les temperatures-GREATER TNAN SATURATION TEMPERATURE FOR SG F tESSURE.

OUESTION 7.26 (1 00)

During a natural cireviation cooldown, the preferred method of NC cystem depressuri=ation is by

c. opening the normal spray valve.

b. opening the NV auxiliary spray valve.

> c. opening the pressuri=er PORVs.

de opening the reactor vessel vent valves.

QUESTION 7.27 (1.00)

If the following colors appeared on the control room critical cafety function status tree display, which color coded path

]

hos the highest priority?

a. RED

b. ORANGE

c. YELLOW

d. GREEN f

1 .

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~~~~EI616E665C [~66NTR6L GUESTION 7.28 (1 00)

If the following critical safety functions were all displayed orange, which one has priority?

o. Suberiticality.

b. Heat Sink.

c. Integrity.

d. Inventory.

QUESTION 7.29 (1.00)

A hydrogen bubble formed in the reactor vessel is eliminated by

c. increasing pressurizer temperature above core thermocouple readings.

b. injecting oxygen into the reactor coolant system via the chemical and volume control system.

c. maximizing coolant flow by running all reactor coolant pumps, increasing letdown flow to 120 spm, and placing the cation bed deminerali:er in service in parrallel with the mixed bed demineralizer.

d. venting the reactor vessel head to the PRT.

QUESTION 7.30 (1.50)

If a reactor trip occurs and the ' Safety Injection Actuated

light is NOT lit, what conditions must the operator immediately verify?

(***** CATEGORY 07 CONTINUED ON NEXT PAGE anums)

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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 39

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R 65bl655E5L E5NT55L QUESTION 7.31 (1.00)

Prior to commencing a Natural Circulation Ccoldown per EP/1/A/50/1A1, it is necessary to borate the NC system to 100 ppm greater than that required to provide a Xenon free Shutdown Margin (SDM) of 1300 pcm at the minimum expected cold shutdown temperature. Which of the follow-ins is the reason why the shutdown baron concentration should be 100 ppm greater than necessary to achieve the 1300 pcm SDM?

c. Due to improper NC system mixing with natural circulation flow.

b. Due to possible errors in NC system temperature indications with natural cireviation flow.

c. Due to changes in baron worth with natural circulation flow.

d. Due to restricted reactor vessel head cooling with natural circulation ficw.

QUESTION 7.32 (1.00)

Durins a reactor coolant leak identification process, which of the

following sives the requirement for shifting centrifugal charging pump suction from the VCT to the FWST?

a. Pressurizer level decreases to 17%.

b. VCT level decreases to 20%.

c. Letdown isolated and pressurizer level decreasins.

d. Prior to stating a second centrifusal charging pump.

QUESTION 7.33 (1.00)

On a steam senerator overpressuri=ation casualty, which of the followins 4

is the correct priority (in order of highest to lowest) of the methods used to depressuri e the affected S/G?

a. MSIV bypass, PORV, MSIV, Turbine driven CA pump

b. PORV, MSIV bypass, MSIV, Turbine driven CA pump

c. MSIV bypass, PORV, Turbine driven CA pump, MSIV

d. PORV, MSIV bypass, Turbine driven CA pump, MSIV (xxxxx CATEGORY 07 CONTINUED ON NEXT PAGE *****)

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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 40

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~~~~656f6L66fCdL 66 TR6L QUESTION 7.34 (1.50) ,

List ALL the immediate operator actions for a loss of offsite power.

QUESTION 7.35 (1.00)

If a 500 arem pocket dosimeter reads ______ or more, it shouldtde returned to HP for re:eroing. ,

s. 120 mtem ,

! b. 200 stem

c. 250 mrem

d. 300 mren QUESTION 7.36 (1.00)

If you recieve no whole body radiation exposure for'the first seven weeks of a calendar quarter, then you may recieve up to ______

durin3 the eight week without exceeding the normal whole body 2

administrative exposure limit. ,

a. 100 mrem

b. 500 mren

c. 800 mrem ,

d. 1000 mrem 2

QUESTION 7.37 (1.00)

Which of the following is a 10 CFR ?0' exposure limit? ;

a. 5 ' rem / year-whole body.

b. 1 rem / quarter-whole body.

c. 18.75 rem / quarter-hands.

d. 7 rem / quarter-skin of whole body.

(xxxxx CATEGORY 07 CONTINUE 0 ON NEXT PAGE xxxxx) t 3 e b

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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 41

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~~~~E 655L6Eid L 5ENTEUL l

QUESTION 7.38 (1.00)

Which of the following radiation exposures would inflict the greatest biological damage to man?

a. 1 Ren of GAMMA.

b. 1 Rem of ALPHA.

c. 1 Rem of NEUTRON.

d. NONE of the above; they are all equivalent.

QUESTION 7.39 (1.00)

When an individual has received ______ of an administrative exposure limit, he/she must inform his/her supervisor.

c. 75%

b. 80%

c. 85%

d. 90%

i t

(xxxxx END OF CATEGORY 07 xxxxx)

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 42

QUESTION 8 01 (1.00) ,

Match the following Tech Spec defined leakage with the associated loak rate limit. Assume normal operating reactor coolant system tamperature and pressure. The limits may be used more than once.

COLUMN A COLUMN B

1. Pressure boundary leakage. a.'O spm.

2. Controlled leakage. b. 1 spm.

3. Identified leakage. c. 10 spm.

4. Unidentified leakage. d. 40 spm.

5. Primary-to-secondary e. 500 spd.

leakage throu3h one SG.

QUESTION 8.02 (1.00)

To prevent entering a Technical Specification action statement, the Quadrant Power Tilt Ratio shall not exceed ______ when reactor !

power is above 50%.

a. 1.00 ,

b. 1.02

c. 1.05

d. 1.09 QUESTION 8.03 (1.00)

If the reactor coolant system pressure exceeds 2735 psig when in Mode 3, Technical Specifications requires the pressure to be reduced to within the limit within

! a. 5 minutes.

b. 15 minutes.

c. 30 minutes.

d. one hour.

(**x** CATEGORY 08 CONTINUED ON NEXT PAGE **xxx)

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 43

^

QUESTION 8.04 (1.00)

If control power is lost to a pressurizer power operated relief velve while in Mode 1,

e. no action is required by Tech Specs provided another PORV is operable and all pressurizer code safety valves are operable.

b. Tech Specs' require the the power supply to be removed from the associated block valve after verifying it to be open, if the PORV is not-made operable within one hour and continuous operation is desirable.

c. Tech Specs require the associated block valve to be shut and its' power removed if the PORV is not made operable within one hour and continuous operation is desirable.

d. Tech Specs require action to be initiated within one hour to place the plant in at least HOT STANDBY within the fc11owing hour if the PORV is not made operable.

QUESTION 8.05 ( .50)

Toch Specs does NOT require any action if one control rod is icmovable, provided the immovable rod is within 12 steps of its group step counter demand position. TRUE or FALSE?

l QUESTION 8.06 ( .50)

The shutdown bank demand position indication CANNOT be used as the one required Rod Position Indication system when in Mode 3 with the shutdown banks withdrawn. TRUE or FALSE?

(xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx) i 9

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 44 GUESTION 8.07 (1.00)

If the lowest operating loop Tave drops below -_____

, Tech Specs allow ______ to restore Tave within the limit or proceed to place the unit in hot standby.

a. 557 de3rees F, 30 minutes

b. 557 degrees F, 15 minutes

c. 551 degrees F, 30 minutes

d. 551 degrees F, 15 minutes QUESTION 8.08 (1.00)

If one overtemperature delta T instrument channel is inoperable and placed in the trip condition, then

a. the plant must be shutdown when the survie11ance testing for the operable channels become due, if it is not restored to an operable status,

b. the operable channels must be tested one at a time to prevent a reactor trip from occurring.

c. the inoperable channel may be bypassed for a limited time in order to perform the survie11ance testing on the operable channels.

d. the surviellarice test interval may be extended up to 100% to allow time to restore the inoperable instrument to an operable status.

(xxurr CATEGORY 08 CONTINUED ON NEXT PAGE *****)

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 45 QUESTION 8.09 (1.00)

Turbine overspeed protection is required by Technical Specifications because an overspeed condition could cause

a. turbine components to become missles which may penetrate the turbine casing, allowing higher than allowable off-site doses with an assumed maximum allowable fuel cladding and primary-to-secondary leakage.

b. turbine components to'become missiles, which may damage safety-related equipment.

c. the reactor thermal power to exceed the limits in the Unit's License.

d. the reactor to 30 Pro *Pt critical.

