Text

Exam Rept 50-338/OL-87-01 on 870209-12.Exam Results:Five of Five Reactor Operators & Six of Six Senior Reactor Operators Passed Exam.Nrc Exams & Comments on Exams Encl
	ML20206F466
Person / Time
Site:	North Anna
Issue date:	04/07/1987
From:	Moorman J, Munro J; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML20206F444	List: ML20206F440; ML20206F466;
References
	50-338-OL-87-01, 50-338-OL-87-1, NUDOCS 8704140261
	Download: ML20206F466 (130)
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. o ENCLOSURE 1 EXAMINATION REPORT 338/0L-87-01 Facility Licensee: Virginia Electric and Power Company Facility Name: North Anna Power Station Facility Docket Nos.: 50-338 and 50-339 Written, oral, and simulator examinations were auiinistered at North Anna Power Station near Mineral, Virginia.

Chief Examiner: // MW [

J. h . Moorman, III Y-h-87 Date Signed Approved by: 240 N ,n# M/7f7 {

John F.'Munro, Section Chief ~Date Signed Summary:

Examinations on February 9-12, 1987.

Written, oral, and simulator examinations were administered to eight candidates; all of whom passed. Three candidates were administered simulator re-examinations; all candidates passed.

Based on the results described above, five of five R0's passed and six of six SR0's passed.

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8704140261 870408 PDR ADOCK 05000338 V PDR L

REPORT DETAILS

1. Facility Employees Contacted:

R. Buck, Supervisor, Operations Training

M. Crist, Training Staff

L. Edmonds, Superintendent, Nuclear Training

R. Enfinger, Superintendent, Operations

D. Fellows, Training Staff

W. Harrell, Station Manager

W. Shura, Training Staff

R. Stevens, Training Staff

Attended Exit Meeting

2. Examiners:

J. Arildsen N. Jensen, INEL

J. Moorman B. Picker, INEL

Chief Examiner

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

a. R0 Examination - Analogous SR0 questions in parenthesis (1) Question 1.03 (5.17)

NRC Resolution:

Utility comment accepted. The answer has been modified to allow full credit if "95%" is not mentioned. However, the answer must provide assurance that the candidate understands that maintaining DNBR 2:1.3 will not always prevent DNB.

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(2) Question 1.04(3) (5.10b)

NRC Resolution:

Utility coment accepted. RCS Temperature is a parameter and will be accepted for full credit. Tu will also be accepted; however, T will not be accepted sint'e it remains essentially constant dt7 ring operation.

(3) Question 1.05 (5.18)

NRC Resolution:

Utility comment not accepted. The non-existence of superheating in a system is not a condition that must be present for natural circulation to exist. Rather, the non-existence of superheating promotes natural circulation.

(4) Question 1.08b(5.07b)

NRC Resolution:

Utility comment accepted. The MTC becomes "more negative" with core age and " increase" will be an acceptable answer. The answer key has been modified.

(5) Question 1.11 (5.19)

NRC Resolution:

Utility comment not accepted. The utility coment is based on a dynamic situation (i.e., the difference between two points).

The question clearly states the initial conditions of the problem.

(6) Question 1.19 NRC Resolution:

^

Utility comment not accepted. This type of knowledge should be basic to anyone who has completed an R0 or SR0 training program.

Errors caused by miscalculation will not be compounded through the remainder of the question.

(7) Question 1.23a NRC Resolution:

Utility comment not accepted. It is clearly stated in the question that all parameters are equal (i.e., source strength of each reactor is equal).

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(8) Question 1.23b NRC Resolution:

Utility comment not accepted. Knowledge of how rod worth varies over core life is supported by NUREG-1122, Knowledge and Abilities Catalog for Nuclear Power Plant Operators:

Pressurized Water Reactors. The question was used to check operator knowledge of how rod worth varies over core life. To assume equivalent rod worth would imply a trivial answer.

(9) Question 2.06 NRC Resolution:

Utility comment not accepted. Knowledge of the power supplies for pressurizer heaters is supported by NUREG-1122. The point value of the question has been changed to 1.25 points.

(10) Question 2.10e NRC Resolution:

Utility comment accepted. The utility's comment is an equivalent answer. This clarification will be added to the answer key.

(11) Question 2.11a NRC Resolution:

Utility comment not accepted. Reference material provided states that the floating seal ring limits leakage to 50 gpm.

Since no further reference material is available to support other limits, this is the value that will be used.

(12) Question 2.15 NRC Resolution:

Utility comment accepted. Discharge canal and bladder tank are equivalent answers and either will be accepted for full credit.

The answer key has been modified.

(13) Question 2.17 NRC Resolution:

Utility comment accepted. The answer key was changed to require only the minimum values asked for in the question.

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4 (14) Question 2.23 (6.07)

NRC Resolution:

Utility comment accepted. Due to the poor quality of the

-question, Part A has been deleted. Total value of the question is changed to. 75 points.

(15) Question 3.06 Utility comment accepted. Due to poor wording of the question and the clarification given during the exam, both A and C will be accepted as correct answers.

(16) Question 3.12c (6.17c)

Utility comment accepted. Due to a typographical error in the question, this part of the question has been deleted. The question value is now 1.0 point.

(17) Question'3.16 (6.14)

Utility comment accepted. Recommended answers are equivalent to answers 3 and 1. The answer key has been modified to recognize these clarifications.

(18) Question 3.18 NRC Resolution:

, Utility comment accepted. The answer key has been modified to accept this additional answer.

(19) Question 3.24a NRC Resolution:

Utility comment accepted. The answer has been changed to include N44 as an input to the main feed bypass valve controller and to give .25 points for its inclusion in the answer. The total value of the question is now 1.25 points.

(20) 4.01 (7.01)

NRC Resolution:

Utility comment not accepted. It is acknowledged that the question is not specific as to whether seal injection flow or No.1 seal leakoff flow is to be used. However, the reference material provided by the utility states that the minimum seal injection flow is six gpm and the minimum No.1 seal leakoff

5 flow is .2 gpm. If a candidate answered the question that there are no correct answers using the six gpm minimum seal injection flow as a premise, this answer will be accepted for full credit.

If the candidate answered the question assuming .2 gpm for No. 1 seal leakoff flow and 15 psig minimum VCT pressure (answer 2),

this will be accepted for full credit.

(21) Question 4.13 (7.13)

NRC Resolution:

Utility comment accepted. The answer key has been modified to distinguish between the separate conditions,

b. SR0 Examination (1) Question 5.05 NRC Resolution:

Utility comment accepted. The answer key has been revised to reflect the correct answer.

(2) Question 5.14b

-NRC Resolution:

Utility comment not accepted. Although conductive heat transfer does occur in this instance, it is not the primary cause for the increase in heat transfer.

(3) Question 6.10a NRC Resolution:

Utility comment accepted. The answer key has been modified to reflect all poscible answers. The point value of the question has been revised to reflect the additional answers. Part C was deleted by a clarification during the exam.

(4) Question 7.09d NRC Resolution:

Utility comment accepted. The answer key has been changed to reflect the correct answer.

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(5) Question 7.14 (4.14)

NRC Resolution:

Utility comment accepted. The answer has been modified to accept " Inject the BIT" as a separate step. The point value of the question has been increased by .25 points.

(6) Question 7.20 NRC Resolution:

Utility comment accepted. The requirement for the flow paths to be placed in an order has been deleted and the value for each response has been raised to .365 points.

(7) Question 8.09 NRC Resolution:

Utility comment accepted. The answer key has been changed to reflect the requirements of the North Anna Power Station.

c. After further review of the examinations, the following changes were made.

(1) Question 1.18 NRC Resolution:

Distractors 'A' and 'C' are basically the same statement.

Either answer will be accepted for full credit.

(2) Question 2.20 NRC Resolution:

This question only solicits answer (2). This answer will be accepted for .75 points. Part one of the answer has been deleted.

The parenthesis have been deleted to require " chloride stress corrosion" to be in the answer.

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7 (3) Question'2.21 (6.03)

NRC' Resolution:

The answer will be modified .to accept "to aid in warming up the RHR system" and "for overpressure protection anytime." The answer " pressure equalization" is equivalent to " allow for expansion."

-(4) Question 3.19 NRC Resolution:

The answer will be changed as follows:

Normal: 125 vdc vital bus (125 vdc/120/vac static inverter)

Alternate: 480 V ac emergency panel (1H1,1J1) - 480/120 ac-transformer Emergency: battery (to 125 v dc vital bus)

(5) Question 4.23 (7.24)

NRC Resolution: '

Acceptable as a reason for using steam pressure mode will be "because. Tavg mode can not be used to cooldown below 547 F (no ,

. load Tavg)." Reasonable wording will also be accepted.

(6) -Question 3.17 (6.20)

NRC Resolution: .

The answer key has been amended to clarify the " Breaker 86 and 87 protective relays" as overcurrent and phase-differential.

4. Exit Meeting At the~ conclusion of the site visit the examiners met with representatives

-of the plant staff to-discuss the results of the examination.

There was one generic weakness noted during the simulator examination.

The area of below normal performance was imprecise communications.

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

facility for under-instruction examiners-and examiner audit personnel were 1

-noted and appreciated.

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

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p- 1 U. S. NUCLEAR-REGULATORY COMMISSION

, REACTOR OPERATOR LICENSE EXAMINATION FACILITY: NORTH ANNA 1&2 REACTOR TYPE: PWR-WEC3 DATE ADMINISTERED: 87/02/09 EXAMINER: MOORMAN. J.

CANDIDATE: AiAS M

. INSTRUCTIONS TO CANDIDATE:

Use ' separate paper for.the answers. Write answers on one side only.

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

% OF ,

. CATEGORY % OF CANDIDATE'S CATEGORY

.__VALUE TOTAL SCORE VALUE CATEGORY t S. G 30.00 00- I. PRINCIPLES OF NUCLEAR POWER PLANT. OPERATION, THERMODYNAMICS, HEAT TRANSFER AND: FLUID FLOW

28. L5 'L%\

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY-SYSTEMS

! 13.i5 15.i

. '30.00- 25.00- 3. INSTRUMENTS AND CONTROLS

[ 30.tr-30.00 i

d. 5 2. 9 .

-25.2v0 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL

j. . g g.tr .. CONTROL-L iii. is' Totals

'd20.00- %

I Final Grade

' All work'done on this examination is my own. I have neither given

nor. received aid.

1 Candidate's Signature l'

gt .

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

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

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

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

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

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

6. Use only the paper provided for answers.

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

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

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

10. Skip at least three lines between each answer.

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

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

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

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

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

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

1 (Y

18. When you complete your examination, you shall:

a. Assemble your examination as follows:

(1) Exam questions on top.

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

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

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

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

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

[ 1A__EBIOGIELES_9E_NUGLE88_EgWEB_ELONI_QEEB811gN 2 PAGE -2

'IBEBUggyN8dIC@2_UE81_IB8NSEEB_8NQ_ELUIQ_ELQW QUESTION 1.01 (2.00)

Given two. pumps of equivalent design, operating at the same, constant speed:

A. WF will'be the effect of placing the.two pumps in s er i s . =, swith respect to flow and head)?

B. What will be the effect of placing the two pumps in parallel (with respect to flow and head)?

OUESTION 1.02 (1.00)

Given: Three reactor coolant (RCP) pumps operating in parallel, each with a flow rate "m" and a combined flow rate "M". Out of the four possibilities below, choose the one that best fits if one RCP is secured,

a. The_resulting core flow (M) will increase.

b. The resulting core flow (M) will increase along with individual operating RCP flow (m).

c. The resulting core flow (M) will decrease as individual operating RCP flow (m) increases.

d. The resulting core flow (M) will not change due to decrease in RCP back pressure.

