ML20235T582

Exam Rept 50-269/OL-87-02 on 870713-23.Exam Results:Seven of Ten Reactor Operator (RO) & Eight of Nine Senior Reactor Operator (SRO) Candidates Passed Written Exam.All RO & SRO Candidates Passed Operating Exam

ML20235T582
Person / Time
Site:	Oconee
Issue date:	09/18/1987
From:	Bill Dean, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20235T500	List: ML20235T582
References
50-269-OL-87-02, 50-269-OL-87-2, NUDOCS 8710130136
Download: ML20235T582 (134)

Text

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ENCLOSURE 1 REQUALIFICATION EXAMINATION REPORT 269/0L-87-02 Facility Licensee: Duke Power Company 422 South Church Street Charlotte, NC 28242 I

Facility Name: Oconee Nuclear Station Facility Docket Nos.: 50-269, 50-270 and 50-287 Written and operating examinations were administered at the Oconee Nuclear Station near Senec9, Sout Ca rolina.

Chief Examiner: (t 2(I V ' 9 Vh?

William Dean ' Date Signed Approved by: fDate

//'[/7 Signed J6% F. Munro, Section Chief Sumary:

Examination on July 13-23, 1987 Written and operating examinations were administered to ten Reactor Operators (R0) and nine Senior Reactor Operators (SRO). Seven of ten R0s and eight of nine SR0s passed the written examination. Ten of ten R0s and nine of nine SR0s passed the operating examination. Based on the results described above, seven of ten R0s and eight of nine SR0s passed.

i 8710130136 870923 PDR ADOCK 05000269 V PDR s

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REPORT DETAILS 1-. Facility Employees Contacted:

• P. Stovall, Plant Operations Instructor

• D. Tidwell, Instructor

• R. Bugert, Training Supervisor

• L. Hindman, Instructor

• J. Pr' ice, Shift Operating Engineer

• T. Campbell, Operations

• T. Barr, Manager, Oconee Training D. Sweigart, Superintendent of Operations M. Tuckman, Plant Manager

• Attended Exit Meeting

2. Examiners:

• W. Dean, NRC C. Casto, NRC J. Huenefeld, PNL B. Gore, PNL L. Lawyer, NRC L. Wert (NRC RI - attended exit)

• Chief Examiner

3. Examination Review Meeting At the conclusion of the written examinations, the examiners provided Mr. Stovall 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. The comments made by the Facility Educational Specialist are not addressed, since they provide additional information for examination development only. Eight of 22 (36.4%) changes to the answer key were made as a result of inadequate or insufficient training material provided to the NRC for examination development.

a. R0 Examination (ApplicableSR0questionsareinparentheses)

Question 1.02(5.02): Agree with facility comment. Answer "c" will also be accepted.

Question 1.11(b): Do not agree with facility comment. No information was provided to indicate that the magnitude of change in rod worth over core life is as significant as the change in l

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moderator temperature coefficient, which is j nearly 0 delta k/k/ degree F at B00. The  !

referenced lesson plan assumes that the maximum dropped rod worth is 0.3% delta k/k.

For a given rod, its worth will change relatively little over core life.

Question 1.12: Agree with facility comments. Xenon oscillations will not be required as part of the answer.

Question 1.14(5.14): Agree with facility comment. Typographical error in the answer key will be corrected.

Question 2.06a(6.09a): Facility comment acknowledged. The question l is clearly asking for another source of water for SSF-ASW System. Any alternate answers will be evaluated based on the assumption that all Intake Water is ,

unavailable.

l l Question 2.06b(6.09b): Facility comment acknowledged. Though it is not listed as the reason for the installation of the eductor in the referencs lesson plan, the device does minimize the air injection into the S/Gs when the SSF-ASW System is in service.  ;

Alternate answers will be evaluated for '

applicability.

Question 2.10: Ac ee with facility comment. Lesson plan should be changed to reflect current plant conditions.

Question 2.11: Agree with facility comment. The recommended additional answer will be incorporated and six of seven responses will be required for full credit. The lesson plan should be updated to reflect the  ;

correct crder of preference.

Question 2.12(6.14): Agree with facility comment. Full credit will be given if candidates list four l correct loads which lose cooling water.

Question 2.13b(6.15b): Facility comment acknowledged. The question was not specific enough to elicit the system manipulations required. The facility recommended answer is trite. The question will be deleted.

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-3 Question 2.14(6.17): Facility comment acknowledged. The recommended additional answers will be i accepted as additional correct answers.

The referenced lesson plan should be updated to reflect this information.

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-Question 3.05(6.05): Facility comment acknowledged. The material l is covered in the referenced lesson plan, l and NUREG 1122, "Knowledges and Abilities- 1 Catalogue for Pressurized Water Reactors" l

supports the validity of this question.

However, since there is no requalification learning objective to' support this topic and the fact that the importance rating is  !

l low relative to , requal related knowledge, l

the question will be deleted.  ;

l Question 3.07(6.08): Agree with facility comment. Additional recommended answer will also be accepted.

Question 3.08a: Agree with facility coment. Answer key j will be changed as recommended. Facility should update their lesson plan.

Question 3.13: Agree with facility comment. Answer key will be modified as recommended.

Question 3.14: Facility comment acknowledged. The only terminology that is not expressly utilized in the facility training material is

" Coincidence Logic contacts". This part of the required answer will be deleted and the point values will be redistributed among the

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remainir,g answers to require knowledge of the ordering of the components only.

Question 4.01(7.01): Facility comment acknowledged. The question was not specific enough with regards to existing plant conditions to elicit the desired response. The EPG is an excellent reference source for background information associated with emergency procedures, and may serve as a source of information if facility lesson plans are inadequate.

Question will be daleted.

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Question 4.03b: Agree with facility comment. Due to the -

conflicting information. in the Health I Physics Manual and the Emergency Plan either answer will be accepted. The Health Physics 3 Manual should be updated to agree with i current procedural guidance, l Question 4.06: Agree with facility comment. Recommended additional answers will also be accepted.

Question 4.11: Agree with facility coment. Note that the corrected version of the lesson plan had not been provided to the NRC examiners.

b. SR0 Examination i Question 5.05: Agree with facility comment. Answer "d" i will also be accepted.

, Question 7.13: Agree with- facility comment. The )

) recommended additional answer will be accepted. Six correct responses will. be 1 required.

Question 7.15: Agree with facility comment. The referenced I AP is incomplete with regards to the automatic actions that occur for the given l RIA alarms and should be updated.  !

Question 8.01: Agree with facility comment. Question was i not specific enough to elicit the desired response. The recommended additional answer will also be accepted.

Question 8.06: Facility comment acknowledged. The recommended answer will be accepted only if the candidate expressly states the required assumption.

Question 8.07: Agree with facility comment. Additional ,

recommended answers will also be accepted.

Question 8.13: Facility recommended answer is equivalent to the answer key. No change to key required.

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

There were some generic weaknesses noted during the ; operating examinations, particularly inadequate usage .of annunciator . response procedures and poor knowledge'among SR0s of the Standby Shutdown Facility.

The examiners expressed concern over the use of two senior operators to direct the actions of the reactor operators during situations involving usage of the. Emergency Procedures. - There' appeared to be a tendency 'to forego applied usage of ;'ese procedures and defer Lto' the' directions of the Unit Supervisor, who directed actions based upon his memorization of the procedures. In some instances, the SR0 reading the procedures could not keep track'of where they were in.the procedures due.to multiple transitions within the emergency procedure framework in a very short time frame. It was also noted in some instances that little, if'any, consultation between the-two SR0s in the control room took place with respect to utilization and application of. the emergency procudures. . Use of the procedures _as a tool to put the plant in a safe and stable condition should be emphasized more strongly in requalification training, to avoid situations where the plant is being operated solely on an individuals recall of the_ emergency procedures. - The procedures, if they are being implemented correctly I and in a timely _ manner, should be the controlling element in mitigating. i emergency events. j l

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

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

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hE5 U. S. NUCLEAR REGULATURY COMMISSION REACTOR OPERATOR REQUALIFICATION EXAMINATION FACILITY: OCONEE 1, 2&3 _ _.

REACTOR. TYPE: PWR-B&W177 DATE ADMINISTERED: _87/07/13 EXAMINER: CASTO. C.

CANDIDATE:

INSTRUCTIONS TO CANDIDATE:

Read the attached instruction page carefully. This examination replaces the current cycle facility administered requalification examination.

Retraining requirements for f aile,tre of this examination are the same as for failure of a requalification examination prepared and administered by your training staff. 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 four (4) hour 3 after the examination starts.

% OF CATEGORY  % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 18.00 25.00 i. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW l% V betec 25.00 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS

1. 76 0 JJL4'd k 25.00 3. INSTRUMENTS AND CONTROLS 18.00 25.00 4 PROCEDURES - NORMAL. ABNORMAL.

EMERGENCY AND RADIOLOGICAL CONTROL 2-7 . 2 3 -  % Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.

Candidate's Signature l

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, NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the adminir,tration 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 _1 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. 'l I

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.

l9. Number each answer as to category and number, for example, 1.4, 6.3.

'10. Skip at least threq lines between each answer.

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

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

l13. 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.

L l15. Partial credit may be given Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

l16. If parts of the examination are not clear as to intent, ask questions of the examine.r_ 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.

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

a. Assemble your examination as follows: i (1) ' Exam questions on top.

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

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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, PAGE 2 '

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

Which one of the following correctly describes the behavior

.of RCS pressure if a Small Break LOCA which was not large enough to actuate the ECCS were to occur, without Feedwater available?

a. Pressure initially decreases slowly, then rapidly drops l

'i when the OTSGs are boiled dry.

b. Pressure decreases slowly until it levels off somewhere above ECCS actuation pressure.

c. Pressure initially increases, then rapidly drops when I the OTSGs are bciled dry,

d. Pressure initially decreases, then rapidly increases when the OTSGs boil dry. j l

e. Pressure initially decreases, then when OTSGs boil dry, continues to decrease, but at a much slower rate. 3 i

e QUESTION 1.02 (1.00) )

Which one of the following instrument failures would cause the behavior of the parameters shown on attached drawing ,

OC-TA-NT-15?

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a. Delta Tc Failure "A" Side LOW

b. Delta Tc Failure "A" Side HIGH

c. Delta Tc Failure "B" Side LOW )

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d. Delta Tc Failure "B" Side HIGH l

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~ 1. . PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.03 (1.00)

Refer to the attached handout OC-RT-SM-3 "NI Response for Subcritical and Critical Reactor",.for each point labelled A-D' match following descriptions with the response of the NI.

1. Stable SUR

2. Equilibrium value of suberitical multiplication

.3 . Withdraw an increment of control rod

4. Withdraw an-increment of control rod to make the reactor critical.

QUESTION 1.04 (1.00)

Which one of the following is correct concerning differential control rod worth (DRW)?

a. It is a measure of reactivity due to rod position.

l l b. With a normal cosine flux shape, DRW . reaches a maximum value at a rod index of less than 28%.

c. Rod Group Overlapping maintains a constant DRW.

d. Its unit is delta K/K/% index.

QUESTION 1.05 (1.00)

Which one of the following is an advantage associated with the use of soluble poison in the RCS?

a. Soluble poison aids in reactor control by its effects on the moderator temperature coefficient, the void I

coefficient and the pressure coefficient.

l l b. Soluble poison can be injected quickly enough to overcome all positive reactivity insertions without control rod '

movement. .

c. Soluble poison control provides for a smoother flux shape.

d. Soluble poison at lower temperatures crystallizes.

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,12 - PRINCIPLES:OF NUCLEAR' POWER. PLANT OPERATION. PAGE- -4 THERMODYNAMICS.' HEAT TRANSFER AND FLUID FLOW

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<QUESTIOND 1.06 (1.00)

Assuming.that the plant is operating at ' full power, what-will be the difference.in.the.following parameters, if OTSG z l

tub 2'fou11ng1has. occurred to a significant degree?

a)' OTSG Level j

b)~.Superheat Temperature  !,

.- 1 QUESTION 1.07 (1.00) 'l

-1 An/ECP~is calculated for a startup following a reactor trip from 50% power, with equilibrium xenon in the carefat MOL.

Indicate if the ACTUAL critical rod positi'an will'be HIGHER, LOWER or the SAME compared to the calculated position for each of-the following situations. Use attached: curves as i appropriate.and treat each situation individually.

a)' Xenon reactivity curve for trip from 100% power is used j to calculate conditions for a startup 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after the trip. (Computer is'not in service to give this info)  ;

b) The differential boron worth'at an EOL condition is used. Assume no change in boron concentration is 1 desired' prior to achieving criticality.

QUESTION 1.08 (1.00)

For EACH of the following conditions state whether the j actual Shutdown Margin would be greater than/ less than/-the

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i camp as calculated Shutdown Margin. Assume BOC, Mode 1 full 'l power.  :

.1. Actual poison burnup exceeds that value used in the j calculated SDM.

2. The actual worth of the maximum stuck rod assumed in the calculation is lower than predicted. 'l

. QUESTION 1.09 (1.00) i What indication tells the operator when all nitrogen.has been vented from the Pressurizer, when forming a steam  !

bubble in accordance with OP/0/A/liO3/OS? i I

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[. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION 1 PAGE 5 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.10 (2.00)

Refer to the attached drawing OC-TA-NT-10, "FDW Main Control Valve Fails Open at 50% Power", to answer the following:

a) Why does FWPT Speed increase, then level off at point (2)?

b) What is causing reactor power to increase starting at point (3)?

c) After reactor power and MWe stabilize, prior to the reactor trip, what is the relationship between Loops A and B "T Cold"?

OUESTION 1.11 (2.00)

Unit 2 has just restarted following a refueling outage while Unit 3 is near EOC. Answer the following regarding the differences in plant response between the units (Explain your answer):

a) At a steady power level of 10EE(-8) amps during a startup, equal reactivity additions are made. Which i Unit will have a higher startup rate?

b) At 50% power, with ICS in MANUAL, a control rod drops.

Assuming NO RUNBACK and NO OPERATOR ACTION, which Unit I

will have a lower steady state Tavg?

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i OUESTION 1.12 (1.00)  !

l Recently, Inconel Axial Power Shaping Rods (APSRs) have been installed on all 3 units. What is the main physical change in these rods and how has this change improved the APSR's ,

capability to control axial flux imbalances? l l

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}. POINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 6 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.13 (2.00)

How does each of the following parameter changes affect the DNBR (INCREASE, DECREASE or NO EFFECT)? Briefly explain your answer and DO NOT consider the transient effects or any control system or operator actions.

a) Pressurizer temperature increases 5 degrees.

b) Mass flow rate in the core increases 10%.

QUESTION 1.14 (1.00)

The pressurizer PORV is leaking by during operation at 85%

l power. Assuming a Ouench Tank pressure of 20 psia and l saturation conditions in the pressurizer corresponding to l 2240 psia, what is the quality of steam on the downstream side of the PORV7 Show all calculations.

QUESTION 1.15 (1.00)

The reactor is producing 100% rated thermal power at a core i delta T of oO degrees and a mass flow rate of 100% when a i station blackout occurs. Natural circulation is established and core delta T goes to 40 degrees. If decay heat is 2%,

what is the core mass flow rate (in %)?

a. 1.3

b. 2.0 C' 30 .__.__ _ _._. - . . _ _ _ - - -- - - - - - - - - ~ - - - - - - - - -----

~-d. 4.2 1

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g. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 7 QUESTION 2.01 (1.00)

Which one of the following should enable MS-93, TDEFWP steam supply valve, to open if it fails to open on an automatic signal?

a. Verify proper operation of the DC Oil Pump.

b. Line up backup service air to the valve operator,

c. Isolate instrument air to its reducer and bleed the air off the reducer.

d. Take the control switch to "Off" to remove power from l the solenoid valve.

QUESTION 2.02 (2.00)

Match the compenents listed in Column A with the correct location where they penetrate the RCS. Answers may be used more than once.

Column A Column B

a. Unit 1 PZR Spray 1. Al RCP Suction  !

b. Unit 3 Normal HPI Line 2. B1 RCP Suction

c. Unit 2 Letdown Line 3. Al RCP Discharge  ;

d. Unit 2 Decay Heat Removal Line 4. B1 RCP Discharge j

5. 82 RCP Discharge

6. A Hot Leg

7. B Hot Leg QUESTION 2.03 (1.00)

The attached drawing, Figure 7.29 shows the LPI system  !

aligned for which one of the following modes of operation?

a. Switchover mode on Unit 1  ;

b. Swi tc hover mode on Unit 2 {

c. Normal decay heat removal on Unit 2

d. Normal decay heat removal on Unit 3 I

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2 RLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 8 QUESTION 2.04 (1.00)

Which one of the following correctly describes the response of the Type "B" RZ Module to a contol power loss?

a. If vital control power to the Auto / Manual ES logic is interrupted - position indication and manual control power will not be available to the manual control P/Bs.

b. If the vital power source is lost with no emergency signal present, no effect will be seen on the Auto / Lamp /Pushbutton,

c. If the vital power source is lost, digital control logic remains operable.

d. If vital power is restored while the emergency signal is present operation to the emergency signal present logic must be manually restored.

QUESTION 2.05 (1.00)

A valid EFW start signal exists. Unless otherwise specified all appropriate controls are in automatic. Which one of the following conditions would prohibit the injection of EFW l into the OTSG by the Turbine driven pump?

a. The low oil pressure switch PS-301 has failed low.

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b. EFWPT control in Pull-Te-Lock and a failure of KV1D Brk 6 (solenoid power supply) which trips open.

l c. While selected t o P r i ma r.y. _ .l e v e l _. con t erL_Aloss_.oLpnwnr_ __ _ __ __ --_ -- J to the Primary channel occurs,

d. The valve position limit switches for MS-93 Steam Supply Valve fail to recognize the valve opening.

QUESTION 2.06 (1.00)  !

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a) Aside from CCW Intake water, how else Can water be I supplied to the CCW Intake area if it is required for j operation of the SSF Auxiliary Service Water System. i l

b) How is air injection into the S/Gs minimized when using l the SSF ASW System? l

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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 9 OUESTION 2.07 (1.00)

What is the basis for the following administrative controls that were recently instituted on the Main Steam to Auxiliary Steam interconnect 1ons' Only one unit s Main Steam is used to supply Auxiliary Steam via only the 2" reducer MS-129 The other two units' Main to Aux Steam reducers, MS-126(6") and MS-129 are totally isolated.

QUESTION 2.08 (1.00)

Describe the two (2) intersystem ties between the HPI system and the SSF Makeup system. Include in your answer the source and discharge flow path.