QUESTION 8.10 (1.00)

Indicate whether each of the following statements is TRUE or FALSE.

a. When in Mode 6, Tech Specs requires the Source Range Neutron Flux Monitors to have their Alarm Setpoints at 0.5 decade above the steady-state count rate, to perform Core Alterations.

b. When in Mode 6, Tech Specs requires only visual indication of neutron flux from the Source Range Monitors in the Control Room, and only audible Source Range indication in the Containment, to perform Core Alterations.

QUESTION 8.11 (1.00)

If the Containment Purge System Noble Gas Activity Monitor (EMF-39) is declared inoperable, then

~

a. no-pursin3 e- ' - t i..5 can be done via this pathway.

b. two independent containment air samples must be analy=ed and two qualified people must verify the purse lineup to allow purging via this pathway.

c. the flow rate must be verified at least once every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to allow purging via this pathway.

d. grab samples must be taken and analy=ed at least once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to allow purging via this pathway.

(xxxxx CATEGORY 08 CONTINUED ON-NEXT PAGE xxxxx)

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 46 N

OUESTION 8.12 (1.00)

A quarterly surveillance requirement of Tech Specs may be extended up to ______ days without declaring the component inoperable due to the surveillance testing not being performed.

a. 9

b. 23

c. 32

d. 41 QUESTION 8.13 (1 00)

Provide the minimum number of individuals required by Tech Specs for the follswing positions to operate both Units at full power.

c. ______ Shift Supervisor (s).

.b. ______ SR0(s) (not including the Shift Supervisor (s)).

c. ______ R0(s).

d. ______ NEO(s).

e. ______

STA(s).

QUESTION- 8.14 (1.00)

Leakage at the high pressure seal at the incore instrument seal table is defined in Tech Specs as

a. Unidentified Leakage.

b. Uncontrolled Leakage.

c. Identified Leakage.

d. Pressure Boundary Leakage.

(xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE *****)

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 47 QUESTION 8.15 (1.50)

List the requirements to satisfy the definition of Reactor Building Integrity.

QUESTION 8.16 (1.00)

The difference between Cold Shutdown and Refueling Modes is

o. the thermal power level from decay heat.

b. the average coolant temperature.

c. the value of keff.

d. the status of the vessel head closure bolts.

QUESTION 8.17 (1.00)

If the-high setpoint flux trip setpoint is measured and determined to be

]

between the Tech Spec trip setpoint and allowable value, Tech Specs

m. still allows operations to continue,

b. requires a shutdown within one hour,

c. requires the plant to be in a cold shutdown condition within one hour.

d. requires an immediate manual reactor trip.

QUESTION 8.18 ( .50)

Tech Spec Surveillance Requirements do not have to be performed on inoperable equipment. TRUE or FALSE?

QUESTION 8.19 (1.50)

List the three acceptable baron injection flow paths that may be used to meet the LCO on operable boron injection flow paths for Mode 1.

(xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx) l l

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 48 GUESTION 8.20 (1.00)

The indicated AFD is considered to be outside its target band as soon as

______ operable excore channel (s) is/are indicating outside the target br;nd.

a. one

b. two

c. three

d. four QUESTION 8.21 (1.00)

What is the bases for all MSIVs to be operable in Made 1?

o. It ensures only one SG will blowdown in the event of a steam line rupture.

b. It minimi=es the metallurgical effects associated with a steam line rupture cooldown on the reactor coolant system.

c. It ensures that SG transients from a steam line rupture do not exceed the maximum allowable pressure limit.

d. It ensures that we do not rely on the SG safety relief valves to relieve pressure normally relieved throvsh the MSIVs.

QUESTION 8.22 (1.00)

Which of the following is NOT required to meet the operability requirements of Tech Specs for an Emergency Diesel Generator?

a. A day tank containing a minimum required fuel volume.

b. A Fuel Storage System containing a minimum required fuel volume.

c. A 125 VDC battery and charger connected to the diesel generator control loads.

d. An operable fuel transfer pump.

(xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx)

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 49 QUESTION 8.23 ( .50)

The reactor is required by Tech Specs to be suberitical for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to movement of irradiated fuel in the reactor vessel.

TRUE or FALSE?

QUESTION 8.24 (1.00)

The concentration limits of radioactive material released in liquid Offluents to unrestricted areas are found in

a. Tech Specs. .i

b. 10 CFR 20.

c. 10 CFR 50.

d. 10 CFR 55.

QUESTION 8.25 (1.00)

Which of the following is the responsibility of the Shift Supervisor in regard to the use of operatins Procedures performed during the shift?

a. Ensures the Working Copy File has a sufficient number of unused copies available.

b. Initiates all procedure changes required during the shift.

c. Reviews and approves all completed procedures.

d. Compares the Working Copy to the Control Copy to ensure all changes are entered prier to use.

(xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx)

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 50 GUESTION 8.26 (1.00)

The ' Buddy System' for entering the Reactor Building is required as soon as

a. the Reactor Coolant System temperature exceeds 200 degrees F.

b. the Reactor Coolant System temperature exceeds 350 degrees F.

c. the reactor power level exceeds 5%.

d. the reactor power level exceeds 10%.

QUESTION 8.27 (1.00)

When transferrin 3 initials from a performed part of a procedure to a rotype of the procedure, the signature required indicating the initials have been transferred is that of

a. the operator initia11in3 the original procedure.

b. a supervisor.

c. any on shift licensed operator.

d. any on shift SRO licer, sed oper ator , but not an R0 licensed operator.

QUESTION 8.28 (1.00)

To deviate from the sequence of steps of an operating procedure, the operator must get verbal approval from

a. any two licensed operators.

b. any two licensed operators, of which one must be SRO licensed.

c. any two licensed operators, of which one must be an SRO licensed supervisor.

d. any two SRO licensed operatcrs.

(xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx)

l I

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 51 '

GUESTION 8.29 (1.00)

A Procedure Discrepancy Form, when used, is normally filled out for t

a. Periodic Test Procedures.

b. Operating Procedures.

c. Abnormal Operating Procedures.

d. Emer3ency Operatin3 Procedures.

QUESTION 8.30 (1 00)

When temporary jumpers are installed in a cabinet, the cabinet identifies this by an attached

a. orange tag.

b. white tas..

c. yellow tag.

d. brown tag.

QUESTION 8.31 (1.00)

When required to shutdown and cooldown the plant due to exceeding a Tech Spec ' ALLOWABLE VALUE', it must be done at a minimum rate of

a. 10%/ MIN and 10F/HR.

b. 10%/HR and 10F/ MIN.

c. 10%/HR and 10F/HR.

d. 10%/ MIN and 10F/ MIN.

(xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx)

-, _ - - - - - g , ,,

i

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 52 l QUESTION 8.32 (1.00)

Safety ta3 stubs are not removed until the !

a., work supervisor signs for issue.

b. shift supervisor signs for issue approval.

c. safety tag is affixed to the required operatins device.

d. safety tas is removed from the operatins device for clearance.

QUESTION 8.33 (1.00)

What is the maximum amount of consecutive time that the Shift Supervisor can.be relieved by an eligible STA?

a. 10 minutes.

b. 15 minutes.

c. 30 minutes.

d. one hour.

QUESTION 8.34 (1.00)

Prior to 8 a.m. each Monday, the Shift Supervisor will collect all completed Round Sheets and route them to the

a. Station Manager.

b. Superintendent of Operations.

, c. Shift Operating Engineer.

d. Performance Engineer.

(xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx) k

)

l

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 53

. QUESTION 8.35 (1.00)

Fuel handling interlocks may be bypassed when not directed by an approved written procedure if approval is given by

a. any two licensed operators, provided one is an SRO.

b. any two licensed operators, provided one is the Fuel Handling Supervisor

c. any two Operations Group Supervisors, provided one is the Superintendent of Operations.

d. any two Operations Group Supervisors, provided one is the Shift Supervisor.

QUESTION 8.36 (1.00)

It is permissible to briefly exceed the ' full steady state licensed power level

by as much as ______ for as long as ______.

a. 2%, 15 minutes

b. 2%, 30 minutes

c. 3%, 5 minutes

d. 5%. 2 minutes GUESTION 8.37 (1.00)

When the Station Manager, Operating Superintendent, Technical Services Superintendent and Maintenance Superintendent are all absent from the station

a. the person filling in for the Station Mana3er shall approve all procedures and procedure changes.

b. the Shift Supervisor shall approve all procedure and procedure changes.

c. the person filling in for the Operating Superintendent shall approve procedures and procedure changes.

d. then'no approvals of procedures or procedure changes can be done.

(xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx)

- - - e ,

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 54 GUESTION 8.38 (1.00)

List the two Evacuation Sites.

QUESTION 8.39 (1.00)

The highest allowable planned whole body exposure that may be received, with appropriate approval, by a volunteer to save a life is

a. 25 rems.

b. 75 rems.

c. 150 rems.

d. 375 rens.

- QUESTION 8.40 (1.00)

During a Site Area Emer3ency, unaccounted personnel will be searched for by

a. the responsible supervisor.

b. Health Physics.

c. the Fire Brigade.

d. Security.

QUESTION 8.41 (1.00)

The Evacuation Coordinator is the individual in charge at the ______,

during a Site Evacuation.

a. Site Assembly Point

b. TSC

c. Charlotte Corporate Office

d. Evacuation-Relocation Site (xxxxx END OF CATEGORY 08 xxxxx)

(xxxxxx3xxxxxx END OF EXAMINATION xxxxxxxxxxxxxxx) i i

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EQUATION SHEET Where og = m2

( dens t ly) g ( vel oci ty) g ( a rea )1 = ( dens i ty ) 2 ( vel oc i ty ) 2 ( a re a )2 KE = mv2 PE = mgh PEg + keg +PgVg = PE2+KE 2+P Y22 where V = specific

~lf volume ,

P = Pres)ure Q = icp(Tout.Tj,) Q = UA (T,y,-Tsta) Q = ilhg.h2 I P = P,10(SUR)(t) p . p ,t/T SUR = 26.06 T = (B.p)t I P delta K = (K,gg-1)/K,gg CRg(1.K,ggi) = CR 2 (I-Keff2) CR = S/(1.K ,gf)

M = (1-K,ggi) SDM = (1.X,gg) x 100%

K (1.Keff2) eff decay constant = In (2) = 0.693 A = Ag e-(decay constant)x(t) t t 1/2 1/2 Water Parameters Miscellaneous Conversions 1 gallon = 8.345 lbs 1 Curie = 3.7 x 1010 dps 1 gallon = 3.78 liters 1 kg = 2.21 lbs 1 ft3 = 7.48 gallons 1 hp = 2.54 x 103 Btu /hr Density = 62.4 lbg/f t 3 1 MW = 3.41 x 106 Btu /hr Density = 1 gn/cm 1 Btu = 778 f t-lbf ,

Heat of Vaporization = 970 Btu /lbe Degrees F = (1.8 x Degrees C) + 32 '

Heat of Fusion = 144 Btu /lba 1 inch = 2.54 centimeters 2

1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 f t-lbm/lbf-se: I 5

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TABLE D Ib '

Properties of Dry Saturated Steam (mminuedJ Temperature E,rithalp> Entrop>

Spec &e volume l

$at $ar $at

F P""' $at $at Sat. Essp Evap ,,,,,

psia vapor houid vapor hquid hquid h h s s s, hr a e si a u a y vi 3306 0.00 1075.8 1075.8 0.0000 2.1877 2.1877 32 0.08854 0.01602 2.1709 2.1770 2947 3.02 1074.1 1077.1 0.006 t 35 0.09995 0.01602 2.1435 2.1597 2444 8.05 1071.3 1079.3 0 0162 40 0.12.170 0.01602 2.1429 20 % 4 13.% 1068.4 1081.5 0 0262 2.1167 45 0.14752 0 01602 2.1264 1703.2 18.07 1065.6 1083.7 0.0361 2.0903 50 0.17811 0 01603 28.06 1059.9 1088.0 0.0555 2.0393 2.0948 60 0.2563 0.01604 1206.7 38.04 1054.3 1092.3 0 0745 1.9902 2.0647 70 0.3631 0.01606 867.9 48.02 1048.6 1096.6 0.0932 1.9428 2.0360 80 0.5069 0.01608 633.1 57.99 1042.9 1100.9 0.1115 1.8972 2.0087 90 0.6982 0.01610 468.0 67.97 1037.2 1105.2 0.1295 1.8531 1.9826 100 0.9492 0.01613 350.4 77.94 1031.6 1109.5 0.14I7 1.8106 1.9577 110 1.2748 0.01617 265.4 87.92 1025.8 1113.7 0.1645 1.7694 1.9339 120 1.6924 0.01620 203 27 97.90 1020.0 t il7.9 0.1816 1.7296 1.9112 130 2.2225 0.01625 157.34 1014.1 1122.0 0.1984 1.6910 1.8894 2.8886 0 01629 123.01 107.89 140 1.6537 1.8685 0.01634 97.07 117.89 1008.2 1126.1 0.2149 150 3.718 127.89 1002.3 1130.2 0.2311 l.6174 1.8485 160 4.741 0.01639 77.29 137.90 996.3 1134.2 0.2472 1.5822 f.8293 170 5.992 0.01645 62.06 147.92 990.2 1138.1 0.2630 1.5480 1.81M 180 7.510 0 0I651 50.23 984.1 1142.0 0.2785 1.5147 1.7932 190 9.339 0.01657 40.% 157.95 167.99 977.9 1145.9 0.2938 I.4824 f.7762 200 11.526 0.01663 33.64 178.05 971.6 1149.7 0.3090 1.4508 1.7598 210 14.123 0.01670 27.82 970.3 1150 4 0.3120 1.4446 1.7566 212 14.6 % 0.01672 26.80 180.07 188.13 %5.2 1153.4 0.3239 1.4201 I.7440 220 17.186 0.01677 23.15 958.8 1157.0 0.3387 1.3901 1.7288 230 20.780 0.01684 19.382 198.23 208.34 952.2 1160.5 0.3531 1.3609 I.7140 240 24.% 9 0.01692 16.323 945.5 1864 0 0.3675 1.3323 1.6998 250 29 825 0.01700 13.821 216.48

' 228.64 938.7 1167.3 0.3817 1.3043 I.6860 260 35 429 0.01709 11.763 931.8 1170.6 0.3958 1.2769 1.6727 270 41.858 0.01717 10.061 238.84 249.06 924.7 1873.8 0.4096 12501 I.6597 280 49.203 0.01726 8.645 7.461 259.31 917.5 1176.8 0.4234 1.2238 f.6472 290 57.556 0.01735 910.1 1179.7 0 4369 1.1980 1.6350 300 67.013 0.01745 6 466 269.59 902.6 1182.5 0.4504 1.1727 1.6231 310 77.68 0.01755 5.626 279.92 894.9 1185.2 0.4637 f.1478 1.6115 320 89.66 0.01765 4.914 290.28 887.0 1187.7 0.4769 1.1233 1.6002 330 103 06 0.01776 4.307 300.68 0.01787 3.788 311.13 879.0 1190.1 0 4900 1.0992 1.5891 340 118.01 I

l l

)

TABLE D-Ib Properties of Dry Saturated Steam (continuedi Temperature g $pecific solume Enthalpy Entrop)

Temp.

I"

'F ^ sat. Sat $at Sat Sai Sat.

legent sapor hqun$ EssP sapor Esap heu d ,,po, a y s, e, h, h, h, s, s ,, s, 350 134 63 0.01799 3.342 321.63 870.7 1192.3 0.5029 1.0754 1.5783 360 153.04 0 01811 2 957 332.18 852.2 1194 4 0.5158 1.0519 1.5677

