ML20205P311

Exam Rept 50-395/OL-86-02 on 861117-20.Exam Results:All Eight Reactor Operator Candidates,One Senior Reactor Operator Candidate & One Instructor Certification Candidate, Passed All Portions of Exams.Exams & Answer Key Encl

ML20205P311
Person / Time
Site:	Summer
Issue date:	03/23/1987
From:	Arildsen J, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20205P271	List: ML20205P268 ML20205P311 ML20205P365
References
50-395-OL-86-02, 50-395-OL-86-2, NUDOCS 8704030234
Download: ML20205P311 (138)

Text

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ENCLOSURE 1 EXAMINATION REPORT 395/0L-86-02 Facility Licensee: South Carolina Electric and Gas Company P. O. Box 764 Columbia, SC 29218 Facility Name: V. C. Summer Nuclear Station Facility Docket No.: 50-395 Written, oral, and simulator examinations were administered at the V. C. Summer Nuclear Station near Jenkinsville, South Carolina.

Chief Examiner: YMM -

sites (. ,198'?

Date Signed

. Arildsen Approved b : Neb hn 4. MunrE Sg6 tion Chief

/m.sem 23,196;t Date Signed Summary:

Examinations on November 17-20, 1986.

Oral examinations were administered to ten candidates, ten of whom passed.

Simulator examinations were administered to ten candidates, ten of whom passed.

Written examinations were administered to nine candidates, nine of whom passed.

Based on the results described above, eight of eight R0s passed, the one of one SRO passed, and the one of one Instructor Certification passed.

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REPORT DETAILS 4

1. Facility Employees Contacted:

• K. Woodward

• J. Heilman

• T. Matlosz i

• Attended Exit Meeting

2. Examiners:

• J. Arildsen, RII L. Defferding, PNL F. Jaggar, EG&G J. Moorman, RII

• Chief Examiner

3. Examination Review Meeting .

At the conclusion of the written examinations, the examiners provided Mr. Terry Matlosz with a copy of the written examination and answer key for review. The comments made by the facility reviewers are included as Enclosure 3 to this report, and the NRC Resolutions to these comments are listed below,

a. R0 Exam (SRO Exam)

(1) Question 1.16d NRC Resolution:

Agree. The second part of the question has been deleted and the point value has been adjusted.

(2) Question 2.22 (6.12)

NRC Resolution:

Disagree. Credit has been given only for the correct information. For full credit, the candidate must state the answer according to current plant design.

In the future, the facility is requested to perform timely revisions and updates to operator training materials to reflect design changes and to provide the most correct and complete training materials to the examiners for examination preparation.

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' Physics Manual. The. an wer'of "75 REM" .was accepted only when 't ( lib

tne candidate specifically rdferenced EFP-011. e' ,

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3 Agree. An.' 3 (f' cur of the ten listed respqnses.Nere accepted for ';

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, b. Examinatjon Answer Chang 2s Made By' Grader

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- 3 j Added addf tional correct re'sponse. . "More" power is produced in e th'e, upper' portion of the core than f n the bottom portion.N ,,

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(3) -Question 1.11 (5.14)

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(f Answer changed from a to c.

Reason:

Answer was incorrect due to distractor being rearranged during fg' exam formatting.

j-(4) Question 1.18 (5.19)

Resolution:

(. Answer changed from #1 to #3.

Reason:

Answer changed to reflect a correct answer that would more commonly occur during normal operation.

(5) Question 1.19 (5.20)

Resolution:

. s Accepted either a or d as the correct answer.

Reason:

Question did not state if the pump in question was a "REAL OR IDEAL" pump.

(6) Question 1.20 Resolution:

i No change will be seen in the system curve.

/ Reason:

i The slight decrease in steam generator pressure will not be seen in the performance of the main feed pumps.

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l (7) Question 2.01a (6.09a)

Resolution:

References to " limiting effects of an ejected rod" in place of I rod misalignment are NOT acceptable.

Reason:

Technical Specifications clearly state that one reason for having RIL is to limit the effects of rod misalignment.

(8) Question 2.03a Resolution:

Answer changed from true tb false.

Reason:

The question wording was changed during examination assembly, but the answer was not changed.

(9) Question 2.06 Resolution:

Accepted any one of the following for full credit: H0H Mix Bed; Non-Lithiated Mixed Bed; Cation Demineralizer.

Reason:

All demineralizers used a mixed ' bed resin with one being Lithiated.

(10) Question 2.08b I Resolution:

Flow will be directed to the VCl a3 PRT.

Reason:

Examiner error in reading the drawing in reference.

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(11) Question 2.08c Resolution:

Accepted VCT in place of SW HX.

Reason:

This is. common terminology for the destination of fluid from the Excess Letdown Heat Exchanger.

(12) Question 2.13 Resolution:

Points will be redistributed.

Reason:

The time delay is not an event as asked in the question. It is common practice to have both SW pumps running at all times; this information will not have as high a point value.

(13) Question 3.04 (6.14)

Resolution:

Accepted either c or b as the correct response.

Reason:

Examiner error in reading the reference document. Both responses correctly answer the question.

(14) Question 3.05 Resolution:

Changed to state that only five stationary gripper coils can be-supported simultaneously on-the DC Hold Cabinet.

Reason:

Typographical error.

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(15) Question 3.13a (6.17a)

Resolution:

Accepted "RCS temperature / pressure transient."

Reason:

Obtain a more complete / correct response.

(16) Question 3.15b Resolution:

Deleted No. 2 " Block start of start-up feed pump."

Reason:

Typographical error.

(17) Question 3.21 Resolution:

Deleted -- 2/3 -- 1977 psig from P-11.

Reason:

These are light indications only which make the operator aware that he may block P-11.

(18) Question 4.06 Resolution:

Accepted either a or c as th'e correct answer.

Reason:

If only one RCCA is dropped, the " Rods on Bottom" annunicator will not be received. Both responses answer the question.

(19) Question 4.13 Resolution:

Accepted "During initial E0P usage" as a correct response.

Reason:

The CSFs may be monitored during the performance of the I0As but must be started after the IOAs.

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4. Exit Meeting l

f At the conclusion of the site visit the examiners met with representatives of'the plant staff to discuss the results of the examination.

The areas . of below normal performance included: understanding of the Technical Specifications operability for independent ECCS subsystems with respect to VCT outlet valves; basic knowledge of the pressurizer's sealed reference legs; and, understanding of two and three white electrical potential indication lights.

t In general, the candidates exhibited noteworthy strength in awareness to indications and communications.

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: CU,MMER REACTOR TYPE: PWR-WEC3 DATE ADMINSTERED: 86/11/17 EXAMINER: JAGGAR, F.

CANDIDATE LJ AU:3MDV INSTRUCTIONS TO CANDIDATE:

Use separate paper for the answers. Write answers on one_ side only.

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

% OF CATEGORY  % OF CANDIDATE'S CATEGORY

'VALUE TOTAL SCORE VALUE CATEGORY 29.50 24.69 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

30.00 25.10 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS l

30.00 25.10 3. INSTRUMENTS AND CONTROLS 30.00 25.10 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL ~

119.5' Totals All work done on this examination is my own. 'I have ne'ither given

nor received aid. ,

Candidate's Signature-T

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

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

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

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

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

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

6. Use only the paper provided for answers.

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

8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer sheet.

9. Number each answer as to category and number, for example, 1.4, 6.3.

10. Skip at least three lines between each answer.

11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.

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

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

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

15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

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

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

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18. When you complete your examination, you shall:

a. Assemble your examination as follows:

(1) Exam questions on top.

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

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

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

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

d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still.

in progress, your license may be denied or revoked.

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,1.~\ PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Pcca~ 4

. ' THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.01 (1.00)

The reactor is at equilibrium Xenon. Boron is 890 ppm, rods.are in manual with Tavg on program and the turbine is loaded to 475 MW.

Turbine load suddenly reduces to 360 MW, but the steam dump fails to activate. Assuming no protective action occurs, what will be the new steady state reactor power and Tavg?

a. Reactor power 50%, Tavg 587 F

b. Reactor power 50%, Tavg 594 F

c. Reactor power 38%, Tavg 587 F

d. Reactor power 38%, Tavg 594 F QUESTION 1.02 (3.00)

a. Of the coefficients that contribute to the power defect, which coefficient contributes most to the change of power defect over core life? What causes the change in the coefficient you chose?

b. Explain why power defect is desirable for reactor operation at power.

c. Of the coefficients that contribute to power defect, which coefficient acts first to affect reactivity on a sudden power change due to rod movement? Explain why the coefficient you chose reacts before the others.

(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

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.1 . _ PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Paco 5 '

'. " TEERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW i

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QUESTION 1.03 (1.50)

a. Explain why fission product gas build-up in the gap between the fuel and clad causes Doppler (power only) Coefficient to become more negative over the life of the core. (1.0)

b. Does the effect of " clad creep" cause the Doppler (power only)

Coefficient to become MORE or LESS negative, over the life of the core? No explanation required. (0.5)

QUESTION 1.04 (1.00)

The reactor is operating at 100% power with all rods out, near EOL with equilibrium Xenon conditions when power is to be reduced to 50%.

The operator observes that AFD is within its band and decides to lower power by borating, leaving rods fully withdrawn. Actual Tavg follows programmed Tavg. Describe the change that will occur in AFD and why it occurs, prior to changes in Xenon having a noticeable effect.

QUESTION 1.05 (2.00)

a. If a specific amount of reactivity is added to TWO suberitical reactors that are identical, except for Shutdown Margin:

1. Which reactor will undergo the greatest change in count rate?

2. Which reactor will take a greater amount of time to reach a stable count rate?

b. For a reactor startup, how does the initial source count rate affect:

1. Rod height at criticality?

2. (Count rate) power level at criticality?

(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

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1. '_ PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Paco 6

'. 'THEBMQP_XEAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.06 (1.50)

a. Does Beta Effective Increase, Decrease, or Remain the Same, from BOL to EOL7 EXPLAIN YOUR CHOICE. (1.0)

b. For TWO equivalent positive reactivity additions to a critical reactor, will the SUR be the Same, Larger, or Smaller at EOL as compared to BOL? NO EXPLANATION IS NECESSARY. (0.5)

QUESTION 1.07 (3.00)

a. Determine the amount of reactivity (in pcm) required to increase Keff from 0.97 to 0.985.

b. By what factor will the COUNT RATE change as a result of increasing Keff from 0.97 to 0.9857

c. What would be the condition of the reactor after increasing Keff from 0.97 to 0.985, if the same amount of reactivity were added again?

QUESTION 1.08 (1.00)

a. Why are the neutrons used in the fissile fuel " slowed down" prior to use in the fuel?

b. Why are these same neutrons slowed down "quickly" prior to use?

QUESTION 1.09 (1.50)

TRUE OR FALSE

a. Rod Worth increases significantly as power increases because Rod Worth is proportional to flux and flux is proportional to power,

b. Rod Worth is higher at lower temperatures because fewer neutrons leak from the core and more neutrons are present in the moderator.

c. The HZP Total Rod Worth increases over core life due to less competition being offered by soluble boron and by the burnable poison rods.

(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

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_ , 1.  ; PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Pcsa - 7.

. ' THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.10~ (1.00)

Estimate the change in Tavg resulting from an inadvertant dilution of 10 ppm boron at EOL with rods in MANUAL and the turbine in AUTO.

State any assumptions of values used.

QUESTION 1.11 (1.00)

When performing a reactor S/U to full power that commenced five hours ,

after a trip from full power equilibrium conditions, a 0.5%/ min ramp was used. How would the resulting Xenon transient vary if-instead a-2%/ min ramp was used?

a. The Xenon dip for the 2%/ min ramp would occur SOONER and the magnitude of the dip would be SMALLER.

b. The Xenon dip for the 2%/ min ramp would occur LATER and the magnitude of the dip would be SMALLER.

l c. The Xenon dip for the 2%/ min ramp would occur SOONER and

the magnitude of the dip would be LARGER.

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d. The Xenon dip for the 2%/ min ramp would occur LATER and the magnitude of the dip would be LARGER. .

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! QUESTION 1.12 (1.50)

Indicate TRUE OR FALSE for the following statements concerning Xenon i behavior following a reactor trip.

! a. Xenon peaks later if the reactor trip occurs at high' power as l compared to low power.

! b. The Xenon concentration decreases following the peak because l the half-life of Xenon is shorter than the half-life of Iodine.

c. If Xenon and Iodine had the same half-life value, Xenon would

still peak following a reactor trip.

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1. 'PRINCIPLER OF NUCLEAR POWER PLANT OPERATION. Peca 8

'TEERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.13 -(0.50)

State whether the situations below will generate the greater tensile stress on the INNER or the OUTER wall of the reactor vessel.

Consider each situation seperately,

a. Heatup at a rate of 80 degrees F/hr

b. Increasing pressure 250 psig QUESTION 1.14 (1.00)

TRUE or FALSE

a. During a RCS heatup, as temperature gets higher, it will take a smaller letdown flow rate to maintain a constant pressurizer level.

b. Increasing condensate depression (subcooling) will cause BOTH a decrease in plant efficiency AND an increase in condensate (hotwell) pump available NPSH.

QUESTION 1.15 (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.

b. The ratio of the Critical Heat Flux to the actual heat flux,

c. The maximum local heat flux at core elevation (z) divided by average fuel rod heat flux.

(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

,l. ' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Peca 9

.  ; THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW i

4 f..t$2soK'ft' QUESTION 1.16 (W From the words or phrases in parentheses, choose the one which correctly completes each statement below,

a. The ratio of peak to average value of power distribution is known as (peaking factor /or/ thermal limit). (0.3)

b. The value of the Heat Flux Hot Channel factor will (increase /or/

decrease /or/ remain constant) when power is decreased. (0.3)

c. The Heat Flux Hot Channel factor staying below the limit (assures /or/ does not assure) that DNB will NOT occur during normal operation. (0.3)

d. Calculation of Enthalpy Rise Hot Channel factor assumes core power is (uniform /or/ not uniform) and flow through each channel is (the same /or/ different) throughout the core. (0.6)

QUESTION 1.17 (1.50)

a. After operating at 100% power for three months, power is suddenly lost to all of the reactor coolant pumps. Below are three operations that can be done to enhance natural circulation.

Why is each done?

1. Pressurizer level should be maintained at 50% or greater

2. Maintain at least 15 F subcooling in RCS

3. Maintain heat sink

b. Briefly explain how the following parameters will be trending if natural circulation is LOST:

1. RCS differential temperature

2. Steam generator steam pressure

3. Steam generator level (assume constant AFW flow)

(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

.1 . PRINCIPLER OF NUCLEAR POWER PLANT OPERATION. Paca 10

'TIIERM0 DYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.18 (2.50)

a. A variable speed centrifugal pump is operating at 1/4 rated speed in a CLOSED system with the following parameters:

Power = 300 KW Pump delta P = 50 psid Flow = 880 gpm What are the new values for these parameters when the pump speed is increased to full rated speed? (0.75)

b. Choose the answer that most correctly completes the sentence.

"In a CLOSED system, two single stage centrifugal pumps operating in parallel will have ----(choose from below)--- , as compared to the same system with one single stage centrifugal pump operating with one pump isolated."

1. a higher head and higher flow rate.

2. the same head and the same flow rate.

d. the same head and a higher flow rate.

4. a higher head and the same flow rate. (1.0)

c. How is the available NPSH to a centrifugal pump affected by an increase in system flowrate? Assume NO change in pump speed nor pump configuration. (0.25)

d. State TWO reasons why cavitation is undesirable? (0.50)

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

,1. ' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Paco 11

.- THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.19 (1.00)

Which of the following best describes the parameter changes that occur across a centrifugal pump in a closed system?

a. Temperature INCREASES, Enthalpy INCREASES.

b. Temperature INCREASES, Enthalpy DECREASES.

c. Temperature CONSTANT, Enthalpy CONSTANT.

d. Temperature CONSTANT, Enthalpy INCREASES.

QUESTION 1.20 (1.00)

On Figure IC2.6 attached to exam, show the effect on the operating curve AND characteristic curve of increasing Main Feed Pump speed without changing the position of the Feed Water Regulating Valve.

Include the Figure 102.6 with your answer paper.

(***** END OF CATEGORY 1 *****)

2. ' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pcca 12

. ' SYSTEMS QUESTION 2.01 (2.50)

a. What are three reasons, as stated in Tech. Specs., for having Rod Insertion Limits? [0.75]

b. With respect to the Rod Insertion Limits, "...the steamline break accident imposes the highest shutdown margin requirement." Explain why this is a true statement. [1.00]

c. Considering each of the following sets of conditions separately, which condition will make the Steamline Break accident worse? [0.75]

1. BOL or EOL

2. Reactor shutdown or at 100% power

3. Tavg at 350 F or at 547 F QUESTION 2.02 (1.00)

List TWO relief valves that discharge into the Volume Control Tank.

QUESTION 2.03 (1.00)

TRUE or FALSE.

a. During safety injection, a running RHR pump will stop if RWST level drops to </= 18% and its respective suction valve from the RWST (8809 A/B) is open.

b. Following a LOCA, both RHR pumps must operate to provide adequate flow to the reactor for ensuring ECCS design criteria of 10CFR50 are met.

QUESTION 2.04 (1.00)

Which RHR valve and in which direction (Open, Closed) does the operator MANUALLY adjust to reduce the RCS temperature when the RHR system is in service for a normal plant cooldown?