QUESTION 1.03 (1.00)

What in the design basis of having a DNBR > or = to 1.37 OUESTION 1.04 (2.00)

List the four (4) plant parameters observed to insure that CHF or DNDR are not exceeded.

QUESTION 1.05 (2.00)

What are all the conditions that must be present in order e for natural circulation to exist?

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(- 1 It__EBINCIELES_gE_UUg6E88_EgNER P68NI_gEEB8IlgN 3 PAGE 3-

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QUESTION 1.06 (1.00)

With respect to reactor thermal limits which of the following statements is NOT correct.

Ja. The ratio of_the peak linear power density to the average linear power density _in the core at a particular elevation is called the nuclear heat flux hot cl.annel factor.

~b. The average linear power density in the core is expressed in units of.kw/ft_and is the total' thermal power divided by the active length of all the fuel rods.

c. The purpose of limiting the enthalpy rise hot channel factor is to prevent bulk boiling from taking place during normal operations.

d.-The rod bow penalty (RBP) accounts for the bowing of fuel rods as their'burnup increases.

e.'The purpose of the limit on the heat flux hot channel

-. factor is to insure that fuel clad temperature does not exceed 2200 deg F during normal operations.

D'ESTION U 1.07 ( .50)

Consider a fuel pellet at 70 deg F. A 6.7ev neutron coming in will be absorbed. The 6.7ev neutron will be absorbed in the outer part of the fuel. The inner fuel will not even see the neutron (low flux). This phenomenon is called QUESTION 1.08 (1.50)

Write on your answer sheet INCREASES , DECREASES or DOES NOT CHANGE for the following:

The magnitude of the fuel temperature coefficient (FTC):

A. INCREASES / DECREASES / DOES NOT CHANGE with increase in power.

B. INCREASES / DECREASES / DOES NOT CHANGE with core age.

C. INCREASES / DECREASES / DOES NOT CHANGE with decrease in moderator temperature coefficient (MTC).

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1,3_ ' PRINCIPLES OF NQGLEAR POWEB_ PLANT ODERATION s PAGE. 4 ISEBdQQyU801GS1_SE81_IE6MSEEB_6NQ_E6Mlp_E(QW

i QUESTION 1.09 (1.00)

. The' negative reactivity added when f uel temperature increases is primarily caused by ______,.

a.'self shielding of the fuel

b. doppler broadening

c. an increase in the GAMMA heating contribution

d. fuel pellet swell thus decr easing the gap QUESTION 1.10 (1.00)

Which one of the following statements is correct?

At the beginning of the Xe transient on a power decrease following 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months at 100% power:

note: EXe3 denotes xenon concentration

a. Direct EXe] . increases, indirect EXe3 decreases, total EXe]' decreases.

b. Direct EXe3 increases, indirect'EXe] increases, total EXe] increases.

c. . Direct EXe] decreases, indirect EXe] decreases, total EXe] decreases.

d. Direct EXe] decreases, indirect EXe] increases, total EXe3 increases.

e. Direct EXe] decreases, indirect EXe] increases, total EXe3 decreases.

QUESTION 1.11 (1.00)

GIVEN: Two identical control rods, each absorb an equal amount of neutrons. The neutron flux at the center of the core equals that at the edge of the ccre. Why do the control rods in the middle of the core (radially) have a greater effect on Keff than the control rods at the edge of the core (radially).

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11__EBINGIE6ES_gE_NUG6E88_EgyEB_E68NI_gEEB811gN2 PAGE '5

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QUESTION 1.12 (1.00)

What effect does rod shadowing have on the worth of control rods?

DUESTION 1.13 (1.50)

On a reactor startup, what 3 conditions indicate the reactor is critical?

OUESTION 1.14 (1.00)

Give two reasons why.10 exp -8 amps is chosen as a standard reference for critical rod height data.

note: " standard reference" is NOT an acceptable answer

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- it__EBINGIELES_QE_NyGLE68_EQWEB_EL8MI_QEE8811QUt PAGE '6-ISE60QQYN@dlCS2_SE8I_IB8NSEEB_8NQ_ELylp_ELQW

' OUESTION 1.15 .(2.00)

Match the term in column A with the correct. definition in column B.

column A column B a) Specific Entropy 1) BTU /deg F b) DNBR 2) Ratio of local O to to CHF

>1.30 c) Quality 3s Internal energy of a substance

d. Enthalpy 4) % steam mass to total steam

& water macs

5) BTU / lbm-deg R

6) Ratio of critical O to local O

7) Internal Energy plus Flow Energy of a substance

8) % steam volume to total steam and water volume OUESTION 1.16 (1.00)

What effect does adding an 800 ci source yielding 1 X 10 exp 8 neutrons /sec have on the magnitude of Keff in a subtritical reactor?

NOTE: For simplicity assume the microscopic TOTAL cross section of the source equals zero,

a. increase t

b. decrease t'

c. no change

d. insufficient data

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n-1r__EBINCIE6ES_gE_UUC6E88_EgyEB_E68NI_QEgB8IlgN 2 PAGE' 7

-IUEBdgpyd@dICS2 _dESI_IB8NSEEB_8NQ_ELUlp_E698 OUESTION 1.17 (1.00)

Given a SUR of 0.1 dpm, determine the final power P in terms-of the initial power Po after O.1 hr. Show all work.

QUESTION 1.'18 (1.00)

Choose the best answer for the definition'of subcritical multiplication.

a. The process of utilizing source neutrons to sustain the chain reaction for Keff < 1.

b. The phenomenon where by source neutrons are used to measure the fractional curvature change of the flux for Keff < 1

c. The manipulation of neutron sources to sustain the chain reaction until Keff = 1.

d. The phenomenon where by source neutrons are used to stabilize reactor period /startup rate thus ensuring reactivity (rho) is << Beff for Keff < 1.

QUESTION 1.19 (1.50)

Calculate the heat transferred across one U-tube of a steam generator. Show all work.

GIVEN: (for simplicity)

U-tube heat transfer coefficient: 1.565 BTU /(sq ft-deg F)

U-tube height: 25 ft U-tube outer radius: 1/2 inch primary coolant temperature: 550 deg F secondary water temperature: 480 deg F QUESTION 1.20 (1.50)

List three things, that in practice, prevent water hammers from occurring

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22__E68NI_DgglgN_INg6UDINg_S8 Eely _8ND_Edg8GENCy_SYSIgd3 PAGE 9 QUESTION 2.01 (1.00)

Which one of-the following is NOT a source of water to the PRT?

a) Letdown Relief Valve b) RCP Seal Water return line relief valve c) Excess. letdown / loop drain header relief valve d) Reactor Vessel Flange Leakoff Detection Drain e) PDTT relief valve QUESTION 2.02 (1.00)

Which one statement below regarding the Source Range Nuclear Instrumentation System is INCORRECT.

a) P-6 allows the source range high level reactor trip signal to be bypassed manually when one of the two intermediate range instruments is above 10 E-10 ion chamber amps.

b) Placing BOTH source range blocking switches to the BLOCK position de-energi:es the high voltage supply to both source range instruments.

c) The source range high level trip is blocked when P-10 is present.

d) When P-6 is present and P-10 is not present, the source range high level trip is automatically reinstated and the source range high voltage re-energized when one of the two intermediate ranges is below P-6 reset.

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- J1t__EBINC1ELgg_OElNLJCLg68_ POWER PLANT OPERATION 2 -PAGE 8 IUEBdO9XN8dICS,_Hg81_IB8NSEgB_8ND_ELUlp_ELOW

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QUESTION 1.21' (2.00)

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

a. Enthalpy.

b. Pressure

c. Entropy

d. Temperature QUESTION 1.22 (1.50)

A motor driven centrifugal pump is operating at a low flow condition. You then start opening the throttle valve on the discharge side. How will each of the following be affected?

(INCREASE, DECREASE, or NO CHANGE)

a. Discharge Pressure

b. Available NPSH

c. Motor. Amps-4 QUESTION 1.23 (1.00) i Unit A is at EOL while Unit B has just been started up after a refueling.

Assuming a-rod speed of 48 spm, both reactors are taken critical by pulling 50 steps at a time, waiting until counts stabilize then pulling again.

Assuming all systems and parameters are identical at the commencement of the startup, and both units are initially shutdown by 2% (delta k/k):

a) Which Unit will have the highest source range counts when criticality is reached?

b) How will critical rod heights compare in the two Units?

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

'2 t__ELANI_DgSigN_ INCLUDING _S8EgIy_8ND_gdg8GENCY_SYSIgdS PAGE (p QUESTION 2.03 (1.00)

Which one of the following describes the method of NaOH solution addition to the Quench Spray System 7 a)- An eductor utili=ing OS pump discharge draws NaOH solution from the Chemical Addition Tank (CAT) into the QS pump output. '

b) Gravity feed-from the CAT to the RWST near where the OS pumps'take a suction.

c) . Gravity feed from the CAT to the area between the OS pump inlet isolation valve and the suction side of the pump.

d) :The CAT pump discahrges the contents of the tank into the OS pump suction with a pre-determined flow rate set by a manual throttle valve.

QUESTION 2.04 (1.00)

Which location below is the discharge point for the

. pressurizer head vent?

a) Containment fuel canal

b)- Upper region of containment below quench spray rings c) Pressurizer Relief Tank d) Suction side of containment Hydrogen Recombiners OUESTION 2.05 (1.00)

Which valve listed.below is used to throttle auxiliary spray flow?

a)_ FCV-122 (Charging Flow Control Valve) b) HCV-311 (Aux Spray Valve )

c) 'PCV-455B (Loop C Spray Valve) d) PCV-455A (Loop A Spray Valve)

L e) You cannot throttle auxiliary spray f (***** CATEGORY O2 CONTINUED ON NEXT PAGE *****)

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2=__ELGUI_DE@lGN_lGCLUDIN@_S8Egly_8ND_EMEBGENCY_SYSIEMS PAGE ll QUESTION 2.06 (f.zS) 01.M Match the following Pressurizer heater banks in Column A with their proper MCC in Column B.

COLUMN A COLUMN D A) Back-up Heaters

1) Group I (0.25) a) 1A1

2) Group II (0.25) b) 1B1

3) Group IV (0.25) c) 1C1

4) Group V (0.25) d) 1D1 D) Control bank heaters Group III (0.25) e) 1G1 f) 1H1 g) 1J1 QUESTION 2.07 (1.00)

A "High Containment Pressure" Automatic Safety Injection signal will: (Choose one) a) cause a main steam line isolation.

b) be initiated by 2/4 containment pressure instruments greater than 17 psig.

c) be blocked whenever the reactor trip breakers are open.

d) cause a feedwater isolation and a phase "A" isolation.