~~

370 173.37 0.01823 2.625 342.79 853 5 l196 3 0.5286 1.0287 1.5573 380 195.77 0 01836 2.335 353 45 844 6 1898.1 0.5413 1.0059 1.5471 390 220.37 0.01850 2.0836 364.17 835.4 1899.6 0.5539 0.9832 1.5371 400 247.31 0 01864 1.8633 374.97 826 0 1201.0 0.5664 0.9608 1.5272 410 276.75 0 01878 1.6MO 385.83 816.3 1202.1 0.5788 0.9386 I.5174 420 308.83 0 01894 1.5000 3 % .77 806.3 1203.1 0.5912 0.9166 1.5078 430 343.72 0.01910 1.3499 407.79 796 0 1203.8 0.6035 0.8947 I.4982 440 381.59 0.01926 1.2171 418.90 785 4 1204.3 0.6158 0.8730 1.4887 450 422.6 0.0194 1.0993 430.1 774.5 1204 6 0.6280 0 8513 1.4793 460 466.9 0.01 % 0.9944 441.4 763.2 I204 6 0.6402 0.8298 1.4700 470 514.7 00198 0.9009 452 8 751.5 12G4 3 0.6523 0 8083 1.4606 480 566.1 0.0200 0 8172 464 4 739 4 1203.7 0.6645 0 7868 14$13 490 62 4 0.0202 0.7423 476 0 7268 1202.8 0 6766 0.7653 1.4419 500 680.8 0 0204 06749 487.8 713 9 1201.7 06887 0 7438 14325 520 812.4 0.0209 05594 SI1.9 686 4 1198.2 0.7130 0.7006 1.4136 540 %2.5 0 0215 0.4649 536 6 656 6 1193 2 0.7374 0.6568 1.3942 560 1133.1 0.0221 0.3868 562.2 624.2 Ii86 4 0 7621 0 6121 1.3742 ,

580 1325.8 0 0228 0 3217 588.9 588 4 1177.3 0.7872 0.5659 1.3532 600 1542.9 0.0236 0.2668 610.0 548.5 I165.5 0.8131 0.5176 1.3307 620 1786.6 0.0247 0.2201 646.7 503 6 I150.3 0.8398 04664 1.3062 640 2059.7 0 0260 0.1798 678.6 452.0 1130.5 0.8679 0.4110 1.2789 660 2365 4 0 0278 0.1442 714.2 390.2 1104 4 0.8987 0.3485 1.2472 680 2708.1 0.0305 0.1115 757.3 309.9 1067.2 0.9351 0.2719 1.2071 700 3093.7 0 0369 0.0761 823.3 172.1 995.4 0 9905 0.1484 1.1389 705 4 3206.2 0 0503 0.0503 902.7 0 902.7 1.0580 0 1.0580 1

l i

TA BLE D-l.i' Propenses of Dry Satur.iini Steam +

Prenure g Spec (H; volume E nthalp) Entrop)

T emp .

E**

Sat Sat $at Y lequed tapor Isqu ed bE Sat Set Etap San tapor lequid ,,,,,

p s, t s, he h, h, s, s, s, l .0 101.74 0.01614 333 6 69 70 1036.3 1106 0 0.1326 1.84 % l.9782 2.0 126 08 0 01623 173 73 93 99 1022.2 0.1749 l l 16.2 1.7451 1.9200 3.0 141 48 0.01630 118.71 109 37 1013.2 1122.6 0.2006 1.6855 1.8863 40 152.97 0 01636 90 6) 120 86 1006 4 0.2198 1827.3 l.6427 1.8615 5.0 162.24 0 01640 73.52 130.13 1001.0 1131.1 0.2347 2.6094 1.8441 60 170 06 0 01645 61.98 137.% 996.2 l134.2 0.2472 1.5820 7.0 1.8292 176 85 0 01649 53.64 144.76 992.1 1136 9 0.2581 1.5586 8.0 1.8167 182.86 0.01653 47.34 150.79 988.5 1139.3 0.2674 1.5383 1.8057 90 188.28 0 016 % 42.40 156 22 985.2 1I41.4 0.2759 1.5203 1.7962 10 193.21 0 01659 38 42 161.17 982.1 1143.3 0.2835 1.3041 1.7876 14 696 212.00 0.01672 2680 180 07 970.3 I150 4 0.3120 1.4446 1.7566 15 213.03 0 01672 26 29 181.11 %9.7 1850 8 0.3135 1.4415 1.7549 20 227;96 0 01683 20 089 1 %.16 960.1 1856.3 0.33 % 1.3962 25 240 07 I.7319 0 01692 .6.30) 208 42 952.1 1160 6 0.3533 1.3606 1.7139 30 250 33 0 01701 13 746 218.82 945.3 I!64.1 0 3680 1.3313 1.6993 35 259.28 0 01708 11.893 227.91 939.2 1167.1 0 3807 1.3063 1.6870 40 267.25 0.01715 10 498 236 03 933 7 1169.7 0 3919 1.2844 1.6763 45 274 44 001721 9 401 243.36 928 6 I I 72.0 0 4019 1.2650 1.6669 50 281.01 0 01727 8.515 250.09 924 0 1874.1 0 4110 1.2474 1.6584 55 287.07 0 01732 7.787 256.30 919 6 1875.9 0 4193 1.2316 1.6504 60 292 71 0 01738 7.175 262.09 915.5 1877.6 0.4270 1.2168 f.6438 65 297.97 0.01743 6 655 267.50 911 6 18791 0.4342 1.2032 1.6374 70 302 92 0.01748 6.206 272 61 907.9 1180 6 0 4409 1.1906 1 6315 75 307.60 0 01753 5.816 277.43 904 5 l 181.9 80 0 4472 1.1787 1.6259 312.03 0 01757 5.4 72 282.02 901.1 l 183.1 0 4531 1.1676 1.6207 85 316.25 0.01761 5.168 286.39 897.8 1884.2 0 4587 1.1571 1.6154 90 320.21 0.01766 48% 290.56 894.7 1185.3 0 4641 1.1471 f.6112 95 324.12 0.01770 4 652 294 56 891.7 1186 2 0 4692 1.1376 1.6068 500 327.81 0.01774 4 432 298.40 888.8 l187.2 0 4740 1.1286 1.6026 110 334.77 0.01782 4 049 305.66 883.2 1188.9 0 4832 1.1117 1.5948

TABLE D la Properties of Dr3 Saturated Steam tronimurds Pressure Spect.c wolume Entbatr> Entropy g

Temp.,

P'm . Su Sai sa $a Su Su 1 EP Etap

"' i.eund sapor 1. quid *apor liquid ,,po, p t v, e, h, h, h, s, s, s, 312.44 877.9 1190 4 0 4916 1.0962 1.5878 120 341.25 0.01789 3.728 3 455 318 81 872.9 1191.7 0.4995 1.0817 1.5812 130 347.32 0.01796 3.229 324.82 868.2 1193 0 0.5069 1.0682 1.5751 140 353 02 0.0180 330 51 863.6 1894 1 0 5138 I.0556 1.5694 150 358 42 0.01s09 3 015 335.93 859.2 1895.1 0.5204 1.0436 1.5640 ,

160 % 3.53 0.01815 2.834 341.09 854 9 1196 0 0.5266 1.0324 1.5590 170 M8 4I 0 01822 2.675 346.03 850.8 1896.9 0.5325 1.0217 1.5542 180 373 06 0.01827 T532 350.79 846 8 I197.6 0.5381 10116 1.5497 190 377.51 0.01833 2.404 0.01839 2.288 355.M 843.0 1898 4 0.5435 f.0018 1.5453 200 381.79 376.00 825.1 1201.1 0.5675 0.9588 1.5263 250 400.95 0.01865 I.8438 1.5433 393.84 809 0 1202.8 0.5879 0.9225 1.5104 300 417.33 0.01890 0.01913 1.326.) 409.69 794.2 1203.9 0.60 % 0.8910 1.4966 350 431.72 444.59 0.0193 1.16?) 424.0 780.5 12.4.5 0 6214 0.8630 1.4844 400 0.0195 1.0320 437.2 767.4 1204.6 0 63 % 0 8378 I.4734 450 4 %.28 467.01 0 0197 0.9278 449.4 755.0 1204 4 0 6487 0.8147 1.4634 500 0.8424 460 8 743.I 1203.9 0.6608 0.7934 1.4542 550 476.94 0.0199 0.7698 471.6 731.6 1203.2 0 6720 0.7734 1.4454 600 486.21 0.0201 494.90 0 0203 0.7083 481.8 720.5 1202.3 0 6826 0.7548 I.4374 650 0.6554 491.5 709.7 1201.2 0 6925 0.7371 1.4296 700 503.10 0.0205 0.0207 0.6092 500.8 699.2 1200.0 0.7019 0.7204 1.4223 750 510.86 0 0209 0.% 87 509.7 688.9 1198 6 0.7108 0.7045 1.4153 800 518.23 0.0210 0.5327 518.3 678.8 1197.1 0.7194 0 6891 1.4085 850 525.26 0.0212 0.5006 526.6 668.8 1195 4 0.7275 0.6744 1.4020 900 $31.98 950 538 43 0.0214 0.4717 534.6 659.1 1193.7 0.7355 0 6602 1.395' 1000 544.61 0.0216 0 44 % 542.4 649.4 1191.8 0.7430 0.6467 1.3897 0.0220 0.4001 557.4 630 4 I I87.7 0.7575 0 6205 1.37D0 I100 5 %.31