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

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2. PLANT DESIGN INCL

• 11NG SAFETY AND EMERGENCY Pcco 13 '

, SYSTEMS e

QUESTION 2.05 (1.00)

During shutdown Reactor Coolant Pump #3 is started after a seal replacement. After operating for approximately 20 minutes the following is observed.

1. #1 seal delta P >400 psi

2. Standpipe. low level

3. #1 seal leakoff has increased.

ASSUME:

1. Plant pressure is at 400 psi

2. Seal injection at 6 gpm.

Which one of the following is a probable cause for the abnormal indications?

a. #3 seal failure,

b. VCT pressure is low,

c. #1 seal bypass is open.

d. RCDT pressure has increased.

QUESTION 2.06 (1.00)

Which of the letdown demineralizers is used to remove excessive Lithium buildup in the RCS?

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

2. ' PLANT DESIGN INCLUDING SAFETY AND EMERGEHg1 Pcca 14 ,

SYSTEMS t

QUESTION 2.07 (1.00)

When the Reactor Makeup system is in AUTOMATIC, how is the l controller (for FCV-113) programmed for the correct boric acid I concentration?

I

a. The controller has an input from the boronometer. I

b. The latest analysis from Chemistry is automatically inputted into the controller,

c. The controller uses the setting on the manual / automatic potentiometers,

d. The controller uses the integrated charging header flow signal. l l

l QUESTION 2.08 (3.00)

a. What are 2 specific instances when the operator may initiate flow through the RHR System Letdown Control Valve i (HCV .teff)?

b.

M % YSA7 Explain the effect on Seal Return flow, if the Seal Water Heat Exchanger becomes plugged or restricted on the seal water side,

c. State the TWO components that the outlet of each of the heat exchangers below may be directed.

1. Seal Water Heat Exchanger.

2. Excess Letdown Heat Exchanger.

QUESTION 2.09 (1.50)

State the THREE diesel engine / generator shutdown signals that are enabled during an emergency start of the diesel?

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

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,,2. ' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Paco 15 SYSTEMS QUESTION 2.10 (2.00)

Explain the TWO methods which maintain the diesel near operating temperature while the diesel is idle.

QUESTION 2.11 (1.00)

An undervoltage on a 7.2 kV safeguards bus occurs 20 seconds after the receipt of a Safety Injection signal. Which one of the following statements regarding sequencing of loads onto the safeguards bus is correct?

a. All loads except Load Block #1 are stripped and the ESF Loading Sequence is reinitiated once the DG output breaker is closed,

b. Sequencing stops until the DG output breaker is closed at which time it continues from the point at which the under-voltage occurs.

c. Sequencing stops until the DG output breaker is closed at which time only the ECCS-related equipment sequence will be reinitiated.

d. All loads except the ECCS-related equipment are stripped and only the ECCS-related equipment sequence will be continued once the DG output breaker is closed.

QUESTION 2.12 (1.00)

Which trip of the 7.2 KV safeguards busses 1DA and 1DB will cause all loads to be dropped and then seqenced back onto the bus?

a. Overcurrent

b. Undervoltage

c. Phase Differential

d. Ground fault I

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

l l

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pcca 16

'. SYSTEMS i

QUESTION 2.13 (2.00) i As CST level decreases, describe the events that occur to ensure an I adequate supply of water is available to the Emergency Feedwater Pumps. l l

l QUESTION 2.14 (1.00)

Which one of the following choices describes the signal input (s) that is/are used to control the position of the governor on the Turbine Driven Auxiliary Feed Pump?

a. Turbine speed and pump discharge flow,

b. Turbine speed.

c. Steam pressure and pump discharge pressure.

d. Steam flow and pump discharge flow.

QUESTION 2.15 (1.50)

List the THREE RCS leakage detection systems that must be OPERABLE when the plant is operating at 100% power.

QUESTION 2.16 (1.00)

State the motive force for the RCS through the following:

1. Th RTD manifold.

2. Tc RTD manifold.

QUESTION 2.17 (1.50)

If the Rod Control Startup Reset Switch was inadvertently placed in the "Roset" position during mode 1 operation, what THREE components in the rod control system would have to be restored to the proper setting?

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

,2. ' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY' Paga 17

- ' SYSTEMS s

-QUESTION 12.18' (1.50)

Refer.to Table 2-1 attached to this-exam. Complete the Table, remove and include with your answer paper.

QUESTION 2.19 (1.00)

State the TWO functions of the throttle valves on the Cold Leg Injection lines.

QUESTION 2.20 (1.00)

State the purpose of the following, located in the Containment Spray System.

)

a. Orifices in the 12 inch suction lines to the pumps in association with orifices in the NaOH tank outlet lines.

b. Orifices in the spray pump discharge lines.

i QUESTION 2.21 (1.00)

Describe the flow path for the following steam generator main feed lines. (Begin your descriptien where the flow path leaves the Main Feed System),

a. Forward Flush.

b. Reverse Flush.

QUESTION 2.22 (1.50)

Of the signals used to automatically close the Feedwater Isolation Valves (1611 A, B, C), which TWO are specifically designed to prevent water hammer in the feedwater piping and steam generator inlet connections. Include setpoints and coincidence (logic) where applicable.

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

','3. ', INSTRUMENTS AND CONTROLS Pcca 18 l

l l QUESTION 3.01 (1.25) i

! The CONTROLLING pressurizer level channel fails HIGH during 100%

power operation. Assuming NO operator action is taken, which reactor protection signal will cause the reactor to trip?

Include a description of the events, (causes and effects) from j the time of the instrument failure until the reactor trip.

l l QUESTION 3.02 (1.00)

As containment temperature increases from normal operating temperature, what will be the relationship between actual and indicated pressurizer ,

level? Select the most correct answer.

a. Indicated level reads higher than actual level.

b. Indicated level increases regardless of actual level.

c. Indicated level and actual level are the same.

i l d. Indicated level reads lower than actual level.

l QUESTION 3.03 (1.00)

Which one of the following malfunctions would cause a pressurizer level indication of 0%7 l

I a. dP cell diaphragm rupture

b. Reference leg rupture

c. Impulse line rupture

d. Equalizing valve leakage l

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l

,.3 ' INSTRUMENTS AND CONTROLS Paco 19 QUESTION 3.04 (1.00)

-Which one of the following statements about the Digital Rod Position Indication (DRPI) System is correct?

i

a. The DRPI system uses a series of-magnetic switches to i determine rod position.

b. Anytime rods within a group differ by more than 12 steps a ROD DEVIATION alarm is generated.

c. A control bank B rod on the bottom will not generate a

RPI ROD AT BOTTOM alarm unless another bank B rod i or a control bank C or D rod indicates at least 12 steps.

f d. Power to the DRPI system is supplied from the control rod MG sets.

)

i QUESTION 3.05 (1.50) f

a. What consequences could be expected in the Rod Control System's DC Hold Cabinet if 2 or more groups of rod drive mechanisms were placed on hold power (excluding Control Bank D rods)?

Explain your reasoning. (1.0) 1

b. What is the purpose of each power supply (125 VDC and 70 VDC) in the DC Hold Cabinet? (0.50) t h

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

, .3 . ' INSTRUMENTS AND CONTROLS Pc3 20 t

QUESTION 3.06 (1.00)

Match the conditions in Column A with the expected indication provided by the rod speed indication meter given in Column B.

Responses in Column B may be used more than once.

COLUMN A COLUMN B t

a. Immediately before an operator removes the N44 1. O s/m.

instrument fuses because of a failed high detector (N44) that has been failed for FIVE 2 8 s/m.

minutes. Rods in AUTO with no temperature mismatch. 3. 40 s/m.

. b. Rods in MANUAL with a 10 F temperature 4. 48 s/m.

mismatch.

5. 72 s/m.

c. Rods in AUTO with a 1 F temperature mismatch.

(Assume increasing differential). 6, 88 s/m.

d. Rods in AUTO with a 4 F temperature mismatch.

i- QUESTION 3.07 (1.50)

Describe the effect on both Overtemperature Delta T AND Overpower j Delta T setpoints (INCREASE, DECREASE, or NOT CHANGE) for each of the below conditions? Consider each separately.

a. Tavg increases

b. Pressure increases 1

c. N41 upper detector fails high i

QUESTION 3.08 (2.00)

a. In addition to Pressurizer level, list the THREE control / processing circuits that utilize the output of the i T-average auctioneer,

b. State the five protection / alarm signals generated by the T-average signal from each loop.

i

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

l

3.' ,

INSTRUMENTS AND CONTROLS Peca 21 6

QUESTION 3.09 (1.00)

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

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

b. Detector, Log Integrator, Pulse Shaper, Pulse Counter, Meter

c. Detector, Pre Amp, Log Integrator, Discriminator, Meter

d. Detector, Log Amp, Meter QUESTION 3.10 (1.50)

The Detector Current Comparator receives input from all FOUR upper and lower power range detectors. How are these inputs compared, what is the alarm setpoint and when is this circuitry in operation?

QUESTION 3.11 (1.00)

Indicate which, if any, of the Excore Nuclear Instrumentation Ranges (SOURCE, INTERMEDIATE, POWER or NONE), is referred to in each of the following statements. More than one may apply to each.

a. Provides a direct input to the Rod Control System.

b. Has a reactor trip function that is blocked at some time between startup and full power operation.

c. Detector output current is adjusted by a front panel poteniometer.

d. Utilizes a Boron-10 coating in it's detectors.

e. Operates in the " Ion Chamber" region of the " Gas Filled Detector Characteristic Curve".

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

l

.l. ' INSTRUMENTS AND CONTROLS Pcca 22 i

QUESTION 3.12 (2.00)

a. Describe the instrument coincidence and setpoints necessary to insert the Reactor Coolant low flow trips into the protection system for 1 AND 2 loop loss of flow.

b. Explain how and why the undervoltage AND underfrequency low flow reactor trips operate differently.

QUESTION 3.13 (1.25)

The Reactor Protection System is designed so that a turbine trip will cause a Reactor Trip above P-9 (50% power),

a. Why is the Reactor Protection System designed to do this? (0.5)

b. Provide a Reactor Protection signal that would act to give protection in the event that the Turbine Trip / Reactor Trip did not operate on a turbine trip from full power. (0.25)

c. State the TWO ways that the Reactor Protection System senses that a turbine trip has occured? (0.5)

QUESTION 3.14 (1.00)

Which one 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 Tavg is also present.

d. Causes a turbine trip.

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

7 -

o r c;- ., , 4 '. i., 3.

3 t' . \'

4 t

, .3 \lNSTRUMENTS AND CONTROLS -jeg / Paco 23

1. ,1

,s .

\

{

/ M

.s.

ga t </ ~. 'f b QUESTION 3.15 (1.75) >

Provide TWO additional (different/ separate) AUTOMATIC signals (3 s a'.

other than High-High S/G water level (P-14),iwhich will cause the'. reactor protection and logic system to generate'a feedwater isolation valve closure. signal. (Setpoints not required). (0.5)

, b. Provide ALL direct and inimediate indirect automatic actions associated with P-14, other than feedwater isolation valve closure. (1.0) j ,

c. What protection is provided (reason / basis') by the F-14 signal? (0.25) *g~'i

( ( \

'l A k

QUESTION 3.16 (1.00) ,

(

Which one of the following statements concerning the operation'of the Reactor Trip breakers and Reactor Trip Bypass breakers is correct?

a. Tripping is accomplished by an undervoltage relay, normally held closed by 15 VDC power from the logic cabinet.,

s _n,

b. A Train B trip signal will trip both Reactor Trip breaker B and Bypass breaker B.

, c. The alarm "RX TRIP BYP BKR INOP POS" indicates the Bypass breaker is racked to the " Test" position but is NOT closed,

d. Control Power for the Reactor Trip breakers comes from 120 VAC channels I and III for RTA and RTB respectively. ,

\

s t

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

_ _ _ _ _ +

7

,3.'INNTRUMEhTSANDCONTROLS, Pcca 24 4

QUESTION 3.17 (1.25)

a. How will the core cooling monitor of Train A respond if-all RTD

[0.25]

inp'uts are disabled by their disable switches.

'b. What,is indicated by the following colors of lights for RTD and theimocouple temperature indication when they are lit?

1.

• Green

, r; j / 2. Yellow

3. Red [0.75]

, c. Is OPERABILITY of the Core Cooling Monitor required by T.S.?

[0.25]

QUESTION 3.18 *'(2.00)

State all the signals that are used-by the Cold Overpressure Protection system (COPS) to ensure proper operation.

I QUESTION 3.19 (2.00)

! a. State the five signals / conditions that will ALLOW the "C" Charging Pump to start Automatically. Assume transfer switch aligned to "A" Train.

b. State the response of both "A" and "C" Pump breakers if both 4

"A" and "C" Pumps are running on the "A" train and a blackout is received on that train.

QUESTION 3.20 (1.00)

What is the purpose of the delay in the time it takes for the actual Steam Generator Level signal to reach the PI controller where it is compared to Program Level?

i i

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

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3. INSTRUMENTS'AND CONTROLS , Pcca 25 s

QUESTION 3.21 (1.50)

List the four protection signals /permissives provided by the pressurizer pressure protection circuits. Include coincidence (logic) and setpoints.

QUESTION 3.22 (1.50)

The following pertain to indications on the Reactor. Vessel Level Indicating System.

a. How will the upper-range indication respond when.a RCP is-started in the associated loop?

b. What will the narrow-range indication show when any RCP is running?

c. How does wide-range indication change as reactor power is increased from 0 - 100%7 L

" T l

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(***** END OF CATEGORY 3 *****) -l 1

T- 1

, A '. _PROCEDUBES - NORMAL. ABNORMAL. EMERGENCY Paco 26

' AND_BADIOLOGICAL CONTROL s

QUESTION 4.01 (2.00)

Prior to increasing Tavg from mode 5, your heatup procedure (GOP-1) gives the option NOT to withdraw shutdown banks if either of TWO conditions exist. State the TWO conditions.

QUESTION 4.02 (1.00)

Which one of the following statements describing the method of unit shutdown from Mode 2 to Mode 3 is correct?

a. Using manual rod control, insert control banks D, C, B, and A to ZERO steps. Maintain the shutdown banks fully 1 withdrawn.

b. Using manual rod control, insert control banks D, C, B, and A to FIVE steps. Maintain the shutdown banks fully withdrawn.

c. Using manual rod control, insert control banks D, C, B, and A to zero steps. Using group select, insert shutdown banks to zero steps. Open reactor trip breakers. Reset reactor trip breakers. Using group select, fully with-draw shutdown banks.

d. Using group select, insert all rods to zero steps. Open reactor trip breakers. Reset reactor trip breakers.

Using group select, fully withdraw shutdown banks.

1 J

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(***** CATEGORY 4 CONTINUED ON NEXT PAGE-*****)

(

, , -- , - . , . . - , , , , . - - - . , , , n ,- , - ,- ,,r, -,. . . . ,

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, .4 . PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Pega 27

+

AND RADIOLOGICAL CONTROL i 1

QUESTION 4.03 (2.50)

Indicate the numerical value(s) associated with the following precautions as stated in GOP-Appendix A.

a. Maximum differential pressure between RCS and S/G.

b. Maximum differential temperature between RCS loops.

c. Minimum RCS flowrate during RCS dilutions.

d. Maximum RATE of power increase above 20% reactor power without management approval.

e. Maximum RCS temperature during RHR operations.

QUESTION 4.04 (2.00)

a. Indicate the Immediate Corrective Action (s) required, if while operating at 50% power, the following alarms occur simultaneously.

"RCP A #1 SL LKOFF FLO HI/LO" and "RCP A #1 SL dP LO"

b. How much time is allotted to take the above Immediate Corrective Action?

c. For how long may the RCP be operated with a #1 Seal Failure?

QUESTION 4.05 (1.00)

Briefly describe how the trip bistables of a failed power range detector are placed in the trip condition.

i

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

, ,4 . PROCEDURES - NORMAL. ABNOBti&L. EMERGENCY Pcga 28-AND RADIOLOGICAL CONTROL QUESTION 4.06 (1.00)

Which one of the following statements concerning the procedure for a dropped RCCA, EOP-10.0, is correct?

a. Upon starting recovery of the dropped RCCA, an URGENT FAILURE alarm will occur because the lift coils for the other rods in the group have been disconnected,

b. The delta flux target band is not applicable during a dropped RCCA malfunction and recovery.

c. A " RODS ON BOTTOM" alarm will'NOT be received.

d. Recovery from a dropped RCCA will be facilitated if Tavg is higher than Tref prior to commencing withdrawal of the dropped RCCA.

QUESTION 4.07 (1.50)

Which of the following conditions would NOT require re-initiation of safety injection according to EOP-1.2 " Safety Injection Termination"?

RCS PRESSURE SUBCOOLING PZR LEVEL %

a. Stable 25 15

b. Increasing 40 3

c. Decreasing 30 10

d. Stable 35 5 1

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

.4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Paga 29 AND RADIOLOGICAL CONTROL QUESTION 4.08 (1.50)

Match the condition in column A to the action required in column B.

Column B responses are NOT used more than once.

COLUMN A COLUMN B

a. Shutdown Margin in Question. 1. Reduce plant load.

b. 2/3 lo Steamline press. < 675# 2. Reset mech, over-speed.

c. Emergency Deisel trip during SI

3. Emergency borate.

d. Feed booster pump trip

4. Place Rods in

e. PZR pressure control channel fails manual.