QUESTION 2.09 (1.00)

Explain the basis for each of the following system limitations:

1. Maintaining Letdown Storage Tank level above approx. 18 inches.

2. Maintaining Letdown Storage Tank pressure vs. level curve within the operating range.

QUESTION 2.10 (2.00)

List the two auto start signals (including setpoints) for the Emergency Feedwater System and the two setpoints/ conditions at which level will be automatically controlled.

i QUESTION 2.11 (2.00)

List the six different flowpaths of electrical power to the Oconee Nuclear Station, (including the appropriate transformer and buses supplied) in their order of preference. Assume that the applicable main generator is producing > 200 MWe when it is considered as a source of power.

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p. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 10 j I

l OUEST10N 2.12 (1.00)

Provide 4 distinct problems that could result due to an ,

inadvertent ES actuation (consider all channels) if operator action is not taken promptly to correct the situation.

I OUESTION 2.13 (1.50) l a) Explain why AUTOMATIC control of steam header pres,sure on the Turbine Hand / Auto station, with the ICS in TRACK, is not a preferred mode of operation at low power levels?

l b) Dur ng st -tup hat is d ne to ensure the 't will ur ine plac no be s w an th t a .1 OUESTION 2.14 (1.50)

If a loss of Instrument Air to the air-operated valves in the Makeup portion of the HPI System occurred, what 3 l methods / alternate flow paths could be utilized to maintain I pressurizer level, assuming that "A" HPI pump is in service at the time of the failure?

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3. INSTRUMENTS AND CONTROLS PAGE 11 QUESTION 3.01 (1.00)

Which one of the following will NOT cause the ICS to enter the TRACKING mode of operation?

a. Placing the Diamond Control Station in Manual,

b. Placing BOTH Main Feedwater Valves in manual control.

c. Providing the turbine with 45% more power than is being produced by the generator,

d. Feedwater Cross Limits in effect.

QUESTION 3.02 (1.00)

Which one of the following conditions correctly describes the requirements for a 230KV Switchyard Isolation to occur >

a. Undervoltage on 2 of 3 phases on EITHER the Red or Yellow Bus on BOTH Channels of UV protection.

b. Undervoltage on 4 of the 6 phases monitored between the j Red and Yellow Buses on EITHER Channel of UV protection.

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c. Undervoltage on 4 of the 6 phases monitored between the l

Red and Yellow Buses on BOTH Channels of UV Protection. '

d. Undervoltage on 2 of 3 phases on BOTH the Red and Yellow Buses on BOTH Channels.

e. Undervoltage on 2 of 3 phases on BOTH the Red and Yellow Buses on EITHER Channel.

QUESTION 3.03 (1.00)

Which one of the following correctly describes an interlock associated with Component Cooling Discharge Valve CC-8?

a. If CC-8 is closed, NEITHER CC Pump may be started.

b. CC-8 closes on actuation of ES-1 or ES-2.

c. CC-7 (MOV CC Discharge Valve) closes if CC-8 closes.

d. If CC-8 closes any operating CC pumps will continue to run.

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3. INSTRUMENTS AND CONTROLS PAGE 12 QUESTION 3.04 (1.00) h i

Which one of the following correctly describes now control of ES components is obtained on a ~ype "A" RZ Module assuming an emergency signal is still present? j l

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a. Control can NOT be obtained until the ES signal clears.  ;

b. Depress the Manual pushbutton for the component, which will restore the component to its state prior to the ES.

and then control may obtained at the control board switch only,

c. Depress the Manual pushbutton for the component, and then control may be obtained at the control board swi tc h only.

d. Depress the Manual pushbutton for the component, and then control may be obtained locally only.

QUESTION 3. S (1.00) )

Which one of t ' following correctly describes the local swi tc hgea r contr s for the SSF? l I

a. Breaker interlocks revent closure of a break the l manual close pushbut on (located on lowerpf(pt'<by ht hand I portion of the breaker when the breake,pris in the open  ;

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b. Local closing of the breaker up ng the control switches (different from the manual PS can be accomplished as long as the breaker is not Iacke completely out. l

c. Local tripping of th breakers using the control switches (different from t manual PBs) is per.'tted only in the l Test position.

1

d. With a 1 of control power the ability to rip a breake with the manual open pushbutton is los .

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

e 1 INSTRUMENTS AND CONTROL _S PAGE 13 QUESTION 3.06 (1.00)

Which one of the following conditions would result in an Out Inhibit being generated in the Rod Contol Logic?

a. Safety Roo Groups at the out limit,

b. Asymmetric fault with power level at 50%.

c. A startup rate of 2.0 DPM in the Source rance,

d. High neutron error signal (2.5%).

l QUESTION 3.07 (1.00)

Which one of the following is a symptom of a bellows failure on Reactor vessel level instrument LT-5?

a. A lower than actual reading.

b. A higher than actual reading.

c. A false zero level indication.

d. Oscillations between high and low levels, j

OUESTION 3.08 (1.50)

Indicate the INITIAL RESPONSE of the following parameters and components following the failure of "A" S/G Outlet Pressure HIGH: (Assume plant initially at 80% and no trip occurs) a) Feedwater flow to "B" S/G b) "A" Bypass Valves c) Generated MWe

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

l.

3. INSTRUMENTS AND CONTROLS PAGE 14 QUESTION .3.09 (1.00)

Fill in the blanks td' correctly complete the following statements regarding the Pressurizer PORV:

The lower setpoint of 475 psig for NDT protection is selected by use of a . The PORV may be operated manually by the use of a located at the . The PORV is actuated by a pilot valve which is connected to a system.

QUESTION 3.10 (1.00)

Refer to figure DC-EL-EPD-1 " Basic Logic Diagram" attached, for each of the given inputs A,C,D,F determine the resultant output of B.E and G. (state answer in terms of 0 or 1)

OUESTION 3.11 (1.00)

Should a control rod "in" movement signal be generated and the rods run in the out direction, what Diamond Control Panel Lamp would illuminate AND what auto action would result from this condition?

l GUESTION 3.12 (1.00) 1 List the two interlocks that must be satisfied in order to start the SSF ASW pump if an ES-1 or 2 (loadshed) signal is present.

i l DUESTION 3.13 (1.50)

List the 6 modifications made to the Reactor Protection System (either automatically or manually) when the Shutdown Bypass Key Switch is placed in ' Bypass".

l l

l l

I

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

l 1

3 INSTRUMENTS AND CONTROLS PAGE 15 QUESTION 3.14 (1.00)

Place the following RPS components in the order they exist in a typical RPS Channel and indicate the number of each of these components per channel:

1) K Contact Logic Relay (s)

2) Interlocking Test Contact (s)

3) Manual Trip Switch Contact (s)

4) Coincidence Logic Contact (s)

OUESTION 3.15 (1.00)

List two (2) automatic actions which occur affecting the SSF in the event a Unit 2 Channel A load shed signal is received.

QUESTION 3.16 (1.00)

Indicate with which unit (s) the following High Pressure Injection System interlocks are associated:

a) On low seal injection flow, the StanJby HPI pump will start.

]

b) If seal injection is lost and the Component Cooling is lost, the associated Seal Return Valve will close.

QUESTION 3.17 (1.00)

According to AP/1/A/1700/i3 Section B " Dam Failure Without Loss of CCW Intake Canal" an operator is to be dispatched to place the TDEFDWP Cooling Bypass switch to the " Bypass" position. Explain what this action would accomplish with regard to the TDEFDWP cooling system.

I l

1

(***** END OF CATEGORY 03 *****)

1

. PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE 16 RADIOLOGICAL CONTROL QUESTION 4.01 (1.00) {

j Which one of the following methods contained within the Emergency Procedures is the best method for removal of RCS q

voids that are due to the presence of Non-condensible gases? j

a. Reoressurization of the RCS. l 1

-l

b. RCP Restart.

c. RCP Bumping.

]

l

d. Vessel or Hot Leg venting.

QUESTION 4.02 (1.00)

Upon a loss of 1KI AP/1/A/1700/23 directs an operator to the Aux Shutdown Panel to perform various actions. Which one of the following is an action the operator can perform from this pane 17 i I

a. Re-energize TBVs  !

i

b. Bypass INI Inverter 1

c. Control RCS pressure with Pressurizer Heater Banks 1, 3 and 4

d. If conditions warrant trip 1 RCP in 'B' loop I

l J

OUESTION 4.03 (1.50)  ;

l 4

Indicate what the exposure limits are for the following I conditions of exposure: (assume NRC-FORM 4 is current) .

I a) Oconee Maximum yearly permissable Whole Body l b) Maximum Planned Emergency Exposure to the Whole Body to Save Lives c) NRC Ouarterly Skin Exposure Limit  !

I l

(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****) j I

i L___________________________________________________________.___________.____

,4 . PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE 17 RADIOLOGICAL CONTROL OUESTION 4.04 (1.00)

List the three breakers in the correct order in which they must be operated in order to allow the Keowee Hydro Generator to produce voltage output, once the wicket gate position is established.

QUESTION 4.05 (1.00)

Place the following Emergency Procedures in the correct l order of priority in which they are referred to when performing EP/1/A/1800/01, and which are continually monitored until plant conditions stabilize: (Assume that a S/G Tube Rupture was NOT the initial entry condition)

1) Excessive Heat Transfer

2) Loss of Heat Transfer

3) Loss of Subcooling

4) Steam Generator Tube Leak I

OUESTION 4.06 (2.00)

AP/1/A/1700/19 " Loss of 4160v Power and the BWST" has the operator realign system components as a result of the failures. For EACH of the following state the resultant source of elec trical / water supply to the component.

1. HPI pump suction.

2. HPI pump electrical power.

3. Standby bus #1.

4 HPI pump motor cooling.

1 1

OUESTION 4.07 (1.00)

! EP/1/A/1800/01 Section 506 " Unanticipated Nuclear Power l

Production", has the operator " Verify open 1HP-5 (Letdown Isolation)" prior to initiation of Emergency Boration.

Explain the need for this action step.

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

'l 4 PROCEDURES - NORMAL. ABNORMAL,' EMERGENCY AND PAGE 18

~

RADIOLOGICAL CONTROL l

'OUESTION- 4.08 (;1.50)  ;

Aside from the Main FDW Control Valve (FDW-32), list the other six valves which must be closed in order isolate the A OTSG if it is ruptured.  ;

j i

i OUESTION 4.09 (2.00) l Besides tripping the reactor, starting the Keowee Hydro Units and announcing to/ notifying the proper personnel, what are the remaining immediate actions if the control room must be evacuated, as stated in AP/1/A/1700/8? Assume that time  ;

exists to complete actions prior to evacuation, and that the 1 SSF will NOT be required to be put in operation. Include in l your answer where operators are dispatched to and what I materials are required to be taken out of the control room. j l

i OUESTION 4.10 (1.00)  ;

)

What are the two criteria that must be met in order to I utilize OP/1/A/1102/02, " Reactor Trip Recovery".

1 1 QUESTION 4.11 (2.00)  !

If the HPI System has actusted due to a low pressure condition, what are the twt . c ri teria , of which either one must be met, that must be ctq=idered to secure the HPI l

System' 1

QUESTION 4.12 (1.00)

If the RCPs can be effective in cooling the core, even with two phase flow, why must the RCPs be secured when 0 subcooling is noted?

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

l

4 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 19 RADIOLOGICAL CONTROL 4

i OUESTION 4.13 (1.00)

Recently, during testing of the Emergency CCW siphon flow, problems developed resulting in loss of siphon. Among the corrective actions were some procedural changes. Provide the basis for the following procedural steps in AP-11, " Loss i of Power": I a) Steam is isolated to the Condenser Steam Air Ejector

)

first stages, j b) TDEFW Pump suction is aligned to the hotwell from the j UST as level in the UST decreases. 1 1

QUESTION 4.14 (1.00) i Refer to attacheo Enclosures 7.3A and 7.5. i Assume no RCPs are operating and the RCS is in a natural circulation cooling mode. The Steam Generator operating l range indications have failed and the extended startup range  ;

is being used to control level. If a level equivalent to i 85% OR is desired and the following conditions exist, DETERMINE the indicated startup range level to maintain.

Conditions are SG pressure = 900 psig and RB temperature =

l 250 deg. F.

l l

(***** END OF CATEGORY 04 *****)

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

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O l- I  ::: ,::::;I , I.. ' 0 l 0 200 400 600 800 1000 i SG PRESSURE (psig)  ;

1 l

1 1

i l

DIRECTIONS: TO CONVERT O.R. LEVEL TO XSUR i

1) READ SG OUTLET PRESSURE FOR DESIRED SG FROM CONTROL ROOM INDICATION

2) INTERSECT THIS PRESSURE LINE WITH DESIRED O.R. LEVEL

3) READ REQUIRED XSUR INDICATION AT INTERSECTION CAUTION Refer to Enci 7.5 to correct the XSUR level for i high RB temperature as necessary.

DLG/9-10-85 L

EP/1/A/1800/01 Page 1 et 1

e. - ENCLOSURE 7.5 LEVEL CORRECTIONS FOR HIGH REACTOR BUILDING TEMPERATURE Corrections to Indicated Level Reactor Steam Generator Core Building Startup operate Full Flood Temp Pressurizer Range Range Range Tank

('T) (inches) '

(inches) (1) A (ft) 100 6 1 1 1 0.0 150 8 6 2 1 0.2 200 15 . 13 4 3 0.4 250 24 21 6 5 0.7 300 33 30 8 8 1.1 350 45 41 11 '10 1.4 400 58 50 14 14 1.9 Modification Factor to Indicated Level Corrections NOTE: Use the Modification Factor on the temperature compensated scales -

only (Steam Generator Operate Range Level and Pressurizer Level)

Steam Generator Operate Pressurizer Pressurizer Downconer Range Temp (*F) Level Temp ('T) Level 68 1.00 68 1.00 150 1.02 150 1.02 200 1.04 200 1.04 250 1.06 250 1.06 300 1.09 300 1.09 350 1.12 350 1.12 400 1.17 400 1.17 450 1.23 450 1.21 500 1.31 500 1.31 550 1.42 550 1.36 ,

600 1.61 1 650 1.99 Calculation of Actual Level Pressurizer Actual = Indicated level - (Correction x Modification Factor)

Steam Generator Operate Range Steam Generator Actual level = Indicated level - Correction Startup Range Full Range Core Flood Tank EXAMPLE: Reactor Building Temperature 250*F Indicated Startup Level 220" Correction for R. B. Temperature 21" Actual Level = 220" - 21" 199" EXAMPLE: Reactor Building Temperature 300*F Pressurizer Temperature 400*F Indicated Pressurizer Level 100" Correction for R. B. Temperature ' 33" Modification Factor 1.17 Actual Level = 100" - (33" x 1.17) 61"

l J... PflNCIPLES OF. NUCLEAR POWER PLANT OPERATION. PAGE 20 i THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW .

e-ANSWERS - OCONEE 1, 2&3 -87/07/13-CASTO,.C. 1 ANSWER 1.01 (1.00) b Qb d N 4

REFERENCE OP-OC-SPS-PTR-AT pp 13/14; LO la (4.1/4.7) 000074A207 ...(KA'S)

ANSWER 1.02 (1.00) b O(L C.

REFERENCE OP-OC-TA-NT Figure 15; LO lo (3.6/3.8) 016000G015 ...(KA'S)

ANSWER 1.03 (1.00)

A- 3.

B- 2.

C- 4.

D- 1.

REFERENCE Oconee HD OC-RT-SM-2 obj 2m. 3.8/3.8 192008N103 ...(KA'S) I i

l ANSWER 1.04 (1.00) 1 d.

REFERENCE Oconee OP-DC-SPS-RT-IP pp. 28 obj 2a 2.8/3.1 192005K105 ...(KA'S) l i

l 1 C_________________...____ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ __ _ . _ _ _

.y.. RRINCIPLES OF NUCLEAR PO_WER P_LANT OPERATIONi- '

PAGE. 21-

. THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW '

c.

ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C.

ANSWER 1.05 (1.00) c.

REFERENCE Oconee OP-OC-SPS-RT-IP pp. 14 Obj 1.K 3.1/3.4 001000K519 ...(K4's)

ANSWER 1.06 (1.00) a) Increase (+.5) l b) Decrease REFERENCE OP-OC-TA-NT pp 16; LO ik ANSWER 1.07 (1.00) a) Lower (+.5 ea) b) Lower REFERENCE OP-OC-SPS-RT-RBC pp 26/27; LO 2b.4 (3.6/4.2) 001010A207 ...(KA'S)

ANSWER 1.08 (1.00)

1. less than

2. greater than REFERENCE Oconee OP-OC-SPS-APC-T41 3.8/3.9 192002K114 ...(KA's)

-l 1

1 i

l -l w - _ - _- - - _ _-_ _ ____ _ _ - - - -_ - _ _a

1. P_RINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 22 T'HERMODYN MICS, HEAT TRANSFER AND FLUID FLOW O

ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO. C.

ANSWER 1.09 (1.00)

Quench Tank Pressure (+.5) stops increasing (+.5)

REFERENCE OP-OC-SPS-CM-PZR pp 17; LO 11 (2.6/2.8) 007000A206 ...(KA'S)

ANSWER 1.10 (2.00) a) Increase due to the decrease DP across the FDW control valve (+.5) Levels off due to reaching its high speed stop (+.5) b) Tavg has decreased, so power is increased (+.5) c) Loop B TC has increased, Loop A Tc has decreased (+.5)

REFERENCE OP-OC-TA-NT pp 13/14; LO 1h (3.1/3.4) 059000A212 ...(KA'S)

ANSWER 1.11 (2.00) a) Unit 3 (+.5) due to a lower Beta coefficient at EOC(+.5) b) Unit 2 (+.5) due to MTC being less negative, so Tavg must decrease more to add + reactivity (+.5)

REFERENCE l DPC Fundamentals of Nuclear Reactor Engineering, CH 3 )

(2.9/3.2) (2.9/3.1)  !

000003K116 192004K103 ...(KA'S)

ANr,WER 1.12 (1.00) ]

These rods are BRAY (lower absorbtion cross section) and  ;

have a longer effective poison length. (+.5 They have a #

less severe impact on Axi 1 Flux Imbalance and minimize induced xenon oscillations (+.5) l I

,1 . , , PRINCIPLES OF NUCLEAR POWER PLANT OPERATION t PAGE- 23 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW e

ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C.

REFERENCE OP-OC-SPS-THF-PD pp 12; LO 21.4 (3.2/3.5) 192005K114 ...(KA'S)

ANSWER 1.13 (2.00) a) Increases (+.5) Pzr temperature increase will cause a pressure increase, increasing margin to saturation (+.5) b) Increases (+.5) Delta t across the core will be lower to produce the same power. Th wil decrease and the coolant in the upper regions will be farther from saturation.