%7.22 0.0223 0.M19 571.7 611.7 1183.4 0.7711 0.59 % l.% 67 1200 577.46 0.0227 0.3293 585 4 593.2 1178 6 0.7840 0.5719 1.3559 1300

' 1400 587.10 0.0231 0.3012 598.7 574.7 1173 4 0.7%3 0.5491 1.34 54 5%.23 0 0235 0.2765 611.6 5%.3 1167.9 0 8082 0 5269 1.335l 1500 0.0257 0.1878 671.7 463.4 II 35.1 0.8619 0.4230 1.2849 2000 635.82 0.0287 0.1307 730.6 360.5 1091.1 0 9126 0.3197 1.2322 2500 668.13 0.0346 0.0858 802.5 217.8 1020.3 0.9731 0.1885 1.1615 3000 695.M 0.0503 0.0503 902.7 0 902.7 1.0580 0 1.0580 3 X)6.2 705.40 eb

Propertes d Syperhiated Sterm' ;

. l Ahs yena.. Tamperawrs. < F j (Set tIp..*F ) 200 300 400 500 600 100 300 900 6000 1100 1200 1400

,. 392.6 452.3 512 0 571.6 631.2 6908 750.4 809.9 869.5 929.1 988.7 1807.8 I A. l150 4 1895.8 1241.7 1288.3 1335.7 13835 1432.8 1482.7 15335 1585.2 1637.7 1745.7 1 (101.74) s. 2.0512 2.1153 2.1720 2.2233 2.2702 2.3137 2.3542 2.3923 2 4253 2.4625 2 4952 2.5%6 l

,. 78.16 90.25 102.26 l 14.22 126.16 138 10 150 03 161.95 17387 185.79 197.71 221.6 5 A. 1148.8 1895.0 1241.2 1288.0 1335.4 1383 6 1432.7 1452 A 1533 4 1585.1 1637.7 1745.7 (162.24) s, l.8738 1.9370 1.9942 2.0456 2.0927 2.1%I 2.1767 2 214x 2.2509 2.2851 2.3178 2.3792 ,

p. 38 85 45.00 $1.04 57.05 63.03 69 01 74 9s 80 95 m 92 92 88 98.84 110.77 10 A.. II46 6 1193.9 1240 6 1287.5 1335.1 1383 4 1432.5 1482.4 e 51.12 15s5 0 1637.6 1745 6 (193.21) s. 1.7927 I8595 1.9172 1.% 89 2.0160 2.0596 2.1002 2.1384 2 1744 2.2068 2.2413 2.3028

v. 30.53 34 68 38.78 42 86 46 94 51.00 55 07 59 13 63.19 67.25 75.37 14.696 A. 1892.8 1239.9 1287.1 1334.8 1383.2 1432.3 1482.3 1533.1 1584 8 1637.5 1745.5 (212DO) s. I8160 18743 1.9261 1.9734 2.0170 2.0576 2.0958 2.1319 2.1662 2.1989 2.2603

,. 22.36 25 43 28 46 31.47 34 47 37.46 40 45 43 44 46.42 49 di 55 37 20 A. 1191.6 1239.2 1286.6 1334 4 1382.9 1432.1 1482.1 1533 0 1584 7 1637.4 1745.4 (227.96) s. l.7808 1.8M I.8918 1.9392 1.9829 2.0235 2.0615 2.0973 2.1321 2.1648 2.2263

,. II .040 12.628 14 168 15.688 17.19s 18 702 20.20 21.70 23 20 24 69 27.68 40 A. 1186.8 1236.5 1284.8 1333.1 1381.9 1431.3 1481.4 1532.4 1584 3 1637.0 1745.1 (267.25) S. 1.6994 1.7608 I8140 1.8619 1.9058 1.9467 1.9850 2.0212 2.0555 2.0583 2.1498

r. 7.259 8.357 9 403 10.427 11.44I 12.449 13.452 14 454 15 453 16 451 18.446 60 A. . 1181.6 1233.6 1283.0 1331.8 1380.9 1430.5 1480.8 1531.9 15MI M 1636 6 1744.8 (292.71) s.. . l.6492 1.7135 1.7678 1.8162 1.8605 1.9015 1.9400 1.9762 2.0 Hi6 2.0434 2.1049

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 55 ANSWERS -- CATAWBA 1 -85/09/09-TOM ROGERS ANSWER 5.01 (1.00) d REFERENCE HNS OP-HC-SPS-RT-NHF, p.48.

ANSWER 5.02 (1.00) e REFERENCE HNS OP-HC-SPS-RT-SH, PP.23-24.

ANSWER 5.03 (1.00) c REFERENCE HNS TS Bases, 3/4.1.1.2.

ANSWER 5.04 (1.50) 0 0.5 points each for NNS:

1. Shutdown marsin.

2. Limit + reactivity on a rod ejection. @ O.807 #"'(

3. Acceptable power distrubutiion limits. g,,,yd e'4edIW '

G 0.5 points each for CNS* lo. $ F # kJ b M l 1. Shutdown margin. g 9 gg

2. Rod misalignment effects for analyzed accidents.

! 3. Acceptable Power distrubution limits. C. M* D Id# #

REFERENCE HNS Core Performance, p.31. SCC AS Id#I '"'. '

l CNS TS Bases 3/4.1.3.

CMS MI /4 f.3.2 l

L

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 56 g g--------------------------------------

ANSWERS -- CATAWBA 1 -85/09/09-TOM ROGERS

' ANSWER 5.05 (1.00) b REFERENCE MNS OP-MC-SPS-RT-RP, p.24.

ANSWER 5.06 (1.00) c REFERENCE MNS OP-MC-SPS-RT-NMF, p.46.

ANSWER 5.07 (1.00) d REFERENCE MNS OP-NC-SPS-RT-RK, p.12.

ANSWER 5.08 (1.00) b REFERENCE MNS OP-MC-SPS-RT-RK, p.12.

ANSWER 5.09 (1.00) l a REFERENCE HBR, Reactor Theory, Sessions 41 and 42 i

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5.- THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 57 ANSWERS -- CATAWBA 1 -85/09/09-TOM ROGERS ANSWER 5.10 (1.00) c REFERENCE SHNP, RT-HO-1.10, pp. 14-15.

ANSWER 5.11 (1.00) c REFERENCE SHNP RT-HO-1.6, pp. 10-12.

ANSWER 5.12 (1.00) a p=(k2-ki)/k2ki=(1.004-0.92)/1.004(.92)=0.91.

REFERENCE NET, Mod. 3, NUS Corp. p.6.1.

ANSWER 5.13 (1.00) c REFERENCE VCS, RT, p.I-3.10.

ANSWER 5.14 (1.00) a REFERENCE VEGP, Training Text, Vol. 9, p. 21-47 Wastinghouse Reactor Physics, pp. I-3.17 a 19 DPC, Fundamentals of Nuclear Reactor Engineerin3r P. 106 001/000-K5.49 (2.9/3.4)

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PA,GE 58 ANSWERS -- CATAWBA 1 -85/09/09-TON ROGERS i

ANSWER 5.15 ( .50)

TRUE.

REFERENCE NET, Vol. 3, NUS, G-15.

4 ANSWER 5.16 (1.00) a REFERENCE DPC, Fundamentals of Nuclear Reactor Engineerin3' P. 138 i 001/000-K5.02 (2.9/3.4)

ANSWER 5.17 (1.00)

., c

REFERENCE DPC, Fundamentals of Nuclear Reactor ens ineering, p. 170

. 001/000-K5.13 (3.7/4.0)

ANSWER 5.18 (1.00) i b

' REFERENCE NET, Vol. 3, NUS Corp. p. 7.3-3.

i i ANSWER 5.19 (1.00)

C REFERENCE CNS Data Book, Table 6.4.