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2s__ELONI_ DESIGN _INGLUDINg_g8 Eely _8ND_EMEBGENgy_gygIEMS PAGEil QUESTION 2.08 (1.00)

Which of the following does the operator MANUALLY adjust to reduce the RCS temperature when the RHR system is in service for a normal plant cooldown, per OP 14.17 a) Throttle open CCW from RHR Heat Exchanger outlet isolation valve.

b) Throttle open RHR Heat Exchanger outlet isolation valve.

c) Throttle closed RHR Heat Exchanger bypass valve.

d) Throttle closed RHR mini-flow recirculation valve.

QUESTION 2.09 (1.00)

Listed below are valves associated with the Recirculation Spray (RS) System. Indicate whether each of the valves listed are NORMALLY OPEN or CLOSED.

a) MOV-SW-102A and B (Service Water supply header x-connects) b) MOV-SW-105A and B (Service Water B return header isolation valves) c) MOV-RS-101A (Casing Cooling Pump A FIRST discharge valve) d) MOV-RS-155B (Outside RS Pump suction valve)

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2___E60NI_QEg1GN_1NCLUDING_,58Egly_86Q_EUEBGENCy_@y@lgdS PAGE13-QUESTION 2.10 (1.00)

'In regards to the Chemical and Volume Control System (CVCS),

state what position (OPEN, CLOSED, AS IS) the following valves fail upon a loss of air.

.a) Letdown-isolation valves LCV-1460 A/B b) Orifice Isolation valves LCV 1200 A/B/C c) . Pure Grade water supply valve FCV-1114A d) Boric Acid supply to blender'FCV-1113A e) Emergency Dorate valve QUESTION 2.11 (1.50)

Indicate whether the following statements regarding RCP -

seals are TRUE or FALSE.

a) The-floating ring seal, will limit leakage to 50 gpm if the #1 seal fails.

b) o#3 seal is designed to withstand full RCS pressure.

c) Seal water injection from CVCS enters the RCP between

'the seal package and the pump radial bearing.

QUESTION 2.12 ( .50)

TRUE/ FALSE A RED urgent failure alarm light indicates that a major electrical failure has occurred in the logic cabinet.

QUESTION 2.13 (1.50)

The principle driving f orce f or PZR normal spray flow is the differential pressure between _________and the _ ____.

(***** CATEGORY O2 CONTINUED ON NEXT PAGE *****)

2- PL8dI_ DESIGN _ INCLUDING _SOEEly_8ND_ EMERGENCY _SYSIEMS' PAGE I4 QUESTION 2.14 (2.00)

List the 4 flow paths within the reactor vessel which BYPASS the fuel rods.

QUESTION 2.15 (1.50)

List the 3 independent sources of water to the Fire Main System.

QUESTION 2.16 (2.00)

LIST 4 of the 5 Design bases for the ECCS Cooling Performance following a LOCA as stated in 10CFR50.46.

QUESTION 2.17 (2.00)

List 5 parameters associated with the RCP's which are monitored after starting a RCP as stated in OP 5.2 "RCP Operation". Provide the required minimum values which must exist if applicable.

QUESTION 2.18 (1.50)

List 6 Emergency loads supplied by the Service Water System for a period of up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after a LOCA. Include those that require service water as a back up.(Sets are considered as one load)

QUESTION 2.19 (1.00)

Where is the source of power f or the automatic field flash of the Emergency Diesel Generators generated 7 i

a 1

(***** CATEGORY O2 CONTINUED ON NEXT PAGE *****)

2 1__E68NI_pgglGN_INCLUp1NG_gBEEIY_8Np_EUEBGENCY_SYSIEMS PAGE15I QUESTION 2.20 (o 75)

,1.50P Sodium Hydorxide (NaOH) added during the injection phase after a LOCA will eventually be distributed by the Quench Spray System and raises the Containment sump pH approximately 8. What are the two (2) reasons for establishing the elevated pH in the containment?

. QUESTION 2.21 (1.50)

State 3 reasons for having HCV-1142 (RHR letdown penetration from the RHR heat exchangers) kept about 10% open.

QUESTION 2.22 (1.50)

State two purposes for the interlock between the letdown isolation valves, LCV-1460A/B, and the orifice isolation

. valves, HCV-12OOA/B/C.

in QUESTION 2.23 W Concerning the Rod Control System:

Place the following components in their proper flow path d g,\ yf order. Start from the normal power supply and ending at the CRDM's

1) DC hold cabinet

2) Power Cabinet

3) Motor generator set

4) Reactor Trip breaker

5) Automatic Rod Control Unit

6) Rod Position Indication Cabinet

7) Logic Cabinet b) For the components in Part a, above, STATE the number of each present in the system.

(***** END OF CATEGORY O2 *****)

l ih _INSIBUMENIS_8ND_CgNIBOLS 'PAGE #

. QUESTION 3.01 (1.00)

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

a) Allows bypassing a steam dump cooldown interlock.

b) Allows operator block of SI signal.

c) Causes feedwater isolation if low Tavg is also present.

d) Causes a turbine trip.

QUESTION 3.02 (1.00)

Which of the following conditions is NOT required for automatic swapover of the LHSI pumps to the Recirculation Mode following a SI?

a) RWST Lo-Lo Level b) A LHSI pump recirc isolation MOV closed for each pump c) SI signal present d) SI Recirculation Mode signal present OUESTION 3.03 (1.50)

Concerning the Overtemperature Delta Temperature Setpoint (OTSP) describe how (increases, decreases or remains the same) each of the following parameter changes will effect the OTSP.

a) Increase in Tavg b) Decrease in Reactor Pressure c) Increase in Delta Flux Penalty i

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

3 2__INSIBUMENIS_8ND_CQNIBQL@ PAGE11.

QUESTION 3.04 (1.00)

Which statement below regarding the Main Generator Protection-System is INCORRECT.

a) A generator trip always results in a turbine trip when the generator is loaded.

b) Once the generator is loaded, a turbine trip always results in a generator trip, c) A turbine trip above the protection interlock P-7 (10% power) results in a Reactor trip.

d) A reactor trip always results in a turbine trip.

QUESTION 3.05 (1.25)

Describe how the High Steam Line Flow SI input varies and the parameter on which this program is taased.

QUESTION 3.06 (1.00)

Which statement below regarding pressurizer control is CORRECT 7 3

a) All /( channels provide input to the SI low pressure signal.

3 b) All A channels can be utilized ta control the operation of the spray valves.

3 c) All 4 channels send their signals through an Isolation Amplifier after supplying input to their respective protective circuit.

3 d) All >P channels can supply input to PORV Interlock circuitry to prevent PORV's lifting at low pressures.

(***** CATEGORY 03 CONT::NUED ON NEXT PAGE *****)

3:__INSIBUMENIS_AND_CONIBOLS _PAGElb QUESTION 3.07- (1.00)

Which of the following is NOT an input-into the OT Delta T trip point calculator?

.a) Power Range Nuclear Power b) RCS pressure c) Tavg d) AFD QUESTION 3.08 (1.00)'

Which of one the following statements describes the two Delta'T's. measured on the Core Cooling Monitor when the Loop-1' button is depressed?

'a) (Loop-A Th - Loop A Tc) and (Highest core thermocouple - Loop A Tc) b) (Loop A Th - Loop A Tc) and (Highest core thermocouple - Loop A Th) c) '(Average core thermocouple - Loop Th) and

~(Average core thermocouple - Loop A Tc) d) (Loop'A Th - Highest core thermocouple) and

-(Highest core thermocouple - Loop A Tc) e) (Loop A Th - Average core thermocouple) and (Loop a Th - Loop A Tc)

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

i

12__INgI8UdgyIS_6ND_CQNIBQL@ PAGEO '

OUESTION 3.09 (1.00)

With the pressuriner level control selector switch in position I/II, a failure causes the following plant events.

(Assume no operator actions taken.)

1) Charging flow reduced to minimum

2) Pressurizer level decreases

3) Letdown secured and heaters off

/

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

a) Level channel I failed high b) Level channel I failed low c) Level channel II failed high d) Level channel II failed low QUESTION 3.10 (1.00)

List the two RPS design conditions which necessitate the use of 2/4 Reactor trip protection logic vice 2/3 logic.

l l

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

r -

)

i l

3- INSTRUMENTS AND CONTROLS. PAGE 20 1

_i

-QUESTION 3.11 (1.00)

Which of the following would be the INITIAL response of the feedwater flow due to the response of the S/G Water Level Control System if the

~

steam pressure transmitter controlling the SGWLCS failed HIGH while at 50% power?

a.) Feed flow would INCREASE due to the maximum steam pressure input

-to the steam flow signal.

b.) Feed flow would INCREASE due to the level mismatch error between actual and programmed level caused by the pressure instrument failure c.) Feed flow would DECREASE due to the mismatch between steam and feed flow signals caused by the pressure instrument failure.

d.) Feed flow would REMAIN THE SAME due to the dominate of the level error signal over the flow error signal.

e.) Feed flow would REMAIN THE SAME as the steam pressure will not affect the steam flow signal.

l.00 QUESTION 3.12 N Indicate whether each of the statements below regarding the High Head Safety Injection System (HHSI) is TRUE or FALSE.

a) The alternate power source, J Bus, is ONLY used for maintenance on the "B" charging pump, and this pump has no automatic pump start capability when connected to the J bus.

b) Normal lead pumps during a SI are the "A" and "B" HHSI pumps, delltfdef All three pumps get a start signal from a SI g

signal, but the "A" pump is locked out to allow the "C" pump to start on its normal (H bus), if its breaker is racked out.

QUESTION 3.13 ( .50)

.TR1]E or FALSE Pulling the control power fuses when the Source Range level trip switch is in " Bypass" will cause a trip signal to occur.

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

M2__IU_SIBUDENTS AND_CQNTBQLS PAGE 11'

--c -

QUESTION 3.14- ( .50)

TRUE or FALSE Following a loss of offsite power, the. load shedding feature is actuated when the diesel output breaker closes.

OUESTION 3.15 (2.00)

. List 5 protection logic signals generated by the Pressurizer

. Protection System. _ (Include in your answer set points, coincidence and associated interlocks, if any)

-OUESTION 3.16 (1.50)

List 3 purposes of Rod Insertion Limits.

QUESTION 3.17 (1.00)

List the 4 requirements, control manipulations that will make up the logic to manually close the Diesel Generator output breakers (15H2).

QUESTION 3.18 (1.50)

List the 6 reactor trips which are enabled / blocked by the reactor trip system interlock P-7.

, QUESTION 3.19 (1.50)

List the THREE power supplies to the Vital 120 VAC distribution system and identify them by their priority.

QUESTION 3.20 (1.00)

List the TWO conditions that will provide signals to automatically open the ECCS accumulator discharge valves (1865 A/B/C).

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

32 __INSIBUMENIS_8ND_CONIBgLS PAGE Iti QUESTION 3.21 (1.00)

In regards to the PZR Pressure Control System, Unit 2 has an alarm, MANUAL NDT PROTECTION REQUIRED, which annunciates whenever temperature is <340 deg. F and pressure is > 550 psig. What action is required of the operator if such an alarm is received?

OUESTION 3.22 (1.00)

The Reactor breaker shunt trip coils have been modified to also energize upon any trip signal to the Undervoltage coils. What is the reason for this modification?

QUESTION 3.23 (1.00)

The Detector Current Comparator receives input from all 4 upper and lower power range detectors. How are these inputs compared, and what conditions are needed to auto bypass

. circuitry while at power?