(Also higher flow removes bubbles from rod surface)(+.5)

REFERENCE DPC Thermodynamics / Fluid Flow pp 196-198 (3.4/3.6) 193008K105 ...(KA'S)

ANSWER 1.14 (1.00) at 2240 psia, hg = 1115 BTU /lb (+.5) at 20 psia, at saturation conditions, hg = 1156 BTU /lb and hf = 196 BTU /lb calculate: (1156-1115)/(1156-196) = .043 >>> 95.7% quality If use Mollier: 96% quality (+/- 1%)

REFERENCE OP-GA-SPS-THF-STM pp 20/21; LO 2e (3.3/3.4) 193003K125 ...(KA'S) 1 ANSWER 1.15 (1.00)

C REFERENCE DPC Thermodynamics, pp 192-5;  !

(3.1/3.4) 1 002000K501 ...(KA'S) I i

I j

J2..' PLANT' DESIGN INCLUDING' SAFETY AND EMERGENCY SYSTEMS PAGE 24

' ANSWERS.-- OCONEE 1, 2&3 -87/07/13-CASTO, C.

-l ANSWER 2.01 (1.00)

REFERENCE U OP-OC-SPS-SY-EF pp 21/22, 32/33; LO le, if (3.4/3.5) 061000A207 ...(KA*S)

ANSWER 2.02 (2.00) a) 3 (+.5 ea)  ;

b' 3 l c) 2 )

d) 6 REFERENCE OP-OC-SPS-SY-RCS pp 12/13; LO la (4.1/4.1) (3.7/4.0) (4.5/4.6) 002000K106 002000K108 002000K109 ...(KA'S) l ANSWER 2.03 (1.00) i d

REFERENCE I Oconee SY-LPI-5 3.2/3.5 I 005000K402 ...(KA*S)  ;

I i

ANSWER 2.04 (1.00)  ;

b. i REFERENCE Oconee Op-OC-SPS-IC-ES obj . h 3.6/4.2 3.7/4.2 013000A204 013000A205 ...(KA'S)  !

,23, _ PLANT DESIGN INCLUDING SAFETY AND_ EMERGENCY SYSTEflS PAGE 25

'. ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C.

ANSWER 2.05 (1.00) d.

REFERENCE Oconee OP-OC-SPS-SY-EF objs 1.b/1.a/1.o 3.4/3.8 061000A204 ...(KA*S)

ANSWER 2.06 (1.00) a) A Submersible Pump is installed to discharge into the nearest CCW Piping manway. (+.3 ea) b) An Air Efector is valved into the SSF ASW Suction piping when,the SSF is actuated.

[-(or A.T h ,4.%Mc cdd[NCAAS dunutti hr 'qppkab 7 ll5 fo yc4- LAbr bM b f D b)

REFERENC'E G 4 W DA Gir {-hoM AtW 4o CT% ukkr (b))

OP-OC-SPS-SSF-ASW Tracking 87-034/030; LO 2b (3.9/4.2) 061000K401 ...(KA'S)

ANSWER 2.07 (1.00)

Possible overpressurization of Aux Steam and subsequently, steam line to TDEFWP if both control valves were to fail open due to too low of relief capacity in both systems.

REFERENCE Oconee LER 87-003 (3.3/3.4) 030000K107 ...(KA'S)  !

ANSWER 2.08 (1.00)

A letdown line from the Common letdown line (prior to the L/D coolers) thru HP-(426) to the fuel transfer tube.[0.5]

A line from the SSF Makeup Pump which takes suction on the altrenate fuel transfer tube, injects water through j SSF-HP-(398), This line serves as system makeup.[0.5)  !

REFERENCE Oconee OP-OC-SPS-5/-HPI obi 3.a/b 3.3./3.2 004000k405 ...(KA'S) i

- a

2... RLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE '26

• ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO., C.

ANSWER 2.09 (1.00)

1. To insure suction to the HPI pumps'is not lost [0.5]

2. Insure gas does not enter the HPI Pump suction on an ESFAS [0.5]

REFERENCE Oconee OP-OC-SPS-SY-HPI pp. 19 obj S.b 2.8/3.0 3.1/3.4 0040010K01 004020A104 ...(KA'S)

ANSWER 2.10 (2.00)

1) Both MFWPT hydraulic oil pressure < 75 psig (+.5 ea)

2) ,Both MFWPT discharge pressure < 750 psig 3)P X with loss of both MFWPT

4) 240" with loss of both MFWPT & All RCPs REFERENCE )

OP-OC-SPS-SY-EF pp 64; LO ik, 11 (4.5/4.6) 061000K402 ...(KA'S)

ANSWER 2.11 (2.00) l

1) Main generator to IT to ITA/TB and MFB1/2 ( + . 33 ea khr- amyC )

l 2) 230 kV Swyd to CT1 to ITA/TB and McB1/2

3) keowee to CT1 to ITA/TB and MFB1/2 , g gj g gg , g/g g

("e() keowee to CT4 to Standby Bus 1/2 4)d Lee Gas Turbines to CTS to Standby Bus 1/2 gy() 230KV or 525KV Back Charging to IT to ITA/ITB and MFB1/2 REFERENCE OP-OC-SPS-EL-EPD pp 17/18; LO 2 (3.7/4.2) (2.9/3.3) 062000K104 062000K406 ...(KA*S)

,2 . , , PLANT DESIGN INCLUDING SAFETV AND EMERGENCY SYSTEMS PAGE 27 ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C.

ANSWER 2.12 (1.00) 1

1) Excessive Boration of Plant (+.25 ea)

2) Plant over pressurization j

3) Chemical Spray Hazard to RB Components s j

4) Loss of Cooling Water to necessary componen ts h,gp4p od pron #AJ U' S 1 t n d tOrdual (t.m pp.< h t'0(k REFERENCE m g.p. g m gnjeJ OP-OC-SPS-IC-E5 pp 15; LO lj (3.7/4.0) 1 013000A206 ...(K4'5) i ANSWER 2.13 (1.50)

\

a) The Megawatt error is blocked in track, so the header i pressure error is controlling position of the control valves. (+.5) This causes instability in header ,

pressure, hence feed flow oscillations.

b) Al is ICS sta ions into au matic laced in -uto. (+.5) m g befdq the (Jrb e p

REFERENCE OP-OC-SPS-IC-ICS Tracking #86-064; LO lb (3.2/3.2) {

059000K107 ...(KA'5)  !

ANSWER 2.14 (1.50) l 1

1) Use HP-26 Motor operated valve (+.5 ea)hroMi -

2) Use HP-122 (Manual Bypass) gdd"

3) Use "C" HPI pump via HP-27 M y b,aAU'd pd vl REFERENCE ( ) sp*~ U(1-Qt0 OP-CC-SPS-5Y-HPI pp 29; LO 3c (T.4/3.6)  !

078000K302 ...(kA'5)  !

I

. 4

, i i

i o_____________________________________________ ____ __ J

,3 .. , INSTRUMENTS AND CONTROLS PAGE 28

• ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C.

ANSWER 3.01 (1.00) b REFERENCE OP-OC-SPS-IC-ICS pp 23/24; LO 1h (3.2/3.2) 059000N107 . . . ( k' A ' S )

ANSWER 3.02 (1.00) e REFERENCE l OP-OC-SPS-EL-EPD pp 24; LO 3c.5 (2.6/3.2) 1 062000K401 ...(KA'S) l ANSWER 3.03 (1.00) a REFERENCE OP-OC-SPS-SV-CC pp 19: LO 2c (3.2/%,2) ,

008012 301 ...(FA'S) l ANSWER 3.04 (1.00)

\ c t

t REFERENCE gCP-OC-SPS-IC-ES pp 18; LO 1h

• (3.3/3.7) ,

013000K410 ...(KA'S)

ANSWER 3.05 (1.00)

__---._-....-__J

I J ., I.NSTRUMENTS AND CONTROLS PAGE '29

• ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. 1 REFERENCE Oconee OP-OC-SPS-SSF-EPS 2.6/2.7 062000A404 ...(KA*S)

ANSWER 3.06 (1.00)

C

)

REFERENCE l

Oconee OP-OC-SPS-IC-CRI 1.m. 3.5/3.8 s 001000K401 ...(KA'S)

ANSWER 3.07 (1.00)

a. o $L C.

REFERENCE Oconee Op-OC-IC-RCI obj. 3.g. (3.1/3.6) 002000K603 ...(KA'S)

ANSWER 3.08 (1.50)

CDnct1M463 a) h (+.5 ea) b) Opens c) Decreases REFERENCE OP-OC-SPS-IC-ICS. pp 80/81; LO is (3.0/3.1), (3.4/3.6) 016000K303 016000K312 ...(KA'S) i ANSWER 3.09 (1.00) key selector switch; key switch; ICS cabinet 13; Cl,$2 IE DC--- (+ RFea) ~

33 REFERENCE ,

OP-OC-SPS-CM-PZR pp 14; LO in I (3.8/4.1) 010000K403 ...(KA'S) i l

l

_ _ _ _ _ _ _ _ _ . _ _ _ - _ - _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _. _ _ _ _ _ _ _ a

3 ., INSTRUMENTS AND CONTRCLS PAGE 30

. ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C.

ANSWER 3.10 (1.00)

B-1 E-0 G-1 REFERENCE Oconee OC-EL-EPD-1 obj .1. a /1. b 2.5/3.2 194001A107 ...(KA'S) l ANSWER 3.11 (1.00) l the Motor Fault Lamp illuminates [0.2S] and the Diamond station swaps to manual [0.75].

REFERENCE Oconee OP-OC-SPS-IC-CRI 1.1/1.g 3.8/3.8 001402k402 ...(KA'S) i I

ANSWER 3.12 (1.00)

1) the SSF incoming feeder breaker from Unit 2 Main Feeder l Bus #2 (OTSI-1) is open (+.5 ea) l

2) SSF Diesel Generator Breaker OTSI-4 is closed i

1 REFERENCE OP-OC-SPS-SSF-ASW pp 24; LO 2h (4.0/4.2) 061000K406 ...(KA'S)  !

)

ANSWER 3.13 (1.50)

1) Bypasses Low Pressure Trip (+.25 ea)

Variable Low Pressure Trip Power /RCP Trip Flux / Flow /qlmbalance Trip l 2) Inserts 1720 psig aM$ Pressure Trip setpoint

3) High Flux setpoint is reset t o 4*/.

REFERENCE OP-OC-SPS-IC-RPS pp 16; LO le (3.3/3.6)

Ol2000K604 ...(kA~5) o

,33, INSTRUMENTS AND CONTROLS PAGE 31'

• ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C.

ANSWER 3.14 (1.00)'

2-emm- t . is enr nomhme ~ iur currecu vr um .?

LW

=dt ( Sud mcf . 33 Ot- bck $1wW refvtrk!

3 tu,e- 'fo Qgik tk corrit+ DftlW)

REFERENCE OP-OC-SPS-IC-RPS pp 39-45/ fig 24b; LO 2a (3.1/3.5) l 012000k603 ...(KA'S)

ANSWER 3.15 (1.00)

Normal incoming swi tc hgear breaker OT51-1 trips SSF Supply breaker B2T-4 trips

[0.5 ea]

REFERENCE Oconee OP-OC-SPS-SSD-EPS pp. 36 obj 1.1. 2.8/3.1 062000K403 ...(KA'S)

ANSWER 3.16 (1.00) a) All three units (+.5 ea) b) Units 2 and 3 REFERENCE OP-OC-SPS-SY-HPI, pp 36/37; LO 3q (2.8/3.1) 003000K404 ...(KA'S) l ANSWER 3.17 (1.00)

(this would deenergize both SV-209 and -208) causing LPSW-138 to fail open supplying HP5W to EFWPT pump bearings

[0.5) and causing HPSW-184 to open providing cooling water to the oil cooler [0.5]

REFERENCE Oconee AP-13 pp. 8 obj. 1.0, 1.p., 3.1/.3.3 000075k007 ...(kA'S) i

4 PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 32 RADIOLOGICAL CONTROL.

ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C.

ANSWER 4.01 (1.00) d REFERENCE DPC EPG pp 2-77/79; (4.0/4.4) 000074K311 ...(KA'S)

ANSWER 4.02 (1.00) a.

REFERENCE Oconee AP/1/A/1700/23 3.5/3.5 3.2/3.4 000057A105 000057A106 ...(KA'S)

ANSWER 4.03 (1.50) a) 4.5 rem (+.5 ea) b) 7 5 rem o tt E $ (IEM c) 7.5 rem REFERENCE l DPC HPM, pp 9/10; Emergency Plan, pp K-1 l (2.6/3.4) 194001K103 ...(KA'S)

ANSWER 4.04 (1.00)

1) Close the Field Breaker (+.33 ea)

2) Close the Generator Supply Breaker 1

3) Close the Field Flashing Breaker REFERENCE i OP-OC-SPS-CM-kHG pp 20; LO ik l (4.0/4.3) l 064000A401 ...(KA'S)

4

,43,_fROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 33 RADIOLOGICAL CONTROL o

ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C.

ANSWER 4.05 (1.00) 3, 2, 1, 4 ( .25 for each switch to get correct order) l 466 Mif 96 5 REFERENCE j EP/1/A/1800/1, pp 9/10; Reference Document pp 1-8 )

(3.8/3.9) 000007G012 ...(KA'S)

J I

)

ANSWER 4.06 (2.00)

1. From the Spent Fuel Pool

2. Aux. Service Water Switchgear or grA&O6Y Sui #/

3. From CT-5 oc LEE. sr& OK G&Mk wO 4 From the Aux. Service Water System o r HPI u/

[0,5 ea.] ,

l REFERENCE AP/1/A/1700/19 Oconee 3.7/4.1 3.9/4.7 0000554203 000062A211 ...(KA'S)

ANSWER 4.07 (1.00)

Letdown should be established to offset the increase in'RCS inventory due to the initiation of E-boration.

REFERENCE Oconee EP/1A/1800/01 4.4/4.7 000029K312 ...(KA'S)

ANSWER 4.08 (1.50)

1) SU Cntrl Valve (FDW-35) (+.25 ea)

2) EFDW Cntrl Valve (FDW-315)

3) TBV Block Valve (MS-17)

4) MS to SSRH (MS-79)

5) MFW Block Valve (FDW-31)

6) SU Block Valve (FDW-33)

REFERENCE Oconee OMP 2-1, pp i encl 4.4 (4.1/4.2)

E 4 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 34 RADIOLOGICAL CONTROL ANSWERS.-- OCONEE 1, 2&3 -87/07/13-CASTO, C.

000040G010 ...(KA'S)

ANSWER 4.09 (2.00)

1) Set batch size on makeup control to 90,000 gallons and reset (+.25 ea)

2) Open Makeup isolation (HP-16)

3) Open RC Bleed Transfer Pump "A" Discharge CS-46

4) Start "A" Bleed Transfer Pump

5) Dispatch operator to Units 1/2 Waste Disposal Panel

6) Go to Aux Shutdown Penel with: Reactor Log (+.5 total)

Emergency and Abnormal Procedures Removal / Restoration Book Emergency Plan

7) Maintain Hot Shutdown Conditions REFERENCE Oconee AP-8, pp 2/3; OMP 2-1, ENCL 4.4 (4.1/4.2) 000069G010 ...(KA'S)

ANSWER 4.10 (1.00HS ra^hrI-1

1) Reactor startup within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of trip (+.5 ea)

2) Cooldown has not been initiated.

C h D act y f") [d W CW 4 DY REFERENCE Oconee OP/1/A/1102/02, pp 1 (3.3/3.5) 001050G010 ...(VA'S) i l

ANSWER 4.11 (2.00)

(tFN

1) LPI is in operation, with a flow rate > 1000 gpm in each line and stable for > 22F OR r 20 minutes i)1.0 b f'#

2) All RCS hot and co ;d legs at least 50 de rees I ss t'h a n tne saturat ~on tem' for RCS pressu e and securi g HP I is nec pssary t preve t PZR lev 1 from going off s pie high PEFERENCE

( '

l l OP-OC-SPS-IC-ES pp 24; I

(4.1/4.6) 009000A324 ...(KA'S)

I

4 PROCEDURES - NORMAL. ABNORMAL,_ EMERGENCY AND PAGE 35

• ' RADIOLOGICAL CONTROL d

ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. 1 1

ANSWER 4.12 (1.00)

To minimize inventory depletion, such that if the pumps were  ;

lost inadvertently, core juncovery could occur. j j

REFERENCE Oconee EPG Reference Document, pp 4-45/46 ,

(4.1/4.2) j 000074K308 ...(KA 5) l ANSWER 4.13 (1.00) a) Allows the system to keep a vacuum on the condenser .]

b) Allows more UST water available to cool CSAEs (+.5 ea) j

• lL Csedit -for (011 & fuesten 40 theFwb REFERENCE Oconee LER 86-011; AP-11; (4.3/4.6) 000055K302 ...(KA'S)

ANSWER 4.14 (1.00) 284 inches +-3 inches REFERENCE Oconee EPs 3.7/4.1 103000A101 ...(KA'S) i

)

l L

e ,

s' h)$[WA s U. S. NUCLEAR REGULATORY CONMISSION SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION FACILITY: _gggNEE_11_2h3___________

REACTOR TYPE: _PWR-@gW1ZZ______________

DATE ADMINISTERED _@ZigZi13________________

EXAMINER: _g@gTg1_C._______________

2 CANDIDATE: _________________________

l IN@IBUCIlgNS_IQ_C9NpJpSIE1 l l

Read the attached instruction page carefully. This examination replaces the current cycle facility administered requalification examination. i Retraining r 4 4uirements f or failure of this examination are the sama as for failure of a requalification examination prepared and administered by your training staff. Points for each question are indicated in i parentheses after the question. The passing grade requires at least 70% I in each category and a final grade of at least.80%. Examination papers will be picked up four (4) hours after the examination starts.

l

% OF CATEGORY % OF CANDIDATE'S CATEGORY

__206UE_ _Igl@( ___SCgBE___ _y@6UE__ ______________g@IEGQBY_____________

_le z gg__ _ggtgg ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND

,- THERMODYNAMICS 249e@9'__ _29z99 ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION

_19z99__ _22z99 ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL

_183 99__ _29399 ___________ ________ 8. ADMINISTRATIVE PROCEDURES,

- CONDITIONS, AND LIMITATIONS

_4?f79"_ _ ___________ ________% Tot al s Final Grade All work done on this examination is my own. I have neither given nor received aid.

1 Candidate's Signature I

l l

L.

r .

. NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the f ollowing 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.  ;

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3. Use black ink or dark pencil gnly 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). l

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6. Use only the paper provided f or answers. )

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

i section of the answer sheet.

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8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write gnly gn gne 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 tht ge lines between each answer.

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

12. Use abbrevi ations only i f they are commonly used in facility literatute.

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.