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5.- THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 59

~~~~TEER 66Y A 5C5~ ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ l ANSWERS -- CATAWBA 1 -85/09/09-TOM ROGERS i l

7 ANSWER 5.20 (1.00) l 1

b l t

. REFERENCE CNS Data Book, Table 6.2.4. ;

i 4

ANSWER 5.21 (1.00) ,

It is used in conjuction with an equilibrium xenon worth curve which [

is a function of reactor power. (When the appropriate values are added, j the xenon worth is essentially zero in both cases.)

REFERENCE l CNC Data Book, Curves 6.6.1 and 6.6.2. i i

ANSWER 5.22 (1.00) !

b REFERENCE i NRC IRE Westin3 house' Tech Manual, para. 1.1.7.1. -

r ANSWER 5.23 (1.00) c

^

REFERENCE FNP FSAR, Table 15.1-4. f ANSWER. -5.24 (1.00) b r

REFERENCE j

- MNS TS, 8 3/4 4-15.

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLLIDS, AND PAGE 60 l ANSWERS -- CATAWBA 1 -85/09/09-TOM ROGERS t-ANSWER 5.25 (1.00) a REFERENCE MNS Core Performance, p.10.

ANSWER 5.26 (1 00) c REFERENCE '

MNS Core Performance, p.26.

ANSWER 5.27 (1.00)

I a

REFERENCE Steam Tables.

ANSWER 5.28 (1.00) b

, REFERENCE I MNS Core Performance, p.20.

ANSWER 5.29 (1.00) b REFERENCE MNS Core Performance, pp.12-13.

i a

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! 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 61

. THERMODYNAMICS ANSWERS -- CATAWBA 1 -85/09/09-TOM ROGERS ANSWER 5.30 (1.00) d REFERENCE MNS Thermo, Sec. 8.

ANSWER 5.31 (1.00) b REFERENCE General Physics, HTAFF, p.328.

ANSWER 5.32 (1.00) e Throttling process-constant enthalpy.

h=1191.8 for PZR G 1000 psia hsat for 5 psi 3 is 1131 REFERENCE

? team tables.

1 f

ANSWER 5.33 (1.00) a REFERENCE Nuclear Reactor Analysis, Duderstadt & Hamiltor, p.491.

ANSWER 5.34 (1.00) '

! C l Couble speed => (2) squared x head =4 x 20 =80.

REFERENCE HTTFF, Gen. Phys. pp. 322-324. .

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 62 ANSWERS -- CATAWBA 1 -85/09/09-TOM ROGERS ANSWER 5.35 (1.00) 3 REFERENCE TFF&HT for NPP, DPC, p.134.

s ANSWER 5.36 (1.00) 3 REFERENCE Caneral Physics, HT & FF, pp. 139, 148, 156, and 331 ANSWEP 5.37 (2.00)

3. [NCREASE (0.5)

b. INCREASE (0.5)

2. CECREASE (0.5)

d. DECREASE (0.5) 7:EFERENCE Tr.er31 ohysics, HT & FF - Fluid Flow Applications for Systems and Components 00','000-l(5.01 (3.1/3.4)

ANSWE" 5.38 (2.00)

I 1.5 points each:

i 9.: J: in a bank move toseiher within 12 steps.

2. 22ntrol banks have proper sequence and overlap.

T, ?:1 ins?rtion limits are not violated.

2. A..ial power limits are observed.

s e

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6. ?LANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 63 ANSWERS -- CATAWBA 1 -85/09/09-TOM ROGERS ANSWIR. 6.01 (1.00) b

-REFERENCE iiNS OP-MC-SPS-SY-NC, p.24.

CNS NC Lesson Plan, p.24.

ANSWER d.02 (1.00)

REFE?ENCE MNS JP-hC-SPS-SY-NC, p.27.

CNS PSM CH-IC-ILE-22.

A{ISWEF 6.03 (1.00) a REFEFENCE 1ll'] '.' P - M C - S P S - S Y - N C , p.29.

CU^i 'SM 2N-IC-IPE-10.

A :!'3 '4 C S i 04 (1.00)

_ _. . ... .L- e m

.'i t ' 0 JF-+:-SPS-SY-NV, p.10.

TZ . ' ' -CJ-SPS-PS-NV, p.7.

(J:3t!E' C.OS (1.00)

- rm-r e.

m r ->: , .e

_ _ r

' :T .: N C-SPS-SY-HV, p.19.

C "'_ ;'.& - C N - S P S - P S - N V , p.25.

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6 _;NT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 64

';'32Ils -- CATAWBA 1 -85/09/09-TOM ROGERS

( '! ? L E 6.06 (1.00)

EEFERENCE 23 CP-MC-3P3-SY-SSF, p.10.

ANC2E" 6,07 (1.00) a RITERENCC

:? -MC -3P 3 -3Y-ND , p.16.

CNC NO Lesson Plan, p.22.

A:;:WER 6.09 (1.00) e e - -

_w e. nk.- -

4 _

m: ':d 7sd. Mont., MC-IC-RM-3.

CNC PCM r:ad. Mont., CN-IC-EMF-9.

AP:WL- L.?? (1.00)

S ,> :. ; . t s each. s4A/Y FooR.

:- : .densate Stor ge Tank.

. : .>r.c e r. s t e Hotwell.

t h .: ' i s .- 3ervice Water.

T, 9L C, "

_- r ..n r . . , . r.. ,

' 4: 3PI-3Y-CA, p.G.

. - ;, L -: . o r. Flan- p.7.

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't AN~ SYS TEMS DESIGN, CONTROL, AND INSTRUMENTATION 65 6 PAGE AN':WC35 -- C AT AWB A 1 -85/09/09-TOM ROGERS AM 3 W E T. 6,10 (1.00) 2 .'.25 pcints each.

~

1 La- law level in any SG.

~ Trip cf both main feedwater pumps.

. S: 31;n 1.

O.ackcut signs 1.

-- r-'e'nre-- '-

tdC @ -MC -SP3 -3Y-C A , p.9.

CUE A Lesson Plan, p.8.

AS:WEP '

,11 (1.00)

=

-. e. r e. t: .t.%

. eu. r MNC 2r-MT-SPE-SY-VI, p.8.

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,u ': CATAWBA 1 -85/07/07-TOM i;.0GERS e- _ . ,

( ,. . n. n. s

.i 1. - . s -

1: 1. .1 C - 3 P 3 - 3 Y - N F , p.23.

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c. 2F, p.1.

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. 22: _:31 3eq. Lesson Plan, p.14.

- . _ c,

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... _i A'- : '.. 1.00)

- -.:. '! ? ' ' ' : T! A C 'l i t s 1 Instrument & Control Power System Lesson Plan, p.3.

, . , _.v>e.. -- r. . _ 1 .

As ;~ (1 '0)

. i. D e ',e c t . . Price, p.42.

. .s e . ~n. n. s.

f b

~ C :" - : Y - U I ., p. 2 e- .n. A

( , . /1 ,v . ,. 6 .3

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~ : ' T ^ -: - TATAWBA 1 -85/09/09-TOM ROGERS

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.: CATAMBA 1 -85/09/09-TOM ROGERS

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. . _ . . . . . . . . .. n.3

$ _ 5.Y. . . , = . ~.: 2:tcr Inst.rumentation Systems Handbook, Vol. 1, Ch. 2

,.,,,3

~ ~ ~

~- -

'Ic . ' v ae 6, RPS and Question Bank, Section 6 s.

,-._,r_.-

._ . . -s. .

9 . 4 g

d _- T.72?is DESIGN, CONTROL, AND INSTRUMENTATION PAGE 69

. .. : ~:ATAWBA 1 -85/09/09-TON ROGERS

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. . 2 .- Fijures CN-IC-IRE-21 & 22

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7 C M C D E SION , CONTROL, AND INSTRUMENTATION PAGE 70

. .' Cc,7AWEA 1 -05/09/09-TOM ROGERS

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_ ., .i - 2;3TIMS DE3:GN, CONTROL, AND INSTRUMENTATION PAGE 71

! _ n .7. : - - CATAWBA 1 -85/09/09-TOM ROGERS I

l ANSuER 6.35 (1.00) c REFERENCE CAT, OP-CN-SPS-SY-HC, p. 12

! 008/030-A2.02 (2.5/2.8) t ANSWER 6.36 (1.00) 0 0.5 points each;

a. 82.4%.

b. 78%.

REFERENCE CNS PSh CN-CMP-SG, p.1.

ANSWER 6.37 (1.00) b

, REFERENCE CNS PSM CN-CMP-SG, p.1.

i ANSWER 6.38 (1.00) i

e t

[ REFERENCE CNS Unit 2 Operator Letter Dated 4/8/85.

l l ANSWER 6.39 (1.00)

. b a . or C. t

! REFERENCE CNS. Unit 2 Operator Letter Dated 4/8/85.

I i m i f

k i

i

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i

, l 5

E a

-,,..,w +7-~ , . m,-, . - - r. - 4, . - - - .--e--.v, m ,- -n-,m , w - -r n,,m-- -y n,, g,wme - e - ~ ~ - + =

i 1 ,

1 .

l 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 72 ANSWERS -- CATAWBA 1 -85/09/09-TOM ROGERS i

i ANSWER 6.40 ( .50) ,

.TRUE '

i j; REFERENCE CNS Unit 2 Operator Letter Dated 4/8/85.

4 l

i 4

i!

f f 4

i b

I h

i k

i

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s i

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b L

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

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 73

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

~~~~E 55dLUEiE L 55NTEUL ANSWERS -- CATAWBA 1 -85/09/09-TOM ROGERS i

i ANSWER- 7.01 (1.00) b REFERENCE MNS OP/1/A/6150/02A, p. 1.