High.

5. Safety inject.

f. First stage turbine pressure fails High. 6. .Close PCV-444.

7. Reactor trip.

8. Reset SI

[0.25 each]

QUESTION 4.09 '(1.00)

During a natural circulation cooldown following a reactor trip, which one of the following criteria determines the amount of RCS subcooling required?

a. RCS cooldown rate.

b. Reactor power history (decay heat rate).

c. Pressurizer level.

d. Number of CRDM fans running.

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

14.' PROCEDURES-NORMAL. ABNORMAL. EMERGENCY Paco 30

=

AND RADIOLOGICAL CONTROL v

QUESTION 4.10 '(1.00)

.Which one of the following statements describes the cooldown/

depressurization SEQUENCE of EOP-4.0, " Steam Generator Tube Rupture"?

a. Commence rapid cooldown; depressurize maintaining subcooling simultaneously with cooldown.

j b. Cooldown to 520 degrees; depressurize to 1500 psig; complete

.cooldown; complete depressurization.

. c. Cooldown to required temperature; depressurize to required i pressure once cooldown is. completed.

4

d. Commence rapid cooldown and depressurization without limits on subcooling.

i J

QUESTION 4.11 (1.50) i During a small break LOCA (SBLOCA), it is required to trip the RCPLif the trip criteria are met. If forced flow through the core promotes cooling,

why are the RCPs tripped?

i QUESTION 4.12 (1.00)

Functional Restoration Procedure FR-S.1, " Response to Nuclear Power Generation /ATWS", has the operator trip the turbine asLone of the immediate actions. However, one of the major concerns.in the ATWS transient response evaluations is the excessive RCS pressure developed due to significant heatup of the primary coolant. Since: keeping the turbine on the line would mitigate this temperature rise, why is the turbine tripped anyway?

QUESTION 4.13 (1.00)

According to EOP 12.0, Monitoring of Critical Safety Functions, when 4 does monitoring of the Critical Safety Functions begin?

(***** CATEGORY 4. CONTINUED ON NEXT'PAGE *****)

4

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

,~.--.,-r-, ,4 y . .,,..-.m, , . , , ~ , . . _ , - ,. - - . , ..

Pea 31

.4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY

' AND RADIOLOGICAL CONTROL s

.- QUESTION 4.14 (2.50)

During a serious emergency, operators may be called upon to assist in I

search and rescue or recovery operations in the plant.

a. In such cases, according to Health Physics manual, what dose could you receive:

1. To bring an injured worker to safety? [0.5]

2. To get ESF equipment running if the core was uncovered? [0.5]

b. What are the possible effects of receiving radiation exposures of 50 REM? Include short and long term effects. [1.0]

) c. Who must authorize this voluntary radiation exposure up to the emergency limits? [0.5]

i QUESTION 4.15 (2.50)

.1

a. State the legal (10CFR20) quarterly exposure limits for the following.

1. Whole Body.

2. Extremities.

3. Skin,

b. State the V.C. Summer Administrative Exposure.Guidlines for the following.

1. Whole Body.

2. Extremities.

3. Skin.

c. To what value can the Administrative Guideline for whole body be raised to and who (by title) must approve this limit increase?

(***** *****)

CATEGORY 4 CONTINUED ON NEXT PAGE I

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' , ,4 . PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pega 32 AND RADIOLOGICAL CONTRO_L QUESTION 4.16 (2.00)

According to SOP-401, Reactor Protection and Control System, list FOUR of the FIVE Instrument Channels that need not be placed in the trip mode when they are declared to be out-of-service.

QUESTION 4.17 (1.50)

a. Per Special Instruction 8602, what is the time limit for notifying the NRC Emergency Operations Center of an inoperable NRC Control Room Hotline?

b. State the TWO methods that may be used for the notification listed in "a." above.

QUESTION 4.18 (1.50)

a. What qualifications are required to perform the second verification of tag placement, accuracy and equipment position? [1.0]

b. TRUE or FALSE?

Under certain circumstances during electrical maintenance requiring intermittent operation of circuits which preclude installation of a Danger Tag, it is permissible for any qualified danger tagger to be stationed at the electrical device in lieu of a Danger Tag. [0.5]

QUESTION. 4.19 (1.00)

From what TWO procedures is EOP 13.0, Response to Abnormal Power Generation, entered? ,.

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)-

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.4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Pega 33

' AND RADIOLOGICAL CONTROL

\

' QUESTION 4.20 (1.00)

During the recovery phase of a plant Blackout the normal off-site power supplies are placed back in service. Which one of the below listed actions must the operator take to insert speed droop control in the diesel governor system before paralleling the power sources?

a. Depress Emergency Start Reset and select parallel operation on the Voltage Regulator switch.

b. Select parallel operation on the Voltage Regulator switch and turn on the synchronizing scope.

c. Depress Emergency Start Reset and Test Start pushbottons.

d. Reset Blackout sequencer and depress the Exciter Reset switch.

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(***** END OF CATEGORY 4 *****) )

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

'l 1

4

a i FEED PUMP SPEED PROGRAM . -

Head Loss go., e , ie e,oo,,,  ;

200 -

_9 9

' 160 - (215 PSID)

. " Pump

% Head AA p 120 -

e -

80 - (105 PSID) l A

4 l 1 e4 (45 PSID) No Load AP e i I I i i i I I I 10 20 30 40 SO 60 70 80 90100

g 1 Power (%)

g I

. 1 T< .

Feed Flow 173 T22757JDOS

- n. - n m o n 1

i

s Table 2-1.

for Question 2.18 For the following list of RHR valves, identify their normal positions for the indicated modes of operation.

Use "C" for closed or "O" for Open.

Valves Mode 1 CL Recire HL Recirc

a. RWST Suction (8809)

b. RCS Loop C Suction (8702, 8701)

c. Contain Recire. Sump Sections (8811, 8812)

d. RHR Discharge Cross-Connect Valves (8887)

e. Charging /RHR Pump Suction (8706)

Table 2-1 i

Please remove this Table and include with your answer paper.

l I

1 l

i i

i

-, . . ,, , ~. - - - w -

l

, EQUATION SHEET

~.

f = ma

~

v = s/t

,. ,g ,,yg+ g,g 2 Cycle efficiency = E

. E = aC -

a = (vg - v,)/t , ,

KE = hav A = AN f=v o+ at A = A,e v

]

PE = mgh w = 8/t A = In 2/tg = 0.693/tg .

W = v4P' tq(eff) = (t,:)(ts) . . . .

AE = 931Am . 1

+

( )

4 Q=[ncAT

, P , I . I

~

IX o ~

', Q = UAAT I = I . UX , ,

Pur = W'f *n I.I lo

-x h o

y.p to SUR(t) . TVL = 1.3/u y.p .t/T HVL a 0.693/u O

'SUR = 26.06/T _

T = 1.44 DT SCR = S/(1 - K,gg) ph SUR = 26 /A g {f CRg = S/(1 - K,gg )

~

T = 11*/p ) + [(i 'o)/A,ggo ] 1 C

eff}1 " 2 Cl ~~*eff}2 T,= 1*/ (p - D M = 1/(1 - K,gg) = CR g/CR0 T = (I - p)/ A,gg p g " C1 ~ K eff)0 /(1 - K,gg)g p = (K,gg-1)/K,gg = AK,gg/Kaff SDM = (1 - K,gg)/K,gg a= [L*/TKygg.] + [E/(1 + A,ggT )] 1* = 1 x 10 seconds

~

P = E6V/(3 x 1010) g A aff = 0.1 seconds I = No -

Idgg=Id22 WATER PARAMETERS Id =Id2 g

1 gal. = 8.345 lba 2 R/hr = (0.5 CE)/d (,,t,,,)

I gal. = 3.78 liters R/hr = 6 CE/d (feet) -

1 ft = 7.48 gal.

MISCEI.I.ANEOUS CONVERSIONS .

Density = 62.4 lbm/ft 3 1 Curia = 3.7 x 10 dps O 3

Density = 1 gm/cm 1 kg = 2.21 lba Heat of varorization = 970 Etu/lba 1 hp = 2.54 x 103 BTU /hr Heat of fusica = 144 Btu /lbm 1 Hw = 3.41 x 10 Btu /hr 6

1 Atm = 14.7 Psi = 29.9 in. I'g. 1 Stu = 778 f t-lbf

. I ft. H O 2

= 0.4333 lbf/in 1' inch = 2.54 cm F = 9/5*C + 32 "C = 3/9 (*r . 32)

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Paga 34

. THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW s

ANSWER 1.01 ,(1.00).

c.

REFERENCE VCS Reactor Theory, p. I-5.26 ANSWER 1.02 (3.00)

a. Moderator Temperature Coefficient (MTC) [0.5] due to an increase

- (more negative) in MTC as boron concentration:is reduced over core life [0.5].

b. Power defect has a stabilizing influence on reactor operation because it resists power changes. (As power increases, power defect adds negative reactivity and as power decreases, power defect adds positive reactivity). (1.0) 9

c. Doppler (FTC) [0.5]. Fuel temperature changes before the other parameters change, to effect the'other coefficients [0.5].

4 REFERENCE 1

Millstone Reactor Theory, RT-13, Pp 6-7 and RT-12.

VCS Reactor Theory I-5.2;-5.4 ANSWER 1.03 (1.50)

a. The gases contaminate the gap which reduces the thermal conductivity of the helium gas which raises the temperature of the fuel. (1.0)

b. LESS negative. (0.5) l 1

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(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) I

, ,1 . . PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,- Pcca 35

- ' THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW s

REFERENCE G.P. Heat. Transfer and Fluid Flow, Pp 235-240.

Millstone Reactor Theory, RT-13, p2.

ANSWER 1.04 (1.00) .-.

More positive reactivity will be added in the upper core regions, resulting in a more positve (less negative) AFD [0.5]. -Due to the greater decrease in.the temperature of the coolant exiting the core relative to the dpcrease of the inlet coolant. a[ .

09 th.1 1-cr si gro m + in Pne_. g g'<, pu, { ;m Le j C r'

• 6 L'

.,7, b.m bOr , Co.:} %, j d clvl, Tk anuaeco .

Westinghouse Nuclear Training Operations, Ch 8 Surry ND-86.2-LP-8, pp 8.14/15 gf/g/7f,

/

001/000; KS.29 (3.7/3.9)

ANSWER 1.05 (2.00)

a. 1. The reactor with less SDM. Mde ; WM 4 CCEP b d 40M. 6 docui " p .c$<,3 9 c c b 8 5 '-

2. The reactor with less SDM. [0.5 ea.]

b. 1. Does not affect critical rod height.

2. The larger the initial count rate, the higher the power level at criticality. [0.5 ea.]

REFERENCE VCS Theory review text I-4.26-27

(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

~

'[1 . ' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Pega 36 THERMODYNAMICS. MAT TRANSFER AND FLUID FLOW ANSWER 1.06 (1.50)

a. Decrease [0.5] Pu 239 concentration increases (while U 235 concentration decreases) [0.5].

b. Larger SUR at EOL. [0.5]

REFERENCE HO-RTR-23 to-27 VCS Review text I-3.10-3.14 ANSWER 1.07 (3.00)

a. 1570 pcm. +/- 50 pcm

b. 2 (Exactly double).

c. Slightly supercritical. [1.0 ea.]

REFERENCE VCS Theory Review text I-3.2, -4.27 ANSWER 1.08 (1.00)

a. To increase the probability of thermal fission (in the. fissile fuel).

b. To minimize the amount of time the neutrons remain in the resonance energy spectrum thereby minimizing the probability of resonance capture.

[0.5 ea.]

REFERENCE VCS Review Text I-2.20;-2.22 s

(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

.1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, . Pcco 37 THERMODYNAMICS.-HEAT TRANSFER AND FLUID FLOW s

ANSWER 1.09 (1.50)

a. False

b. False

c. False [0.5 ea.]

REFERENCE VCS Theory Review text I-5.38; -5.49-ANSWER 1.10 (1.00)

Delta RHO Net = Delta RHO iso + Delta RHO Boron.

Delta RHO iso = -Delta RHO Boron = -(Delta Boron Conc.)(Alpha Boron)

= -(-10 ppm)(-10 pcm/ ppm)

= -100 pcm Delta T iso = Delta RHO iso Alpha iso

= -100 pcm

-20 pcm/ F

=+5F +/- 1 F Delta Tavg = Delta T iso = + 5 F REFERENCE VCS Curve Book Fig.~ V-8 l

ANSWER 1.11 (1.00) i d'(5 9gf? ljllff8?

l l

(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

l

.i

,' l' - ' PRINCIPLES OF NUCLEAR POWER-PLANT OPERATION, Paca 38 s' THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW s

REFERENCE-CNTO " Reactor Core Control" Section 4 Westinghouse Simulator Trng book, "Rx Theory and Core Physics",

Fig I-5-54 VCS Review text I-5.57-76 001/000; K5.38(3.5/4.1)

ANSWER 1.12 (1.50)

a. True

b. False

c. True [0.5 ea.]

REFERENCE VCS Review Text I-5.57; -5.76 ANSWER 1.13 (0.50)

a. Outer

b. Inner [0.25 ea.]

REFERENCE i

NUS, Vol 4, Unit 10.1  ;

WNTO, " Thermal / Hydraulic Principles and Applications", pp 13-57/58 l l

004/000; K5.09(3.7/4.2) l l

ANSWER 1.14 (1.00) l l

a. FALSE

b. TRUE [0.5 ea.]

• • • • • )

(***** CATEGORY 1 CONTINUED ON NEXT PAGE l

. ,1 - PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Paso 39

- THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW REFERENCE

~

General Physics, HT&FF, pp. 155 and 320 and Subcooled Liquid Density Tables ANSWER 1.15 (1.50)

a. Latent Heat (of Vaporization)

b. DNBR

c. Heat Flux Hot Channel Factor [0.5 ea.]

REFERENCE WBN, HT & FF, pp. 11, 22, and 24 General Physics,-HT & FF, pp. 38 & 228 and VCS, TS, p. B3/4 2-1 002/000-K5.01 (3.1/3.4) ev D O'G ANSWER 1.16 (.L, 60 )

a. Peaking Factor (0.3)

b. Increase. (0.3)

c. Does not assure. (0.3)

d. Not uniform,

[0.3]J.F4an.gpV5f M/ON

- . . . . . a.

REFERENCE GP HTFF pp. 193-198 Thermal-Hydraulic Principles, Pp. 13- 30 thru 36 CNS Exam Bank 4

(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

,,1 . PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Pacs 40 THERMODYNAMICS. HEAT TRANSFER AND FLUID FION ANSWER 1.17 (1.50)

a. 1. To ensure that no vapor pockets form in the loops

2. To prevent steam pocket formation

3. To help thermal driving head [0.25 ea.]

b. 1. RCS dT will increase (exceed 100% full power value)

2. Pressure will decrease.

3. Level will increase. [0.25 ea.]

REFERENCE HTTFF pgs. 356, 357 3.4 000 015 EK 1.01 4.4 ANSWER 1.18 (2.50) 3 3

a. Power (2) = Power (1) * (N2/N1) = 300 * (4) = 19.2 MW 2 2 Delta P(2) = delta P(1) * (N2/N1) = 50 * (4) = 800 psid Flow (2) = Flow (1) * (N2/N1) = 880

• 4 = 3520 gpm [0.25 ea.]

b. Answer: (1.0) ,jjy j)ff[g7

c. DECREASES (0.25)

d. 1. Pump efficiency and

2. Flowrate are reduced and

3. Mechanical pump damage (erosion, pitting and vibration) may result. [any TWO @ 0.25 ea.]

! REFERENCE VCS HTFF pp. 322; 324-326; 319, 320 l

l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

1

,,1. 'LPRINCIPLES~OF NUCLEAR POWER PLANT OPERATION, - Paso 411 THERMODYNAMICS. HEAT TRANSFER AND ' FLUID FLOW i s  ;

ANSWER ~ 1.19 (1.00)

-a.(Ld St}h lldlB7 REFERENCE WC Thermal-Hydraulics Ch.10 GP HTFF pp. 145-152 ANSWER 1.20 (1.00)

See curve B (operating curve) and BB (characteristic curve) on Fig.

IC2.6 ANSWER attached.

REFERENCE VCS IC-2 p. 14; Figure 102.6 I

i I

(***** END OF CATEGORY 1 *****)

j _-

1 i

FEED PUMP SPEED PROGRAM -

i, I

Head Loss go.o eomo te proo,..

200 -

I k iso - (245. PSID)

Pump N B g

j Head N AA 120 -

5 BB -

so (105 PSID)

? A -

Ib 2--,

')

,E 1

l l t 4o4 (45 PSID) No Load AP wl Hesd _

,m Head Loss > Loss o I I I I I I I I I F A j 1 20 so 40 so ao 70 80 soioo S/G A 4

\' '

B Power (%)

Pressure t hs"-------- --

Entering I LossB 1 1

'y 1 2 Feed Flow

[

C eve G - Co.c] Ce,ls Cw-ve) l C G 6 - Cu . cy S$$

1 m cus,se  %

l c s, u c_-,a ce.ca E6S'9fg-n

- n -

i l______________ _-..___ -_ _._________________ _ ____ _ _ __ _

'.,2.' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Peca 42 1

' SYSTEMS ANSWER 2.01 -(2.50)

a. 1N) ensure adequate trip reactivity. (minimum SDM)-

To limit the potential effects of rod misalignment.