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

17. You must sign the statement on the cover sheet that indicates that the j' 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  ;

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18. When you complete your examination, you shall:

a. Assemble your examination as f ollows:

(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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_THER_MO_D_YN_A_M_IC_S

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QUESTION 5.01 (1.00)

Which one of the following correctly describes the behavior of RCS pressure if a Small Break LOCA which was not large enough to actuate the ECCS were to occur, without Feedwater j' available?

a. Pressure initially decreases slowly, then rapidly drops when the OTSGs are boiled dry.

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b. Pressure decreases slowly until it levels cf f somewhere l above ECCS actuation pressure.

I l c. Pressure initially increases, then rapidly drops when the OTSGs are boiled dry.

d. Pressure initially decreases, then rapidly increases .

when the OTSGs boil dry.

e. Pressure initially decreases, then when OTSGs boil dry, continues to decrease, but at a much slower rate.

l QUESTION 5.02 (1.00)

Which one of the following instrument failures would cause the behavior of the parameters shown on attached drawing OC-TA-NT-157

a. Delta Tc Failure "A" Side LOW

b. Delta Tc Failure "A" Side HIGH

c. Delta Tc Failure "B" Side LOW

d. Delta Tc Failure "B" Side HIGH

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QUESTION 5.03 (1.00)

Which one of the following is correct concerning differential control rod worth (DRW)?

a. It is a measure of reactivity due to rod position.

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b. With a normal cosine flux shape, DRW reaches a maximum l value at a rod index of less than 28%. )

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c. Rod Group Overlapping maintains a constant DRW..

d. Its unit is delta K/K/% index.

l QUESTION 5.04 (1.00)

Which one of the following represents the maximum linear power density which would be expected in the core during i

full power operations?

a. Local Power Density multiplied by Nuclear Peaking Factor,

b. Radial Peaking Factor multiplied by the Local Peaking Factor. )

c. Average Kw/ft for the core multiplied by the Nuclear Peaking Factor.

l d. Nuclear Peaking Factor multiplied by the Maximum Local l Power Density.

QUESTION 5.05 (1.00)

Which one of the following conditions would hinder natural circulation flow in the RCS?

a. An increase in thermal driving head.

l b. An increase in the velocity head of the fluid.

c. An increase in the mass of feedwater into the DTSG.

d. A decrease in Two-phase (nucleate boiling) within the Core.

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l QUESTION 5.06 (1.00)

Assuming that the pl ant is operating'at full power, what )

will be the difference in the following parameters, if OTSG l tube fouling has occurred to a significant degree?

a) OTSG Level  !

b) Superheat Temperature j QUESTION 5.07 (1.50)

With the Unit operating at 100% power with all control systems in automatic, a Turbine Bypass Valve fails full ,

open. Indicate how the following parameters will change relative to their initial values when plant conditions stabilizes (INCREASE, DECREASE, REMAIN THE SAME) a) Tavg b) MWe c) Reactor power l QUESTION 5.08 (2.00) l l For each of the following parameters describe the indication i which is used to ensure adequate coupling following a loss of RCP's with one steam generator available.

a. That on both loops

b. Tcold on the operating generator I c. Cooldown rate on both loops l d. Thot - Tcold on the operating steam generator

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QUESTION 5.09 (1.00)

With regards to the Flux / Flow / Imbalance T.*ip function, state whether a Positive or Negative Flux Imbalance would allow a higher trip setpoint (assuming other pertinent parameters are the same). Explain your answer, i

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5'__IME9By_9E_NYCLg68_EgWEB_E68N1_9EE6@llQN1_ELUlp@g_6NQ PAGE 5 IUEBd99%NgDICQ QUESTION 5.10 (1.00)

What indication tells the operator when all nitrogen has been vented from the Pressurizer, when forming a steam bubble in accordance with OP/0/A/1103/05?

QUESTION 5.11 (2.00)

Refer to the attached drawing DC-TA-NT-10, "FDW Main Control Valve Fails Open at 50% Power", to answer the followings a) Why does FWPT Speed increase, then level off at point (2)?

b) What is causing reactor power to increase starting at point (3)?

c) After reactor power and MWe stabilize, prior to the reactor trip, what is the relationship between Loops A and B "T Cold"?

QUESTION 5.12 (1.50)

State three mechanisms that can produce high boron concentrations in the core post LOCA.

QUESTION 5.13 (1.00)

Explain how the use of RCP " bumps" minimizes the effect of gas AND steam accumulation within the RCS,

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QUESTION 5.14 (1.00)

The pressurizer PORV is leaking by during operation at 85% j power. Assuming 4 Quench Tank pres'sure of 20 psia and caturation conditions in the pressurizer corresponding to 2240 psia, what is the quality of steam on the downstream side of the PORV? Show all calculations.

a I

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. QUESTION 5.15 (1.00)-

The reactor..is producing 100% rated thermal power.'at a' core delta T of 60 degrees and a mass flow rate of 100*/. when a station blackout occurs. Natural circulation is' established and core delta T.goes to 40 degrees. If decay heat is 2%,

what is the core mass flow rate- (in %)?

a. 1.3

b. 2.0

c. 3.0

d. 4.2 l

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Az__P(8MI_gIgIEgg_QEgigNz_C9hI6962_8ND_lygI69dENISIlgN PAGE 7 QUESTION 6.01 (1.00)

Which one of thy f ollowing will NOT cause the ICS to enter the TRACKING mode of operati.on?

a. Placing the Diamond Control Station in Manual.

b. Placing BOTH Main Feedwater Valvas in manual control. .

c. Providing the turbine with 45% more power than is being  ;

produced by the generator.

d. Feedwater Cross Limits in effect.

QUESTION 6.02 (1.00)

Which one of the f ollowing should enable MS-93, TDEFWP steam supply valve, to open if it fails to open on an automatic j signal?

a. Verify proper operation of the DC Oil Pump.

b. Line up backup service air to the valve operator.

c. Isolate instrument air to its reducer and bleed the air off the reducer.

d. Take the control switch to "Off" to remove power from the solenoid valve.

QUESTION 6.03 (1.00)

Which one of the following correctly describes an interlock associated with Component Cooling Discharge Valve CC-B?

a. If CC-8 is clositd, NEITHER CC Pump may be started.

b. CC-8 closes on actuation of ES-1 or ES-2.

c. CC-7 (MOV CC Di r.harge Valve) closes if CC-8 closes.

d. If CC-8 closes any operating CC pumps will continue to run.

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6t__P(@NJ_QI@lgyp,Qg@lGNz _ggNI@p(g,@Np_ly@l@UggNI@]lgy PAGE 8 QUESTION 6.04 (1.00)

Which one of the f ollowing correctly describes the response of the Type "B" RZ Module to a contal power loss?

a. If vital control power to the Auto / Manual ES logic is interrupted - position indication and manual control power will not be available to the manual control P/Bs.

b. If the vital power source is lost with no emergency signal present, no effect will be seen on the Auto / Lamp /Pushbutton.

c. If the vital power source is lost, digital control logic remains operable.

d. If vital power is restored while the emergency signal is I present operation to the emergency signal present logic must be manually restored, l

l l QUESTION 6.05 (1.00) )

l e f ollowing correctly describes the local j Which one of '

switchgear con ois for the SSF?

a. Breaker inter ks prevent closure of a reaker by the manual close pu Qbutton (located cn Joder right hand portion of the breaker). When the reaker is in the en position.

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b. Local closing of the re ers using the control switches (different from the ma, al PBs) can be accomplished as long as the breaker 's not racked completely out.

c. Local tripping the breakhrs y using the control switches (different fr m the manual PBb is permitted only in the Test posit n.

d. With oss of control power the ah lity to trip a l bre er with the manual open pushbu ton is lost.

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- 6:__P,L@NI_gygIEDg_pEglGNg _CgNIggL 1_@Np_lNgIgUDENI@IlgN PAGE 9 QUESTION 6.06 (1.00)

A valid EFW start signal exists. Unless otherwise specified all appropriate controls are in automatic. Which one of the following conditions would prohibit the injection of EFW into the OTSG by the Turbine driven pump?

a. The low oil pressure switch PS-301 has failed low.

b. EFWPT control in Pull-To-Lock and a failure of KV1D Brk 6 (solenoid power supply) which trips open.

c. While selected to Primary level control, a loss of power to the Primary channel occurs.

d. The valve position limit switches f or MS-93 Steam Supply Valve f ail to recognize the valve opening.

1 QUESTION 6.07 (1.00)

Which one of the following conditions would result in an Out Inhibit being generated in the Rod Contal Logic?

a. Safety Rod Groups at the out limit.

b. Asymmetric fault with power level at 50%.

c. A startup rate of ?,0 DPM in the Source range.

d. High neutron error signal (2.5%).

QUESTION 6.08 (1.00)

Which one of the following is a symptom of a bellows failure on Reactor vessel level instrument LT-5?

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a. A lower than actual reading.

b. A higher than actual reading.

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c. A false zero level indication.

d. Oscillations between high and low levels.

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QUESTION 6.09 (1.00) i a) Aside from CCW Intake water, how else can. water be I 1

supplied to the CCW Intake area if it is required for operation of the SSF Auxiliary Service Water. System. l f

a b) How is air injection into the S/Gs minimized when using the SSF ASW System?

i QUESTION 6.10 (1.00)

What is the basis for the f ollowing administrative controls l that were recently instituted on the Main Steam to Auxiliary-Steam interconnections?

Only one unit's Main Steam is used to supply Auxiliary The other two I Steam via only the 2" reducer MS-129.

units' Main to Aux Steam reducers, MS-126(6") and '

MS-129 are totally isolated.

l QUESTION 6.11 (1.00)

Describe the two (2) intersystem ties between the HPI system and the SSF Makeup system. Include ir. your answer the source and discharge flow path.

QUESTION 6.12 (1.00)

Explain the basis f or each of the following system limitations:

1. Maintaining Letdown Storage Tank level above approx. 18 inches.

2. Maintaining Letdown Storage Tank pressure vs. level curve within the operating range.

QUESTION 6.13 (1.00)

List the two interlocks that must be satisfied in order to I start the SSF ASW pump if an ES-1 or 2 - (loadshed) signal is present.

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QUESTION 6.14 (1.00)

Provide 4 distinct problems that.coul'd result due to an l -inadvertent ES' actuation (consider all channels) if operator action is not.taken promptly to correct the situation.

QUESTION 6.15 (1.50) a) Explain why AUTOMATIC control of steam' header pressure on the Turbine. Hand / Auto station, with'the ICS in TRACK, is not a preferred mode of operation at low power levels?-

b)' Du ng a tup,. one to ur thel Unit' '

no be nT en the'T Contro is d in auto atic?

QUESTION 6.16 (1.00)

Indicate with which unit (s) the following High Pressure.

Injection System interlocks are associated:.

a) On low seal injection flow, the' Standby HPI pump will start.

b) If seal injection is lost and the Component Cooling is lost, the associated Seal Return Valve will close.

QUESTION 6.17 (1.50)

If a loss of Instrument' Air to the air-operated valves in ,

the Makeup portion of the HPI System occurred, what-3 methods / alternate flow paths could be utilized to maintain pressurizer level, assuming that "A" HPI pump is in service 3 at the time of the failure? ' -

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QUESTION 7.01 (1.00)

Which one of the following methods contained within the Emergency Procedures is the best method for removal of RCS {

voids that are due to the presence of Non-condensible gases? i i

a. Repressurization of the RCS.

b. RCP Restart. ]

c. RCP Bumping.

d. Vessel or Hot Leg venting. ,

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1 QUESTION 7.02 (1.00)

Which one of the f ollowing correctly oescribes the required actions if the reactor achieves criticality 1.5% (deltak/k) below the estimated critical posi ti on7 i

a. Fully insert all regulating rods, but the safety rods may stay fully withdrawn,

b. Fully insert all regulating rods, fully insert all safety rods to Group 1 at 50% withdrawn.

c. Fully insert all regulating and safety rods.

d. No rod insertion is required, but the ECP shall be recalculated.

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e. No action required and startup may continue.  !

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. Be9196991986_QQNIBQL QUESTION 7.03 (1.00)

Upon a loss of 1KI AP/1/A/1700/23 directs an operator to the Aux Shutdown Panel to perform various actions. Which one of the following is an action the operator can perform from this panel?

a. Re-energize TBVs

b. Bypass 1KI Inverter

c. Control RCS pressure with Pressurizer Heater Bands 1, 3 and 4

d. If conditions warrant trip 1 RCP in 'S' loop QUESTION 7.04 (1.00)

The attached drawing, Figure 7.28, shows the LPI system aligned for which one of the following modes of operation?

a. Switchover mode on Unit i

b. Switchover mode on Unit 2

c. Normal decay heat removal on Unit :2 i

d. Normal decay heat removal on Unit 3 QUESTION 7.05 (1.00)

List the three breakers in the correct order in which they must be operated in order to allow the Keowee Hydro Generator to produce voltage output, once the wicket gate position is established.

QUESTION 7.06 (1.00) a) Which two individuals, by title, may author;.e the bypassing of Main Fuel Bri dge interlocks?

4 b) Where are the bypass switchei ffor these interFocks

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, QUESTION 7.07 (1.00)

Place the following Emergency Procedures in the correct order of priority in which they are referred to when performing EP/1/A/1800/01, and which are continually monitored until plant conditions stabilizes (Assume that a S/G Tube Rupture was NOT the initial entry condition)

1) Excecsive Heat Transfer

2) Loss of Heat Transfer

3) Loss of Subcooling

4) Steam Generator Tube Leak QUESTION 7.08 (1.00) l While perf orming CP-604, " Solid Pl ant Cooldown", following an accident condition, it has you transfer to CP-605, "Subcooled Cooldown", if Pressurizer level gets less than

• 300 inches. What is the basis behind this setpoint being the criteria for this procedural transition?

QUESTION 7.09 (1.00)

Referring to the attached excerpt from OP/1/A/1102/1, Enclosure 4.1 for performing a Unit Startup, what is the significance of the " Bullets" preceding the substeps f ollowing step 2.1?

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l QUESTION 7.10 (1.00) J i EP/1/A/1800/01 Section 506 " Unanticipated Nuclear Power l Production", has the operator " Verify open 1HP-5 (Letdown Isolation)" prior to initiation of Emergency Boration.

Explain the need for this action step.

QUESTION 7.11 (1.50)

Aside from the Main FDW Control Valve (FDW-32), list the l other six valves which must be closed in order isolate the l

A OTSG if it is ruptured.

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l QUESTION 7.12 (2.00)

I Besides tripping the reactor, starting the Keowee Hydro j Units and announcing to/ notifying the proper personnel, what are the remaining immediate actions if the control room must ,

be evacuated, as stated in AP/1/A/1700/8? Assume that time {

exists to complete actions prior to evacuation, and that the I SSF will NOT be required to be put in operation. Include in your answer where opfrators are dispatched to and what materials are required to be taken out of the control room.

QUESTION 7.13 (1.50)

List six different methods of heat transfer recovery / core cooling contained within EP/1/A/1800/1, Section 502, " Loss of Heat Transfer", assuming that the Reactor Coolant Pumps are UNAVAILABLE.

QUESTION 7.14 (1.00)

What are the two criteria that must be met in order to utilize OP/1/A/1102/02, " Reactor Trip Recovery".

QUESTION 7.15 (2.00)

In accordance with AP/1/A/1700/18 " Abnormal Release of Radioactivity", List all the automatic actions associated with each of the following Radiation Monitors:

1. 1RIAs 37 WG Disposal Normal /High

2. RI As 33 Liquid Waste Normal /High

3. 1RIA 49 RB Gas i l

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4. 1RIA 45 Unit Vent Gas Normal 3

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9 QUESTION 8.01 (1.00) 1 i

Which one of the following correctly describes the required i actions to retrieve a Red Tag Stub, if the Work Supervisor {

responsible for the job is not on site? )

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a) Another Work Supervisor in that group may authorize retrieval, as long as he has phone approval from the _

responsible Work Supervisor, b) The Group Superintendent is the only individual who may sign authorizing removal, and he must inform the Work Supervisor responsible for the work when he returns to j the site. q

c. The Group Superintendent must approve tag retrieval, but l he may authorize this based on verbal approval and having another individual sign his name and initial authorizing tag removal.

d. The Shift Supervisor is the only individual who can authorize tag retrieval in this situation. I I

J QUESTION 8.02 (1.00) j I'

Assume full power operations and Refer to the attached Tech Spec excerpts for the Reactor Building Cooling Units (RBCUs) to answer the following: In the event that degraded performance (below acceptance limits) on all RBCUs was identified, which one of the following actions would be correct in accordance with Tech Specs?

a. If the RBC system is not restored to meet the i requirements of Speci fication 3.3.5.b (1) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, l the reactor shall be placed in a condition with RCS pressure below 350 psig and RCS temperature below 250 i

deg'. F within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. One at a time, each RBC train can be taken out of ser vi c e for 7 days to undergo maintenance.

c. Reduce reactor power to a point below which the RBCUs can safely meet design analysis conditions,

d. Place the affected unit in at least Hot Shutdown within l the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and in at least Cold Shutdown within l the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, in accordance with section 3.0.

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Refer to the attached Tech Specs for Radioactive Liquid Effluents 3.9 and answer the following. As a result of routine li quid releases, members of the public in unrestricted areas received a calendar year dose of 12 mrem whole body. Which one of the following would be the required action (s) in accordance with Tech Specs?

a. Implement the Oconee Emergency Plan, and submit a report to the NRC.

b. Submit a report to the NRC.

c. Submit a report to the NRC, and implement the provisions of Tech Spec section 3.0.

d. Restore the concentration of radioactive material released in liquid ef fluents to unrestricted areas below established limits, and implement the provlsions of Tech Spec section 3.0.

QUESTION B.04 (1.00)

Which one of the following has an associated Tech Spec Limiting Condition for Operation under 3.10 Radioactive Gaseous Effluents?

a. Waste Gas Holdup Tank oxygen concentration,

b. Auxiliary Building Exhaust System gaseous effluent.

c. Contaminated oil incineration.

d. Gaseous effluent air dose due to particulate.

QUESTION O.05 (1.00)

Indicate whether the following would require the Operations Duty Engineer to contact the Superintendent of Operations immediately or on a delayed basis, as outlined in Oconee )

OMP 1-3, " Operations Duty Engineer". l a) A Tech Spec Violation that is not a result of exceeding a Safety Limit or a Limiting Safety System Setting.

b) Contaminated Injury requiring offsite transportation.

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QUESTION 8.06 (1.50)

As Unit Supervisor on Unit 2, which is shutdown for ongoing refueling operations, a maintenance request to work on RIA-43 (Vent Particulate RM), which will require its deenergization, is given to you for approval. Will you allow this maintenance to take place? Explain the reasoning for your answer.