I, 2f QUESTION 3.24 LLAfr1 a) What are the inputs to the Main Feedwater Bypass Valve controllers?

b) To place the Main Feedwater Bypass Valves in AUTOMATIC control while at low power, what controller-related conditions must be established?

QUESTION 3.25 (1.75) a) What consequences could be expected in the Rod Control System's DC Hold Cabinet if 2 or more groups of rod drive mechanisms were placed on hold power (excluding Control Bank D rods)? Explain you reasoning. (1.0) b) Why is there both a 125 VDC and a 70 VDC power supply in the DC Hold Cabinet?

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

c 5:__INSIBUdEUIS_8MD_CQUIBg6S PAGE c3-QUESTION 3.26 (1.50)

Sketch the rod speed program by indicating rod speed versus error signal.-

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

-4 1__E8QCEQUBES_;_NgBUG6t_GENgBd662_gdgBGENCy_6ND PAGE li B6 Dig 6gGIC66_CgNIBQL

& f .o O QUESTION M - ' 1. :,G ) -

Prior to operating Reactor Coolant Pumps in accordance with OP-5.2, Reactor Coolant Pump Operations, the minimum seal flow should be ____ gpm and VCT pressure should be a minimum of ____ psig.

1. O , 10

2. O.2 , 15

3. 2.0 , 30

4. 5.0 , 20 OUESTION 4.02 (1.00)

List FOUR indications of one dropped rod at 75% power.

QUESTION 4.03 (1.00)

What operator actions are required upon evacuating the control room if the reactor could not be tripped before exiting the control room?

QUESTION 4.04 (1.00)

Which of the following describes a temporary change which alters the INTENT of a procedure?

a. A change that corrects an incorrect valve lineup.

b. A change that modifies the criteria by which a system's operability is determined,

c. A change that allows partial use of a procedure to test a subtrain without affecting remaining equipment in that train.

d. A change that allows you to change incorrectly specified instruments for data taking.

l

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

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2 o, ATLANTA, GEORGIA 30323

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'4 __PROCEQURES 2 - NORMAL _ABrjgBM8L t _EMEBGENQY_ANQ t PAGE' If 88D1960GIC66_QQNISQL QUESTION- 4.05 (1.00)

If you are in a 100 mrad / hour gamma field'for 45 minutes, what is.your dose in MREM after 45 minutes?

a. 45

b. 75

c. 450

d. 750

, QUESTION 4.06 (1.00)

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

a. controlling turbine load, b, taking manual control of individual control rod banks.

c. taking manual control of individual control rod groups.

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

QUESTION 4.07 (2.00)

Prior to a reactor startup, with the RCS at normal operating pressure and temperature, the following RCS leakages exist. For each leak listed below, indicate whether you could STARTUP or would have to remain SHUTDOWN.

(Treat each leak below as an independent event) a) A leak f rom an unknown source of 1.5 GPM.

-b) 6.0 GPM_from a manual valve packing gland, c) 0.4 GPM from one S/G.

d) 0.1 GPH from the reactor vessel head INNER seal.

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

4 t__EROCEDURES - NORMAL _8BNQBMBL 1 _ sEMERGENCY _BNQ PAGE 16 08D1969GIC86_ggNIBQL t.

QUESTION 4.08 ( .50)

'You are releaseing radioactive liquid waste in accordance with 1-OP-22.11, Releasing Radioactive Li quid Waste, when orie of the operating circulating water-pumps trips. LYou may continue the release for up to 5 minutes while attempting to restart the pump. TRUE/ FALSE QUESTION 4.09 ( 1'. 00 )

List all conditions that require the Control Rod Drive Mechanism Shroud

. Cooling Fans to be in operation.

QUESTION 4.10 (1.50)

Match ~the action listed in Column A with the approximate power level in Column B at which this action is taken on a unit startup to 100% power.

COLUMN A COLUMN B

a. Place a second Main Feed pump in service 1) 15%

2) 30%

b. Stop increasing power and check for a 3) 50%

chemistry hold 4) 60%

5) 70%

c. Perform a calorimetric 6) 90%

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

m

42__EBgCEQUBES_ _NQBM86t_8@NQBM861_EMEBGENCY_8NQ PAGE t'7

bed 196901G86_GQNIBQL CUESTION .4.11 . ( 2:. 50) l Match the terms in column-A to the values in column B for the radiation exposure guidelines. Assume whole body dose unless otherwise stated.

CAUTION: Some answers could be used more than once. (0.5 ea)

COLUMN A- COLUMN B

a. NRC limits /qtr 1. 0.5 REM

b. ' Virginia. Power limits /qtr 2. 1.25 REM

c. . NRC' pregnant woman limit / gestation 3. 1.0 REM

d. NRC-general public limit / year 4. 0.75 REM

e. NRC quarterly limit with a Form 4 5. 5 REM

6. 3 REM

. QUESTION 4.12 (1.00)

List the 4. methods given in the S/G Tube Rupture EOP to identify which S/G is' ruptured.

QUESTION 4.13 (1.50)

Following a valid reactor trip and saf ety injection, what are the Reactor Coolant Pump Trip Criteria? (Assume normal containment conditions)

QUESTION 4.14 4(.00)/,2 9

~ List-the 4mmediate operator actions to initiate emergency baration if it is requiret on an Anticipated Transient Without Trip condition. Assume Safer Injection has not accuated and is not desired.

OUESTION 4.15 (1.50)

List the SI. termination criteria following a LOCA.(Include all appropriate values)

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

S2__EBgCEQQBgS_;_NQBd86t_8BNQBd862_gdgBGENQY_8ND PAGE 16 ~

86DI969GIC66_CQNIBg6 QUESTION 4.16 (1.00)

List the 4 DISTINCT hazards to which personnel are exposed when an entry into the reactor' compartment is made.during reactor operations.

QUESTION 4.17 (1.00)

' List four of the critical conditions required to be recorded during a etartup when 1 X 10E-8 amps is attained.

QUESTION 4.18 (1.00)

List ALL immediate operator actions required by 1-AP-14, Low Condensor

' Vacuum, if condensor vacuum lowers, but does not increase above 9.5" HG absolute.

QUESTION 4.19 (1.00)

List all of the immediate operator actions if a valid Reactor Coolant Pump Vibration DANGER Annunciator is received while at 30% power?

QUESTION 4.20 (2.50)

List FIVE indications of a loss of Component Cooling Water in accordance with AP-15, Loss of Component Cooling.

l l^

OUESTION 4.21 (3.00) 1 l List ALL'immediate actions required by 1-AP-14, Loss of Reactor Coolant System Pressure. List only those items that would bbe verified, not items from the " Response Not Obtained" category.

I l

QUESTION 4.22 (1.00)

Briefly explain the effect that placing an " unsaturated" mixed bed domineralizer in service will have on the reactor coolant system and on control of the reactor.

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

4 z__EBQgEDUBES_ _UQBM861_GENQBM961_EMEBGENgY_8ND PAGE 19 ,

BODI969 GIG 66_GOUIB06

{

l QUESTION 4.23 (1.00)

During a natural circulation cooldown, it is desired to cooldown using the steam dumps. Which MODE is the steam dump system operated in and WHY?

1 l

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4

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

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

1:__EB1hGIELESi9E_UUG6E88_EgWEB_EL8NI_gEgB8IIgN2 PAGE ~9

-IMEBUggyN8dlCS2_Ug81_IB8NSEEB_8UQ_ELylp_E6gW ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J ANSWER 1.01 (2.00)

A. It doubles (or increase) the head for a given mass flow rate.

B. It will double (or increase) the mass flow rate capacity for a given head.

REFERENCE Surry lesson plan ND-83-LP-8, Rev 1, p8.18; NA NCRODP-83 191004; K1.09/1.10(2.4/2.4)

ANSWER 1.02 (1.00)

C REFERENCE Surry lesson plan ND-83-LP-8, Rev 1; NA NCRODP-83 191004; K1.14(2.4)

ANSWER 1.03 (1.00)

With a DNDR of 1.3, during normal operation and anticipated operational occurrences, there is(a 95%) confidence that DN8 does not occur. When > 1.3 likelihood of DNB occurring decreases.

REFERENCE Surry lesson plan ND-86.3-LP-2, p2.10; NA NCRODP-83, ARR-13 193008; K1.10(2.9) i I

-1t__EBIBGIELES_RE_UUGLE68_EQWEB_EL6MI_QEEB6I1RNS PAGE 10 ISEBdQDYNBdlGS2_bE8I_IB8NSEEB_GND_ELUID_E6QW ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J ANSWER 1.04 (2.00)

1. reactor power

2. coolant flow rate __

3. RCS W temperature of d 4 RCS pressure.

REFERENCE Surry lesson plan ND-86.3-LP-2, p2.10; NA TS 2.1 193008; K1.05(3.4)

ANSWER 1.05 (2.00)

1. Density difference (or DELTA T) created by heat addition by the heat source and heat removal by the heat sink.

2. The heat sink must be elevated pisywically duuve Lise iveat source.

REFERENCE Surry lesson plan ND-86.3-LP-4, p4.5; NA NCRODP-83, ARR-12 193008; K1.21(3.9)

ANSWER 1.06 (1.00) e REFERENCE Surry lesson plan ND-86.3-LP-3, pp3.4, 3.5, 3.7, 3.10, 3.12 193009; K1.05(3.1)

REFERENCE NA TS 3.1 PWG-5(2.9/3.9) t l

[

L

s-ic__EBINCIE6EQ_QE_NQCLE88_EQWEB_E68NI_QEEB8IlgN 1 PAGE 11 IHEBdQQYN8dlCS 1 _Hg81_IB8NSEEB_8NQ_E6Q1Q_E6QW ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J ANSWER 1.07 ( .50)

Self shielding / Self shielding of the fuel pellet.

REFERENCE Surry lesson plan ND-86.2-LP-1, pl.7; NA NCRODP-86.1 192001; K1.08(2.3)

ANSWER 1.08 (1.50)

a. DECREASES b . -L E CitC A C E 5 MANU

c. DOES NOT CHANGE REFERENCE Surry lesson plan ND-86.2-LP-1, pl.4, 1.11; NA NCRODP-86.1 192004; K1.07(2.9)

ANSWER 1.09 (1.00) b REFERENCE Surry lesson plan ND-86.2-LP-1, pl.16; NA NCRODP-86.1 192004; K1.05(2.3)

ANSWER 1.10 (1.00) d REFERENCE Surry lesson plan ND-86.2-LP-4, p4.4, 4.8; NA NCRODP-86.1 192006; K1.06(3.4)

la__EBING1ELES_QE_UUGLd88 EQWEB_EL8UI_QBEB8IlQN S PAGE 12 ISEBdQQ188diQS 2_UE8I_IB8NSEEB_8NQ_E6UlQ_ELQW ANSWERS -- NORTH ANNA 18<2 -87/02/09-MOORMAN, J ANSWER 1.11 (1.00)

Neutrons at or near the edge of the core have a higher probability of leaking out than the ones at the center which

~

have a higher probability of causing fission. (Hence: DRW at center is > than at edge).

REFERENCE Surry lesson plan ND-86.2-LP-6, p6.12; NA NCRODP-86.1 192005; K1.14(3.2)

ANSWER 1.12 (1.00)

The presence of adjacent control rods may cause a significant change in an individual control rod worth.