To assure power distribution limits are met. [0.25 ea.]

b. Due to the large value of positive reactivity inserted by MTC during the resulting uncontrolled RCS cooldown. (1.00)

c. 1. EOL

2. Shutdown

3. 547 F [0.25 ea.]

REFERENCE Technical Specifications, pg B 3/4 1-3

• • • CAF ***

K&A 001-000-K5.08 / IF 3.9 001-000-K5.04 / IF 4.3 ANSWER 2.02 (1.00)

1. L/D Hex (tube side)

2. Seal Water Hex (tube side)

3. Letdown Reheat Hex (shell side) [2 @ 0.5 ea.]

REFERENCE VCS, AB-3, CVCS, p. 22 004/000-K6.03 (2.4/2.5)

ANSWER 2.03 (1.00)

a. -True- M.s e @ ///5/87

b. False [0.5 each]

REFERENCE VCS AB-7 pp. 5,16

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

., 2 - . PLANT DESIGN INCLUDING-SAFETY AND EMERGENCY Paca 43 SYSTEMS s

ANSWER 2.04 (1.00)

> RHR Heat Exchanger Outlet Valve (HCV-603 A/B) [0.5]; OPEN [0.5]

REFERENCE VCS AB-7 p.23 PWG K/A 29 (3.6/3.9)

ANSWER 2.05 (1.00) b.

REFERENCE VCS Plant System Descriptions, AB-4, pp 29-31 ANSWER 2.06 (1.00) 40 E+

Cation Ped Demineralizer O(R REFERENCE E 7a '- . ; 1 A i.g gg g " ygi ," y ry, b VCS AB-3 p. 17

///r/B7 ANSWER 2.07 (1.00) c 4

REFERENCE VCS AB-5 p. 23 004/010-K6.05 (3.1/3.3) i

)

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

j

2.' PLANT DESIGN INCLUDING' SAFETY AND EMERGENCY' Paga 44 I

'. '8.YSTEMS t

.i 1

ANSWER 2,08 (3.00)

a. 1. To purify coolant when on RHR cooling.

2. Provide additional letdown during heatup or bubble formation.

3. RCS solid plant pressure control. [Any 2, 0.5 ea.]

b. Continued flow is assured by a relief valve which directs flow to the VCT. (1.0) i l L.%A Per 'A f/jfg7/

c. SW HX: 1. Chg. pump suct.

2. VCT [0.25 ea.]

Ex. LD HX: 1. SW HX (vcTD @ /

2. RCDT ad f I [0.25 ea.]

efaltri E%MV REFERENCE VCS AB-7 p.18; AB-3 Fig. AB3.8, AB3.7 ANSWER 2.09 (1.50)

1. Diesel Overspeed.

2. Generator Differential.

3. Low Lube Oil. [0.5 ea.]

REFERENCE VCS Plant System Descriptions, IB-5, p. 33 ANSWER 2.10 (2.00)

1. The lube oil prelube system circulates oil through a heater to normal lubrication components.

2. The cooling water keep-warm system circulates water through a heater to normal water-cooled components.

[1.0 ea.]

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

.,2.' PLANT DESIGN INCLUDING SAFETY AND-EMERGENCY Peca 45

' SYSTEMS REFERENCE VCS, IB-5, Diesel Generator, pp. 34 & 35 064/000-K1.02 (3.1/3.6)

-A1.01 (3.0/3.1)

ANSWER 2.11 (1.00) a REFERENCE VCS, GS-2, Safeguards Power System, p. 36 ANSWER- 2.12 (1.00) b.

REFERENCE VCS Plant System Descriptions, GS-2, p. 19 ANSWER 2.13 (2.00) 34 As the CST level decreases, the EFS pump suction pressure decreases M. At a pressure of (10.4 psig) the Service Water Backup supply M,gtWand . isolation valves automatically open Lfg$7 (after a 42 sec.8' time delay) M Also, one SW pump per train wN 1 auto. start 18<5T to ggjkad'23da..Its' ensure the water supply availability.

f*55 REFERENCE VCS IB-3 pp.9,10 f

ANSWER 2.14 (1.00) b.

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

1

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

2.'

PLANT DESIGN INCLUDING SAFETY AHQ_EMERQENCY Pac 6 46

- t

~ SYSTEMS

~

REFERENCE VCS IB-3-p. 14 ANSWER 2.15 (1.50)

-A Reactor building atmosphere particulate RMS

-Reactor building sump level Monitoring System

-Reactor building atmosphere gaseous RMS OR the reactor building cooling unit condensate flow rate.

i REFERENCE VCS TS 3/4.4.6 072/000; PWG-8 (3.2/4.0)

ANSWER 2.16 (1.00)

1. Th - Delta-P across the S/G

2. Tc - Delta-P across the RCP [0.5 each]

REFERENCE VCS, AB-2, P. 19 ANSWER 2.17 (1.50)

Bank overlap unit ,

P-A converter Group step counters [0.5 each]

l REFERENCE VCS, IC-5, P. 38 )

I

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

2. , PLANT-DESIGN INCLUDING SAFETY AND EMERGENCY Paca 47

' SYSTEMS 7

' ANSWER 2.18 (1.50)

Valves Mode 1 CL Recirc HL Recirc

a. 8809 O C C

$ b. 8702, 8701 C C C

c. 8811, 8812 C O 'O

d. ~8887. O C- 'O-i

e. 8706 C O O

.; [0.1 each].  ;

REFERENCE VCS, AB-10, PP. 44-4,7 ANSWER 2.19 (1.00)

1. Ensure an equal flow to each branch line.

I

2. Limit total flow to the maximum capacity of one charging pump.

[0.5 ea.]

2 REFERENCE VCS, AB-10, P. 22 1

'i i

l ANSWER 2.20 (1.00) J

'. I

a. Ensure that the RWST and NaOH tanks empty.at the same rate.

b. Ensure even flow distribution to the spray: headers. [0.5 each]

REFERENCE , l l

VCS, AB-8, PP. 8, 13 ~

i

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

i h

+3..

' . 2 .' ' BLANT DESIGN INCLUDING SAFETY AND EMERGENCY Paco 48

' SYSTEMS ANSWER 2.21 (1.00)

a. Forward Flush - Upstream of FIV -> Common S/G header recirculation line upstream of (Valve 1679) -> Dearater or Condenser.

b. Reverse Flush - Downstream of FIV -> Steam generator blowdown system (downstream of blowdown containment isolation valves).

[0.5 ea.]

REFERENCE VCS, TB-7, PP. 22-23 ANSWER 2.22 (1.50) p )17 97

/. 4-. Le Lv-Lv C/Gievel [0.2] -- 2/3 LU 13 - L_G[l5'I 2.1].

-2. "'h, sigr.nl which is une cuiuvidouvo vf: [0.1] -

ja /b h[_gPiessure/Eg.zj - ggLu. Agnocu vyAs g .

/% W llDf87 3*/ ;'3. y}2""TY.M".uv UJWP/ Y"" W '"

• U h mu Ae vv 4 5LU%CD'.M REFERENCE-A

n. L ~ oA A , - e t

,Y L 'i '

E. , A, ,i, . . ,_ _ y '

m VCS, TB-7, P. 22 ( :h ,o-%o--o

,e i 7u i<ve5 L9 2J -

i - ~ , , . - - -t , <g

~/3 u 'J ~ ~" w rn~,y v

%e %

• Ltcy m \ k & t L We a,mca%ce eS : &.&

<g 4 lo fed Irqvab Co.3'] 42a.C [o.( ,

j frlg(f 0. Lo 900 N00 C0'3] <13"k (0.Q 58 V47

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

l

• 3.,. INSTRUMENTS AND CONTROLS

, Paca'49 i

. ANSWER '3.01 (1.25)

Charging flow decreases [0.25], pressurizer level decreases

[0.25], latdown isolates [0.25), and pressurizer level-increases

[0.25]. High Pressurizer leyel trip (92%) [0.25].

REFERENCE ..

VCS IC-3 pp. 38-40 ANSWER 3 . 0,2 (1.00)

REFERENCE

% /f/[M VCS 10-3 pp. 34-36 s

ANSWER 3.03 (1.00) s c

REFERENCE VCS IC-3 pp. 34-36 ,

000/028-K2.02 (2.6/2.7)n\

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

ANSWER 3.04 (1.00) c on N f lll 0

REFERENCE u

VCS IC-4 p. 15 g,. -

's y 'b 4.

y

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

S

.,3 . . INSTRUMENTS AND CONTROLS Paco 50-ANSWER 3.05 (1.50) j$f

[

a. Cabinet has the capacity to support up to 6'e stationary gripper coils simultaneously [0.5]. So with 2 groups or more, would overload / heat the cabinet [0.5]. (This could lead to a rod or rods being dropped).

b. 125 VDC-Latching Rods

-70 VDC-Holding Rods (0.25 ea.)

REFERENCE VCS IC-5'p.26 i 001/050; PWG-1(3.6/4.1)

ANSWER 3.06 (1.00)

a. 1

b. 4

c. 1

d. 3 [0.25 ea.]

REFERENCE t

VCS IC-5 p. 18; Fig. 105.6 ANSWER 3.07 (1.50)

OTdT OPdT

a. Decrease Decrease

b. Increase No change

c. Decrease No change [0.25 each]

REFERENCE VCS IC-6 pp. 18-24 K&A 012000K4.02/IF 3.9

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

-l

3.'. INSTRUMENTS'AND' CONTROLS Para 51' 4

-ANSWER. 3.08 (2.00)-

-a. 1. RIL Programmer.

2. Rod Control System.

3. Steam Dump. [0.25 ea.]-

.b. - 1. - OTdT

2. OPdT

~3. Lo Tavg

4. Lo Lo Tavg '

5. Hi Tava alarm [0.25 ea.]

REFERENCE VCS IC-6 pp. 16, 26 ANSWER 3.09 (1.00) a REFERENCE VCS IC-8 Fig. IC8.6 ANSWER 3.10 .(1.50)

The highest reading upper / lower detector is compared to the~ average of the upper / lower detectors [0.5]~and generates an alarm at 1.02 increasing [0.5]. The circuit is in operation above 50% power on any channel [0.5].

REFERENCE VCS IC-8 p. 42' 015/000; K6.04 (3.1/3.2) & A1.04 (3.5/3.7)

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

1

'*3 . INSTRUMENTS AND CONTROLS Pago 52 ANSWER 3.11 (1.00)

a. INTERMEDIATE and POWER [0.2]

b. SOURCE, INTERMEDIATE and POWER [0.3]

-c. NONE [0.1]

d. INTERMEDIATE and POWER [0.2]

e. INTERMEDIATE and POWER [0.2]

REFERENCE VCS IC-8 p. 12; Figs. 108.9, .12, .13 ANSWER 3.12 (2.00)

a. 1. Single loop loss of flow will insert if 2/4 power ranges are above 38% (P-8). (0.5)

2. Two loop loss of flow will insert lf 2/4 power ranges >10%

[0.2] OR; [0.1] 1/2 impulse pressures > 10% [0.2]

b. The undervoltage trip is to provide a trip signal on loss of power to the RCP's. [0.2] The reactor will trip and RCP's contin-ue to provide coastdown flow. [0.2] The underfrequency trip pro-vides protection for a grid disturbance. [0.2] The RCP breakers and reactor are tripped [0.2] to prevent deceleration and loss of coastdown flow. [0.2]

REFERENCE VCS IC-9 pp. 46, 47

q. /, OR w tl c%se a Lp i7 "Is flu elemenh m one /=p sense.

I4ss A n to % Flow s A powe-- l oce. P 4. 4 [p e)

)' O it w.\\ cwe a ko 1%{$ rf\& e & rd m cach k "P5 seme. le ss b p % (2 at m,+w po+ eboue_ P4. Co3]

91 5] I,l # # ?

b. W il alto acceph 7A e Ploa bpt peuA pAb L loss af4/u a me eb /=p4.C i]

T1'e. 49 a4w kps pnuulunor oqApud,m %I ss s f-Stua In Aowwe. Leos. n.g 7kflM bps cEono+ pwud[pkhkaguidONd/f,Miig, uV,,4, P"A O' g / ff f

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

,3.

. INSTRUMENTS AND CONTROLS Pasa 53' ANSWER 3.13 (1.25)

a. Because the turbine serves as the' heat' sink to the reactor,-

a reactor trip follows a turbine trip to minimize the RCS temperatur fransient (and/or resulting safety valve operation).

(0.5) NSWG gg ;

b. - High~pzr pressure

- High pzr level

- OT Delta T

- S/G lo-lo level [1 required @ 0.25]

c. - All turbine stop valves shut

- Emergency oil (Emer. Trip Fluid. System) pressure low

(< 800 psig) [0.25 each]

REFERENCE VCS IC-9 pp. 58, 59 ANSWER 3.14 (1.00) a REFERENCE VCS IC-9 pp. 52, 80 012/000-K6.10 (3.3/3.5)

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

3. INSTRUMENTS AND CONTROLS Peca 54 ANSWER 3.15 (1.75)

a. 1. SI

2. Rx trip coincident with Low Tavg [0.25 ea.]

b 1. Feedwater pumps trip ,g

2. * '

'- ' cf stai-t up fced r*'=r

• F ,

3. Turbine trip fj/jdP tQf

4. Reactor Trip (>P9) o .:u,

5. FRV and bypass _ valves shut LDc2'each]

c. Protect the Turbine from S/G moisture carryover. [0.25]

REFERENCE VCS IC-8 pp. 57, 50 ANSWER 3.16 (1.00) c REFERENCE VCS IC-5 p. 29 012/000; K2.01 (3.3/3.7) & K6.04 (3.3/3.6) & A4.06 (4.3/4.3)

ANSWER 3.17 (1.25)

a. The analog meters will go to the invalid position for the disabled sensor type, (however, the unit will process the other sensor inputs). [0.25]

b. 1. Green - Margin to saturation greater than both alarm and caution setpoints.

2. Yellow - Margin to saturation between caution and alarm setpoints.

3. Red - Margin to saturation less than alarm setpoints.

[0.25 each]

c. Yes (Accident Monitoring Instrumentation) [0.25]

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

,3- . INSTRUMENTS AND CONTROLS P;gs 55

• l REFERENCE VCS, IC-12, PP. 16-17 ANSWER 3.18 (2.00)

.Th Wide Range RTDs (3)

Tc Wide' Range RTDs (3) 4 MVG-8701 A & B (open)

MVG-8702 A & B (open) [0.5 each]

REFERENCE

~VCS, IC-3, PP. 32-33 ANSWER 3.19 (2.00)

a. 1. "C" Pump Breaker on "A" Train racked up.

2. No T.D. overcurrent, i

r 3. MCB switch in normal condition.

4. "A" Pump breaker racked down OR No ESFLS output 5.present.

, t-

5. SI signal. [0.2 each] - C V ^ c:x [ '" T# D

b. "A" Pump Breaker remains closed.

M wer5.^ W //fff "C" Pump Breaker trips.open. [0.5 each]

REFERENCE VCS, GS-2, PP. 38-40 ANSWER 3.20 (1.00)

)'

Minimize the fluctuations on the control system due to shrink and swell.

I

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

i 4

.3.

. INSTRUMENTS AND CONTROLS- -

.Pca . 56

.: e 4

' REFERENCE ~

VCS, IC-2, P. 11 ANSWER 3.21 (1.50)

Low pressure reactor trip -- 2/3 - 1870 psig. .

Low-pressure safety injection -- 2/3 -- 1850 g.j gg7 P -2/3 1277 psig: 9I cute unblock 2/3 -- 1985 psig.

High pressure reactor trip -- - 2/3 -- 2380 psig. [0.1 each]

REFERENCE VCS, IC-3, PP. 16-17 ANSWER 3.22 (1.50)

a. Upper-range will indicate minimum level,

b. Narrow-range will indicate maximum level.

c. Wide-range will increase from 100% to approximately 110%.

[0.5.each]

REFERENCE VCS, 10-13, PP. 12-13 i

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

j

r- 3

,4 . PROCEDURES - NORMAL,-ABNORMAL, EMERGENCY Paga 57

• - ' AND RADIOLOGICAL CONTROL ANSWER 4.01 (2.00)

1. The RCS is borated to the cold shutdown concentration (and verified by sample).

2. Tavg is 557 F and the Res is borated to the hot shutdown, Xenon free concentration (and verified by sample).

[1.0 ea.]

REFERENCE VCS GOP-1, Appendix A, p. 4 ANSWER 4.02 (1.00) a.

REFERENCE VCS GOP-5 pp. 5-14 ANSWER 4.03 (2.50)

a. 1600 paid

b. 25 degrees F

c. 3000 gpm

d. 3%/hr

e. 350 degrees F [0.5 ea.]

REFERENCE VCS, GOP-Appendix A, Generic Operating Precautions, pp. 2, 3, & 5

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

. . l

.', 4 . PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Peca 58 4

• l

' AND RADIOLOGICAL CONTROL

. ANSWER 4.04- (2.00)

a. Close the RCP A seal leakoff. valve (PVT-8141A). (1.0)

b. '5 min. '(0,5)

c. 30 min. (0.5)

REFERENCE VCS, SOP-101, Reactor Coolant System, pp. 51 &'52 ANSWER 4.05 (1.00)

Removing-the applicable control and instrument power fuses on the power range drawers.