CNS OP/1/A/6100/01, Encl. 4.1,2.17.

ANSWER 7.02 (1.00) a REFERENCE

, MNS OP/1/A/6100/01 p. 19.

CNS OP/1/A/6100/01, Encl. 4.1, 2.95.10.

ANSWER 7.03 (1.00) l a

REFERENCE MNS OP/1/A/6100/01, pp.14-15.

l CNS OP/1/A/6100/01, Encl. 4.1, 2 70 and 2.75.

l ANSWER 7.04 (1.00) d REFERENCE MNS OP/0/A/6100/17, Encl. 41, p.5.

l ANSWER 7.05 (1.00) {

5 c j REFERENCL ;

CNS OP/1/A/6100/02, p.1.

i r

i i

i

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 74

~~~~R d656[66f6d[~66NT66[~~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- CATAWBA 1 -85/09/09-TOM ROGERS ANSWER 7.06 (1.00) d REFERENCE CNS OP/1/A/6200/01, Enc 1 4.1, 2.8.1.

ANSWER 7.07 (1.00) c REFERENCE CNS TP/1/A/2650/01, p.3.

ANSWER 7.08 ( .50)

TRUE.

i REFERENCE CNS TP/1/A/2650/01, p.3.

ANSWER 7.09 (1.00) b I

REFERENCE CAT, OP/1/A/6100/03, p. 5 PWG-429 (3.7/4.0)

ANSWER ,

,7.10 (1.00) a REFERENCE MNS AP/2/A/5500/14, Case I, p.2.

CNS AP/1/A/5500/15, Case I, p.2.

1 i

1

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 75

~~~~R d6f6E66f6 E~66 TR6E~~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- CATAWBA 1 -85/09/09-TOM ROGERS ANSWER 7.11 (1.00) b REFERENCE MNS AP/2/A/5500/16, Case IV, p.9.

CNS AP/1/A/5500/16, Case IV, p.14.

ANSWER 7.12 (1.00) d REFERENCE .

NNS AP/2/A/5500/18, p.2.

CNS AP/1/A/5500/18, p.4.

ANSWER 7.13 (1.00) b REFERENCE CNS TP/1/A/2650/01, p.3.

ANSWER 7.14 (1.00) b REFERENCE CAT, AP/1/A/5500/13, p. 2 000/024-K3.02 (4.2/4.4)

ANSWER 7.15 (1.00) c REFERENCE CAT, AP/1/A/5500/04, p. 4 000/017-PWG424 (3.5/4.1)

4

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 76

~~~~Rd656E66ECdE 66 TRUE~~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- CATAWBA 1 -85/09/09-TOM ROGERS ANSWER 7.16 (1.00)

J b d hw & 4. Tesl$k.

REFERENCE CAT, AP/1/A/5500/17, p. 2 and Enclosure 10 000/068-K3.12 (4.1/4.5)

ANSWER 7.17 (1.00)

1. 10 minutes (0.5)

2. From the SSF (0.5)

REFERENCE CAT, AP/1/A/5500/08 000/015-PWG421 (2.7/2.8)

-PWG428 (4.4/4.4)

ANSWER 7.18 (1.50) ,

1. Ensure Containment Evacuation Alarm is sounded (if EMFs are in alarm) (0.5)

2. Terminate any waste Sas release in progress (0.5)

3. Stop Containment Purge (VP) (0.5)

REFERENCE CAT, AP/1/A/5500/25, p. 3 000/036-K3.03 (3 7/4.1)

ANSWER 7.19 (1.00) a REFERENCE CAT, AP/1/A/5500/19, p. 2 000/025-K3.03 (3.9/4.1)

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 77

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

R 6 6L 656dL"66 TE6L ANSWERS -- CATAWBA 1 -85/09/09-TOM ROGERS ANSWER 7.20 (1.25) 9 0.25 points each.

1. Manually exercise reactor trip train A & B switches.

2. Verify a reactor trip has occurred.

3. Verify a turbine / generator trip has occurred.

4. Verify ETA a ETB are energized.

5. Verify SI is not required.

REFERENCE MNS AP/2/A/5500/01, p.3.

CNS AP/1/A/5500/01, p.3.

ANSWER 7.21 ( .75)

'O 0.5 points each NHS:

1. Stop both DEH pumps.

2. Place the turbine in manual and close the governor valves in FAST action.

3. Locally trip the turbine.

CNS!

1. Locally trip turbine.

d

2. Unload turbine with Standby Load Set potentiometer.

3. Depress ON and Bypass pushbuttons on Turb. CNTL. PNL.

REFERENCE MNS AP/2/A/5500/01, p.3. -

CNS AP/1/A/5500/02, p.2.

i i

b 0

e

t

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 78 '

RADIOLOGICAL CONTROL ANSWERS -- CATAWBA ~1 -85/09/09-TOM ROGERS ;

4 f ANSWER 7.22 (1.00) ,

9 0.25 points each for NNf3 :

5 i 1. Manually insert rods.

2. Locally open the trir breakers.

3. Locally open the MG sets output breaker.

4. Locally open the MG sets motor breaker. ;

)

9 0.25 points each for CNS:

1. Manually insert control rods. 3

2. Locally open trip and bypass breakers.

3. Trip control rod drive MG sets A and B. !

4. Open control rod drive MG set feeder breakers. i REFERENCE l MNS EP/2/A/5000/11.1, I p.2.

! CNS EP/1/A/5000/01, p.2.

ANCHER 7.23 (1 00) !

I d

REFERENCE ;

MNS EP/2/A/5000/01, Foldout page. -

CNS EP/1/A/5000/01, p.8. ,

[

ANSWER 7.24 (1.00) l F

REFERENCE i l MNS EP/2/A/5000/01, p.10.

CNS EP/1/A/5000/01, pp. 16-17.

ANSWER 7.25 (1.00)

r REFERENCE MNS EP/2/A/5000/01, p. Encl. 1 Foldout.

i CNS EP/2/A/5000/1A1, Encl. 1, p.15. i I

i i

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

4 ..

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 79

~~~~Rd656E6G56dE~66NTEEE~~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- CATAWBA 1 -85/09/09-TOM ROCERS j

ANSWER 7.26 (1.00) b REFERENCE MNS EP/2/A/5000/1.1, p.5.

CNS EP/1/A/5000/1 Air p.4.

1 i ANSWER 7 27 (1.00) t 4

o REFERENCE MNS EP/2/A/5000/10, p.2.

CNS EP/1/A/5000/02, p.2.

~ ANSWER 7.28 (1.00)

, a ,

REFERENCE i

MNS EP/2/A/5000/10, p.2.

CNS EP/1/A/5000/02, p.3.

l ANSWER 7.29 (1.00) d REFERENCE MNS EP/2/A/5000/16.3 CNS EP/1// A/5000/2F3 r p.7.

I r i

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 80

~~~~R d656[6616d[~6d TRdE~~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- C.iTAWBA 1 -85/09/09-TOM ROGERS ANSWER 7.30 (1.50) 9 0.5 points each:

Check ift

1. SGs <725 psis.

2. PZR <1845 psis.

3. Containment > 1.2 psis.

REFERENCE CNS EP/1/A/5000/01, p.3.

ANSWER 7.31 (1.00) a REFERENCE CAT, EP/1/A/5000/1A1, p. 2 PNG-424 (3.5/4.0)

ANSWER 7.32 (1.00) d REFERENCE 000/009-K3.04 (4.1/4.3)

-K3.21 (4.2/4.5) .

CNS AP/1/A/5500/10, p.2.

ANSWER 7.33 (1.00) b REFERENCE CAT, EP/1/A/5000/2C2, pp. 3-4 i

o 9

l .

l

.i

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 81

~~~~

R D56E6656dE~66NT66E~~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- CATAWBA 1 -85/09/09-TOM ROGERS

! ANSWER 7.34 (1.50)

1. Verify at least one diesel running (0.5)

2. Verify B/0 sequencer actuated (for running diesel) (0.5)

3. Ensure CA' pump #1 is running (0.5)

REFERENCE CAT, AP/1/A/5500/07 000/056-K3.02 (4.4/4.7)

ANSWER 7.35 (1.00) d REFERENCE MNS Orientation Manual, p.27.