REFERENCE

-Surry lesson plan ND-86.2-LP-6, p6.19; NA NCRODP-86.1 001/000; K5.05(3.5)

ANSWER 1.13 (1.50)

Start up rate is positive and constant, reactor power is increasing, and there is no outward rod motion.

REFERENCE Surry lesson plan ND-86.2-LP-7, p7.51; NA NCRODP-86.1 192008;K1.11(3.G)

la__EBINGIELES_QE_NQQLE88_EQWEB_E68NI_QBEB8IlgN 3 PAGE' 13 IUEBdQQYN8d1Ggi_ME8I_IB8NSEEB_8NQ_ELQ1p_E6QW ANSWERS -- NORTH ANNA 1842 -87/02/09-MOORMAN, J ANSWER' 1.14 (1.00)

(two of the three answers below required) i 1. Neutron production is relatively high, so power is constant when the reactor is critical.

2. Below 10 exp-8 amps the output of the intermediate range may not be directly proportional to the neutron population.

.3. Reactivity has not yet been changed by the moderator or fuel temperature.

REFERENCE Surry lesson plan ND-86.2-LP-7, p7.57; NA NCRODP-86.1 192008; K1.12(3.5)

ANSWER 1.15 (2.00) a) 5 b) 6 c) 4 d) 7 REFERENCE Surry lesson plan ND-83-LP-(1-10); NA NCRODP-83 193008; K1.10/K1.06(2.9/2.8) l ANSWER 1.16 (1.00)

C

{ REFERENCE Surry lesson plan ND-86.1-LP-6, p6.35; NA NCRODP-86.1 000/015; K5.06(3.4) k

li__EB10CIELEE_QE_MUGLE88_EQWE8_EL8MI_QEEE811QNt PAGE I4 IHEBMQQYN8dlG@x_dg8I_IB8NSEEB_8NQ_ELylp_ELgW ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J

-ANSWER 1.17 (1.00) 0.1hr = 6 min P = Pa 10 exp SUR(t) (+.5)

P = Pa 10 exp O.1 dpm(6 min) (+.25)

P = Pa 10 exp O.6 (+.25)

P = 3.98 Po REFERENCE Surry lesson plan ND-86.1-LP-8, pB.12; NA NCRODP-86.1 192003; K1.09(2.3)

ANSWER 1.18 (1.00) a se C.

REFERENCE Glasstone & Sesonske. Nuclear reactor engineering third ed. New York: Van Nostrand Reinhold Co., 1981.

192003; K1.01(2.7)

ANSWER 1.19 (1.50)

D = UA DELTA T (+.5)

A = 25' (2 pi r) (+.25)

DELTA T = 70 deg F (+.25) 0.5" = .042' (+.25) t O= 1.565 BTU /(sq ft-deg F) x 25ft x 2 pi r x 70 deg F (+.25)

O = 723 BTU REFERENCE Surry lesson plan ND-83-LP-1, p1.27; NA NCRODP-83 193007; K1.08(3.1) k

{

L

p'

-it__EBINQlELES_QE_UQQ6E88_EQWEB_E68NI_QEEB811QN1 PAGE 15

INEBdQQYN8MIQ@i_BE81_IB8NSEEB_8NQ_E6Qlp_E(QW ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J b ANSWER 1.20 (1.50)

1. Gradual warm up of steam lines

2. Proper venting of tanks and components during warm up and operation.

3. Steam traps

4. Lines kept full (Others as appropriate)

REFERENCE Surry lesson plan ND-83-LP-8, pB-36; NA NCRODP-83 193006; K1.04/1.10(3.4/3.3)

ANSWER 1.21 (2.00)

a. REMAIN THE SAME

b. DECREASE

c. INCREASE

d. DECREASE REFERENCE Surry lesson plan ND-83-LP-(1-10); steam tables; NA NCRODP-83 193004; K1.15(2.0)

ANSWER 1.22 (1.50)

a. Decrease

b. Decrease

c. Increase REFERENCE Surry lesson plan ND-83-LP-8; NA NCRODP-83 191004; K1.15(2.6) l I

' PAGE 16 iz_iE81NCIELES_QE_UUCbE8RPQWE8_EL8BI_QEE88I1QN 1 -

-IBEBdQDYN8d1GSz_Ug8I_IB8NSEEB_6NQ_ELylD_E(QW ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J ANSWER 1.23 (1.00) a) Wi11.be the same (+.5 ea) b) Unit B will be higher REFERENCE Westinghouse Reactor Core Control, pp 6-23/26 Westinghouse Fundamentals of Nuclear Reactor Theory, pp 8-48/60 001/010; K5.08(2.9/3.2) & 001/000: K1.05(4.5/4.4)

2c__E66NI_DE@lGN_INC6MDING_S8EEIy_6ND_EUEBGENCy_SYSIEUS PAGE 24 ANSWERS -- NORTH ANNA 1842 -87/02/09-MOORMAN, J ..

ANSWER 2.01. (1.00) d (1.0) ,

REFERENCE NA NCORDP 88.1 Reactor Coolant p 2.24 007/000; A3.01 (2.7/2.9)

ANSWER 2.02 (1.00) d (1.0)

REFERENCE NA NCROPD-77 RPS p 39 015/000; K4.01 (3.1/3.3)

ANSWER 2.03 (1.00) b 1.00 REFERENCE NA NCORDP 91.1 ESF-OSS O26/000; K4.01 (4.2/4.3) r ANSWER 2.04 (1.00) a aaaaaaaaaaaaaaaa REFERENCE NA NCRODP 88.1 "RCS-PZR and Press. Relief" 002/000; K4.03 (2.9/3.2) i ANSWER 2.05 (1.00) a (1.0)

REFERENCE NA NCRODP 93.8 PZR Press. Control and Protect.

010/000; A4.01 (3.7/3.5)

2i__E60NI_QEgl@N_lNCLUDINQ_@@EEIY_8Np_EUE8@gNCY_SYSIEUS PAGE 25

' ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J 2 5~

ANSWER 2.06 ' .(,. 5 0 ) -

A) 1) 1J1 (a) (0.25)

2) 181 (b) (0.25)

3) 1H1 (d) (0.25)

4) 1C1 (c) (0.25)

B) 1A1 (e) (0.25)

REFERENCE NA NCRODP 93.8 010/000 K2.01 (3.0/3.4)

ANSWER 2.07 (1.00) d REFERENCE NA NCRODP 91.1 ESF p.2.29 013/000 K1.01 (4.2/4.4)

ANSWER 2.08 (1.00) b bbbbbbbbbbbbbbbb REFERENCE NA NCRODP 88.2 RHR System 1-OP-14.1 ANSWER 2.09 (1.00) a) Open (0.25 ea) b) Closed c) Open d) Open REFERENCE NA NCRODP 91.1, "ESF-RSS" 026/000; K1.02 (4.1/4.1)

2i__ELBUI_DE@l@N_ INCLUDING _@8E@ly_8ND_Ed@BGENCY_@Y@l@d@ PAGE 26 ANSWERS -- NORTH ANNA 1842 -87/02/09-MOORMAN, J ANSWER 2.10 (1.00) a) Fails closed- (0.2 ea) b)- Fails closed c) Fails closed Fails open #'

Fails as is'[er his valv' 55 ^ "Og"" #f'*ggj go,_od ddnW h O d) e)

REFERENCE NA NCRODP 80.3 Chemical and Volume Control 004/000; A2.04 (3.6/4.2)

ANSWER 2.11 (1.50) a) True (0.5 ea) b) False c) False REFERENCE NA NCORDP 88.1 "RCS-RCP" 002/000; K1.13 (4.1/4.2)

ANSWER 2.12 ( .50)

TRUE' (0.5)

REFERENCE NA NCRODP 93.5 Rod Control 001/000; K4.03 (3.5/3.8)

ANSWER 2.13 ( 1. "a0 )

RCP discharge and the PZR or DP across the core, and water level in the PZR (0.75 ea)

REFERENCE NA NCRODP 08.1 RCS p. 2.9 002/000; K1.09 (4.1/4.1)

2 t__EL8MI_ DESIGN _INCLUDINQ_Q8Egl%_8NQ_EMESGENCY_@y@lgd@ PAGE. 27 ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J

. ANSWER 2.14 (2.00)

1) To upper head plenum via no::les in core barrel flange.

(0.5)

2) Between hot leg discharge nozzles and upper core barrel outlets. (0.5)

3) Between baffle plates and core barrel. (0.5)

4) Around inserts in guide thimble tubes in the. fuel assemblies. (0.5)

REFERENCE ND 80.1-LP-2 p. 2.32 002/000; A1.05 (3.4/3.7)

ANSWER 2.15 -(1.50)

-Service Water Reservoir (0.5 ea)

-Lake Anna

- -Discharge Canal or BloMerl'04K REFERENCE NA NCRODP 89.4, Feedwater Systems-AFW 086/000; K1.03 (3.4/3.5)

ANSWER 2.16 (2.00)

(any 4 of 5 at 0.5 ea)

1) Max. Fuel Element Cladding Temp. < 2200 Deg. F

2) Cladding Oxidation < 17% thickness

3) Hydrogen generated by Zirc-Water reaction <1% of max.

possible.

4) Core remains in a coolable geometry

5) Provides for long term decay heat removal i REFERENCE 10CFR50.46 NA NCRODP 91.9 ESF p.2.6/2.7

! 006/050; PWG 4 (4.2/4.3) i f

i

! i

2t__EL6NI_DE@lQU_INGLyQ1NQ_@@EEIY_6ND_EME8QEUQY_SygIEMS PAGE 28 ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J ANSWER 2.17 (2.00)

1) Motor current (0.25 ea)

2) Bearing temperature

3) Seal injection flow, 6-/ gpm

4) ' Seal leak off flow, 0.2-g gpm

5) Seal differential pressure > or = 200 psid REFERENCE NA NCROPD 80.1 p. 3.22 OP 5.2 p. 8/9 003/000; PWG-7 (3.5/3.9)

ANSWER 2.10 (1.50)

1) 4 Recirculation spray heat exchangers (0.25 ea)

2) Dack-up for containmont Recirculation air coolers (if needed)

3) Charging pumps oil

4) Charging pumps water coolers

5) One compressor per unit

6) One control room air conditioner system per unit REFERENCE NA NCRODP 92.2 Service Water 076/000; K1.19 (3.6/3.7)

ANSWER 2.19 (1.00)

From its own (0.5) 125 VDC Olutribution system (0.5)

REFERENCE NA NCRODP 90.4 Print ESIC 11C sh 6 064/000; K1.04 (3.6/3.9)

2 i__E68NI_Dgg1GN_ INCLUDING _S@EEIy_@ND_EMEBGENCY_SYSIEd$ PAGE 29 ANSWERS -- NORTH ANNA 1 E<2 -87/02/09-MOORMAN, J (0.75)

ANSWER 2.20 41.307 il Tu ,c.-.c c mdicactive i uJ i .m f , v.. , thc c c r,t s i r,m u nit a4-merspWGP e . - (C . 7C '

2) To control the pH of the water that collects in the containment sump. XA basic pH helps to prevent Chloride stress corrosion.X (0.75)

REFERENCE NA NCRODP 91.1 ESF-Quench Spray System 026/000; K4.02 (3.1/3.6)

ANSWER 2.21 (1.50)