REFERENCE-VCS, SOP-404, Excore NI System, p. 16 015/000-A4.03 (3.8/3.9)

ANSWER 4.06 (1.00) a o " C.

REFERENCE l}//$8Y VCS EOP-10.0 p.14 AN3WER 4.07 (1.50) d.

RF7ERENCE VCS EOP-1.2, pp 4,-5

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

1

• 4.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY P.ca 59

', ' AND RADIOLOGICAL CONTROL ANSWER 4.08' (1.50)

a. 3

b. 5

c. 2

d. 1

e. 6

f. 4 [0.25 each]

REFERENCE VCS SOP 102- , 210-33, 306-52, 401-35, 57 ,

AOP 11.0-1 EOP 1.0 ANSWER 4.09 (1.00) d REFERENCE VCS, EOP-1.3, p. 7 000/009-K3.26 (4.4/4.5)

ANSWER 4.10 (1.00) c REFERENCE VCS, EOP-4.0, pp. 9 - 13

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

,4.' PROCEDURES - NORMAL. ABNORMAL, EMERGENCY Pcc; 60

611D RADIOLOGICAL CONTROL ANSWER 4.11 (1.50)

To prevent excessive depletion of RCS water inventory through a small break in the RCS~which might lead to severe core uncovery if RCP's are not tripped within the proper time frame.

REFERENCE Westinghouse Background Info. "RCP Trip / Restart" 000/009; EK3.23(4.2/4.3)

ANSWER 4.12 (1.00)

Turbine is tripped so that the heat sink ( S/G levels) will be maintained as long as possible [0.5] on a total loss of feedwater ATWS [0.5].

REFERENCE Westinghouse Background Info on FR-S.1, pp 7, 76/77 EPE 029; EK3.12(4.4/4.7)

ANSWER 4.13 (1.00)

After Immediate Operator Actions 4of any other EOP are comp 1 ted.

REFERENCE (o- dem3 %6l EoPwage) $ (( / /I[?

EOP 12.0 p. 2 PWG-11: Performing Immediate Actions (4.3/4.4)

ANSWER 4.14 (2.50)

a. 1. 100 REM A o "'"P+ T R Ih O// 1 4W 54

2. 25 REM [0.5 each] ),g , %j.7

b. Increased liklihood of cancer, particularly 1 kemia. Short term somatic effects include blood changes. [1.0]

c. Emergency Director (or authority as delegated by the ED) [0.5]

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

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• 4.

, *FROCEDURES - NORMAL, ABNORMAL, EMERGENCY Pcsa 591

. ' AND RADIOLOGICAL CONTROL REFERENCE

'VCS HP Manual pp. 5-13, 14; 4-12,15 RPP-ot\ ; p.\ ) 6??-030 p. % 3g j j.V l ANSWER 4.15 (2,50)

a. 1. 1.25 R/qtr.

2. 18.75 R/qtr.

3. 7.5 R/qtr. .[0.25 ea.]

b. .1. 1 R/qtr.

2. 12 R/qtr.

c.

$s-f?/g% Y 2 R/qtr not to exceed 5 R/yr.

f %Mw  ?.pg@f M ' #y-

. Plant "-- cer [0.5 ea.] -

O m %~- % d<ur RmA C P *ed*M /

. REFERENCE 9['//j///7

/

VCS HP Manual pp. 5-7; 5-13 t4 PP- /r2 p .1 awe. M6 ad . I YI ///f #7 ANSWER 4.16 (2.00)

1. Source Range Excores

2. Intermediate Range Excores

3. Containment Pressure (High-3)

4. RWST Levels 5.

REFERENCE 4/so (, , Qc,,5 Emergency Feedwater Ou3 krdec Suction Pressure Low. h[any 4 9 0.5 g, u n ii (%de-udtge.)

VCS, SOP-401, P. 4 g, Q& ( g q79_

%. >\ i\  % 'I

,j,flg7

/0. b un n.e r hre.s.5%re b M

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

?,4 . . PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pra 62

,' AN_D RADIOLOGICAL CONTROL ANSWER 4.17 (1.50)

a. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> [0.5]

b. 1. Bell System Land Line

2. Relay message (via dispatcher, LLEA, or other available method). [0.5 ea.]

REFERENCE VCS, SI 86-02, P. 5 ANSWER 4.18 (1.5,0) 4ttri 44.- ov y g jjf

a. A qualiflod Dauger Taggcr with a current RO or SRO license.

[1.0]

b. FALSE (Must be an Electrical Maintenance person). [0.5]

REFERENCE VCS SAP-201 pp. 9, 5 ANSWER 4.19 (1.00)

EOP 1.0 ReactorTrip/ Safety Injection Actuation EOP 12.0 Monitoring of Critical Safety Functions [0.5 ea.]

REFERENCE VCS EOP 13.0 p. 1 ANSWER 4.20 (1.00) c.

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

'r 4 e -~.FROCEDURES - NORMAL, ABNORMAL, EMERGENCY P c3 63.

, Y AND RADIOLOGICAL CONTROL 6

REFERENCE VCS Plant System Descriptions, IB-5, p. 32' i

(***** END OF CATEGORY 4 *****) .-

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

u_______________________________---__ ____

l, U. S. NUCLEAR REGULATORY COMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: SUMMen REACTOR TYPE: PWR-WEC3 DATE ADMINSTERED: 88/11/17 EXAMINER: JAGGAR. F.

CANDIDATE k / A CTM A ^;m ,

INSTRUCTIONS TO CANQIDATE: N

• U* ' V Vi j Use separate paper for the answers. Write answers on one side only.

Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing crade 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.

X OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 29.50 24.69 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS,AND THERMODYNAMICS 30.00 25.10 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 30.00 25.10 7. PROCEDURES - NOIDIAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 30.00 25.10 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 119.5 Totals All work done on this examination is my own. I have neither given nor received aid.

Candidate's Signature N

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

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

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

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

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

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

6. Use only the paper provided for answers.

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

8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer sheet.

9. Number each answer as to category and number, for example, 1.4, 6.3.

10. Skip at least three lines between each answer.

11. Separate answer sheets from pad and place finished answer sheeta face down on your desk or table.

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

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

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

15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

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

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

[.

18.-When you complete your examination, you shall: ,

a. Assemble your examination as follows:

(1) Exam questions on top.

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

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

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

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

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

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5. THEORY OF NUCLEAR POWER PLANT OPERATION. P20a 4 ,

FLUIDS.AND THERMODYNAMICS QUESTION 5.01 (1.50)

You have just completed a reactor startup and power level is at the point of adding heat. For the following situations, indicate if final stable power level will be HIGHER THAN, LOWER THAN, or THE SAME AS the power level before the situation occurred. Assume core is at mid-life. Treat each situation separately.

a. Steam dump pressure setting is raised by 20 psig while dumping steam.

b. A 1% steam leak develops outside of containment.

c. An inadvertant 20 ppm bo) a addition is made.

QUESTION 5.02 (1.00)

Which ONE of the below choices does a sufficient Shutdown Margin NOT ensure?

a. The reactor can be made suberitical from all operating conditions,

b. The reacter can be made critical with the Reactor Coolant System average temperature greater than 551 F.

c. The reactivity transients associated with postulated accident conditions are controllable within acceptable limits,

d. The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

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

5. THEORY OF NUCLEAR POWER PLANT OPERATION,_ P2ca 5 4 FLUIDS,AND THERMODYNAMICS l'

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. QUESTION 5.03 (2.25) l For the following conditions, will the CALCULATED ECP for a startup performed 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after a trip from a 60-day.

100 % power run,-be-HIGHER THAN, LOWER THAN, or the SAME'as-1 the ACTUAL control rod position at criticality.

I Treat each condition separately. Briefly explain your answers.

I a. BOL Rod Worth Curves were incorrectly used to. calculate ECP when EOL conditions exist.

-b. The startup is delayed TWO hours.

3

c. One reactor coolant pump is stopped three minutes i prior to criticality.

i I QUESTION 5.04 (1.00) i i During natural circulation cooldown, pressurizer level. increases

abnormally subsequent to the initiation of spray to. reduce pressure.

Explain what is occurring.

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) QUESTION 5.05 (1.00) l Assume a small LOCA results from the rupture of a pressuriser level

transmitter sensing line. Compare the severity of the accident (in '

l terms of mass loss) if the ruptured line is the upper (reference) or lower (variable) sensing line.

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QUESTION 5.06 (2.00) l i How and why is fuel centerline temperature affected by the following?

l Consider each independently:

i

.; a. Corrosion and crud buildup on the fuel cladding.

! b. Fuel densification.

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

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5. THEORY OF NUC MAR POWER PLANT OPERATION. Pqo 6 FLUIDS.AND THERMODYNAMICS

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QUESTION 5.07 (1.00)

The reactor is at equilibrium Xenon. Boron is 890 ppa, rods are.in manual with Tavg on program and the turbine is loaded to 475 MW.

Turbine load suddenly reduces to 360 MW, but the steam dump fails to activate. Assuming no protective action occurs, what will be the new steady state reactor power and Tavg?

a. Reactor power 50%, Tavg 587 F

b. Reactor power 50%, Tavg 594 F

, c. Reactor power 38%, Tava 587 F

d. Reactor power 38%, Tavg 594 F QUESTION 5.08 (1.75)

a. Explain why fission product gas build-up in the gap between,the fuel and clad causes Doppler (power only) Coefficient to become more negative over the life of the core. (1.0)

. b. Does the effect of " clad creep" cause the Doppler (power only)

Coefficient to become HORE or LESS negative, over the life of the core? No explanation required. (0.75) i QUESTION 5.09 (1.50)

I 1

I The reactor is operating at 100% power with all rods out, near EOL with equilibrium Xenon conditions when power is to be' reduced to 50%.

The operator observes that AFD is within its band and decides to lower i

power by borating, leaving rods fully withdrawn. Actual Tavg follows programmed Tavg. Describe the change that will occur in AFD and why it occurs, prior to changes in Xenon having a noticeable effect.

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5. LTHEORY OF NUCLEAR POWER PLANT OPERATION, Paga 7 FLUIDFuAND THERMODYNAMICS QUESTION 5.10 (1.50)

a. Does Beta Effective Increase, Decrease, or Remain the Same, from BOL to EOL? EXPLAIN YOUR CHOICE. (1.0)

b. For TWO equivalent positive reactivity additions to a critical reactor, will the SUR be the Same, Larger, or Smaller at EOL as compared to BOL? NO EXPLANATION IS NECESSARY. (0.5)

QUESTION 5.11 (3.00)

a. Determine the amount of reactivity (in pcm) required to increase Keff from 0.97 to 0.985.

b. By what factor will the COUNT RATE change as a result of increasing Keff from 0.97 to 0.985?

c. What would be the condition of the reactor after increasing Keff from 0.97 to 0.985, if the same amount of reactivity were added again?

QUESTION 5.12 (1.50)

TRUE OR FALSE

a. Rod Worth increases significantly as power increases because Rod Worth is proportional to flux and flux is proportional to power.

b. Rod Worth is higher at lower temperatures because fewer neutrons leak from the core and more neutrons are present in the moderator.

c. The HZP Total Rod Worth increases over core life due to less competition being offered by soluble boron and by the burnable poison rods.

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

5. THEORY OF NUCLEAR POWER PLANT OPERATION. Peca 8 FLUIDS.AHD THERMODYNAMICS QUESTION 5.13 (1.50)

Estimate the change in Tavg resulting from an inadvertant dilution of 10 ppm boron at EOL with rods in MANUAL and the turbine in AUTO.

State any assumptions of values used.

QUESTION 5.14 (1.00)

When performing a reactor S/U to full power that commenced five hours after a trip from full power equilibrium conditions, a 0.5%/ min ramp was used. How would the resulting Xenon transient vary if instead a 2%/ min ramp was used?

a. The Xenon dip for the 2%/ min ramp would occur SOONER and the magnitude of the dip would be SMALLER,

b. The Xenon dip for the 2%/ min ramp would occur LATER and the magnitude of the dip would be SMALLER.

c. The Xenon dip for the 2%/ min ramp would occur SOONER and the magnitude of the dip would be LARGER.

d. The Xenon dip for the 2%/ min ramp would occur LATER and the magnitude of the dip would be LARGER.

QUESTION 5.15 (1.50)

Indicate TRUE OR FALSE for the following statements concerning Xenon behavior following a reactor trip.

a. Xenon peaks later if the reactor trip occurs at high power as compared to low power,

b. The Xenon concentration decreases following the peak because the half-life of Xenon is shorter than the half-life of Iodine.

c. If Xenon and Iodine had the same half-life value, Xenon would still peak following a reactor trip.

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

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, Pcca 9 FLUIDS,AND-THERMODYNAMICS QUESTION 5.16 (0.50)

State whether the situations below will generate the greater tensile stress on the INNER or the OUTER wall of the reactor vessel.

Consider each situation seperately,

a. Heatup at a rate of 80 degrees F/hr b.- Increasing pressure 250 psig QUESTION 5.17 (1.50)

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

, a. The amount of heat required to change i lba of water into 1 lba of steam at a constant temperature.

4 l b. The ratio of the Critical Heat Flux to the actual heat flux.

1

c. The maximum local heat flux at core elevation (z) divided by average fuel rod heat flux.

i QUESTION 5.18 (1.50)

a. After operating at 100% power for three months, power is suddenly lost to all of the reactor coolant pumps. Below are three operations that will enhance natural circulation. Why is each done?

! 1. Pressurizer level should be maintained at 50% or greater

! 2. Maintain at least 15 F subcooling in RCS-

3. Maintain heat sink

b. Briefly explain how the following parameters will be trending if natural circulation in LOST:

1

1. RCS differential temperature

2. Steam generator steam pressure

3. Steam generator level (assume constant AFW flow) i l

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

5. -THEORY OF NUCMAR POWER' PLANT OPERATION. Posa 10 FLUIDS,AND.THERMODYNAMIQS 1

. QUESTION 5.19 (2.00)

a. A variable speed centrifugal pump is operating at 1/4 rated speed-in a CLOSED system with the following parameters:

Power = 300 KW i Pump delta P = 50_ paid

. Flow = 880 rpm

-What are the new values for these parameters when the pump speed

- is increased to full rated speed? (0.75)

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b. Choose the answer that most correctly completes the sentence.

ll "In a CLOSED system, two single' stage centrifugal pumps operating in parallel will have ----(choose from below)--- , as compared to the same system with one single stage centrifugal Pump operating with one pump isolated i j 1. a higher head and higher flow rate.

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2. the same head and the same flow rate.

3. the same head and a higher flow rate.

4. a higher head and the same flow rate. (1.0)

$ c. How is the available NPSH to a centrifugal pump affected by an i increase in system flowrate? Assume NO change in pump speed nor  !

i pump configuration. (0.25) i

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5. THEORY OF NUCLEAR POWER PLANT OPERATION,. Paco 11 FLUIDS,AND THERMODYNAMICS QUESTION 5.20 (1.00)

Which of the following best describes the parameter changes that occur across a centrifugal pump in a closed system?

a. Temperature INCREACES, Enthalpy INCREASES.

. b. Temperature INCREASES, Enthalpy DECREASES.

c. Temperature CONSTANT, Enthalpy CONSTANT.

d. Temperature CONSTANT, Enthalpy INCREASES.

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(***** END OF CATEGORY 5 *****) l t

.6. ' PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Pra 12

,s QUESTION 6.01 (2.00) ,

Indicate whether the following valves are normally OPEN or CLOSED with the reactor operating in Mode 1 at 100% power.

2 a. RWST-to-RHR pump suction valves (MVG-8809A & B) t-

b. CCW-to-RHR HX isolation valves (MVB-9503A & B).

c. RHR system train cross-connect valves (MVG-8887A & B) '

d. RHR bypass flow control valves (FCV-605A & B)

I QUESTION 6.02 (1.00)

The purpose of the interlock that prevents the' letdown isolation ,

valves from opening or shutting unless all three orifice isolation valves are shut is to prevent: (Choose one of,the following) 1

a. exceeding design flow rates of the demineralizers.

I b. excessive heatup rates across the regen heat exchanger.

! c. excessive pressure on the shell of the regen. heat exchanger.

d. unnecessary lifting of relief valves downstream of orifices.

I QUESTION 6.03 (1.00)  ?.

Which one of the following will result in the highest gain produced by the Variable Gain Unit of the Rod Control System?

) a. A LOWER power level sensed by NI-44 i s

b. A LOWER power level sensed by Turbine Impulse Pressure

c. A HIGHER power mismatch error signal )

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d. A LOWER power mismatch error signal i l

e. A HIGHER (Tref-Tava) error signal

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6. PLANT SYSTEMS' DESIGN, CONTROL, AND INSTRUMENTATION Pass 13 QUESTION 6.04 (1.00)

Match-the following conditions with the expected indication provided.by the rod speed indication meter.'The selections may be used more than once.

1. Immediately before an operator removes-the N44 a. O s/m.

fuses because of a failed high detector. Rods in AUTO with no temperature mismatch. b.;8 s/m.

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2. Rods in MANUAL with a 10 F temperature c. 40 s/m.

mismatch.

4 d.~48 s/m.

! 3. Rods in AUTO with a 1 F temperature mismatch.

i e. 72 s/m.

4.. Rods in AUTO-with a 4 F temperature mismatch.

f. 88 s/m.