QUESTION 8.07 (1.00)

Answer the following regarding use of procedures:

a) If a requirement is not met during the conduct-of a procedure, where should this discrepancy be documented?

b) Within how many days should a working copy be verified by comparison with the controlled copy of a procedure?

QUESTION B.08 (1.00) i What type of personnel are evacuated in each of the three categories of Station Evacuation (Category 1, Category 2 and Category 3)?

QUESTION B.09 (1.00)

Refer to the attached Tech Spec excerpts on Operation Safety Instrumentation and answer the following. In the event of a malfunction in two of the Turbine Stop Valve Closure channels that rendered the channels UNTRIPPED and INOPERATIVE, what would be the required action?

QUESTION 8.10 (1.00)

Refer to Figure 3.1.10-1 " Limiting Pressure VS Temperature Curve..." and explain the basis for operation in the acceptable region of this graph.

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92__9901NigIBSIlyE_EBgCEQUBES2 _CgNpillgN@g_@Np_LidlISIlgN@ PAGE 19 QUESTION 8.11 (1.00)

In order to utilize a Human Red Tag in lieu of a Safety Tag, what 4 individuals must agree that the evolution can be cone is a safe manner without compromising the safety tagging l

program?

I QUESTION 8.12 (1.50)

List 5 situations when the NRC Operations Center shall be '

contacted, following initial notification of an emergency.

i QUESTION B.13 (1.00)

For the Reactor Core saf ety limit (Section 2.1) list the three combination of plant parameters which ensure power j peaking is limited for DNBR considerations?

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i QUESTION 8.14 (2.00) i List five conditions which would result in declaring a 8 movable control assembly inoperative in accordance with Tech Specs. Do not include possible causes for the condition in 1 your answer. j l

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,Br__8901NigIggIlyg_EBQQggu@ggg_ggNQlIlgN@g,@NQ_61DlI@llgN! PAGE- 20 QUESTION 8.15 (2.00)

The following conditions exist:

Unit 1 operating at full power Unit 3 at cold shutdown conditions One Keowee Emergency Start Circuit channel has been out of service for maintenance for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

A fuse blows in the tripping coil for the Standby Bus Keowee Feeder Breaker SK1 Explain the Tech Spec consequences of this scenario on the continued operation of Unit 1 AND the potential operation of i Unit 3.

!

• Note: no other Tech Spec related components are out of service, refer to attached Tech Spec excerpts, ir.clude in your answer all associated time restrictions.

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(***** END OF CATEGORY 08 *****)

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

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- OP/1/A/1102/01 ENCLOSURE 4.1 UNIT STARIUP FROM COLD SHUTDOWN T0 RCS TEMPERATURE AND PRESSURE

__ 0F 250'F AND 350 PSIG

~

Verify Date Date  !

Init./ Time Init./ Time 1.0 Initial Conditions 1.1 Procedure Limits and Precautions have been reviewed.

2.0 Procedure I

NOTE: Enclosure 4.1A (Flowchart) should be used as a guide (2.0) by the SRO/RO to aid in maintaining the big picture.

.) 2.1 Complete Enclosure 4.4 (Pre-heatup Checklist).

• If unit startup is following a Refueling Shutdown, complete Euclosure 4.6 (Pre-heatup Checklist Following l 1

a Refueling Shutdown).

• Plot an RCS Boron versus RCS Temperature for 1*.AK/K S/D margin curve per PT/1/A/1103/15 (Reactivity Balance Calculation).

Verify that RCS boron will be adequate for a 1*.aK/K shutdown margin when rod group 1 is withdrawn to 50*.wd at approximately 250*F.

Unit Supervisor l

• Form a steam bubble in the pressurizer per OP/0/A/1103/05 (Pressurizer Operation).

Q Unit Supervisor NOTE: Maintain pressurizer level at - 100 inches until RCS 0: concentration < 7 ppb.

M4 m.=*

3. LIMITING CONDITIONS FOR OPERATION .

3.0 LIMITING CONDITION FOR OPERATION I Speci fics,t 1,og In the e9;it a Limit'ing Condition for Operation (LCO) and/or associ-ated Action requirements cannot be satisfied because of circumstances in excess of those addressed in the specification, the affected unit shall be placed in at least Not Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and in at least Cold Shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless corrective measures are completed that permit operation under the permissible Action statements for the specified time interval as measured from initial discovery or until the reactor is placed in a mcde in which the specification is not applicable. Exceptions to these requirements shall be-stated in the individual specifications.

l Bases This specification delineates the ACTION to be taken for circumstances not directly provided for in the ACTION statements of existing LCOs and whose occurrence would violate the intent of the specification.

For example, Specification 3.3.1 requires that two independent trains of the High Pressure Injection (KPI) System be operable and provides explicit Action requirements if one train of the KPI System is inoperable.

Under the terms of Specification 3.0, if more than one train of the HPI System is inoperable, the affected unit is required to be in at least Hot Shutduwn within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in at least Cold Shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. It is assumed that the unit is brought to the required mode within the required times by promptly initiating and carrying out the appropriate Action statement.

3.0-1 A 89/89/86 12/10/80

i i

b, The 3VSI shall contais a =1r.imua ;evel of 46 fee of -ster

. havi:g a sicisu testentre:::: :i ; 3 3 5'* ppe bo rou' s : a 9:::,u j

e=pera;;r; of 50'T The 'a:;2; val <e _*? 05, on the '.s:narie

. l li e shal'. :' e Io:ked open 'l 'he'e :-qu;teStat8 at: ' ' * *:

)

the 8*.37 she!! be cor.s:Jeref 2 3 4 v i t i e b '. e ani 4:*:on ...*.a.;f -

J iccardsc:e t;h Specif ;s t:an 3 2.

3.3.5 Reactor Butiding Cooliog (RBC) System.

• a . Prior to in ::ating maintenance on any component of the R3C system, the reduodant componest shall be tested to assure opera *ailt:y.

b. Ubec the RCS, with fuel in the core, ts in a conditio: v::h l pressure equal to or greater than 350 pstg or temperature equa' to or greater than. 250'T and subert: teal:

i (1) Two independen: RSC trains, each comprised of an RSC f an, assoc:ated cooling u:::, and assocta:ed IST valves sna;. be operstle.

(2) Tests or maintenance shall be allowed on any compenen: of :ne D3C systes provided ose trato of the R3C and one tra:: of the  !

R35 are opriapie. If ;be RSC system :s o: res!.orec :o :ee:

ne requirements of Spectf tcatten L 3.5.b(:) as:ve .::::: 2:

hours, the reartor shall be placed is a condt :en v::n RCS pressure below 350 pstg and RCS temperature be'o . 252 T vt:5-

n an add:::enal 2t hours.

c. When the reactor is critical:

(!) In additten to the requirere::s of Specifi:attens 3.3.5.bi'.'

above, the remain:cg RSC fan, associated cool::g un::, and assottated EST valves shall be operable.

(2) Tests or matnienanceoshall be allowed on one R5: ::a:n under et;ter of the following conditions:  !

4 (a) One RBC train may be out of serv:ce for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (b) One RSC train may be out of serv:ce f ar ' days prov:ded both RBC trains are operable.*

(c) If the inoperable RBC train is not restored to mee: the requirements of Specification 3.3.5.c(1) wt;nin ::e time permitted by Specification 3.3.5.:(2; ,a' or s:), tre reactor snall be placed := a hot shutdown :rnt: .: .;;t.- 1 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If the requirements of Specification 3 2 5.cs: '

a r- met within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following hot  !

-shusu .a. the reactor shall be placed in a :ond:::on w::n j RCS pressu:e below 350 psig and RCS tempera;ure belaw i 050'T with n an add:ttonal 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. I

• Fer the "3A" RBC train, a one-time extension of inoperability is granted in order to ,ellow for repair, provided both R35 trains are operable and that the "JA" RBC train is returned to servlee no later than#11:59 p.m., April 20, 1985.

~

• • 2010 ppa boron for Ucit 3, ' Cycle to ooty.

Amendment No. 155., 155, 152 3/19/87 3,3-3 .

l l

e 3.3 RADI0 ACTIVE LIQUID ETTI.UINTS l

~  ?

Aeolicabilitv Applies at all , times to the controlled release of all liquid vaste discharged from the site which may contain radioactive materials, except as noted.

Appendix I dose limits for radteactive liquid effluent releases (T.S. 3.9.2) i are applicable only during normal operating conditions wnich include expected l operational oc'eurrences, and are not applicable durtng unusual operating con-ditions that result in activation of the Oconee Emergency Plan.

Ob _ieetive l To establish conditions for the controlled release of radioactive liquid effluents. To implement the requirements of 10 CFR 20.10 CIR 50.3ea.  ;

Appendix A to 10 CTR 50, Appendix I to 10 CTR 50. 0 CTR lti anc -0 :TR 190, i r

Specification '

3.9.1 . Concentration l t

a. The concentration of radioactive material released at anytime  ;

from the site boundary for liquid effluents to Unrestr:cted i Areas (denoted in Figure 2.1-4(a) of the Oconee .Vuelear Station ,

Final Safety Analysis Report) shall be limited to the concen- '

tration specified in 10 CTR Part 20, Appendix 3 Table II, j j Coluem 2 for radionuclides other than dissolved or entrainec l noble gases. For dissolved or entrained noble gases'the con-centration shall be limited to 2 x 10 4 aC:/ml';otal activt:y.

l

b. If the concentration of radioactive material released tn 1:cu:f l

l effluents to Unrestricted Areas exceeds the scove Specifiec l limits, without delay restore the concentra;;an to within :ne j

)

above limits, d 3.9.2 Dose

a. The dose or dose ccamitment to a Memoer ?f The puo;;: fr:m radioactive materials in liquid effluents to '.' ares:::::ed '

Areas shall be Italted to:

l

1) during any calendar quarter:

1 4.5 mres to the total body i 15 mrem to any organ and;

2) during any calendar year:

1 9 mrem :: :ne totsi becy 1 20 mrem :o any organ.

l l

3 c.g A 125/125/122 1/16/84

-- -~

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b. If the calculated dose from the release of radioactive materials !I in liqu:d effluents exceeds any of the above limits, except l 1 during unusual' operating conditions that result in act vation  !

of the Oconee Emergency Plan, and in lieu of any~ other report '

required by Section 6.6.2, a report shall be submitted within 30 days from the end of the quarter during which the release occurred, to the regional .VRC Office which includes the following:

f

1. Cause(s) for exceeding the limit (s) i j

2. A description of the program of corrective action initi-  !

ated to: reduce the releases of radioactive materials i in liquid effluents, and to keep these levels of radio- l active materials in liquid effluents in compisance with the.above limits, or as low as reasonably achievable. '

)

3. Results of radiological analyses of the drinking water I source and the radiological impact on finished de:nn:ng I i water supplies with regard to tne requ:rements of a0 Cy3 l 141. 1 j

2.9.3 Liquid '=~aste Treatment

a. The appropriate subsystems of the liquid radeaste treatment system shall be used to reduce the radioact2ve materials in 1 liquid waste prior to their discharge, if tan projected dose l due to liquid efflucnt releases to unrestricted areas, wnen ,

averaged over 31 days would exceed 0.18 mrem to the total body l or 0.6 mres to any organ.

l

b. If radioactive 11guld waste is discharged w tneut treatment and in excess of the above itmit, a report snail be suomitted wits-l in 30 days to the regional .YRC Office which includes the following:

1. Cause of equ:pment or subsystem nepersb:1:ty.

2. Corrective action to vestore equipment ahd prevent re-currence.

l 3.9 a Chemical Treatment Ponds (CTP 1 and 2) d

a. The quantity of radioactive material in tte Chemi::1 Treatment Ponds (CTP) shall be limited so that, for all rad:enuclides identified, excluding noble gases and t:It:um, the sum of the ratios of activity (in curies) to the 1::::s :n *.?

i Appendix B. Table :I, Column 2 sna11 ::: ex:ee: ; * :TR < ' :!?*.

l

< ..  :( *05 s .>

3,g.2 A 125/125/122 1/16/84

where Aj = pond invento:7 limit for single radionuclides *j'

. (curies)

Cj s 10 CTR 20, Appendix B, Table II, Column 2, concentration for single radionuclides 'j' (curies)

b. Af ter a primary to secondary leak is detected, the initial batch of used Powdez resin shall not be transferred to the CTP.

.Vo batch of used powdex resin shall be transferred to the CTP

,unless the sum of the ratios of the activity of the radionuclides .

identified in the preceeding batch froc any powdez cell in the same unit is less than 0.1* of the limit identified in 3.9.4.a.

IJ S1 Aj

< 1.0 x 102 where Qj = radionuclides activity in the batch j Aj = pond inventory 1Lmit for radionuclides 'j ' .

t

. The radionuclides inventory per batch of used powdex resin transferred, averaged over the transfers of the previous 13 .

weeks, shall not exceed 0.0l* of the pond radionuclides inven- I tory limit. If this average exceeds 0.01* of the pond radionu- '

clide inventory liatt, then a report will be submitted within 30 days to the Regional SRC Office describing the reason or reasons for exceeding the objective and plans for future .

j operation. Decay of radionuclides may be taken into account in derarmining inventory levels.  !

Qj g

• Qj2 * * '

• Sd ( n- 1 )

• Qj n 1 .01% x Aj  ;

a wnere QJ = activity or radi:nuclide 'j' :n the batch n = number of batenes transferred to the chemical i treatment ponds during tne previous 13-week period. I i

3.9.5 izqu d Holdup Tanks

a. The quantity of radioactive material contained in each aut- {

side temporary tank shall be limited to less than or equal  ;

to 10 curies, excluding tritium and dissolved or entratned

• noble gases. Tanks included in :nis specification are : hose outdoor tanks that are not surrounded by 1:ners, dixes, or .

walls capable of holding the tank contents and :nat do not '

have tank overflows and surrounding area drains cor.ne::ed .

to the liquid radvaste treatment systes. '

. The quantity of rad::act:ve mar.er al : r.ta:ned ;n eac: :f he outs de temporary tangs snal'. ce determined :: :'e .::::

ne amove 1: mat Oy ana'y::ng 4 representative samp:.e of :ne tanns contents at . cast :nce per 1ays nen raft:a:::ve t.a:ertals are being ac:ec to :n : ant. l 6

l l

A 125/125/1" 3.;-3 '

1/16/84

1

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c. If the quantity of radioactive material in any outside tes-

. porary tank exceeds the above limit, suspend all addi: ions i

to radioactive mate, rial to the tank without delay.

3.9.6 The provisions of Technical Specification 3.0 do not apply. 3 -

Bases l

The concentration specification is provided to ensure that the concentration of radioactive materials released in liquid wasta effluents from the si:e to unrestricted areas will be less than the concentration levels specified in 10 CTR Part 20, Appendiz 8. Table II. The concentration limit for noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and !:s MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protec-tion (ICRP) Publication 2.

The dose specification is provided to assure that the release of raduac::.ve material in liquid effluents will be kept "as low as is reasonably ach e.ab'e. .

• Also, for fresh water sites with drinking water supplies which can be potent:. ally affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclides concentrattens :n the itnisned drinxing water that are in excess of the requirements of 40 CTR 141. The dose

calculations in the ODCM implement the requirements in Section !!I.A of Appendix :.

. hat conformance with the guides of Appendix I is to be shown by calcula::.cnal procedures based on models and data such that the actual exposure of an indivt-dual through appropriate pathways is unlikely to be substantially underest: mated.

Section IV of Appendix I of to CTR 50 states that the licensee is permitted the flexibility of operation during unusual operating cond:: ions, :o assure the public is provided with a dependable source of power when corepatible with

ensiderations of health and safety of the public. Section I of Appendix ! j of 10 CTR 50 states that this appendix provides specific numerical gus' des j for cesign objectives and limiting conditions for operat:.on, :o assis noi:ers i of licenses for light-water-cooled nuclear power reactors in meet:ng :ne re-quirements to keep releases of radioactive material to unrestr:.cted areas as low as practical, and reasonably achievable, during normal reactor oeerat:.ecs.

neluding exoected operational occurrences. Using tse flexio:11:7 grantec 1 during unusual operating conditions, and the stated applicab:.' :y of :ne :e-s:gn objectives for the Oconee .Vuelear Statten. Appencix I dose 1:n::s for i radioactive liquid effluent releases (T.S. 3.9.2), are concluded :o be ne applicable during unusv.al operating conditions tnat result in :ne ac : vat on of the Oconee I.mergency Plan.  ;

Tor units with shared radwaste treatment systems' i.he liqu:d effluen.s fr:m :ne snared systekn are proportioned a: song the units snaring that system.

The requirements that the appropriate pertions of this system be used -nen spect-f ed provides assurance that the releases of radioactive materials in 1:qu:d ei-

luent.s w:ll be kept "as low as is reasonably aca:evrole." This spe::f:catan

.mplements the requirements :f '.0 OTR Par: 50.264 Seneral es;;n .::er::n i:

f Appenc:x A to ;0 :TR ?ar: 50 and des:gn :n;e::.ve 5e::a n ::.: :: -:-an:a -

a '. 0 TR Part 50.

3.9-4 A 125/125/122 1/16/84

I 4

The inventory liatts of the chemical treatment ponds are based on. limiting the I consequences of an uncontrolled release of the pond inventory. The short term cate limit (2 ares /hr) of 10CTR20.105 is applied to 10CTR2'0.106 in the following expression: ,

Ai z 108 WCi x sal <* 2 mres/hr x 3760 hr

~

1.3 x 10' sal curie- 3 765 mi $00 area /yr yr j

C)

Ai i 1.7'x 103 Cj q ll where AJ = pond inventor.y limit for radionuclides 'j ' (ruries)

CJ = 10CFR20 Appendix 3 Table II, Column : cancentrat:on for radionuclides 'j' .

i 1.3 x 10' gal = estimated volume of smaller :nemical :reatment pend j The bat.ch limits provide assurance that act:vity input to the CTP v 11 be mis-Imazed.

I

! I l

l l

l 3.9-5 A 125/1:5/122 1/16/84

l 1

.- i 3.5 INSTRUMENTATION SYSTEMS, 3.5.1 ,

operation Safety Instrumentation Applicability Applies to unit instrumentation and control systems.

Objective To delineate the conditions of the unit instrumentation and safety circuits necessary to assure reactor safety.

Specifications l

1 3.5.1.1 The reactor shall not be in a startup mode or in a critical state unless the requirements of Table 3.5.1-1, Column C are met.

3.5.1.2 In the event that the number of protective channels operable falls below the limit given under Table 3.5.1-1, Column C; operation shall  !

be limited as specified in Column D.  !