CNS Directive 3.8.6, para. 2.9.

ANSWER 7.36 (1.00) c REFERENCE MNS HP Manual, p.2 1-3.

CNS Directive 3.8.5, Encl. 7.3.

ANSWER 7.37 (1.00) c t

REFERENCE 10 CFR 20.101 ANSWER 7.38 (1.00) d i

a I

i

e i

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 82

~~~~Ed6f6666fddE~660 TEEL""~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- CATAWBA 1 -85/09/09-TON ROGERS >

t REFERENCE !

10 CFR 20.

i i

ANSWER 7.39 (1.00) !

b .

I REFERENCE CNS Directive 3.8.6, p.2. l

?

1 I

[

i a ,

4 i

f t

i l

f i

I i

I l

l

, 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 83 ANSWERS -- CATAWBA 1 -85/09/09-TOM ROGERS ANSWER 8.01 (1.00)

, 0 0.2 points each 1-a.

2-d.

3-c.

4-b.

5-e.

REFERENCE MNS TS, 3.4.6.2.

CNS TS, 3.4.6.2.

ANSWER 8.02 (1.00) i b

] REFERENCE MNS TS, 3.2.4.

CNS TS, 3.2 4.

i ANSWER 8.03 (1.00)

, a i

REFERENCE MNS TS, 2.1.2.

CNS TS, 2.1.2.

i ANSWER 8.04 (1.00) c REFERENCE i MNS TS, 3.4.4.

CNS TS, 3.4.4.

4 i

I i

i

8. . ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS RAGE C4 ANSWERS -- CATAWBA 1 -85/09/09-TOM ROGERS ANSWER 8.05 ( .50)

FALSE.

REFERENCE MNS TS, 3/4.1.3.

CNS TS, 3.1.3.1.

AN' '

8.06 ( .50)

TRUE.

REFERENCE MNS TS, 3.1.3.3.

CNS TS, 3.1.3.3.

ANSWER 8.07 (1.00) d REFERENCE MNS TS, 3.1.1.4.

CNS TS, 3.1.1.4.

ANSWER 8.08 (1.00) c REFERENCE i MNS TS, Table'3.3-1. -

CNS TS, 3/4 3-6.

ANSWER' 8.09 (1.00) b

, REFERENCE MNS TS, Bases 3/4.3.4.

CNS TS, 3/4.3.4.

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 85 ANSWERS -- CATAWBA 1 -85/09/09-TOM ROGERS ANSWER 8.10 (1.00) 9 0.5 points each. ,

c. TRUE.

b. FALSE.

REFERENCE MNS TS, 3.9.2.

CNS TS, 3.9.2.

ANSWER 8.11 (1.00) e del ++t. .

REFERENCE MNS TS, 3.8.2.1.

CNS TS, Table 3.3-13.

ANSWER 8.12 (1.00) b REFERENCE MNS TS, 4.0.2 & 4.0.5.

CNS TSr 4.0.2.

ANSWER 8.13 (1.00) 9 0.2 points each 9 MNS:

a. 1. -

b. 1.

c. 3.

d. 3.

o. 1.

For CNS: CAF.

REFERENCE MNS TS, Table 6.2-1.

! CNS TSr Table 6.2-1.

i em b_ _

i l

~8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 86 l ANSWERS -- CATAWBA 1 -85/09/09-TOM ROGERS ANSWER 8.14 (1.00) d REFERENCE .

CNS TS, 1.21.

ANSWER 8.15 '1.50)

(

0 0.5 points each

1. Each access door is shut (except during personnel passage when at least one of the two is shut).

2. Annulus Ventilation System operable. (WI)=

3. Sealing mechanism for each penetration is operable.

REFERENCE CNS TS, 1.26.

ANSWER 8.16 (1.00) d REFERENCE CNS TS, Table 1.2.

ANSWER 8.17 (1.00) a REFERENCE CNS TS, B 2-3.

ANSWER 8.18 ( .50)

TRUE.

REFERENCE CNS TS, 4.0.3.

6

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 87 ANSWERS -- CATAWBA 1 -85/09/09-TOM ROGERS ANSWER 8.19 (1.50) i

e 0.5 points each

1. From BAT to CCP.

2. From RHST to CCP.

3. Redundant path from RWST to CCP.

REFERENCE CNS TS, 3.1.2.2.

ANSWER 8.20 (1.00) b REFERENCE .

,CNS TS, 4.2.1.1.

S ANSWER 8.21 (1.00) 1 a

REFERENCE CNS TS, B 3/4.7.1.4.

I ANSWER, 8.22 (1.00) d i REFERENCE CNS TS 3.8.1.1.

ANSWER 8.23 ( .50)

FALSE (72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months )

I REFERENCE CNS TS 3.9.3.

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, .AND LIMITATIONS PAGE 88 ANSWERS -- CATAWBA 1 -85/09/09-TOM ROGERS ANSWER 8.24 (1.00) b REFERENCE 10 CFR 20, App. B.

ANSWER 8.25 (1.00) c -

REFERENCE MNS OMP 1-2, p.2. -

CNS OMP 1-4, p.1.

ANSWER., 8.26 (1.00) ,

fh .

REFERENCE i

MNS Sta. Dir. 3.1.8, p.1.

CNS Directive 3.1.2, p.3.

ANSWER 8.27 (1.00) b .

4 REFERENCE CNS OMP 1-4, p.9.

1. T ANSWER 8.'8-2 (1.00)

~

! c 414.k ',

REFERENCE CNS OMP 1-4, p.8.

J t

r

' .L. . , , , . -- - ..~v , _ , , , n

4 J

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 89 ANSWERS -- CATAWBA 1 -85/09/09-TOM ROGERS ANSWER 8.29 (1.00) o etN.

REFERENCE CNS OMP 1-4, p.13. .

, ANSWER 8.30 (1 00) d REFERENCE .

CNS Directive 4.4.3, p.3. -

ANSWER 8.31 (1.00) c REFERENCE CNS Directive 3.1.19, p.2.

ANSWER 8.32 (1.00) c REFERENCE CNS Directive 3.1.1, 6.11. l ANSWER 8.33 (1.00) b r

REFERENCE ~ [

CNS OHP 1-8, p.5. 1 t

ANSWER 8.34 (1.00) j c

i i

l i

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

. . ~ . _ _ .._

8. ' ADMINISTRATIV

E. PROCEDURE

S, CONDITIONS, AND LIMITATIONS PAGE 90 ANSWERS -- CATAW8A 1 -85/09/09-TOM ROGERS REFERENCE CNS OMP 2-19, p.2.

ANSWER 8.35 (1.00) ,

i d

REFERENCE CNS Directive 3.1.17, p.1. . . ,

ANSWER 8.36 (1.00) 8

- REFERENCE CNS Tech Spec Interpretation of 1.25, Dated 10-4-84. ,

ANSWER 8.37 (1.00) d REFERENCE CNS TS 46.8.2, Interpretation Dated 12-26-84.

ANSWER 8.38 (1.00) 9 0.5 point each for NNS:

1. Cowan's Ford Dam.

2. Training & Technolo3y Center.

9 0.5 points each for CNS:

1. DPC Transmission Line Maintenance Warehouse. (ppd 6 apert) '

1

2. DPC Allen Steam Station.

! REFERENCE MNS Directive 3.8.1.

CNS Directive 3.0.7, p.9.

em

~

d

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 91 ANSWERS -- CATAWBA l' -85/09/09-TOM ROGERS ANSWER 8.39 (1.00) b REFERENCE MNS HP Manual, 18.4, p.2.

CNS Directive 3.8.5, p.4.

ANSWER 8.40 (1.00)

.d REFERENCE CNS Directive 3.0.7, p.2.

ANSWER 8.41 (1.00) d RE'F ERENCE- -

CNS Directive 3.0.7, p'.8.

OO O

d 9

4 l

ML20133H829

Text

E. PROCEDURE

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