. 5. Rods in AUTO with one of the Tave control instruments failed low.

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j QUESTION 6.05 (1.50)

Indicate which, if any, of the Excore Nuclear Instrumentation Ranges I (SOURCE, INTERMEDIATE, POWER or NONE), will correctly match with

! the following statements. More than one may apply to each.

}

a. Uses an opposing current technique to compensate for-gamma radiation.

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l. b. Operates in the Proportional region of the Gas-filled. Detector j curve.

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! c. Detector operation is unaffected by gamma radiation.

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d. Covers eight (8) decades of neutron flux.

l j e. Covers twelve (12) decades of neutron flux.

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l f. Shares a common instrument thimble but only covers the lower core area.

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

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6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PcG2 14 s \

s QUESTION 6.06 (2.00) ,

i The following concern RCS instrumentation.

a. Wide Range temperature instruments have a control. input to the Cold Overpressure' Protection System (COPS). Why are the Narrow Range I temperature instruments NOT used in the COPS? [0.6]

i

b. List the FIVs protection / alarm signals that are generated by the Tavg signal AND the TWO protection TRIP signals generated by the j dT signal. [1.4]

1 4 QUESTION 6.07 (2.00) -

During operation at 50% power the CONTROL.RTD in the cold leg of loop one malfunctions and its signal output INCREASES by 6 F. . Explain the effects on the following indications, parameters or control systems independently. Assume no operator actions and all control systems are in automatic,

a. Rod Control

b. Loop #1, OT deltaT setpoint Ni c. Rod insertio,n limits setpoint

d. Pressurizer level 3

QUESTION 6.08 (1.75)

a. List the FOUR plant parameter input signals to"the Overtemperature Delta-T (OTdT) protection circuit. (1.0)

b. What core protection is provided by the OTdT proSect[De '

circuit? (0.25) 4

c. State TWO additional functions'(control / protection) other than reactor trip that the OTdT protection channel provides. (0.5) i 1

(***** CATEGORY 6 CONTIN'UEDONNEXTPAGE*****h b-Wg .,

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6& PLANT SYSTEMS DESIGN. CONTROL, AND INSTRUMENTATION Paca 15 QUESTION 6.09 (2.50)

a. What are three reasons, as stated in Tech. Specs., for having Rod Insertion Limits? [0.75]

b. With respect to the Rod Insertion Limits, "...the steamline break accident imposes the highest shutdown margin requirement." Explain why this is a true statement. [1.00]

c. Considering each of the following sets of conditions separately, which condition will make the Steamline Break accident worse? [0.75]

1. BOL or EOL

2. Reactor shutdown or at 100% power

3. Tavg at 350 F or at 547 F QUESTION 6.10 (2.00)

Explain the TWO methods which maintain the diesel near operating temperature while the diesel is idle.

QUESTION 6.11 (1.00)

State the motive force for the RCS through the following:

1. Th RTD manifold.

2. Tc RTD manifold.

QUESTION 6.12 (1.50)

Of the signals used to automatically close the Feedwater Isolation Valves (1611 A, B, C), which TWO are specifically designed to prevent water hammer in the feedwater piping and steam generator inlet connections. Include setpoints and coincidence (logic) where applicable.

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

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6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Pcco 16 QUESTION 6.13 (1.00)

As containment temperature increases from normal operating temperature, what will be the relationship between actual and indicated pressurizer level? Select the most correct answer.

a. Indicated level reads higher than actual level.

b. Indicated level increases regardless of actual level.

c. Indicated level and actual level are the same.

d. Indicated level reads lower than actual level.

QUESTION 6.14 (1.00)

Which one of the following statements about the Digital Rod Position Indication (DRPI) System is correct?

a. The DRPI system uses a series of magnetic switches to determine rod position.

b. Anytime rods within a group differ by more than 12 steps a ROD DEVIATION alarm is generated.

c. A control bank B rod on the bottom will not generate a RPI ROD AT BOTTOM alarm unless another bank B rod or a control bank C or D rod indicates at least 12 steps.

d. Power to the DRPI system is supplied from the control rod MG sets.

QUESTION 6.15 (1.00)

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

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

b. Detector, Log Integrator, Pulse Shaper, Pulse Counter, Meter

c. Detector, Pre Amp, Log Integrator, Discriminator, Meter

d. Detector, Log Amp, Mater l

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

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6 PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION Pcca 17 I

QUESTION 6.16 (2.00)

a. Describe the instrument coincidence and setpoints necessary to insert the Reactor Coolant low flow trips into the protection system for 1 AND 2 loop loss of flow.

i b. Explain how and why the undervoltage AND underfrequency low flow reactor trips operate differently.

d t (1.25)

QUESTION 6.17 The Reactor Protection System is designed so that a turbine i trip will cause a Reactor Trip above P-9 (50% power).

a. Why is the Reactor Protection System designed to do this? (0.5)

b. Provide a Reactor Protection signal that would act to give protection in the event that the Turbine Trip / Reactor Trip-did not operate on a turbine trip from full power. (0.25) t

c. State the TWO ways that the Reactor Protection System senses I

that a turbine trip has occured? (0,5) i i QUESTION 6.18 (2.00)

a. State the five signals / conditions that will ALLOW the "C" Charging Pump to start Automatically. Assume transfer switch aligned to "A" Train.

b. State the response of-both "A" and "C" Pump breakers <if both "A" and "C" Pumps are running on the "A" train and a blackout is received on that train.

i QUESTION 6.19 (1.00)

What is,the purpose of the delay in the time it takes for the actual Steam Generator Level signal to reach the PI controller where it is compared to Program Level?

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6. PLANT SYSTEMS DESIGN. CONTROL, AND INSTRUMENTATION Paco 18 l l

l QUESTION 6.20 (1.50)

The following pertain to indications on the Reactor Vessel Level Indicating System.

a. How will the upper-range indication respond when a RCP is started in the associated loop?

b. What will the narrow-range indication show when any RCP is running?

c. How does wide-range indication change as reactor power is increased from 0 - 100%?

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(***** END OF CATEGORY 6 *****)

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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Pero 19 AND RADIOLOGICAL CONTROL QUESTION 7.01 (1.00)

Which one of the following is NOT a requirement for using EOP-1.4 "Rediagnosis".

a. SI is in service or required.

b. Operator has determined rediagnosis is necessary.

c. Immediate Operator Actions for EOP-13 " Response to Abnormal Nuclear Power Generation" have been completed.

d. Immediate Operator Actions of EOP-1.0 have been completed.

QUESTION 7.02 (1.00)

When starting a Reactor Coolant Pump, WHAT is the:

a. minimum RCS Loop Pressure?

b. minimum seal leakage across #1 Seal?

c. minimum pressure in VCT7

d. maximum number of starts in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />?

QUESTION  ?.03 (1.50)

The following pertain to the Component Cooling Water System.

a. When is high speed operation of the Component Cooling Pumps permitted? [0.5]

b. What is the maximum CCW Heat Exchanger outlet temperature permitted? [0.25]

c. Explain why an out-of-service CCW pump breaker must be racked out. [0.75]

(***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)

7-

7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pcco 20 AND RADIOLOGICAL CONTROL QUESTION 7.04 (1.00)

The following pertain to precautions as stated in SOP-102, CVCS.

a. At what decreasing RCS pressure is the " SEAL WATER RETURN ISOLATION" (MVT-8100) closed?

b. What action is required if letdown flow is lost during Mode 3 operation?

QUESTION 7.05 (1.00)

After a manual start of Diesel Generator B (DG B) and, with the Diesel Generator carrying BUS 1DB, why is it necessary to synchonize the DG with either the normal or alternate power source prior to shutting down the diesel?

QUESTION 7.06 (1.00)

State the FOUR Immediate Operator Actions as listed in EOP-7.0.

Potential Fuel Assembly Damage.

QUESTION 7.07 (2.00)

Prior to increasing Tavg from. mode 5, your heatup procedure (GOP-1) gives the option NOT to withdraw shutdown banks if either of TWO conditions exist. State the TWO conditions.

.)

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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Pcco 21 AND RADIOLOGICAL CONTROL QUESTION 7.08 (1.00)

Which one of the following statements describing the method of unit shutdown from Mode 2 to Mode 3 is correct?

a. Using manual rod control,. insert control banks D, C, B, and A to ZERO steps. Maintain the shutdown banks fully withdrawn.

b. Using manual rod control, insert control banks D, C, B, and A to FIVE steps. Maintain the shutdown banks fully withdrawn.

c. Using manual rod control, insert control banks D, C, B, and A to zero steps. Using group select, insert shutdown banks to zero steps. Open reactor trip breakers. Reset reactor trip breakers. Using group select, fully with-draw shutdown banks.

d. Using group select, insert all rods to zero steps. Open reactor trip breakers. Reset reactor trip breakers.

Using group select, fully withdraw shutdown banks.

QUESTION 7.09 (2.50)

Indicate the numerical value(s) associated'with the following precautions as stated in GOP-Appendix A.

a. Maximum differential pressure between RCS and S/G.

b. Maximum differential temperature between RCS loops.

c. Minimum RCS flowrate during RCS dilutions,

d. Maximum RATE of power increase above 20% reactor power with-out management approval.

e. Maximum RCS temperature during RHR operations.

(***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)

7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY' Paca 22 ,

AND RADIOLOGICAL CONTROL QUESTION 7.10 (2.00)

a. Indicate the Immediate Corrective Action (s) required, if while operating at 50% power, the following alarms occur simultaneously.

"RCP A #1 SL LKOFF FLO HI/LO" and "RCP A #1 SL dP LO"

b. How much time is allotted to take the above Immediate Corrective Action?

c. For how long may the RCP be operated with'a #1 Seal Failure?

QUESTION 7.11 (1.00)

Briefly describe how the trip bistables of a failed power range detector are placed in the trip condition.

i .

QUESTION 7.12 (1.50) 1 Which of the following conditions would NOT require re-initiation of i safety injection according to EOP-1.2 " Safety Injection Termination"?

RCS PRESSURE SUBCOOLING PZR LEVEL %

a. Stable 25 15

b. Increasing 40 3

c. Decreasing 30 10

d. Stable 35 5

(***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)

7. PROCEDURES - NORMAL,- ABNORMAL, EMERGENCY Pcca 23 AND RADIOLOGICAL CONTROL QUESTION 7.13 (1.50)

Match the condition in column A to the action required in column B.

Column B responses are NOT used more than once.

COLUMN A COLUMN B

a. Shutdown Margin in Question. 1. Reduce plant load.

b. 2/3 lo Steamline press. < 675# 2. Reset mech. over-speed.

c. Emergency Deisel trip during SI

3. Emergency borate.

d. Feed booster pump trip

4. Place Rods in

e. PZR pressure control channel fails manual.

High.

5. Safety inject.

f. First stage turbine pressure fails High. 6. Close PCV-444.

7. Reactor trip.

8. Reset SI

[0.25 each]

QUESTION 7.14 (1.00)

Which one of the following statements describes the cooldown/

depressurization SEQUENCE of EOP-4.0, " Steam Generator Tube Rupture"?

a. Commence rapid cooldown; depressurize maintaining subcooling simultaneously with cooldown,

b. Cooldown to 520 degrees; depressurize to 1500 psig; complete cooldown; complete depressurization.

c. Cooldown to required temperature; depressurize to required pressure once cooldown is completed.

d. Commence rapid cooldown and depressurization without limits on subcooling.

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7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Paco 24 AND RADIOLOGICAL CONTROL QUESTION 7.15 (1.50)

During a small break LOCA (SBLOCA), it is required to trip the RCP if the trip criteria are met. If forced flow through the core promotes cooling, why are the RCPs tripped?

QUESTION 7.16 (2.50)

During a serious emergency, operators may be called upon to assist in search and rescue or recovery operations in the plant.

a. In such cases, according to Health Physics manual, what dose could you receive:

1. To bring an injured worker to safety? [0.5]

2. To get ESF equipment running if the core was uncovered? [0.5]

b. What are the possible effects of receiving radiation exposures of-50 REM? Include short and long term effects. [1.0]

c. Who must authorize this voluntary radiation exposure up to the emergency limits? [0.5]

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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Pers 25 AND RADIOLOGICAL-CONTROL QUESTION 7.17 _(2.50)

a. State the legal' (10CFR20) quarterly exposure limits for the following.

1. Whole Body.

2. Extremities.

3. Skin. [0.25 ea.]

b. State the V.C. Summer Administrative Exposure Guidlines for the following.

i -1. Whole Body.

1 2. Extremities.

1 i 3. Skin. [0.25 ea.]

c. To what value can the Administrative Guideline for whole body be raised to and who (by title) must approve this limit increase?

(1.0)

QUESTION 7.18 (2.00)

According to SOP-401, Reactor Protection and Control System, list FOUR of the FIVE Instrument Channels that need not be placed in the trip mode when they are dealared to be out-of-service.

1 QUESTION 7.19 (1.50)

a. Per Special Instruction 8602, what is the time limit for notifying the NRC Emergency Operations Center of an inoperable NRC Control Room Hotline?

b. State the TWO methods that may be used for-the notification listed in "a." above.

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7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pcco 26 AND RADIOLOGICAL CONTROL QUESTION 7.20 (1.00)

During the recovery phase of a plant Blackout the normal off-site power supplies are placed back in service. Which one of the'below listed actions must the operator take to insert speed droop

-control in the diesel governor system before paralleling the power sources?

a. Depress Emergency Start Reset and select parallel operation on the Voltage Regulator switch,

b. Select parallel operation on the Voltage Regulator switch and turn on the synchronizing scope,

c. Depress Emergency Start Reset and Test Start pushbottons,

d. Reset Blackout sequencer and depress the Exciter Reset switch.

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(***** END OF CATEGORY 7 *****)

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, Paco 27 AND LIMITATIONS -

QUFGTION 8.01 (3.00)

According.to EPP-001, Activation and Implementation of Emergency Plan, what is the criterion for breaching each of the following fission product barriers?

1. Reactor Coolant System.

2. Fuel Cladding.

3. Containment System.

QUESTION 8.02 (1.00)

a. State the THREE initial notifications, as listed in EPP-002, Communication and Notification, that must be made by the Shift Supervisor upon declaration of any emergency classification.

Alternates not required.

b. Within what time limit must the assigned communicator notify the state and local authorities?

QUESTION 8.03 (1.00)

a. Who is responsible for classifying emergencies based on the number of fission product barriers breached.

b. If the individual in part a. is unavailable, who assumes this responsibility?

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(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)

8. ADMINISTRATIVE PROCEDURES. CONDITIONS, Paco 28 ,

AND LIMITATIONS QUESTION 8.04 (1.50)

Answer TRUE or FALSE to the following.

a. Entry into an Operational Mode may be performed even if the conditions for the Limiting Condition for Operation (LCO) are NOT met provided the ACTION requirements are subsequently satisfactorily completed.

b. If a LCO is NOT met and the ACTION statements are NOT appli-caple, then the Senior Reactor Operator has the authority to disregard that particular LCO.

c. Failure to complete a Surveillance Requirement on operable equipment within the specified time interval (plus any allow-able extension) shall constitute a failure of the component to meet its operability requirements.

QUESTION 8.05 (1.00)

Which one of the following is the basis for the high pressurizer water level reactor trip?

a. Prevents solid operations while the reactor is critical.

b. Prevents exceeding containment design pressure in event of a LOCA with all RCS fluid flashing to steam.

c. Prevents loss of pressure control due to spray nozzle being submerged.

d. Protects the pressurizer safety valves against water relief.

QUESTION 8.06 (1.00)

When a system is determined to be inoperable solely because its normal power supply is inoperable, it may be considered operable for the pur-pose of satisfying the requirement of its applicable LCO provided two conditions are met. List these TWO conditions.

(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, Pega 29 ,

AND LIMITATIONS j QUESTION 8.07 (3.50)

a. What is the MINIMUM number of operable excore channels indicating AFD outside the target band before AFD is considered outside its target band by Technical Specifications? (0.5)

b. Assume the plant is operating at full power and the Axial Flux Difference (AFD) has been outside the target band for the last 5 minutes. What are the TWO actions specified which you may choose between to meet the Technical Specification requirements? Include time limitations. (1.0)

c. Assume that it is 0310 on 05/13/86 and the plant is presently at 45% power. Considering the AFD penalty history below,-at what date and time may power be increased above 50%? EXPLAIN.

(Show all work.) Assume no deviation outside the band after 0310 on 05/13/85.

TIME WENT OUT TIME BACK OF BAND IN BAND POWER

^

DATE 05/12/86 0310 0318 85%

-05/12/86 1557 1837 65%

05/13/86 0148 0310 45% (2.0)

QUESTION 8.08 (2.50)

The RCS is heating up at 50 F per hour with the RCS presently at 325 F. Maintenance reports that Charging Pump B repairs will not be completed for one hour but that Charging Pump A is operable.

Technical Specifications Action Statement allows 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to repair an inoperable pump in Mode 3. Assume Charging Pump C is inoperable.

What action, if any, should be taken?

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8. 1 ADMINISTRATIVE PROCEDURES. CONDITIONS. Peo 30 AND LIMITATIONS QUESTION- 8.09 (1~50)

The' concentration of the boric acid. solution in-the Refueling Water Storage Tank.(RWST) shall.be verified once per 7 days in accordance with Technical Specification 3.5.4. LThe chemist sampled the RWST on the'following schedule. '(All samples taken at 1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br />.)