To provide a path to keep the RHR system full (0.50) and to allow for expansion of the system during heat up of the RCS (0.5) and thus ambiently heating up RHR (0.50). or .4o cod in WareM}vrIhr OtR. y @ m (,5) er 4 {cr overp<tsiut prciadfor o q bt- (ii)

REFERENCE NA NCRODP 88.2 RHR 004/000; K1.01 (3.4/3.9)

ANSWER 2.22 (1.50)

1) Provents shocking the regenerative heat exchanger and the orifices (0.75)

2) Keeps the regenerative heat exchanger and associated piping pressurized to prevent flashing (0.75)

REFERENCE NA NCRODP 88.3 CVCS p. 6 004/020; K4.03 (3.0/3.4) 004/020; K6.12 (2.9/3.1)

2 t __EL6NI_DEglGN_INGLUDING_@@EEIY_6ND_EMERGENGY_SYSlEd5 PAGE 30 ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J

, 7 5' ANSWER 2.23 ' 1. 5 0 F-

,af' (3) motor generator set b) 2 reactor trip breaker 2 f; gjL (4)(2) power cabinet 4 (7) logic cabinet 1 (6) rod position indication cabinet 4 (5) automatic rod control unit 1 (1) DC hold cabinet 1 40.-75 f er -a) fully correct, 0.75 fcr b) fully correct,

-^1 _

f s,- caui, switt' needed te-p4-ee e a c ainp uiient in nennnr nederr REFERENCE NA NCRODP 93.5 Rod Control System 001/000 K4.01 (3.5/3.8) l

3 t__INSIByMENIS_AND_CQUIBOLS. PAGE 31' 1 ANSWERS 1- NORTH ANNA 1&2 -87/02/09-MOORMAN, J 4

ANSWER 3.01 (1.00) a

. REFERENCE NA NCRODP 93.1 012/000;.K6.10 (3.3/3.5)

ANSWER 3.02 (1.00)

C CCCCCCCCCCCCCCCC REFERENCE NA NCRODP.91.1 "ESF" 005/000; K4.11 (3.5/3.9)

ANSWER 3.03 (1.50) a) STSP decreases- (0.5 EA) b) STSP decreases C) STSP decreases REFERENCE NA NCRODP 77.RPS p 25 012/000; A1.01 (2.9/3.4)

ANSWER 3.04 (1.00) a (1.0)

REFERENCE 1 NA NCRODP 93.9 Main Generator Control & Protection 045/010; K1.11 (3.6/3.7)

ANSWER 3.05 (1.25) 40% setpoint from 0-20% (0.5) Turbine power (0.25) and linearly from 40-110% as Turbine Power goes from 20-100%

(0.5) 8 L

3t__10SIBUMENIQ_8UD_CQNIBQLS PAGE 32 ANSWERS -- NORTH ANNA i t<2 -87/02/09-MOORMAN, J REFERENCE NA NCRODP 91.1 "ESF-SI of ECCS" 013/000; K1.01 (4.2/4.4)

ANSWER 3.06 (1.00) a or C REFERENCE Westinghouse PWR Systems Manual, Sect 10.2 PZR Pressure Control 010/000; K4.03 (3.8/4.1) & K6.01 (2.7/3.1) t< PWG-4 (3.6/3.7)

ANSWER 3.07 (1.00) a (1.0)

REFERENCE NA NCRODP 93.10 RPS 012/OOO;K6.11 (2.9/2.9) & A2.05 (3.1/3.2)

ANSWER 3.08 (1.00) c (1.0)

REFERENCE NA NCRODP 93.4 " CORE COOLING MONITOR" NA NCRODP 93.4 Learning Objective; Section I, 2.4 ANSWER 3.09 (1.00) a (1.0)

REFERENCE NA NCRODP 93.8 PZR Level Protection and Control 011/000 A2.10 (3.4/3.6)

I l

m-i 3.t__INSIBUMENI@_8ND_GQNIBOLS PAGE 33 ANSWERS -- NORTH ANNA i t<2 -87/02/09-MOORMAN, J ANSWER 3.10 (1.00)

1) Trips not backed up by another protection circuit.

2) The channel is also being used f or control purposes.

REFERENCE NA NCRODP 93.10 RPS 012/000; K4.09 (2.8/3.1)

ANSWER 3.11 (1.00) a REFERENCE Westinghouse PWR System Manual, "SGWLC" 035/010; A2.03 (3.4/3.6) t ,0 0 ANSWER 3.12 -t 1. CO M a) TRUE. (0.5 EA) b) FALSE delgkd _ c) TOUC-REFERENCE NA NCRODP 91.1 p.2.18 013/000; K1.11 (3.3/3.8)

ANSWER 3.13 ( .50)

TRUE REFERENCE NA NCRODP 93.2 Excore Instrumentation System 015/000 K1.01 (4.1/4.2)

ANSWER 3.14 ( .50)

FALSE

3 2__IN$189MENIQ_8ND_CQNIBQLS- PAGE 34 ANSWERS -- NORTH ANNA 18<2 -87/02/09-MOORMAN, J REFERENCE NA NCRODP 90.4 p2.45 062/000 K1.02 (4.1/4.4)

ANSWER 3.15 (2.00)

1) PZR Hi Press. Trip (0.2) 2385 psig(0.1), 2/3(0.1)

2) PZR Lo Press. Trip (0.2) 1870 psig(0.1), 2/3(0.1)

3) PZR Lo-La Press. SI(0.2) 1765 psig and not blocked (0.1),

2/3(0.1)

4) P-11(0.2)- <2000 psig(0.1), 2/3(0.1)

5) Press. input to the OT Delta T(0.4)

REFERENCE NA NCRODP 93.8 010/000; K1.01 (3.9/4.1)

ANSWER 3.16 (1.50) 4dcon**j^

1) Compensate f or power def ect or q0 gidlotn (0.5 #$M'""

ea) 6b

2) To minimize the amount of positive reactivity inserted during a rod ejection accident, and

3) To minimize radial flux tilt (peaking) or -jo maigIqi,s occeg bble power REFERENCE NA NCRODP 77 RPS 001/000; K5.04 (4.3/4.7)

ANSWER 3.17 (1.00)

1) Control switch to close (0.25)

2) Synchronizing selector switch is ON (0.25)

3) DG terminal voltage is 95% (0.25)

4) Breaker 86 and 87 protective relay's are reset (0.25)

(,66- h a<rarn~<~+ 0 7- Tus, s. h,.,wa )

REFERENCE NA NCRODP 90.4 EDG p2.57 064/000; A4.01 (4.0/4.3) i i

3 2__INSIBUMENI@_8ND_CQNIBQLS PAGE 35 ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J ANSWER 3.18 (1.50)

-PZR high water level (0.25 ea)

-PZR la pressure

-Lo primary coolant flow

-RCP breakers open (two pumps)

-Under voltage on both 4 KV buses

-Turbipe trip

- Un deriv 9eacy m beth %KV to nV5 REFERENCE NA NCRODP 93.10 012/000; K4.06 (3.2/3.5)

ANSWER 3.19 (1.50)

-Normal: " S '- V ^,C ' 'i t d (0.5 ea) (lf V dc V ! D*5 4 I2 f V 'I'/ l?O v Ac. Clnlic iaw odktd ]Lu .dby: 125-120 VOC Octtery W v g meu ency pno l (_hu, VAC bucq W m de vaA) y e

tJL)-7 ydO/I20 4 c. Un"5 [c"a fr don 34ty51F'CO3MC'e:

REFERENCE NA NCRODP.90.3 Vital and Emergency Distribution 062/000; K4.09 (2,4/2.9)

ANSWER 3.20 (1.00)

1) RCS pressure >2000 psig

2) SI REFERENCE NA NCRODP 91.1 p2.15 006/000 A3.01 (4.0/39)

ANSWER 3.21 (1.00)

The operator must manually (0.5) initiate over pressure protection (0.5). ( Manually open PZR PORV's)

REFERENCE NA NCRODP 93.8 PZR Press. Control and Protection 010/000; K4.03 (3.8/4.0)

u s

3z__INSIBydENIS_6ND_CONIBOL@ PAGE 36 ANSWERS -- NORTH ANNA 1&2- -87/02/uy-NOORMAN, J ANSWER 3.22 (1.00)

The design change resulted because of experiences where the q undervoltage trip signal alone was not sufficient to trip the breaker. (1.0)

REFERENCE NA NCRODP 77 RPS p.35 012/000; K6.03 (3.1/3.5)

ANSWER 3.23 (1.00)

The highest reading upper / lower detector is compared to the average of the upper / lower detectors (0.5). The circuit auto defeats below 50% power on ALL channels (0.5).

REFERENCE NA NCRODP 93.2 Excore Instrumentation Sys.

015/000; K6.04 (3.1/3.2) & A1.04 (3.5/3.7)

I.25 ANSWER 3.24 L1-rGES a) Inputs are the S/G 1evel instruments (0.25), Turbine first stage pressure (0.25) /l/u c /Mr .143 /n4,= r,th l// V'O (0,25-)

b) The output and demand signals must be approximately 0 (0.5)

REFERENCE NA NCRODP'93.12 SGWLC OP 31.0 MAIN FEEDWATER 059/000; A4.08 (3.0/2.9)

3?. IN@IBUMENI@iANQ_QQNT8QL@ PAGE 37

' ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J ANSWER 3.25 (1.75) a)- Cabinet has the capacity to support up to 6 stationary gripper coils simultaneously (0.5). So with 2 groups or-more, would overload / heat the cabinet (0.5).

b) 125 VDC-Latching Rods 70 VDC-Holding Rods (0.5 for reasons, 0.25'for correctly associating voltages)

REFERENCE NA NCRODP.93.5 Rod Control 001/050; PWG-1(3.6/4.1)

ANSWER 3.26 (1.50)

See attached sketch REFERENCE NA NCRODP 93.5 4 001'/000; K4.03 (3.5/3.8) f

r 3.2(o I.so (RRODP.93. 5/7- 2. 4 -

Eco speso Roos "Ou r *

(1O 72 - .

70 - -

to - -

so- -

10 -

kbero2n no- -.-.,.s i Jo- - g I l T4v4 > Tegy l 20 - i d3 - -

6 ' 65) 5 t

e J, 2 I/

, I i

r 4 ' ' ' 72me,

' ' ' l c' Rect if y

1.10 --a 2 5 h J-

,zgy

&CK.itP TR&v 7 TAV4

,. 20 (Re s e r) l ocxu? -

30 (kG5ET) a -go -.

.--so

-60

- - 70

~

~72 (oLS) bos la u ,,

A) TROL hod bpGEO N hpEerruce Eratoc.

( AMTONIATIL Foo Cea reou)

I

F St__E8QGEQUEEQ_ _UQBM8Ls_GENQBd86t_EME8@ENQY_8NQ PAGE 8 88DIOLOGIG86_G9 NIB 06 ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J ANSWER <1 At ' i . : :4 2 GL sM LtVLOT x} Q jn /19Y /

hff REFERENCE VCS, SOP-101 p1 NA OP-5.2 p4 SUR OP-5.2 p 2,4 ANSWER 4.02 (1.00) four 3 0.25 points each:

1. Rod bottom light

2. Computer alarm, power range tilt, rod deviation / sequence

3. Flux deviation alarm (s)

4. Rapid drop in Tavg and power level

5. Rapid drop in przr level and pressure

6. Power range negative rate alarm REFERENCE NAPS 1-AP-1.4, p.3.

001/050 PWG-10 4.3/4.5 ANSWER 4.03 (1.00) 3 0.5 points each:

1. Trip turbine locally.

2. Manually open reactor trip breakers or the rod drive MG output breakers.

REFERENCE NAPS 1-AP-20, p.3.