3.5.1.3 For on-line testing or in the event of a protective instrument or channel failure, a key-operated channel bypass switch associated with each reactor protective channel may be used to lock the channel trip relay in the untripped state. Status of the untripped state shall be indicated by a light. Only one channel bypass key shall be accessible for use in the control room. Only one channel shall be locked in this untripped state or contain a dummy bistable at any one time.

3.5.1.4 For on-line testing or maintenance during reactor power operatton, a key-operated shutdown bypass switch associated with each reactor protective channel may be used in conjunction with a key-operated l

channel bypass switch as limited by 3.5.1.3. Status of the shutdown bypass switch shall be indicated by a light.

3.5.1.5 During startup when the intermediate range instruments come on i scale, the overlap between the intermediate range and the source q

! range instrumentation shall not be less than one decade. If the 1 overlap is less than one decade, the flux level shall not be greater i than that readable on the source range instruments until the one  ;

decade overlap is achieved, l

I I

i I

3.5-1 ,

A 148,148,'.45 8/20/86 I

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TABLE 3.5.1-1 INSTRUME.VTS OPERATING CONDITIONS (coct'd)

NOTES:

(a) For channel testing, calibration, or maintenance, the sinimum of three operable channels may be maintained by placing one channel in bypass and one channel in the tripped condition, leaving an ef fective one out of two trip logic for a maximum'of four hours.

(b) When 2 of 4 power range instrument channels are greater than 10% rated power, hot shutdown is not required.

(c) When 1 of 2 intermediate range instrument channels is greater than 10 ~10 amps, hot shutdown is not required.

(d) (Deleted) *

(e) If minimum conditions are not set within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> af ter hot shutdown, the unit shall be in the cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(f) 1. Place the inoperable Reactor Trip Module output in the tripped condition within one hour or

2. *emove the power supplied to the control rod trip devices associated with the inoperable Reactor Trip Module within one hour.

(g) ( Deleted)

(h) ' The RCP monitors provide input to this logic. For operability to be met either all RCP monitor channels must be operable or 3 operable with the remaining channel in the tripped state. 4

\

(t) 1. The power supplied to the control rod drive mechanisms through the failed CRD Trip Breaker shall be removed within one hour or

2. With one of the CRD Trip Breaker diverse features (undervoltage or, shunt trip device) inoperable, restore it to OPERABLE status in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or place the breaker in trip in the next hour.

(j) 1. With one SCR Control Relay inoperable in logic channel C or D, restore the inoperable SCR Control Relay to OPERABLE status in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or remove power from the CRD mechanisms supplied by the inoperable channel's SCR Control Relay within the next hour.

2. With two or more SCR Control Relays inoperable in logic channel C or D, remove po er fr:m the CR0 mechanisms supplied by the tnoperable enannel's SCR Control Relay within one hour.

3. 5 - 5 c A 148,148,145 8/20/86

• E08

. ,e 1000 1600 Im1400 J  ;

a E

1200

_C A

5 C.

a 1000 e

3 u

800 C

Y i

=

= -

E00 3

8 4 00 r

200  : 1 0

100 200 300 400 500 600 700 l Indicatea Reactor Coolant System Temperature,*F LINiilNG PRES $URE VS TEMPER ATURE CURVE FOR 100 STD CC/LliER H 2O

/ OCONEE NUCLEAR STATION 4

. Figure 3.1.10 1 3 1-22 7/19/74 L

.' j 3.7.3 In the event tnat ene conditions of Specificatzens 3.7.' are not me within the time specified in Specification 3.7.2. except as noted below in Specification 3.7.4 3.".5, 3.7.6, 3.7.7. and 3.7.3. :te reactor shall be placed in a het shutdown condition w: thin 12 sours. If t sse requirements are not met within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be placed in the cold snutdown condition within t hours.

3.7.4 In the event that all conditions in Specification 3.7.1 are met ex-cept that one of the two Keowee hydro units is expected to be un-l available for longer than the test or maintenance per:od of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the reactor may be heated above 200'T if prev:ously sautdown or be permitted to remain critical or be restarted provided :ne following restrictions are observed. l (a) Prior to heating the reactor above 200*F or prior to the re-start of a shutdown reactor or within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the loss of one Keowee hydro unit, the 4160 volt standby buses shall be .

energized by a Lee gas turbine enrougn the 100 kV eircuit. I The Lee gas turbine and 100 kV transmission c;rcutt snall be j el'ectrically separate from the system grad and offsite non-f safety-related loads.

(b) The remaining Keowee hydro unit shall be connnected to the under- I ground feeder circuit and this path shall be verified operable j within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and weekly thereaf ter. '

1 (c) The remaining Keowee hydro unit sna11 be available to :ne j overhead transmission circuit but generation of :ne system  ;

grid shall* be prohibited except for periods of test. 1 (d) Operation in this mode is tstricted to periods not to exceed

. 45 days and the provision, of this spec:fication may be utiltzed without prior NRC approval only once in three years for each  !

Keowee hydro unit. The U.S. NRC Regional Office, Region II, vill be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.7.5 In the event that all conditions of Specification 2.7.1 are met ex-cept that all 230 kV transmission lines are lost, the reactor shall be permitted to remain critical or be restarted provided the following i restrictions are observed:

(a) Prior to the restart of a shutdown reactor or within I hour of losing all 230 kV transmission 1:nes for an operating reac-tor, the 4160 volt standby buses sna11 be energi:ed by one of the Lee gas turbines through the 100 kV transmission circuit.

The Lee gas turbine and the 100kV transmission circuit shall be c~ompletely separate from the system grid and offs :e non-safety-related loads.

(b) The reactor coolant T shall be above 525'F. Reactor coolant pump power, may be use$"Io elevate the temperature from 500*F to 525' in the care cf restarn. If T decreases below 500*F. restart is not permitted by this specE!ication.

A 127/127/124 3.7-5 3/2/84 1 . .

l L______-_--__ - - - - _ _ _ _ _ _ -- --

,f . ,7, 7 y -e ,: , -

x-T 4

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i)

M3LE3.7-1 i) 0; ,,

OPERABILIS REQUIREMENTS !CR T/E . ,

EEttGENCY 1973 SWITCHING Lod?LlCIRC2IM  !

[

. u t

s. .

i:

. i Minimum 49 0 erable Circuits / Channels

/'4 7#

functional Unit ,

9' .'

k $Dr*ms/ f Operation. #' Degraded Operation l 3 Thx 4 pec 3. 71.1(c) - Per Spec 3.7.2(b) H

't > ,

1. Normal Source. Voltage c 3 4 2 Sensing Circuits (Ong per Phase) \*

7,, ]

2. Startup Sour::e Voltage 3 ,}3 s 2 i Sensdng .Circu..'ts (One per Phase) ( } (.

e , a i,' ,

DJStandsy us W ltage 6

g

• k*

Seasing Circuits (On'.( per Phase

[g -)

on each bus) ,

, j :/ .(

s

4. 34fi F4 fer 6 4a'

- Relafs (Th;gb'us Undervoltage ee per bus) ' *

. 4 Fg Load Shed and Transfer to St:ndby

5. 2 1 l Circuits (Channels A and B) s

6. Keowee Emergency Start Circuit 2 /

i cichnnels A and B) s

i. . ,

t t

)) ~

b

7. Normal Source Breakers N1' a o

J ;c and N2 Control Circuitry

, , ' % ( ,

! 3. Startup Source Breaker 3 UI ;i \ ,,

) and E2 Control Circuitry < (4I 2*

%' b c

9. Standby $us to Main Feeder Bus , 4 2 Brn h6rs, S1 and S2, Control "

Cidcuitry (Including Retransfer I

,4 to Startup Circuits) i 4 .<

I

> 10.1 Standby Bus Keowee Feeder Breakers,

, 4 b <

c a

( o'3' SX1 and SK2, Control Circuitry N

• y,

(

4 1  !

Notes: 2 per bus. ,.

t eb.a J / 1 primar7 a'tWfg s;econdary* for each breaker. j

c. 1 primaryind 1 econdary* on the same breaker. '

\ a ~ *A primary circuit inclu es the closing coil and one trip coil. a secondary circuit includes only one trip coil.

A 127/127/124 3.7-14 3/2/84  !

e

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

hi_IHgg81_g{_Nyg6[$6_ggWE8_E(@NI_ggEB911gh_E6piph_@Ng _PAGE 21 IHEBdQQyN@ digs ANSWERS -- OCONEE 1, 2&3' -87/07/13-CASTO, C.

k 02 L '

ANSWER 5.01 (1.00) ggt

.d REFERENCE OP-OC-SPS-PTR-AT pp 13/14; LO la (4.1/4.7)

OOOO74A207 ...(KA'S)

ANSWER 5.02 (1.00) b C A C-REFERENCE OP-DC-TA-NT Figure 15; LO lo [sp-M- APJ- JC-JCI ~ Pv8N (3.6/3.8) 016000G015 ...(KA'S)

(

ANSWER 5.03 (1.00) d.

REFERENCE Oconee OP-OC-SPS-RT-IP pp._28 obj 2a 2.8/3.1 192OO5K105 ...(KA'S)

ANSWER 5.04 (1.00) c, REFERENCE Oconee OP-OC-SPS-THF-PD pp. 10 obj. 2d 2.9/3.3 193OO9K107 ...(KA'S)

ANSWER 5.05 (1.00)

b. Of h u8

J.

j 5.* THEORY OF NUCLEAR POWER PLANT OPERATION g _% UIDSt_AND PAGE 22 IHE6dODyN@dlCS ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C.

REFERENCE Oconee Duke Thermo pp. 194 3.9/4.2 op-oC . FPS- Fryt.- Mtl rpJ1 0 193OOOK121 ...(KA'S)

ANSWER 5.06 (1.00) a) Increase (+.5) J b) Decrease REFERENCE OP-OC-TA-NT pp 16; LO ik ANSWER 5.07 (1.50) l l

a) Decrease (+.5 ea) b) Decrease c) Increase i

l REFERENCE OP-OC-TA-NT pp 7/83 LO 16 (3.6/3.9) 041020A202 ...(KA'S) i ANSWER 5.OB (2.00)

a. about equal {UyM(pfykeW 3 bt SIkN

b. will be equal to Tsat in the [caf3

c. the isolated generator (l oop ) will l'ag the operating. loop

d. should not be greater than 50 deg F.

REFERENCE Oconee OP-OC-SPS-PTR-AM-1 obj 9 4.2/4.3 OOOO11A209 ...(KA'S) i ANSWER 5.09 (1.00)

Negative (+.5) Due to colder water nearer bottom of the core (Higher allowable Kw/ft) (+.5)

REFERENCE OP-OC-SPS-IC-RPS pp 12, LO lb (3.1/3.3)

i

5.

• THEORY OF NQGLEAR_PQWER_PL@NT_QPERATigNt _FLylpSz_@NQ. PAGE 23 h IdEBdQQyN@dlCS ANSWERS -- OCONEE 1, 2h3 -87/07/13-CASTO, C.

i 012OOOK502 -...(KA'S)

ANSWER 5.10 (1.00)

Quench Tank Pressure (+.5) stops increasing (+.5)

REFERENCE l OP-OC-SPS-CM-PZR pp 17; LO 11 ,

(2.6/2.8) )

OO7000A206 ...(KA'S)  ;

ANSWER 5.11 (2.00) i a) Increase due to the decrease DP across the FDW control valve (+.5) Levels off due to reaching its high speed stop (+.5) {

b) Tavg has decreased, so power is increased-(+.5) {

c) Loop B TC has increased, Loop A Tc has decreased (+.5)

REFERENCE '

OP-OC-TA-NT pp 13/14; LO 1h (3.1/3.4) 059000A212 ...(KA'S)

ANSWER 5.12 (1.50)

1. High injection water boron concentration

2. boiling off of RCS coolant in the core due to decay heat

3. Low thru core flow rate REFERENCE Oconee OP-OC-SPS-PTR-AM-1 pp. 37 obj 19 ANSWER 5.13 (1.00)

The intent of their use is to restore natural circulation.

In doing this, gases in the RCS should be mixed with RCS liquid to eliminate gas pockets. [0.53 Bumping the RCPs condenses the steam bubble by cooling it (and starting reflux boiling). CO.53

5i-_IME98Y_9E NyC6g88_EgyEB_E69NI_ggEB9IlgN g_{691pgz_@Np PAGE- 24 THERMODYNAMICS I

ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. j i

REFERENCE Oconee Op-DC-SPS-PTR-AM-2 obj 1.f. 4.0/4.4 OOO74K311 ...(KA'S)

ANSWER 5.14 (1.00) i at 2240 psia, hg = 1115 BTU /lb (+.5) at 20 psia, at saturation conditions, hg = 1156 BTU /lb and hf = 196 BTU /lb- j calculate: ' (1156-1115)/(1156-196) = .043 >>> 95.7% quality j If use Mollier: 9 5% quality (+/- 1%)

l REFERENCE OP-GA-SPS-THF-STM pp 20/21; LO 2e (3.3/3.4) 193OO3K125 ...(KA*S) l 1

ANSWER 5.15 (1.00) $

i C

REFERENCE DPC Thermodynamics, pp 192-5; (3.1/3.4)

OO2OOOK501 ...(KA'S)

-)

i I

1 l

j i

i j

i

l '

6. PLANT SYSTEMS DESIGN g_ CONTROLt_AND. INSTRUMENTATION. PAGEL 25 .

ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C.

i 1

ANSWER 6.01 (1.00)

J b

REFERENCE OP-DC-SPS-IC-ICS pp 23/24; LO 1h (3.2/3.2) 1 i

059000K107 ...(KA'S)

-)

ANSWER 6.02 (1.00)

C REFERENCE OP-OC-SPS-SY-EF pp 21/22, 32/33; LO'le, if (3.4/3.5) 061000A207 ...(KA'S)

ANSWER 6.03 (1.00) a REFERENCE OP-DC-SPS-SY-CC pp 19; LO.2c (3.2/3.2)

OO9010A301 ...(KA'S)

ANSWER 6.04 (1.00)

I

b. )

REFERENCE-Oconee Op-OC-SPS-IC-ES obj.h 3.6/4.2 3.7/4.2' 013OOOA204 013OOOA205 ...(KA'S) l 1

J ANSWER .05 (1.00) i

{ } 'N q

u l

')

J

h!._EbeNI_SI@IEDS_QE@l h _QQNISQ6_@NQ_lNSIBUdENI@IlgN PAGE 26 ANSWERS -- OCCNEE 1, 2a3 -87/07/13-CASTO, C.

REFERENCE

.Oconee OP-DC-SPS-SSF-EPS 2.6/2.7 062OOOA404 ...(KA'S)

ANSWER 6.06 (1.00) d.

REFERENCE Oconee OP-OC-SPS-SY-EF objs 1.b/1.a/1.o '3.4/3.8 061000A204 ...(KA'S)

ANSWER 6.07, (1.00I C

REFERENCE Oconee OP-DC-SPS-IC-CRI 1.m. 3.5/3.8 OO1000K401 ...(KA'S)

ANSWER 6.08 (1.00) a.OF C REFERENCE Oconee Op-OC-IC-RCI obj. 3.g. (3.1/3.6)

OO2OOOK603 ...(KA'S)

ANSWER 6.09 (1.00) a) A Submersible Pump is installed to discharge into the nearest CCW Piping manway. (+. 5 ea) 6*lu64f crfter- #40afd awau# WMM4> taw /4f b) An Air Ejector is valved into the S F ASW Suction piping.

when the SSF is actuated. [gg gi(gy-@ g[-4.!P t6koved REFERENCE OP-OC-SPS-SSF-ASW Tracking 87-034/03Og LO 2b (3.9/4.2) 061000K401 ...(KA'S)

6:__P(@NI_@Y@ led @_QEgl@N t _CgNI6g(z_@NQ_lNgIBUDENI@IlgN 'PAGE 27' ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C.

ANSWER 6.10 (1.00)

Possible overpressurization of Aux Steam and subsequently, steam line to TDEFWP if both control valves were to fail open due to too low of relief capacity in both systems.

REFERENCE f Oconee LER 87-003 3 1

(3.3/3.4) 03OOOOK107 ...(KA'S) d ANSWER 6.11 (1.00)

A letdown line from the Common letdown line (prior to the L/D coolers) thru HP-(426) to the fuel transfer tube.CO.53 .

A line from the SSF Makeup Pump which takes suction on the i altrenate fuel transfer tube, injects water through SSF-HP-(398). This line serves as system makeup. [0.53 ,

l REFERENCE Oconee OP-DC-SPS-SY-HPI obj 3.a/b 3.3./3.2 OO4000K405 ...(KA'S)

ANSWER 6.12 (1.00)

1. To insure suction to the HPI pumps is not lost CO.53

2. Insure gas does not enter the HPI Pump suction on an ESFAS [0.53 REFERENCE Oconee OP-DC-SPS-SY-HPI pp. 19 obj 5.b 2.8/3.0 3.1/3.4 OO40010KO1 OO4020A104 ...(KA'S) i l

ANSWER 6.13 (1.00) 1

1) the SSF incoming feeder breaker from Unit 2 Main Feeder Bus #2 (OTS1-1) is open (+.5 ea)

2) SSF Diesel Generator Breaker DTS1-4 is closed REFERENCE OP-OC-SPS-SSF-ASW pp 24; LO 2h (4.0/4.2)

__ ________.i____

~

6:l__PLgNI_@ygIgdg_pg@l@yt_CQUIBQLt _@yp_ly@IBUDgyl@Ilgy PAGE 28 ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C.

061000K406 ...(KA'S)

ANSWER 6.14 (1.00)

1) Excessive.Boration of Plant (+.25 ea)

2) Plant over pressurization

3) Chemical Spray Hazard to RB Components

4) Loss of Cooling' Water to necessary components [6snu4 oF T4/*M conpnes eack coad REFERENCE u one cc3 pow)

OP-DC-SPS-IC-ES pp 15; LO ij (3.7/4.0) 013OOOA206 ...(KA'S)

ANSWER 6.15 (1.50) a) The Megawatt error is blocked in track, so the header pressure error is controlling position of the control val ves. (+.5) This causes instability in header pressure, hence feed flow oscillations.

-b-M14-1 GG-stati a n . i o tro-automati e mode-before-the-turbine i s -phrewd in aut W V REFERENCE OP-OC-SPS-IC-ICS Tracking #86-064; LO lb (3.2/3.2) 059000K107 ...(KA'S)

ANSWER 6.16 (1.00) i

\

a) All three units (+.5 ea) )

b) Units 2 and 3 REFERENCE OP-OC-SPS-SY-HPI, pp 36/37; LO 3q (2.8/3.1)

OO3OOOK404 ...(KA'S) l 1

- - - - - _ - .-_---__m-___ _ _

6:__P!;$NI_@? IEdg_QEgl@N g_CQN18 h _@NQ_1NgIBUDENI@llgN PAGE 29 i

I ANSWERG -- OLONEE 1, 2&3 -87/07/13-CASTO, C.