April-1 ----April 8 --- April 161--- April 24 ----April 31

a. EXPLAIN why or why not surveillance time interval requirements were exceeded on April.16. (0.75) -

b. EXPLAIN.why or why not surveillance time interval requirements

were-exceeded on April 24. (0.75) i QUESTION 8.10 (2.00)

The plant is operating at 50% load and you are'the Shift Super-

} Explain what actions, if any, per Technical '

visor. .

i Specifications (TS), would need to be taken for the following-

, conditions. s CONSIDER EACH CASE SEPARATELY. State all possible (TS) options and assume minimum manning requirements currently exist.

a. The BOP operator breaks his leg in'the plant and you send i him'to the hospital for treatment.

! b. The on-coming STA calls and says he can not make it for his assigned shift.

l QUESTION 8.11 (1.50) 4 During Mode 3. operation, the VCT outlet valve (LCV-115 C&E) motors are found to be faulty. The motor breakers are-racked out and the valves

! are verified to be open. Can Mode 3 operation continue? Explain your response.

)!

QUESTION 8.12- (1.00) i i Under what conditions (Modes) may the " Operator at the Controls" '

acknowledge an alarm on the HVAC Control Board?

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'8. ADMINISTRATIVE PROCEDURES, CONDITIONS, HPeca 31 1 AND LIMITATIONS _ l l

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QUESTION 8.13 (1.25)

The following pertain to ,rocedure usage as stated in SAP-139, Procedure Development, Rev '.ew, Approval and Control.

a. How does the Shift Supervisor mark the non-applicable steps for operating procedures that must be entered with initial conditions other than those specified in the procedure? [0.25]

b. For how long is a Temporary Approval of a temporary change l to an existing procedure valid? [0.25]

c. List the THREE items the Shift Supervisor is denoting

~

approval of when he signsSection IV of the PDF-A (Temporary Approval of a change to an existing

! procedure). [0.75]

4

! QUESTION 8.14 (1.25)

The following pertain to SAP-132, Off-Normal Occurrence Evaluation, Reporting, and Resolution.

i

a. List the FOUR personnel (by title) that must be in agreement as to the cause of a Reactor Trip prior to restart and return to power. [1.0]

b. If the cause of the trip cannot be determined whose permission must be obtained prior to entry into Mode 27 [0.25]

QUES :ON 8.15 (1.25)

a. List the Critical Safety Functions (CSF) in order of importance from most to least important. [0.6]

b. State the FOUR fission product barriers whose integrity is assured against radiation release by monitoring the CSF. [0.4]

c. Which CSF solid path condition DOES NOT mean that the associated fission product barrier is to be assummed to be breached. [0.25]

(***** CATEGORY 8 CONTINUED ON NEXT FAGE *****)

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, Peca 32 AND LIMITATIONS 4 QUESTION 8.16 (1.50)

The following pertain to Technical Specification Action Statements in effect when one power range excore nuclear instrument channel is removed from service.

a. Within what time limit must the inoperable channel be placed in the trip condition?

b. TRUE or FALSE?

Another channel (other than the one removed from service) may be removed from service for surveillance testing if the inoperable channel is first bypassed.

c. What action is required if it is NOT desired to restrict Thermal Power or lower the Power Range neutron flux trip setpoints?

QUESTION 8.17 (1.00)

According to EOP 12.0, Monitoring of Critical Safety Functions, when does monitoring of the Critical Safety Functions begin?

QUESTION 8.18 (1.75) 4 a. What is the objective of the Reactor Core Safety Limits in the Technical Specifications?

b. What are the Shift Supervisors notification responsibilities if a Safety Limit is violated?

(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)

1

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, paco 33',

'AND LIMITATIONS QUESTION 8.19 (1.50)

a. What qualifications are required.to perform the second verification of tag placement, accuracy and equipment position? [1.0]

4

b. TRUE or FALSE 7 Under certain. circumstances during electrical maintenance requiring intermittent operation of circuits.which preclude 4 installation of a Danger Tag, it is permissible for any qualified danger tagger to be stationed at the electrical

device in lieu of a Danger Tag. [0.5]

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

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

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EQUATION SHEET

~.

f = ma v = s/t

,'= ms s=vt+

ac 2 Cycle efficiency = * ' ~

E r E = aC -

e = (vg - v,) / c KE = hav v +rt A = AN A = A,e f=va pg = msh to - e/t A = in 2/tg = 0.693/tg ,

W = VAP- . . . .

AE = 9314m .

tq(eff) = (t ,)(ts) i

( 4 )-

I}=[ncAT

, P , I=Ieo 4X

', Q = UAAT I = I,e"#X , ,

Pwr ='W' s -x h 2 - I=I o.

10 P=P 10 (t). TVI. = 1.3/u y,p ,t/T HVI.

• 0.693/u o

'SUR = 26.06/T ~

T = 1.44 DT SCR = S/(1 - K,gg) fA

• o SUR = 26 g ff )l CR x = S/(1 - K,gg )

T = '(1*/o ) + [(s 'o)/A,ggo ] CR 1 (1 - Keff)I = CR 2 (1 - Keff)'2 ,

T = t*/ (, _ g

, M = 1/(1 - K,gg) = CR g/CRO T = (I - p)/ Aeffo M = (1 - K,gg)O/II ~ Eeff)1 8"I eff-1) eff " #eff/Esff SDM = (1 - K,gg)/K,gg a= -5

[1*/TKfgg -] + [E/(1 + A,gft )] ,

t* = 1 x 10 seconds P = I4V/(3 x 10 0) A,gg/e 0.1 seconds

~

E = Na Idgg=Id22 '

WATER PARAMETERS Id =I022 g

1 gal. = 8.345 lbm R/hr = (0.5 CE)/d (meters) 1 gal. = 3.78 liters R/hr = 6 CE/d (feet) -

1 ft = 7.48 gal.

MISCEI.I.ANEOUS CONVERSIONS .

Density = 62.4 lbm/ft 1 Curia = 3.7 x 10 dps 10 Density = 1 gm/cm 1 kg = 2.21 lba i Heat of vaporization = 970 reu/lbm I hp = 2.54 x 103 BTU /hr

! Heat of fusica = 144 Btu /lbm 1 Mw = 3.41 x 10 Stu/hr  !

1 Atm = 14.7 psi = 29.9 in. Ig. 1 Btu = 778 f t-lbf .

l 1 ft. H 2O = 0.4333 lbf/in 1 ' inch = 2.54 cm

• F = 9/5 C + 32 "C = 3/9 (*r - 32)

. _ . _ _ _ _ -. L_ _ _ __ _ _ _- - _ . _ -_ _ .-

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. 5. THEORY OF NUCLEAR POWER PLANT OPERATION. Prgs 34 FLUIDS.AND THERMODYNAMICS ANSWER 5.01 (1.50)

a. LOWER THAN

b. HIGHER THAN

c. LOWER THAN [0.5 ea.]

REFERENCE VCS Reactor Theory I-5 ANSWER 5.02 (1.00) b.

REFERENCE VCS Technical Specifications, B 3/4 1-1.

ANSWER 5.03 (2.25)

a. HIGHER [0.25]

Rod worth increases over core life. More rod withdrawal would be necessary to insert the same reactivity addition if BOL conditions were assumed. Thus, actual criticality at EOL would occur sooner than predicted. [0.5]

b. LOWER [0.25]

Xenon poison concentration will be greater _ requiring rods to be further withdrawn to achieve criticality. [0.5]

c. SAME [0.25]

Insignificant temperature change to cause any significant changes in reactivity. [0.5] Note: Small temperature changes may create small effects.

REFERENCE VCS Curve Book 1

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

7 5- THEORY OF NUCLEAR POWER PLANT ~ OPERATION, P ea:35 FLUIDS,AND THERMODYNAMICS ANSWER 5.04 (1.00)

Due to the decrease in pressurizer temperature / pressure the system is voiding elsewhere [0.5] and forcing coolant into.the pressurizer. [0.5]

REFERENCE GP Theory, Pp. 355-358 ANSWER 5.05 (1.00)

The lower line rupture is more severe [0.5] as the mass loss of water is greater than the mass loss of steam. [0.5]

~ REFERENCE GP Theory, Pp. 355-356 ANSWER 5.06 (2.00)

a. FCT increases [0.5]. Corrosion and crud buildup will increase the overall resistance.to heat transfer [0.5].

b. FCT Increases [0.5]. Densification results in fuel shrinkage and an increase in the gap between the fuel pellet and the clad, OR; Migration of fission gasses from pellets to the gap thus decreasing the heat transfer coefficient. [0.5]

REFERENCE GP HT & FF, P 237-239 ANSWER 5.07 (1.00) c.

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• 5 . THEORY OF NUCLEAR POWER PLANT OPERATION. Page 36 FLUIDS.AND THEBMODYNAMICS REFERENCE VCS Reactor Theory, p. I-5.26 ANSWER 5.08 (1.75)

a. The gases contaminate the gap which reduces the thermal cenductivity of the helium gas which raises the temperature of the fuel. (1.0)

b. LESS negative. (0.75)

REFERENCE G.P. Heat Transfer and Fluid Flow, Pp 235-240.

Millstone Reactor Theory, RT-13, p2.

ANSWER 5.09 (1.50)

More positive reactivity will be added in the upper core regions, resulting in a more positve (less negative) AFD [0.75]. Due to the greater decrease in the temperature of the coolant exiting the core relative to the decrease of the inlet coolant. [0.75]

6k Wre Power 15 9mbcdW O t w[f4 v' p*hn o k Ne. COC MW" M REFERENCE M // ((

Westinghouse Nuclear Training Operations, Ch 8 Surry ND-86.2-LP-8, pp 8.14/15 001/000; K5.29 (3.7/3.9)

ANSWER 5.10 (1.50)

a. Decrease [0.5] Pu 239 concentration increases (while U 235 concentration decreases) [0.5].

b. Larger SUR at EOL. [0.5]

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

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' 5. THEORY OF NUCLEAR POWER PLANT OPERATION, Peo 37 l FLUIDS,~AND THERMODYNAMICS l 1

I REFERENCE a HO-RTR-23 to-27 VCS Review text I-3.10-3.14 ANSWER 5.11 (3.00)

a. 1570 pcm. +/- 50 pcm

b. 2 (Exactly double).

e c. Slightly supercritical. [1.0 ea.]

REFERENCE VCS Theory Review text I-3.2, -4.27 4

i ANSWER 5.12 (1.50) i

a. False

b. False

c. False [0.5'ea.]

REFERENCE VCS Theory Review text I-5.38; -5.49 i.

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5. THEORY OF NUCLEAR POWER PLANT OPERATION. Pcco 38 FLUIDS.AND TB_ERMODYNAMICS ANSWER 5.13 (1.50)

Delta RHO Net = Delta RHO iso + Delta RHO Boron Delta RHO iso = -Delta RHO Boron : -(Delta Boron Conc.)(Alpha Boron)

= -(-10 ppm)(-10 pcm/ ppm)

= -100 pcm Delta T iso  : Delta RHO iso Alpha iso

= -100 pcm

-20 pcm/ F

=+5F +/- 1 F Delta Tavg = Delta T iso = + 5 F REFERENCE VCS Curve Book Fig. V-8 ANSWER 5.14 (1.00)

/~ syy, ;)f s1 REFERENCE CNTO " Reactor Core Control" Section 4 Westinghouse Simulator Trng book, "Rx Theory and Core Physics",

Fig I-5-54 VCS Review text I-5.57-76 001/000; KS.38(3.5/4.1)

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

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, .P @ 39 FLUIDS.AND THERMODYNAMICS ,

-ANSWER 5.15 (1.50)-

a. True

b. False

, c. True [0.5 ea]

REFERENCE VCS Review Text I-5.57; -5.76

'1 ANSWER 5.16 (0.50)

a. Outer

b. Inner [0.25 ea.]

REFERENCE NUS, Vol 4, Unit 10.1 WNTO, " Thermal / Hydraulic Principles and Applications", pp 13-57/58 I

004/000; K5.09(3.7/4.2)

ANSWER 5.17 (1.50)

a. Latent Heat (of Vaporization)

b. DNBR

c. Heat Flux Hot Channel Factor [0.5 ea.]

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r a a

5. ' THEORY OF NUCLEAR POWER PLANT OPERATION, Paca 4 ELUIDS.AND THERMODYNAMIGS REFERENCE WBN, HT & FF,.pp. 11, 22,'and 24 General Physics, HT & FF, pp. 38 & 228 and VCS, TS, p. B3/4 2-1 002/000-K5.01 (3.1/3.4)

ANSWER 5.18 (1.50)

a. 1. To ensure that no vapor pockets form in the loops

2. To prevent steam pocket. formation

3. To help thermal driving head [0.25 ea.]

b. 1. RCS dT will increase (exceed 100% full power value)

2. Pressure will decrease.

3. Level will increase. [0.25 ea.]

REFERENCE HTTFF pgs. 356, 357 3.4 000 015 EK 1.01 4.4 ANSWER 5.19 (2.00) 3 3

a. Power (2) = Power (1) * (N2/N1) = 300 * (4) = 19.2 MW 2 2 Delta P(2) = delta P(1) * (N2/N1) = 50 * (4) = 800 psid Flow (2) = Flow (1) * (N2/N1) = 880

• 4 = 3520 gpm [0.25 ea.]

-b. Answer: g (1.0) i

'Rt sdh I g/g

c. DECREASES (0.25)l I REFERENCE VCS HTFF pp. 322; 324-326; 319, 320

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

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5. THEORY'OF NUCLEAR POWER PLANT OPERA ION, ' *- Pca 41 FLUIDS.AND THERMODYNAMICS .,

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6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION' Pcco 42

-ANSWER 6.01 (2.00)

a. OPEN

b. CLOSED on active loop, OPEN on inactive loop

c. OPEN id . -CLOSED [0.5 ea]

REFERENCE VCS, AB-7, RHR System, p. 21 and SOP-118, Attachment 1A, pp. 4&9 006/020-A3.02 (3.9/4.2)

ANSWER 6.02 (1.00) c.

REFERENCE VCS Plant System Descriptions, AB-3, p. 10 and Review Questions ANSWER 6.03 (1.00) b REFERENCE VCS IC-5 pp. 17-18; Fig. IC5.5 001/000; K4.08(3.2/3.4)

ANSWER 6.04 (1.00)

1. a.

2. d.

3. a.

4. c.

5. a. [0.2 ea.]

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION Paca 43 REFERENCE VCS IC-5 p. 18; Fig. 105.6 NAPS Rod Control Lesson Plan.

ANSWER 6.05 (1.50)

a. Intermediate

b. Source

c. None

d. Intermediate

e. Source, Intermediate, Power

f. Source [0.25 each letter]

REFERENCE VCS IC-8 pp. 11, 12, 21, 28; Fig. IC8.4, 8.5 ANSWER 6.06 (2.00)

a. When the COPS is in service, the RCPs are not operating, therefore, the Narrow Range instruments which are located in the bypass manifold would not be reliable. Will also accept narrow range instruments range does not extend to below 450 F. [0.6]

b. Tavg 1. Low Tavg

2. OPdT

3. OTdT

4. LoLo Tavg

5. Hi Tavg dT 1. OTdT

2. OPdT [0.2 ea.]

REFERENCE VCS IC-6 pp. 14-16 010000K403 l

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1 l

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I

6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Pcco 44 ANSWER 6.07 (2.00)

a. The rod control system will see an increased Tave-Tref deviation and rods will insert.

b. NONE-- The OT deltaT setpoint is generated using protection RTD's.

c. NO EFFECT-- Rod insertion limits are power dependent. Decreased delta T implies lower power. RIL computer uses auct. Hi delta T, so there would be no change in RIL's.

4

d. The increased auct. Tave will cause pressurizer program level setpoint to increase. [0.5 ea.]

REFERENCE VCS IC-6 pp. 25-27 ANSWER 6.08 (1.75)

a. Tavg, dt, Pressure, dI [0.25 each]

b. Prevent exceeding DNB [0.25]

c. Turbine runback Blocks automatic and manual rod withdrawal [0.25 each]

REFERENCE VCS IC-6 pp.23-25 1

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

i

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION Pcco 45 ANSWER 6.09 (2.50)

a. To ensure adequate trip reactivity. (minimum SDM)

To limit the potential effects of rod misalignment.

To assure power distribution limits are met. [0.25 ea.]

b. Due to the large value of positive reactivity inserted by MTC during the resulting uncontrolled RCS cooldown. (1.00)

c. 1. EOL

2. Shutdown

3. 547 F [0.25 ea.]

REFERENCE Technical Specifications, pg B 3/4 1-3

• • • CAF ***

K&A 001-000-KS.08 / IF 3.9 001-000-K5.04 / IF 4.3 ANSWER 6.10 (2.00)

1. The lube oil prelube system circulates oil through a heater to normal lubrication components.

2. The cooling water keep-warm system circulates water through a heater to normal water-cooled components.

[1.0 ea.]