.SUR 1-AP-20, p5

p f

ds__BBQGEQUBES_ _NQBd@bt_@RUQBd@bt_EMEBGENGy_ANQ PAGE 9 l 86 Dig 6QQ1G86_GQUIBQL ANSWERS . -- NORTH ANNA it<2 -87/02/09-MOORMAN, J

' ANSWER 4.04 (1.00) b REFERENCE-NA ADM 5.8, pp 2/3

~-Sur SUADM-ADM-21 p 21 PWG-23: Plant Staffing and Activities (2.8/3.5)

' ANSWER 4.05 (1.00) b.

OF=1 for gamma 100 (45/60) (1 ) =75 REFERENCE 10 CFR 20.

-PWG-15: Radcon Knowledge (3.4/3.9)

TANSWER 4.06 (1.00) a REFERENCE MNS, AP/2/A/5500/14, Case I, p.2.

CAT, AP/1/A/5500/15, Case I, p.2.

Surry AP-1.OO pp 2,3 NA, AP-1.0 p4 001/050; PWG-11(4.4/4.4)

ANSWER 4.07 (2.00) a) Shutdown (+.5 ea) b) Startup c) Shutdown d) Startup L_

4; -PROCEDURES - NORMALt.AENQBMAbs_EMEB@ENGy_ANQ PAGE 10 B6D196001G86_GQUIBOL

. ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J REFERENCE

' SON TS 3.4.6.2-j -NA'TS 3.4.6.2 SUR TS 3.1 002/020; PWG-8 (3.5/4.4)

ANSWER 4. 08. ( .50)

-FALSE.

' REFERENCE NA 1-OP-22.11,'p 4 ANSWER 4.09 (1.00)

Whenever-CRDM's are. energized.

or-if CRDM's are energized when primary plant temp is 100F-to 350F:

REFERENCE.

~ NA OP-21.1 p4 _ NA 1-OP-1.2 p5 SUR OP-21.3 . p .9 ANSWER 4.10 -(1.50) a) -4 (+.5 ea)

-b). 2

c). 6 (5 for Surry)

REFERENCE NA OP-2.1, pp 9-13 Surry OP-2.1.1 pp 14-19 PWG-12: Perform-Integrated Plant ops (3.5/3.4)

4 .- PROCEDURES - NORMAL t_@BNQBMAbi_gMEBQENCY_8ND PAGE- 11

- 1BGDIQLQQ1C66_CQNIBQL

~

I ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J.

r LANSWER 4.11 (2.50)

.a- 2 b- 4 c l'

'd. .1

.e 6 REFERENCE

-Virginai Power GET pp 14-17 PWG-15 3.4/3.9 ANSWER 4.12 (1.00)

-Unexpected rise in S/G 1evel (+.25 ea)

-High radiation on a S/G blowdown line

-High radiation on an MS line monitor

-High radiation as determined by sampling and analysis REFERENCE Surry EP-4.00, pp 2 NA 2-EP-3, pp 2 EPE-038; EA2.03 (4.4/4.6)

ANSWER 4.13 (1.50)

.1). Verify Charging /SI flow (+.5 ea) ( 1,1 or 3)

2) RCS Pressure < 1230 psig

3) If component cooling water to any pump is lost REFERENCE-

'SQNP Foldout Page NA Foldout page for 2-EP-O Surry' Foldout page'for EP-1.OO 003/000; PWG-10 (4.1/4.4)

4t-_E80CEQUBgg_n_UQBd86t_GBNQBMGLt_EMEBGENCY_GNQ PAGE 12 88DIRLQGIGGL_QQNIBQL ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J ANSWER

( lilf )

4.14 Liv {+ttr Surry (+.25 ea) North Anna (+.25 ea)

Verify SI/CHG pumps running / flow 1. Verify 2 SI/CHG pumps running / flow Check RCS pressure <2335 psig 2. Switch BATP to fast speed Switch BATP to fast speed 3. Open MOV 2350 or --In; c c t thu DIT-Open MOV-( )350 4. Check p r press <2335 REFERENCE S' "d #' ' E' Surry FRP-S.1 p3 NA FRP-S.1 p4 EPE-029; PWG-11 (4.5/4.7)

ANSWER 4.15~ (1.50)

! NORTH ANNA (+.3 ea)

-RCS Press > 2000 psig & increasing l-RCS Subcooling > 50 Deg F

-PZR Level > 50%

l-SG Level > 10% or > 30% Advrs Cntm

OR l-AFW Flow > 730 GPM REFERENCE NA EP-O Faldout page 006/050; PWG-7 (3.8/4.2)

. ANSWER 4.16 (1.00) icinizing radiation; heat stress; differential pressure; O2 deficiency

(+.25 ea)

REFERENCE Surry SUADMO-19 p 3

.NA ADM 20.9, pp 1 PWG-18: Knowledge of Safety Procedures (3.0/3.1)

7- _.

r;

'3hIPROCEDURES - NORMALi_8BL49BM86t_EMEgggNQY_AND' PAGE 13

.609196991G86_ggNIBg6' ANSWERS - NORTH ANNA 18<2 - -87/02/09-MOORMAN, J

ANSWER _ '4 . 17 ('1. 00 )

'l

.Any 4'O O.25 points each:

North _ Anna Surry 1.-Bank::C position.. 1. Date Critical

12. Bank D position. 2. Time Critical

3. Auct. High Tavg. 3. Average RCS temp.-

4.cIR N35. 4. RSC Baron concentration 5.1.IR N36. 5. Bank C Control rod position

6. RCS baron concentration. 6. Bank D Control rod position

7. Actual critical position within admin. requirements

REFERENCE

' NAPS 1.OP-1.5, p.12.

SUR 1-OP-1C'_-App. A p 10 of 10 001.010; K5.08 (2.9/3.3) fANSWER 4.18 (1.00)

1. . Reduce generator load until vacuum stabilizen 2.' Check' vacuum breaker (MOV-AS-100) closed (.25) and a water' seal present _ (.25)

REFERENCE NA .1- AP-14,_p 3 i

ANSWERI 4.19 (1.00)

Manually trip the-reactor (.50) and the affected-RCP (.50)

REFERENCE NA-AP-29, p2

~

000/015- PWG-10 4.2/4.5

E

4. -PRQC DURg@_ _NQRMALi_8DNQBd8Lt_EMEB@ENGY_8NQ PAGE 14-

80 Dig 69@lG86_QQN1896-

/ ANSWERS.-- NORTH ANNA 1&2 -87/02/09-MOORMAN,-J (ANSWER 4.20 (2.50)

CC' surge tank low level' alarm

'CC pump auto trip alarm CCW low flow discharge header alarm CCW low pressure discharge header alarm Reactor coolant pump low flow /high temp alarm

--Excess letdown HX. low flow /high temp Non-regenerative HX high temp

. REFERENCE tW4 AP-15, p2 008/030' PWG-10-3.8/4.2.

ANSWER- 4.21 (3.00)

1. Verify pzr porv's closed 2.: . Verify master controller PC-1-444J not f ailed

3. Verify p r spray valves closed 4.; Verify aux spray valve closed 5.. Verify all-p r heaters on 6 .~ Verify RCS pressure stable or increasing REFERENCE NA AP-44 p3

-ANSWER 4.22 (1.00)

An unsaturated mixed bed demineralizer will remove baron from the reactor coolant system (.50) and add positive reactivity (.5)

(Reasonable wording accepted)

REFERENCE

.VCS, sop-102 p i FW4 OP-8.2, p5 SUR OP'8.2,-p 2

.004/000 K6.02 2.5/2.1 I

. . = . -

g- ,

- 4 .' PROCEDURES - NORMAL _AENQBMA63._Et]EB@ENQY_ONQ t PAGE '15 80010600lGOL_GQNIEQL

ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J ANSWER 4.23 (1.00)

Steam pressure-mode EO.253 Tavg. input to the steam' dump control is not valid without forced flow in the loops. E.753 or TAv mode to%ot y usta b ootlown detoa Syy (no go.jf,g .

REFERENCE 3 'If 'i*

NA'1-AP-10, Att. 2, p2 6

'SUR AP-39, p4

n.

1 U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: NORTH-ANNA 1&2 REACTOR TYPE: PWR-WEC3 DATE ADMINISTERED: 87/02/09 EXAMINER: MOORMAN. J.

CANDIDATE: MASTER INSTRUCTIONS TO CANDIDATE:

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

Staple question sheet on top of the answer . sheets. Points for each

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

% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY thy 30.00 4NLAN4 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 29 0 14 f 30.0C -25.00- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 30,tD t%:r LM 7

-30.00 25.00- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL ts. H 30.00 -25.00- 8. ADMINISTRATIVE PROCEDURES, ug,ty- CONDITIONS, AND LIMITATIONS i Cn uT i -120.00 % Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.

Candidate's Signature i

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

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

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

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

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

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

6. Use only the paper provided for answers.

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

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

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

10. Skip at least three lines between each answer.

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

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

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

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

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

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

- I

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

a. Assemble your examination as follows:

(1) Exam questions on top. -

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

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

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

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

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

)

52__IUEgBy_gE_N9CLE88_EgWEB_ELONI_gEEBBIlgN2_ELU1gS2_8ND PAGE 2

'IBEBdggyNBUICS QUESTION 5.01 ( 1. 50) -

With respect to reactor thermal limits, indicate whether each of the following statements are TRUE or FAUSE.

a'. The average _ linear power density in the core is expressed in units of kw/ft and is the total thermal power divided by the active length of all the fuel rods.

b. The purpose of limiting the enthalpy rise hot channel factor is to prevent bulk boiling from taking place during a LOCA.

c. The purpose of the limit on the heat flux hot channel factor is to insure that fuel clad temperature does not exceed 2200 deg F during normal operations.

QUESTION 5.02 (1.00)

Concerning subcritical multiplication, which one of the following statements is NOT correct?

a. The neutron behavior per generation can be stated mathematically.

b. The neutron population will reach and maintain an equilibrium value.

- c. The. fuel in the core effectively multiplies the source neutrons.

d. As the source strength is increased, the magnitude of Keff is increased.

(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)

"~

Dr THEOBy_DE_UUCLE8B_EOWEB_ELONI_QEEBSIlgN 2 _ELUIDS,_8ND PAGE 3" ISEBM9DXUOMlG5 QUESTION 5.03 (1.00)

Given: Three reactor coolant (RCP) pumps operating in parallel, each with a flow rate "m" and a combined flow rate i "M"..Out of the four. possibilities below, choose the one that best fits if one RCP is secured.

a. The resulting core flow (M) will' increase.

b. The resulting core flow-(M) will increase along with individual operating RCP flow (m).

c. The resulting core flow (M) will decrease as individual 1 operating RCP flow (m) increases.

d. The resulting core flow (M) will not change due to decrease in RCP_back pressure.