\

l l

ANSWER 6.17 (1.50) )

1) Use HP-26 Motor operated valve (+.5 eahr W 3

2) Use HP-122 (Manual Bypass)

3) Use "C" HPI pump via HP-27 N P V 'gp yto )

j ~

. l REFERENCE g-({tu) bli Yl' Cd"^ ]

OP-OC-SPS-SY-HPI p 273 40 3c b N"#

(3.4/3.b) OP f/A t / toy /G g) op; gyp-Wl7f 078000K302 ...(KA'S)

J l

1 i

i

1 21L_889EE998EE_!_U9806ht 6EU98dOb1_gdgB@gygy_ANQ PAGE 30 R_AD_IO_L_O_G__IC_A_L__C_O_N__TR_O_L_

' ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. I 1

ANSWER 7.01 (1.00) d REFERENCE DPC EPG pp 2-77/79; (4.0/4.4) _

.j OOOO74K311 ...(KA'S) ,

-l ANSWER 7.02 (1.00) b-REFERENCE OP/2/A/1102/01, Enclosure 4.3, pp 4 (3.6/4.2) .1 OO1010A207 ...(KA'S)

ANSWER 7.03 (1.00)

A.

4 REFERENCE I 3.5/3.5 3.2/3.4 I Oconee AP/1/A/1700/23 OOOO57A105 OOOO57A106 ...(KA'S) i ANSWER 7.04 (1.00) d l l

REFERENCE Oconee SY-LPI-5 3.2/3.5 OO5000K402 ...(KA'S) 2

Zi__E8QQEDWBEE_:_NQsd@Lg_@gNQgdeLi_geggggNQY_9ND PAGEL 31

.86 Dig 6QGIC@L_CQNI6QL ANSWERS --'OCONEE 1, 2&3 -87/07/13-CASTO, C.

l l

ANSWER 7.05 (1.00)

1) Close the Field Breaker (+.33 ea)

2) Close the Generator. Supply Breaker

3) Close the Field Flashing Breaker i

REFERENCE OP-OC-SPS-CM-KHG pp 20p'LO'1k (4.0/4.3) 064000A401 ...(KA'S)

ANSWER 7.06 (1.00) a) Refueling SRO or Shift Supervisor'(+.5ea) b) In a cabinet below the operator's. console on the bridge REFERENCE l OP-OC-SPS-FH-FHB pp 23; LO la (3.0/3.0) (2.3/3.9)  !

034000 GOO 1 034000 GOO 9 ...(KA'S) i i

ANSWER 7.07 (1.00) ci- v uf f dT91N 3, 2, 1, 4 ( .25 for each switch to get correct order)

REFERENCE EP/1/A/1 BOO /1, pp 9/10; Reference Document pp 1-8 (3.8/3.9) ,

OOOOO7G012 ...(KA'S)  ;

1 i

1 ANSWER 7.08 (1.00)  !

300 inches ensures that a large enough bubble exists in the pressurizer for pressure control.

REFERENCE l Oconee CP-604; EPG Reference Document pp 3-96/97 (4.2/4.5)

OOOOO9K321 ...(KA'S) l l

- _.u. ._. -- .-_~ _ __ .

2ti_E89EEEMOEE_2_U98U6L _@gNQB[@L1 t_[d[@GENgy,@NQ PAGE 32 RA_D_IO_L_O_G_I_CA_L__C_O_N__TR_OL_

ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C.

ANSWER 7.09 (3.00)

This identifies steps that may be done in parallel with the critical path (numbered) step.

REFERENCE Oconee OP/1/A/1102/01, pp 3 (4.1/3.9) 194001A102 ...(KA'G)

ANSWER 7.10 (1.00)

Letdown should be established to offset the increase in RCS inventory due to the initiation of E-boration.

REFERENCE Oconee EP/1A/1800/01 4.4/4.7 OOOO29K312 ...(KA*S)

ANSWER 7.11 (1.50)

1) SU Cntel Valve (FDW-35) (+.25 ea)

2) EFDW Cntrl Valve (FDW-315)

3) TBV Block Valve (MS-17)

4) MS to SSRH (MS-79)

5) MFW Block Valve -( FDW-31 )

6) SU Block Valve (FDW-33)

REFERENCE Oconee OMP 2-1, pp i enci 4.4 (4.1/4.2)

OOOO40G010 ...(KA'S)

I l

1

o .

It '_E8QCEQUBE5_;_NQBd@6t_@BNggd@(t_EDE8@ENCY_@NQ PAGE 33 8091969 Gig 86_CQNIBQ6 ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C.

ANSWER 7.12 (2.00)

1) Set batch size on makeup control to 90,000 gallons and l reset (+.25 ea) l 2) Open Makeup isolation (HP-16)

3) Open RC Bleed Transfer Pump "A" Discharge CS-46

4) Start "A" Bleed Transfer Pump

5) Dispatch operator to Units 1/2 Waste Disposal Panel

6) Go to Aux Shutdown Panel with: Reactor Log (+.5 total)

Emergency and Abnormal Procedures Removal / Restoration Book Emergency Plan.

7) Maintain Hot Shutdown Conditions REFERENCE Oconee AP-8, pp 2/3; OMP 2-1, ENCL 4.4 (4.1/4.2) 1 OOOO69G010 ...(KA'S)  !

I ANSWER 7.13 (1.50)

1) Use of FWPs (+.25 ea[*(69$

2) Use of EFWPs from the affected or other units

3) HPI Cooling through PORV

4) HPI Cooling through Head Vents

5) SSF-ASW to S/Gs

6) CBPs to S/Gs 1

7) Dump steam from S/Gs

8) ASW % b @

REFERENCE Oconee EP/1/A/1800/1 (4.0/4.4)

OOOO74K311 ...(KA'S)

ANSWER 7.14 (1.00)

1) Reactor startup within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of trip (+.5 ea)

2) Cooldown has not been initiated.

REFERENCE Oconee OP/1/A/1102/02, pp 1 (3.3/3.5)

OO1050G010 ...(KA'S)

' '7 . ' PROCEDURES - NORMAL _ABNQRMAL t _tEMERGENCY _ANQ PAGE 34 BBR1969GIC06_cgNI696 ANSWERS -- OCONEE 1, 2&3 -B7/07/13-CASTO, C.

i l

1 ANSWER 7.15 (2.00)

1. will isolate the Waste Gas Tanks I #it u nsif u s ft V4v1 Mit

2. Teminate a liquid waste release dag pesP (MT fMW/ q

3. Isolate the RB normal sump ,/sout4fv4 N frou 4 42#

4. Stop the RB purge fan, mini-fan, and isolate RB purge system CO.5 ea3 REFERENCE-l-

l Oconee AP-10 pp. 1 3.3/3.4 OP-St-SPJ-IC-FM PPITi M 070000K114 ...(KA*S) 1 l

)

l 1

'Bc'_6901NISIB@IIME_BBQQEQQBgSg_CQNQlligN@g_@NQ,(IBlI@IlgNS PAGE 35

' ANSWERS --.OCONEE 1, 2&3 -87/07/13-CASTO, C.-

ANSWER B.01 (1.00)

C' 0 4 4-REFERENCE Oconee Station Directive 3.1.1, pp 11 (3.7/4.1) 194001K102 ...(KA'S)

ANSWER 8.02 (1.00) d.

REFERENCE Oconee Tech Spec 3.3.5 3.0/3.7 022OO5KOO5 ...(KA'S)

ANSWER 8.03 (1.00) b.

REFERENCE Oconee Tech Specs 3.9 3.1/3.6 073OO5KOO5 ...(KA*S)

ANSWER 8.04 (1.00)

C.

REFERENCE Oconee' Tech Specs 3.10 2.8/3.4 194001K103 ...(KA'S)

ANSWER O.05 (1.00) a) Delayed (+.5 ea) b) Immediate REFERENCE Oconee OMP 1-3, Enci 5.1/5.2

((2.5/3.4)

[,

D !

' 92'__0201Nig16@IlyE_P6gCEQQ6EQg_CQNQlligN@t,@NQ_Ll611@IlgN@ PAGE 36 ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. )

i 194001A103 ...(KA'S)

ANSWER 8.06 (1.50) l NO (+.5) This will result in deenergizing RIA-45, which is required to be in service to meet TS (3.8.10) requirements ,

for the Reactor Building Purge System (+1.0) l e < L/[.$(f. T) (2 ik-V3 thr/+00WfAdDuf os,KV, NfA /f4 ff ti8 s / ,

j REFERENCE Oconee LER 86-005; TS 3.8.10 (3.1/3.6) 073OOOGOO5 ...(KA'S) l l

1 ANSWER 8.07 (1.00) a) Procedure Discrepancies Process Record g(+.5 ea)  ;

b) 14 days g /g g g REFERENCE ~ E *H M #1" 5** & suctrewa ga Oconee Station Directive 2.2.1, pp 4/5 (2.5/3.4) OMA /~cf 194001A103 ...(KA'S)

ANSWER 8.08 (1.00)

Category 1: General Publ i c /non-occupati onal (+.33 ea)

Category 2: Non-essential to operations but Rad workers Category 3: Personnel identified in the Emergency Response Organization REFERENCE Oconee RP/0/B/1000/10, pp 3 (3.1/4.4) 194001A116 ...(KA'S)

ANSWER 8.09 (1.00)

Column D Bring the unit to hot shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

REFERENCE' Oconee Tech Specs Table 3.5.1-1 3.4/4.3 012OO5KOO5 ...(KA'S)

8 __89d1NigI6611ME_P6QCEQWBESt_CQNQlllQN@t_@NQ,(1611@IlQN@ PAGE 37 ANSWERS -- DCONEE 1, 2&3 -87/07/13-CASTO, C.

i ANSWER 8.10 (1.00)

By maintaing the reactor coolant temperature and pressure in the acceptable region, any dissolved gases in the RCS are maintained in solution.

REFERENCE Oconee-Tech Specs 3.1.10 2.9/3.8 OO1006KOO6 ...(KA'S)

ANSWER 8.11 (1.00)

1) Operational Responsible Supervisor (+.25 ea)

2) Work Supervisor

3) Human Red Tag

4) Individual doing the work REFERENCE l

Oconnee Station Directive 3.1.1, pp 14 l (3.7/4.1) 194001K102 ...(KA'S)

ANSWER 8.12 (1.30) i

1) Change of Classificatition (+.3 ea for any 5) l

2) Termination of Event

3) Further degradation of level of plant safety

4) Effectiveness of protective measures taken

5) Results of evaluations / assessments

6) Plant behaviour that is not understood REFERENCE Oconee OMP 1-10, pp 2 (3.1/4.4) 194001A116 ...(KA'S)

ANSWER 8.13 (1.00)

1. Thermal Power Level

2. Number of RCPs operating

3. Reactor Power Imbalance i . _ _ _ _ _ _ _ _

r .-

'@i__e951Ni@I@@llyE_E69CEQUBE@t_CQNQlligN@g_@NQ_(1glI@IlgN@ PAGE 38 1

6 ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. j l

REFERENCE l Oconee Tech Specs 2.1 2.6/3.8 OO2OO6KOO6 ...(KA'S)

I ANSWER 8.14 (2.00)

Cannot be moved Cannot be located Control rod is misaligned with its group average by more than 9 inches Control rod does not meet the exercise requirements l

Control rod does not meet the rod trip insertion times Control rod does not meet the rod program verification i any 5 of 6 ,

1 REFERENCE Oconee Tech Specs 3.5.2.2 3.7/4.1 l OO1005KOO5 ...(KA'S) j

)

l ANSWER 8.15 (2.00)

I .i Unit 1 shall be placed in a hot shutdown condition within 12

, hours. If these requirements are not met within an f"" additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be placed in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. C1.03 Unit 3 The inoperable circuit / channel shall be restored to operability and the conditions of Table 3.7-1 for normal operation shall be satisfied for all functional units before l

the reactor is returned to criticality. C1.03 l

' REFERENCE Oconee TS 3.7-1 3.1/3.8 062OO5KOO5 ...(KA'S)

e - e.

ENCLOSURE 3 Attachment 1 SPECIFIC COMMENTS REGARDING THE REACTOR OPERATOR LICENSING EXAMINATION Category 1.0 Principles of Nuclear Power Plant Operation, Thermodynamics, Heat Transfer, and Fluid Flow 1.02 Another acceptable answer to this question is "c. Delta Tc Failure "B" Side Low". The Delta Tc circuit does a comparison of the two loop Tc's and will reratio feedwater accordingly in order to balance the TC's. "B" Side Low will generate the same system response as "A" Side High in order to achieve this balance.

i

Reference:

OP-OC-SPS-IC-ICS pages 44 of 87 I

1.06 Another acceptable answer to this question is "b.

The Suberitical Multiplication Factor (M)". It is '

true that the effective half life of the delayed i neutron precursors play a role in the length of time required to reach an equilibr:nn sub-critical countrate. However, the time reIuired for a given reactivity addition increases as the reactor approaches criticality. This is a result of the magnitude of change in "M" which will require more neutron generations to achieve the new equilibrium l countrate given essentially no change in a neutron generation lifetime.

Reference:

Fundamental of Nuclear Reactor Engineering Page 117 OP-OC-SPS-RT-SM pages 14-15 of 22 1.11 Another acceptable answer to this question is "d.

A decrease in Two-phase (nucleate boiling) within the core". A decrease in nucleate boiling within the core can be interpreted to be a decrease in the heat source. A decrease in the heat source would reduce natural circulation flow due to a decrease in the delta T reducing the driving head of the fluid.

Page 1

V .' I I

i l

An examiner clarified answer d." to at least one candidate as increasing subcooled margin which supports the interpretation of a decrease in the heat source.

Reference:

Thermodynamics, Fluid Flow, and' Heat  !

Transfer for Nuclear Power Plants J Page 195 OP-OC-SPS-PTR-AM1 Pages 11, 12, and 1 13 of 50 j 1.19b Another acceptable answer to this question is j

" Unit 3 because rod worth increases over core  !

life resulting in a. larger negative reactivity l addition on Unit Three and therefore a largar i decrease in Tave." The change in rod worth competes i with the change in the MTC.and without additional )

information the overriding effect can not be i determined.

Reference:

OP-OC-SPS-RT-IP Page 22 of 30 l 1.20 Other acceptable answers to thj.s question include: ,

1 l If the overfill were to result in filling the  !

steam lines the weight of the water could cause j damage to the steam line supports;and stanchions. l (Filling steam lines is acceptable at Oconee if pipe hangers have been blocked / braced).

Water carryover to the Turbine-Generator could i result in damage to the blading of the turbine.

Flooding the aspirating ports in the steam generator would cause a loss of preheat and could possibly lead to thermal stresses on the lower tube sheet and/or water carryover to the Turbine-Generator.

Reference:

OP-OC-SPS-PRT-AT Pages 30 of 44 OP-OC-SPS-CM-MT Pages 97 of 113 OP-OC-SPS-CM-SG Pages 11 of.22 Page 2

_ _ - _ _ _ - _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ = _ _ - _ _ _ _ _ _ _ _ - _ _ _ _

. = _ - __ _ _ _ _ _ _ - _ -_- --- .-

1.21 The phrase "in the primary" contained:in this question could be misleading. The: Zinc-Boric acid reaction is not "in the primary" and would be omitted as a source'of hydrogen if'the candidate -

is mislead. Instead of the Zinc-Boric acid reaction some candidates may list'the Steam-Steel' reaction which would occur "in the primary".

Also, an examiner clarified'this question to'at least one candidateLas a " gross failure." This clarification led the candidate'to believe a greater than 1% zirc-water reaction occurred;(as l

assumed.in OP-OC-SPS-SY-HDC) .and therefore changed l

the order of significance of the different sources.-

Reference:

OP-OC-SPS-SY-HDC Pages 11 and 12 of 25 OP-OC-SPS-PTR-AM2: Drwg.'OC-PTR-AM-2 1.22 Most. examinees probably will not address the effect on induced xenon oscillation-from APSR's, since our stress is mainly in the area.of operationally related axial imbalance control via the part-length control rods. While it is recognized that-any ,

pertubation of the flux will perturb the xenon. j distribution, the principal. indicator and concern ')

for the operator when dealing with APSR's is axial i imbalance.

1.25 The answer key is incorrect.when using Mollier.

1 Moisture content is 5% / Quality is 95%. ..

Reference:

Mollier Chart i

Page 3

Category 2.0 Plant Design' Including Safety and Emergency Systems 2.07b All units have an' indication on the computer display for HP-409/410 not being fully closed.

Digital points have been installed on all units which provide indication / input to SPDS.  !

Reference:

OP-OC-SPS-IC-SPDS Page 39 of 50/ 1 Drwg. OC-IC-SPDS-21 j 2.08a The' wording on this question is misleading. The submersible Pump is used to supply water to the CCW l Intake Piping and not the CCW Intake Area. j b The wording on this question is misleading. The l Air Ejector is used'to evacuate any air which may  !

collect in the common SSF Service Water suction line  ;

and possibly cause cavitation of pumps taking j suction from this source. Minimizing air injection j into S/Gs is not the reason for valving in the Air Ejector.

l

Reference:

OP-OC-SPS-SSF-ASW Tracking 87-034/030  !

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2.10 This question may also be interpreted as asking to explain the cycling of the TBVs. If interpreted l this way another acceptable answer would be to l address steam entering the condenser at a rate where I vacuum could not be maintained. After-the TBVs close vacuum.is regained such that the TBVs can once 3 l again open. The answer key only addresses why J l

vacuum is* low and does not address the cycling of I the TBVs. 1

Reference:

AP/1/A/1700/11 2.13 The referenced objectives given for this question require knowledge of the function and basic operation of the RCP Monitor. These objectives are in the OP-OC-SPS-CM-CPM lesson along with a basic description of the RCP Monitors necessary to meet the objectives. The Contact Monitor Auxiliary Power Supply is not addressed in the OP-OC-SPS-CM-CPM lesson since it is not required to satisfy the objectives.

Page 4 j

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l The Contact Monitor Auxiliary Power Supply is addressed in the OP-OC-SPS-IC-RPS lesson but no l requirement is made (objective)'of the operator to l describe the contact Monitor Auxiliary Power Supply. I This question should be deleted.

Reference:

OP-OC-SPS-IC-RPS ,

OP-OC-SPS-CM-CPM 2.14 The Emergency Feedwater System now controls SG level at 30 inches when actuated with at least one RCP running.

Also, since loss of both MFWPTs is required for i actuation of the Emergency Feedwater System the condition which determines the controlling SG level l is whether or not a RCP is on.