REFERENCE VCS, IB-5, Diesel Generator, pp. 34 & 35 064/000-K1.02 (3.1/3.6)

-A1.01 (3.0/3.1)

ANSWER 6.11 (1.00)

1. Th - Delta-P across the S/G

2. Tc - Delta-P across the RCP [0.5 each]

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

'6. $LANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION P c 46 REFERENCE VCS,-AB-2, P. 19 ANSWER 6.12- (1.50)

1. Le-re-Le S/C lovel [0.2] - 2/3-[0.1] -- 5% [0.i]. f[ / /E

2. The--,iguel hich is the coincidence of: JA+i C Lc S/C prescuio [0~zj -- z M [0.1J

-- e2a psig tr1]. '

_L Le f redwater- tompu7ar,ure [G.2] ' 226-deg. F [G &

=c. Lc fccd flow [0.2] < 20%-EO.1].

" ^I 7 REFERENCE - 41. 0 2 c_ b :ce? E'e 9 01 i t 1, it: Lo Lo -Lo a/sicue,l (p,p}, Af [0.T} -S*/,C05]

VCS, TB-7, P. 22

% .kn !dduN J d4 ' Co I3 Sf q. Lo S u b s W . L< u pt n k ee. Cp & < h *p C.0 4 3

/fr/;7 h, Lo ha Slow E.o.3] < ia % Con] gjfp ANSWER 6.13 (1.00)

YJ l l 07 REFERENCE VCS IC-3 pp. 34-36 l

ANSWER 6.14 (1.00) c o-b gg g}(

REFERENCE VCS IC-4 p. 15 I

ANSWER 6.15 (1.00) a

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6. PLANT SYSTEMS-DESIGN, CONTROL, AND INSTRUMENTATION Pcca 47 REFERENCE VCS IC-8 Fig. IC8.6 a, /, O R s.ll cause ch o if */s Dow clwenh m one, jag s.ense. l<a %n 90 % 9toJ & eww y P4. Co&

g 2'. 6C/Rf.g s.ll s secwe.adeie

% 3e_ I m A i En#l.1 9 4 f-lw 8

N Ael-tmE po w <-in co ANSWER 6.16 (2.00) qbove_ 9 4. Co3] g, ffr

a. 1. Single loop loss of flow will insert if 2/4 power ranges are above 38% (P-8). (0.5)

2. Two loop loss of flow will insert lf 2/4 power ranges >10%

[0.2] OR; [0.1] 1/2 impulse pressures > 10% [0.2]

b. The undervoltage trip is to provide a trip signal on loss of power to the RCP's. [0.2] The reactor will trip and RCP's contin-ue to provide coastdown flow. [0.2] The underfrequency trip pro-vides protection for a grid disturbance. [0.2] The RCP breakers and reactor are tripped [0.2] to prevent deceleration and loss of coastdown flow. [0.2]

b. W II a12 a ee-94: he_how O josso4 flod 8%

REFERENCE _ 4 g 6 ps pude on e o 4wo /vo p s Co 43. pre VCS IC-9 pp. 4.6, 47 _ g qqpg. 4 g ,, c g ,,,6,4on Q sI.oss oT % a sn Lwo'oe mo n iso (x .Lo.3

- TA< .flo w 49 5 do 64 p ,TN id e_ th<. (>redec_. on agcu ns f ANSWER 6.17 (1.25) 04 M %e_ W ama u F h ps (he vie /eCo.C

a. Because the turbine serves as the heat sink to the reactor, a reactor trip follows a turbine trip to minimize the RCS temperaturp'*ransient(an/orresultingsafetyvalveoperation).

(0.5) ,

, jfg

b. - High par pressure

- High pzr level

- OT Delta T

- S/G lo-lo level [1 required @ 0.25]

c. - All turbine stop valves shut

- Emergency oil (Emer. Trip Fluid System) pressure low

(< 800 psig) [0.25 each]

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P *I

6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Pass 48 !

l REFERENCF VCS IC-9 pp. 58, 59 ANSWER 6.18 (2.00)

a. 1. "C" Pump Breaker on "A" Train racked up.

2. No T.D. overcurrent.

3. MCB switch in normal condition.

4. "A" Pump breaker racked down OR No ESFLS output 5 present. '

A,.,o n S + k 9cA$

5. SI signal. [0.2 each] g(, ; .

jy

b. "A" Pump Breaker remains closed.

"C" Pump Breaker trips open. [0.5 each]

REFERENCE VCS, GS-2, PP. 38-40 ANSWER 6.19 (1.00)

Minimize the fluctuations on the control system due to shrink and swell.

REFERENCE VCS, IC-2, P. 11 ANSWER 6.20 (1.50)

a. Upper-range will indicate minimum level.

b. Narrow-range will indicate maximum level.

c. Wide-range will increase from 100% to approximately 110%.

[0.5 each]

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P-

6. PLANT SYSTEMS DESIGN, CONTROL, AND' INSTRUMENTATION Para 49 REFERENCE

.VCS, IC-13, PP. 12-13 i

i 4

(***** END OF CATEGORY 6 *****)

-7. =PROCEDUMA - NORMAL'." ABNORMAL'.' EMERGENCY- Pea 50

-AND-RADIOLOGICAL CONTROL

-ANSWER 7.01 (1.00) c REFERENCE VCS, EOP-1.4, P. 1 ANSWER 7.02 _(1.00)

a. 325 psig

b. 0.2 gym

c. 15 psig

d. 3 [0.25 each]

REFERENCE VCS, SOP-101, P. 1 ANSWER 7.03 (1.50)

a. When supplying cooling water to RHR heat exchangers and Non-Essential loads. [0.5]

b. 105 deg. F [0.25]

c. To retain two-loop operability. If the out-of-service pump breaker is not racked out, the auto start feature of the l standby pump is defeated. [0.75]

REFERENCE

! VCS, SOP-118, PP. 1, 2, 4 I ANSWER 7.04 (1.00)

a. 50 psig 8

l l b. Isolate charging. [0.5 each] l i

H

, (***** CATEGORY 7 CONTINUED ON NEXT.PAGE *****) J i

. . . . , , , --% , - - , - , , ,,y,,, y _.,y_, . -,. . .. , . ., . . . , . . . 9,m-- ,e r y y-- -y.. .-

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Peca 51 AND RADIOLOGICAL CONTROL REFERENCE VCS, SOP-102, P. 2 ANSWER 7.05 (1.00)

To prevent an auto restart due to a bus undervoltage signal.

REFERENCE VCS,' SOP-306, P. 27 ANSWER 7.06 (1.00)

1. Announce conditions over plant page system.

2. Stop any refueling operation which could cause further fuel assembly damage or release of radioactive products.

3. Suspend core alteration in a safe, stable position.

4. Evacuate ALL unnecessry personnel from affected area.

[0.25 each]

REFERENCE VCS, EOP-7.0, PP. 9, 10 ANSWER 7.07 (2.00)

1. The RCS is borated to the cold shutdown concentration (and verified by sample). OR

2. Tavg is 557 F and the Res is borated to the hot shutdown, Xenon free concentration (and verified by sample).

[1.0 ea.]

REFERENCE

VCS GOP-1, Appendix A, p. 4

(***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)

rn' ' -

' j

.Pecp 52 i

,7 PROCEDURES - NORMAL. ABNORMAL. EtgRGENCY.

AND RADIOLOGICAL CONTROL ,

I i

ANSWER 7.08 (1.00)

a. ,

i REFERENCE VCS GOP-5 pp. 5-14 ANSWER 7.09 (2.50)

a. 1600 psid

b. 25 degrees F

c. 3000 gpm

d. 3%/hr

e. 350 degrees F [0.5 ea.]

REFERENCE VCS, GOP-Appendix A, Generic Operating Precautions, pp. 2, 3, & 5 ANSWER 7.10 (2.00)

a. Close the RCP A seal leakoff valve (PVT-8141A). (1.0)

b. 5 min. (0.5)

c. 30 min. (0.5) -

REFERENCE VCS, SOP-101, Reactor Coolant. System, pp. 51 & 52 ANSWER 7.11 (1.00)

Removing the applicable control and instrument power fuses on'the power range drawers.

r

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i.

P:

'7.

-PROCEDURES --NORMAL, ABNORMAL, EMERGENCY Pcca 53 AND RADIOLOGICAL CONTROL REFERENCE VCS, SOP-404, Excore NI System, p. 16 015/000-A4.03 (3.8/3.9)

ANSWER 7.12 (1.50) d.

REFERENCE.

VCS E0P-1.2, pp 4, 5 i

ANSWER 7.13 (1.50)

a. 3 I b. 5

c. 2

d. 1

e. 6

f. 4 [0.25 each)

REFERENCE VCS SOP 102- , 210-33, 306-52, 401-35, 57 AOP 11.0-1 EOP 1.0 l

l l

ANSWER 7.14 (1.00) )

c I

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7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Paca 54 AND RADIOLOGICAL CONTROL REFERENCE VCS, EOP-4.0, pp. 9 - 13 ANSWER 7.15 (1.50)

To prevent excessive depletion of RCS water inventory through a small break in the RCS which might lead to severe core uncovery if RCP's are not tripped within the proper time frame.

REFERENCE Westinghouse Background Info. "RCP Trip / Restart" 000/009; EK3.23(4.2/4.3)

ANSWER 7.16 (2.50)

a. 1. 100 REM de 'I f R E M '7 L Eff- Oli isNlnnez), W/l/f97

2. 25 REM [0.5 each]

b. Increased liklihood of cancer, particularly leukemia. Short term somatic effects include blood changes. [1.0]

c. Emergency Director (or authority as delegated by the ED) [0.5]

REFERENCE VCS HP Manual pp. 5-13, 14; 4-12,15 U PP -o s t p.1 s GPP -oxo p , K gr ll/r $7 l

ANSWER 7.17 (2.50) l i

I

a. 1. 1.25 R/qtr.

2. 18.75 R/qtr.

3. 7.5 R/qtr. [0.25 ea.]

b. 1. 1 R/qtr.

2. 12 R/qtr. J48 //f/~

/ 87

p. ,['[pa-m

3. 6 R/qtr. 0.25 ea.]

c.

con 2XRIQs IE ll 2 R/qtr not to exceed 5 R/yr.

Plant Manager- [0.5 ea.]

Ome_6 Adv Plc,Jop<nd%.s Mb i)c/e7

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'7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Paco 55 AND RADIOLOGICAL CONTROL REFERENCE VCS HP Manual pp. 5 /f 99 -753 p. 3 a.,JL 61.I 5-13#//[#/87 0

ANSWER 7.18 (2.00) QQ [q4 %e9%f

1. Source Range Excores 7. Qc_s F/vw fab'b *b '" 1 7.Co,%\e% ne. Tr+ RTG

2. q gng,( c)qnne, Intermediate Range Excores gg7g

3. Containment Pressure (High-3) p. @cea54 M 2Ar pnssgr<_. cMNk .

4. RWST Levels

5. Emergency Feedwater Suction Pressure Low. [any 4 @ 0.5 each]

REFERENCE VCS, SOP-401, P. 4 ANSWER 7.19 (1.50)

a. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> [0.5]

b. 1. Bell System Land Line

2. Relay message (via dispatcher, LLEA, or other available method). [0.5 ea.]

REFERENCE VCS, SI 86-02, P. 5 ANSWER 7.20 (1.00) c.

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7

7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pega 56 AND RADIOLOGICAL CONTROL 4

REFERENCE VCS Plant System Descriptions, IB-5, p. 32 i

i b

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4 4

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b s

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

I[ , ,

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, Pasa 57  !

AND LIMITATIONS  !

ANSWER 8.01 (3.00)

1. One charging pump cannot provide sufficient charging flow at normal operating pressure to maintain a constant pressurizer level.

2. Specific activity of the RCS > 750/E uCi/cc.

3. Leak rate exceeds 1%/ day at design pressure. [1.0 ea.]

REFERENCE VCS EPP-001 pp.3-4 ANSWER 8.02 (1.00)

a. 1. Station Manager (Ollie Bradham).

2. Management Duty Supervisor.

3. NRC. [0.25 ea.]

b. 15 minutes. (0.25)

REFERENCE VCS EPP-002 ATT. III A-D ANSWER 8.03 (1.00)

a. Shift Supervisor (IED/ED).

b. Control Room Supervisor. [0.5 ea.]

REFERENCE VCS EPP-001 p. 5; SAP-200 p. 6 ANSWER 8.04 (1.50)

a. FALSE 4

b. FALSE j

c. TRUE [0.5 ea.] j l

I

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8. ADMINISTRATIVE PROCEDURFS. CONDITIONS. Pcca 58 AND LIMITATIONS REFERENCE VCS TS, p. 3/4 0-1 and B 3/4 0-1 ANSWER 8.05 (1.00) d REFERENCE VCS IC-9 p. 49 ANSWER 8.06 .(1.00)

1. Its emergency power source is operable

2. Its redundant system is operable [0.5 ea.]

REFERENCE VCS TS, p. 1-4 ANSWER 8.07 (3.50)

a. 2 (0.5)

b. Within 15 minutes (0.2)

1. Restore the indicated AFD to within the target band (0.4),

2. Reduce the thermal power to <90% of rated thermal power. (0.4)

~

c. Accumulated penalty over the- past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is 89 minutes. 44ved The penalty will be reduced to 60 minutes at 1618 minutes on 05/13/86 and then power may be increased. M yg pp[qv'/rg 85% 0318-0310 = 8 (0.25) /1k 65% 1637-1557 = 40 (0.25) 45% 0310-0148 = 82/2 = 41 (0.5) 89 min. total penalty ,

05/13/86, from 1557; 81 min left (60 penalty min.- 21 min)

-> 1618 (w/60 penalty min.) 05/13/86 (1.0)

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, Paco 59.

AND LIMITATIONS REFERENCE TS 3.2.1; TS pg. B 3/4 2-2 3.9 015 020 Sys gen 5. 3.9 ANSWER 8.08 (2.50) C Since Mode 3 is about to be entered, (1.0) the heatup must be discontinued (0.75) and Tave held (or reduced) at less than 350 F until Charging Pump B or C is proven operable (0.75).

REFERENCE TS pg. 3/4 0-1; TS pg. 3/4 5-3 3.4 005 000 Sys gen 5 4.0 ANSWER 8.09 (1.50)

a. Interval requirement not exceeded [0.25]. Eight days does not exceed 1.25 times the specified interval [0.5].

b. Interval requirement exceeded [0.25]. The last 3 consecutive intervals exceed 3.25 times the specified interval [0.5].

REFERENCE VCS TS pg. 3/4 0-2 SWPWGK&A 5 3.9 ANSWER 8.10 (2.00)

a. You may operate for up to two hours with one less than the minimum complement (0.5) provided that immediate action is taken to bring the complement up to minimum (0,5)

b. The on-shift STA will have to wait for a relief to come in (0.5); OR if the on-coming SS or SRO meets the STA requirements, the STA position may be vacant (position is not to be left intentionly vacant)(0,5)

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8. -ADMINISTRATIVE PROCEDURES. CONDITIONS. Pcco 60 AND LIMITATIONS REFERENCE VCS TS Table 6-2.1, p 6-4 ANSWER 8.11 (1.50)

No. [0.5] Reactor must be placed in'at least Mode 4. The OPERABILITY statement for ECCS flow. paths is not met (i.e. the valves will not shut upon ESF/SI actuation). OPERABILITY is defined as'all 7

necessary, attendent instrumentation, controls, electrical power, etc. are capable of performing its related support function. [1.0]

l REFERENCE VCS, T.S., PP. 3/45-3 ANSWER 8.12 (1.00)

Plant operations in Mode 5 or 6.

REFERENCE VCS, SAP-200, P. 24 ANSWER 8.13 (1.25)

a. He shall mark "N/A" and initial each step that is not t

applicable. [0.25]

b. 30 days [0.25]

, c. 1. The proposed change.

2. The Safety Review / Safety Evaluation.

3. Selection of review requirements. [0.25 each]

. REFERENCE VCS, SAP-139, PP. 4, 18 3

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, Pcco 61 AND LIMITATIONS ANSWER 8.14 (1.25)

a. 1. STA

2. Shift Supervisor

3. Control Room Supervisor

4. Reactor Operator [0.25 each]

b. Director, Nuclear Plant Operations [0.25]

REFERENCE VCS, SAP-132, P. 10 ANSWER 8.15 (1.25)

a. Suberiticality (S)

Core Cooling (C)

Heat Sink (H) 4 Integrity (P)

Containment (Z) i Inventory (I) [0,1 each]

1 b. Containment Fuel Matrix Fuel Clad RCS [0.1 each]

c. Integrity [0.25]

REFERENCE VCS, EOP-12.0, P. 1, 3 i

^

ANSWER 8.16 (1.50) i a. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1

b. TRUE i

c. Monitor QPTR (once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />). [0.5 each]

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

o .

8. ADMINISTRATIVE PROCEDURES. CONDITIONS. Para 62 AND LIMITATIONS REFERENCE VCS, T.S., P. 3/4 3-6 ANSWER 8.17 (1.00)

After Immediate Operator Actions of any other EOP are completed.

REFERENCE VCS, EOP 12.0, P. 2 ANSWER 8.18 (1.75)

a. Prevent overheating of the fuel and (possible clad perforation) and thus prevent the release of fission products. [1.0]

b. Notify NRC within one hour, notify company management.

(CAF) [0.75]

REFERENCE VCS TS p. B2-1, 6-12 ANSWER 8.19 (1.50)g I / /((

a. A quahfded--Danger--Tegger-wii:h a current RO or SRO license.

[1.0]

b. FALSE (Must be an Electrical Maintenance person). [0.5]

REFERENCE VCS SAP-201 pp. 9, 5

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

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