Reference:

Information Attached OP-OC-SPS-SY-EF Page 64 1

2.15 When listing the flow paths of electrical power, an i option available to the Oconee operators is use of a backup Startup transformer through the Emergency Startup buses supplying power to the 6900 and 4160 volt buses. On Unit 1 the backup Startup transformer would be CT-2. When listed in order of preference this flow path would be listed after the Normal Startup transformer and prior to the Standby transformer CT-4. (Between 3 and 4 on the answer J key.) With this option, back charging the main transformer may be omitted since it. requires considerable time to establish.

Also, ITA/TB may be referred to as the 6900V buses and MFB 1/2 may be referred to as the 4160 V buses.

References:

OP-OC-SPS-EL-EPD Page 17 and 18 of 37 OP/1/A/1107/02 Normal Power Procedure Technical Specification 3.7.1(b)2. i i

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, 2.16 On Unit 1 the RCPs are Westinghouse RCPs which do  !

I not have cooling jacket / seal coolers. Instead CC is used to cool the thermal barrier on the Unit 1 RCPs.  !

A more appropriate answer would be simply the RCP seals when addressing this load.

Reference:

OP-OC-SPS-SY-CC Page 18 of 20 '

OP-OC-SPS-CM-CPS Drwgs. OC-CM-CPS-1 and 2 1 E--_----. "" 5 j

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l 2.17 Some individuals will answer the question by breaking down " Loss of Cooling Water to Necessary Components" to individual components. That is, they will list individually CRD's, RCP Motors, etc. ,

When they have four things listed, they will stop.

Therefore, their answers, although correct, will not reflect all four of the answers given in'the answer key.

2.20b Another acceptable answer would be, "By following procedure." The answer given on the key is a result of following the Unit Startup Procedure. This procedure will insure all ICS stations are in auto except for 6Tc and the second MFP. Neither of these stations will place ICS in track. Therefore, by l following procedure the Unit should not be in Track l when Turbine control is placed in automatic. l

Reference:

OP/1/A/1102/01 Encl. 4.3 OP/1/A/1102/04 Encl. 3.3 2.21 Other acceptable answers to this question include:

- Open HP-410 (this provides a bypass flow path around HP-26 should HP-26 fail to open / HP-115 is normally open)

- Balance letdown flow with seal injection to maintain pressurizer level constant

- Open HP-24/25 (if the question in interpreted to also mean makeup to the LDST / HP98, 99 and 100 are normally open)

Reference:

OP-OC-SPS-SY-HPI Pages 28 nnd 29 of 43/

Drwg. OC-SY-HPI-14 OP/1/A/1104/02 High Pressure Injection l

2.22c MS-97 is closed with its breaker locked open to prevent a loss of vacuum.

References:

OP-OC-SPS-SY-EF Page 67 of 71 Page 6

Category 3.0 Instruments and controls 3.04 While the various types of RZ modules are addressed in OP-OC-SPS-IC-ES, it is not intended that the {

operators be able to classify the RZ modules )

according to " Type". The " Type" designation is used for ease of instruction by grouping the different components according to their RZ control module.

The operators are instructed on recognition of the different modules and how each is to be operated.

Given a drawing of the RZ module, or from the RZ control panel, the Operator can describe the operation of a particular module. This question should be deleted.  !

Reference:

OP-OC-SPS-IC-ES Drwgs. OC-IC-ES-12 thru 17  ;

{

l 3.05 No objective is referenced in support of this question. Operation of local switchgear. controls is not performed by operations and therefore instruction on local switchgear controls is not addressed. This question should be deleted.

3.08 Another acceptable answer to this question is "c.

A false zero level indication". The degree of failure or location of failure will affect the resultant indication. On a bellows rupture a false zero level indication will result.

Reference:

OP-OC-IC-RCI Pages 28 and 29 of 49 Page 7

3.09a The description in OP-OC-SPS-IC-ICS is not accurate.

The BTU limits have been removed above 25% full '

power. With the "A" TBVs open, the "A" S/G will q have an increase in steam flow causing ATc to indicate "A" side cold. With this ATc indication

]

feed flow to'the "B" SG will be increased.

Therefore-the Initial Response for Feedwater Flow to "B" S/G will be to increase.

Reference:

- NSM attached OP-OC-SPS-IC-ICS Pages 44-of 87 l 3.13a&b The pressurizer spray valve (a) and the pressurizer spray block valve (b) are both manually operable i from the control board during a' loss of Auto Power.  !

However, there is no valve position indication .

available at this location. (The pressurizer spray valve can also be operated by a key switch in the ICS cabinet as indicated by the answer key.)

Reference:

AP/1/A/1700/23 Encl. 6.2 3.17a To place the Low Range Cooldown Pressure' instrument into service the operator must contact the instrument department to valve in the transmitter.

The instrument department will have the operator open RC5 (or 6) and RC.7.- l b The correct answer is NO. The PORV Setpoint Swf.tch is used to select the setpoint. The setpoint is not affected by valving the transmitter in or out.

Reference:

OP-OC-IC-RCI Page 21 of 49 -

I OP/1/A/1102/10 Encl. 4.2 l

3.19 The answer key is incorrect in regards to the 1720 psig trip setpoint. The 1720 psig trip setpoint is a HIGH pressure trip. Also another acceptable answer for the administrative high flux setpoint is 5 5% as required by Technical Specifications.

Reference:

OP-OC-SPS-IC-RPS Page 16 of 63 I Technical Specification Table 2.3-1 l

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1 3.20 Terminology was a key factor in answering this question. The candidate could know how the logic and channel operated to produce a reactor trip but  ;

because of an inability to distinguish between the different terms he would be unable to correctly answer this question. The terminology used in this i question is not used by the Oconee operator, 1 therefore this question does not test the objective referenced. This question should be deleted.

3.22 Other acceptable answers to this question include:

I

- (SAM) 3RC-1

- Turbine Bypass Valves

- RCP Seal Flow Indication ,

- Seal Supply Control Valve (HP-31)  !

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Reference:

OP-OC-SPS-EL-VPS Page 20 of 26/

Drwg. OC-EL-VPS-14.

3.26b Other acceptable answers to this question include:

- Control rods withdrawing

- Reduction in Feedwater Clarification provided to several candidates by the examiner related 3.26b back to 3.26a which supports  ;

the additional answers provided. i 1

Reference:

OP-OC-TA-NT Pages 28,-19, and 26 of 29/

Drwgs. OC-TA-NT-14 and 21 1

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4.

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t ,) 4

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7 '

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[, . , , y ,

4 s ,' s i

.- I l Category i.v Proced 6e-Normal, Abnormal, Emerayncy, and l

,1 Radiologfecal Centrol )

' i 1 4.01, This question a'nd answer contain.a'dodole negative i nWC, N@ which makcF, it confusirq. Also answe2' "a" I could'be considered ts sn, incorrect answer due to

'both k;;these limits (in the procedure) being  ;

e directly related to f{egge_ncy of operation. " Rate" l ani =" Frequtacy" can le interchangeable. With this  !

' interpretation'there wo61d be no correct answer. ,

This question should be dsleted.

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l Referepte: OP/0/A/1105/004 )

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4.02 The an'cwer key referedces the DPC EPG for this question. Although olt' lesson plans on the EOP use the'EPG as a basis, we do not instruct dirOttly from the EPG document because it is a very detailed and l i

technical paper that ofte: cordains materill beyond thc scope of what an opehator Enceds.

The particular area addressed by question 4.02 is, howesit, covered with the operator during accident i mitigation lectures, specifically under

>OI'-OC-TA-AM-2, Acaldent Mitigation: Gas / Steam l Bitding. Objective 1.,c requires that the operater Pe able to describe thd "means of dealing with steam / gas, accumulation the the RCS". Attached Are the two p-ages f rsm the lesson plan that address this situation >)from Laese, paces it can be seen that vessel or hot leg venting is, indeed, a method of deal.ing yith non-condensible gases in the RCS, but by ny my ns, the "best" method. RCP Restart or in some cases, " bumps",'is probably.a better method, if RCP's'are available. 'Hewever, the situation will dictate the "best" method for removel cf non-condencibles. Answers b, c, og d should be acceptable fot'guention T.0.1.

i 4.07b Another acceptable answer would be 25 rem. The Maximum Planned Emergency Exposure limit to the Whole Body to Save Lives pe:* th9, Health Physics Manual is 25 rem. However, the amergency Plan lists a limit of 75 rem if approvid by the Emergw.cy Coordinator. Either answer sr.ould be accepttble.

Reference:

Health Physics Finual Page 10 Emergency Plan Page K-1 RP/0/B/1000/11 Encl. 4.2 Page 10 >

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1 4.12 Other acceptable answers would include:  ;

1

2. HPI pump electrical power source could also be listed as Standby bus #1 since this is the power i source to the Aux. Service Water Switchgear. l

3. Standby bus #1 power source could also be listed as the Central Switchyard or Lee steam Station since this is the power source to CT-5. j

4. HPI pump motor cooling could also be listed as HPSW since this is the backup cooling source used until Auxiliary Service Water is started. j l

Reference:

OP-OC-SPS-EL-EPD Drwg. OC-EL-EPD-2 I OP-OC-SPS-SY-HPI Page 24 l 4.17 Another possible answer is; "To cool the LPI fluid after swapping to the emergency sump." Cooling the fluid would accomplish the other two bases by preventing cavitation of LPI and BS pumps and by removing heat from the core.

Reference:

OP-OC-SPS-SY-BS Page 9 I

4.18 Additional answers could include:

1. RIA 37 will also trip the waste gas exhauster.

2. RIA 33 Will also trip both condensate Monitor Tank Pumps. 1

3. RIA 49 will also sound the RB evacuation alarm. l

Reference:

OP-OC-SPS-IC-PRM Pages 19 and 20 of 24 j 1

l 4.19 All RCS hot and cold legs at least 50'F subcooled and j pressurizer level going off scale high is no longer a criteria which must be met in order to secure the HPI System.

This critelia is not listed in the Emergency Plan and has been deleted from the OP-OC-SPS-IC-ES lesson in the last rewrite of this lesson.

Reference:

EP/1/A/1800/01 OP-OC-IC-ES (Last rewrite of OP-OC-SPS-IC-ES attached.)

Page 11

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i This question had misleading terminology. While I

4.22 the term " saturated repressurization" is used in the EPG, the Emergency Plan and lessons do not use this term. Had the phenomena been addressed as-it is j in the Emergency Plan, the candidates would have l understood what was being asked allowing this question to test their understanding of the transient and not-the " term". This question should be deleted. ,

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I Attachment 2 SPECIFIC COMMENTS REGARDING THE SENIOR REACTOR OPERATOR LICENSING EXAMINATION Category 5.0 Theory of Nuclear Power Plant Operation, Fluids and l Thermodynamics l l

I 5.02 Refer to R. O. Licensing Exam 1.02 1

5.07 " " 1.11 l

5.19 " " 1.20 5.20 " " 1.21 .i l

5.22 " "

1.22 l 5.25 " 1.25  !

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\

l Category 6.0 Plant Systems Design, Control, and Instrumentation i l

! 6.06 Refer to R. O. Licensing Exam 3.04 6.07 " " 3.05 6.09a " "

3.09a j

" " 3.13a and b 6.11a and b 6.22b " "

2.20b 6.23 " "

2.21 l l

l 6.25b " " 3.26b l

l Page 13 l

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. .. a Category 7.0 Procedures - Normal, Abnormal, Emergency, and Radiological Control 7.01 Refer to R. O. Licensing Exam 4.01 7.02 " " 4.02 7.12 " " 4.12 7.16 Answers 13 and 14 can be incorporated into one:

HPI Forced Cooling I Answer 12 One student commented that he understood an examiner's clarification to mean that.EFDW from the affected unit was one correct answer and that EFDW from another unit could be counted as another correct answer.

Another answer not listed would be ASW from the Auxiliary Building. Reference - EP/1/A/1800/01 EOP Section 502 Loss of Heat Transfer Step 11.0 7.18 Refer to R. O. Licensing Exam 4.17 7.19 "

4.19 7.22 " "

4.22 l

Category 8.0 Administration Procedure, Conditions, and Limitations 8.01 Answer C is not a correct answer. Two individuals i using Remote Indicator is allowed, while answer d is not allowed and is therefore the correct answer.  ;

Reference:

Station Directive 2.2.2 5.3  !

8.04 Answer C applies if the work supervisor CANNOT BE' LOCATED .to letrieve the red tag stub. i Answer A could also apply if the work supervisor is not on site. I Either answer A or C should be accepted

Reference:

Station-Directive 3.1'1, 6.11.6, and 7.4.3 attached Page 14

8.06 Answer B could also be correct based on the students interpretation of the statement ". . . all EFW Flow instrumentation for 'B' OTSG were inoperable." Page 3.4-3 of this Tech. Spec. states that a S/G level indicator is a sufficient flow indication.

Therefore if "all" EFW Flow instrumentation were

, inoperable that might include level indication also.

If so, answer B is the most limiting LCO and would be a correct answer.

8.13b Refer to R. O. Licensing Exam 4.07b 8.14 It is agreed that the basic premise of the question if for the SRO to know that RIA-45 is required to move fuel. The SRO should also know that if the motor for the pump supplying RIA-43 is de-energized, RIA-45 also becomes inoperable. However, if the'SRO states or assumes that RIA-43, and only RIA-43, is taken out of service (by de-energizing the readout module) then "yes" should be an acceptable answer, since RIA-43 is not required to move fuel.

Also, the intent of Specification 3.8.10 is to insure that the Reactor Building Purge is capable of being automatically isolated in the event of a fuel handling accident. There is no requirement that the Reactor Building Purge be in operation during fuel handling operations. If the Reactor Building Purge is secured and the isolation valves PR 2,3,4, and 5 are closed as required by Specification 3.8.7 then the intent of Specification 3.8.10 is met and RIA-45 could be removed from service.

Reference:

Technical Specification 3.8 Analysis of Technical Specification 3.8.7 (Encl. 7.2 of IIR 086-27-2 attached) 8.15 Although Station Directive 2.2.1, paragraph 10.1 does agree the " Procedure Discrepancies Process Record" form OMP l-9 also addresses alternate locations for documenting procedural discrepancies, such as, R&R book or the " Remarks' section of the procedure process record, These answers should also be acceptable (See Attached sections of OMP l-9).

8.23 Answer 2 - Number of RCPs operating is addressed in the bases of T.S. 2.1 as RC Flow. Therefore, RC Flow should be an alternate acceptable answer for number of RCP's.

Reference:

Technical Specification 2.1 Bases Page 15

Attachment 3 l

SPECIFIC COMMENTS REGARDING THE REACTOR OPERATOR REQUALIFICATION EXAMINATION Category 1.0 Principles of Nuclear Power Plant Operation, ,

Thermodynamics, Heat Transfer and Fluid Flow 1.02- Refer to R. O. Licensing Exam 1.02 1.11b " " 1.19b ,

1.12 " " 1.22 1.14 " " 1.25 Category 2.0 Plant Design Including Safety and Emergency Systems 2.04 Refer to R. O. Licensing Exam 3.04  !

2.05 " "

3.06 2.06 " "

2.08 2.10 " "

2.14 2.11 2.15

]

2.12 2.17-2.13b Refer to R. O. Licensing Exam 2.20b 2.14 " "

2.21 i

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J Category 3.0 Instruments and Controls ]

l 3.04 I 3.04 Refer to R.O Licensing Exam

" 3.05 3.05 " i I

3.07 " 3.08 3.08a " "

3.09a I

3.09 The last part of question 3.09 "The PORV is ]

activated by a pilot valve which is connected to )

a [ class IE] system" is not operator applicable. i Although the statement is, indeed, in the referenced document (OP-OC-SPS-CM-PZR, page 14) the information is not relevant to operations personnel, nor is taught them, nor is there any objective requiring an i operator to know this. We feel that this last portion of the question should be deleted.  !

I 3.13 Refer to R. O. Licensing Exam 3.19 3.14 " " 3.20 Category 4.0 Procedures - Normal, Abnormal, Emergency, and Radiological Control 4.01 Refer to R. O. Licensing Exam 4.02 4.03b 4.07b 4.06 " " 4.12 4.11 4.19 4.13 " " 4.21 Page 17

- c. a.

Attachment 4 SPECIFIC COMMENTS REGARDING THE SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION Category 5.0 Theory of Nuclear Power-Plant Operations, Fluids, and Thermodynamics 5.02 Refer to R. O. Licensing Exam'1.02 5.05 " " 1.11 5.14 " "

1.25 Category 6.0 Plant Systems Design, Control, and Instrumentation 6.04 Refer to R. O. Licensing Exam 3.04 6.05 " "

3.05

" " 3.08 6.08 6.09 " " 2.08 6.14 Refer to R. O. Requal Exam 2.12 6.15b Refer to R. O. Licensing Exam 2.20b 6.17 " "

2.21 1

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Category 7.0 Procedures - Normal, Abnormal, Emergency, and l Radiological Control 7.01 Refer to R. O. Licensing Exam 4.02 7.13 Refer to SRO Licensing Exam 7.16 7.15 Refer to R. O. Licensing Exam 4.18 -l Category 8.0 Administration Procedures, Conditions, and Limitations 8.01 Refer to SRO Licensing Exam 8.04 i 8.06 " " 8.14 )

1 8.07 8.15 8.13 " " 8.23 f i

8.15 Two examinees, Snowden and Chudzik, had already ,

completed their exams and turned them in before it- I was realized that the necessary Tech. Spec. excerpts had not been included with the exam. They did not have the benefit of referring to the excerpts to answer the question.

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1 ENCLOSURE 4 REQUALIFICATION PROGRAM EVALUATION REPORT Facility: Ocoree Nuclear Station Examiners: B. Dean C. Casto J. Huenefeld B. Gore S. Lawyer Dates of Evaluation: July 13-23, 1987 l Areas Evaluated: x Written x Oral x Simulator Examination Results:

R0 SR0 Total Evaluation Pass / Fail Pass / Fail Pass / Fail (S,M,orU)

Written Examination 7/3 8/1 15/4 S*

Operating Test 10/0 9/0 19/0 S Oral 10/0 9/0 19/0 S Simulator 10/0 9/0 19/0 S I 1 Evaluation of facility written examination grading N/A l Overall Program Evaluation i Satisfactory x Marginal Unsatisfactory (Listmajordeficiency areas with brief descriptivecomments)

I

• Satisfactory even though passing rate is 79%. Due to the number of operators i examined, a percentage of 80% was unattainable. This pass rate is  !

essentially 80%.

One individual failed the written exam with a grade of 52.5%. This '

individual's qualifications should be closely examined by the facility, with regards to retraining and licensed activities.

Submitted:p,. Forwarded: Approved:

(. / // [ - ' '?

W. M. Dearr J. F. Mdriro C. A. Julian

.