QUESTION 5.04 (1.00)

The negative reactivity added when fuel temperature increases is primarily caused by __.

a. depletion of U-238

b. doppler broadening

c. depletion of U-235

d. fuel pellet swell thus decreasing the gap

(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)

~~'~ 5 2__IHggBY_gE_NUg6E88_EgWEB_E68NI_gEEB811gN3 _E6UlpS 2_8ND PAGE '41 IUEBdgDYN8 digs e

QUESTION 5.05 ' ( 1. 00) .

Which one of the following statements below is NOT correct regarding senon beh"avior following a power increase?-

note: EXe3 denotes xenon 135 concentration

.a. The' minimum EXe] reached-is independent of the magnitude of the power level-increase and initial power level.

b. . The time .to reach equilibrium is also dependent on the magnitude of the power change and final power level.

c. The time to reach the minimum EXe3 is always < 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months .

d. The time to reach equilibrium is approximately 40-50 hours.

QUESTION 5.06 (1.50)

Indicate whether each of the following will make the moderator temperature coefficient less negative, more negative, or have no effect.

a. increase temperature

b. decrease baron concentration

c. increase core age OUESTION 5.07 (1.50)

Write on your answer sheet INCREASES , DECREASES or DOES NOT CHANGE for the following:

The magnitude of the fuel temperature coefficient (FTC):

A. INCREASES / DECREASES / DOES NOT CHANGE.with increase in power.

B. INCREASES / DECREASES / DOES NOT CHANGE with core age.

C. INCREASES / DECREASES / DOES NOT CHANGE with decrease in moderator temperature coefficient (MTC).

(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)

~

~75t__IugQBy_QE_NUGLE88_EQWEB_EL8NI_QEE8811QUt_ELUIDet_8ND PAGE 5.

IBEBdQQyd8dlGS QUESTION 5.08 (l.00)

A centrifugal pump is started up with its discharge valve open. How would the following parameters differ (INCREASE, DECREASE, or REMAIN THE SAME) if the pump was started with its discharge valve shut?

a. Motor current

b. Discharge pressure QUESTION 5.09 (1.50)

Nuclear reactors are initially loaded with more f uel than is required to bring the reactor critical. The additional fissile raaterial in the core is said to represent built in or excess reactivity. List 3 things that excess reactivity is designed to overcome.

QUESTION 5.10 (2.00)

List the four (4) plant parameters observed to insure that CHF'or DNBR are not exceeded.

t QUESTION 5.11 (1.50)

On a reactor startup, what 3 conditions indicate the reactor is critical?

OUESTION 5.12 (1.50)

List three. things, in practice, that prevent water hammers

+

from occurring 4

OUESTION 5.13 (1.00)

What effect does rod shadowing have on the worth of control rods?

(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)

l L- _

[ D___IMEgBy_gE_UUg6E88_EgWEB_E68NI_gEEB911gN2 _E6UIQS2_6ND

IHEBdggyN901CS PAGE 6' QUESTION 5.14 (1.50)

Attached is a typical boiling curve for water as it approaches, then exceeds,.the DNB point. What are the thermodynamic conditions that cause:

a) The decrease in heat transfer rate in Region III?

b) The increase in heat transfer rate in Region IV?

OUESTION 5.15 (2.00)

A. How does Beff vary over the life of the core?

D. How is Deff affected as plutonium isotopes are produced over the life of the core?

C. How is reactor response af f ected by a lower delayed neutron fraction?

QUESTION 5.16 (2.00)

Given two pumps of equivalent design, operating at the same, constant speed:

A. What will be the effect of placing the two pumps in

. series (with respect to flow and head)?

B. What will be the effect of placing the two pumps in parallel (with respect to flow and head)?

OUESTION 5.17 (1.00)

What is the design basis of having a DNBR > or = to 1.37 OUESTION 5.18 (2.00)

What are all the conditions that must be present in order for natural circulation to exist?

(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)

5 __IUEgBy_gE_i_ LUC [EdR_EOWER_EhdNI_gEEBOIlgN,_ELUlpS1 _6ND PAGE 7 ISEBdggyN8dlCS QUESTION 5.19 (1.00)

GIVEN: Two identical control rods, each absorb an equal amount of neutrons. The neutron flux at the center of the core equals that at the edge of the core. Why do the control rods in the middle of the core (radially) have a greater effect on Keff then the control rods at the edge of the core (radially).

QUESTION 5.20 (1.00)

Give two reasons why 10 exp -8 amps is chosen as a standard reference for critical rod height data, note: " standard reference" is NOT an acceptable answer OljESTION 5.21 (2.50)

What are the purposes of each of the following reactor thermal limits? If a specific accident or condition applies, state this in your answer.

a. Reactor safety limits (1.0)

b. Enthalpy rise hot channel factor (Fn(delta H)) (0.5)

c. Nuclear flux hot channel factor (Fq (z ) ) (1.0)

(***** END OF CATEGORY 05 *****)

GuESTiora 5.m

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.s I

I

4--- I ; C

, j IIdd- M ; 2 E ?

SINGLE- PH ASE INUCLEATE PARTIAL FILM BOILING

$ CONVECTION I BOLLING FILM o f B0iLING 104 ,

1

/

6- /

, ,/. . . . .

103 100 got go2 go3 go4 e

,- bT (*F) --*

(TEMPERATURE DIFFL?tNCE BETWEEN FUEL ROD SURFACE AND SATURATION TEMPER ATURE OF THE COOL ANT)

FIGURE FND-HT-102: BOILING CURVE AND DNB AT VARIOUS PRESSURES (REV.1) ..

cues 7 son 5._1 4 . _ _ _ _ _ - . . - _ _ . _

62__PL6NI_SY@IEUS_Dgg1GU3_CQNIBQL 2 _6ND_INSIBUMENI6IlgN PAGE 9 DUESTION 6.01 (2.00)

LIST 4 of the 5 Design bases for the ECCS Cooling Performance following a LOCA as stated in 10CFR50.46.

QUESTION 6.02 (1.00)

Which valve listed below is used to throttle auxiliary spray flow?

a) FCV-122 (Charging Flow Control Valve) b) HCV-311 (Aux Spray Valve )

c) PCV-455D (Loop C Spray Valve) d) PCV-455A (Loop A Spray Valve) e) You cannot throttle auxiliary spray QUESTION 6.03 (1.50)

State 3 reasonc for having HCV-1142 (RHR letdown penetration from the RHR heat exchangers) kept about 10% open?

DUESTION 6.04 (1.50)

State two purposes for the interlock between the letdown isolation valves, LCV-1460A/B, and the orifice isolation valves, HCV-12OOA/B/C.

a

(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

L 6:__EL6NI_SYSIEMS_ DESIGN2 _CONIBg62_8ND_INSIBUDENIGIlON PAGE lo 1 QUESTION 6.05 (1.00)

Which statement below regarding the Source Range Nuclear Instrumentation System is INCORRECT.

a) P-6 allows the source range high level reactor trip signal to be bypassed manually when one of the two intermediate range instruments is above 10 E-10 ion chamber amps.

b) Placing BOTH source range blocking switches-to the BLOCK position de-energizes the high voltage supply to both source range instruments.

c) The source range high level trip is blocked when P-10 is present.

d) When P-6 is present and P-10 is not present, the source rance high level trip is automatically reinstated and the source range high voltage re-energized when one of the two intermediate ranges is below P-6 reset.

QUESTION 6.06 ( .50)

TRUE/ FALSE A RED urgent failure alarm light indicates that a major electrical failure has occurred in the logic cabinet.

.7f OUESTION 6.07 U 4rtf7 Concerning the Rod Control System:

$0 f Place the following components in their proper flow path order. Start from the normal power supply and ending at the CRDM's

1) DC hold cabinet

2) Power cabinet

3) Motor generator set

4) Reactor Trip Breaker

5) Automatic Rod Control Unit

6) Rod Position Indication Cabinet

7) Logic Cabinet b) For the components in Part a, above, STATE the number of each present in the system.

(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

6:__PLONI_SYSIEUS_ DESIGN2 _CgNIBOL3 _8ND_IN@IBudENI6110N PAGE il QUESTION 6.08 (1.50)

List the design bases for the minimum level requirements of the Emergency Condensate Storage Tank.

QUESTION 6.09 (1.00)

Which of the following is NOT a design basis of the Steam Dump System?

a) Accommodate ramp load increases greater than 10%/ minute.

b) Pass 40% steam flow on a 50% turbine step rejection without a reactor trip occurring.

c) Allow a turbine trip and a subsequent reactor trip from 100% power without lifting the S/G code safety valves.

d) Allow for a smooth shift of plant steam load from the steam dump system to the turbine on a plant startup.

OUESTION 6.10-(l. 76)

-( 1. L01-Match the RCS penetrations in Column A with the appropriate RCS loop segment listed in Column D. (Answers may be used more than once)

Column A Column B a) Excess Letdown 1) Loop A cold leg b) PZR Surge Line 2) Loop A hot leg Nf fj er CVCS Alternate Charging 3) Loop A intermediato leg d) PZR Spray Line 4) Loop B intermediate leg e) RHR Suction 5) Loop C hot leg

6) Loop C cold leg

7) Loop C intermediate leg

(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

- ~

62__PLONI_SYSIEdg_QESIGN2 _CQNIBgt2 _8NQ_INSIBUDENIBIIgN PAGE 11 e

QUESTION 6. 2,1 (2.00) ,,

,~ ; '*

While performing maintenance, it has been determined that the B"charg'ing pump must be tagged out, and the control power fuses for the B charging pump must be removed.

i What two manipulations must be done to prevent the letdown orifice isolhtion valves from closing?

QUESTION 6.12 (2.00)

List 5 protection logic signals generated by the Pressurizer Protection System. (Include in your answer set points, coincidence;nnd associated interlocks, if any) ,

. \

I QUESTION 6.13 (1.00) \

The Reactor trip breaker shunt trip coils nave been modified to-also energize upon any trip signal to the Undervoltage coils. What is the reason for this modification?

OUESTION 6.14 (1.50)

List 3 purposes of Rod Insertion Limits.

QUESTION 6.15 (1.50)

~

Crjncerning the Overtemperature Delta Temperature Setpoint (OTSP) describe how (increases, decreases or remains the same) each of the following parameter changes will effect the OTSP.

a) Increase in Tave

' r'

'b) Decrease in Reactor Pressure f

c)- Increade in Delta Flux Penalty

,7

, Y (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

u

6i__E68NI_SYSIEUS_ DESIGN1 _GONIBQL2_8NQ_INSIBUdENI8IlgN PAGE 13 OUESTION 6.16 (1.00)

Which statement below regarding the Main Generator Protection System is INCORRECT.

a) To prevent a turbine overspeed event, a generator trip always results in a turbine trip when the generator is loaded.

b) Once the generator is loaded, a turbine trip always results in a generator trip.

c) A turbine trip above the protection interlock P-7 (10% power) results in a Reactor trip.

d)- A reactor trip always results in a turbine trip.

1,00 OUESTION 6.17 fi . 50) -

Indicate whether each of the statements below regarding the High Head Safety Injection System (HHSI) is TRUE or FALSE.

a) The alternate power source, J Bus, is ONLY used for maintenance on the "B" char g pump, and this pump has no automatic pump start tapability when connected to the J Bus.

b) Normal lead pumps during a SI are the "A" and "B" HHSI pumps.

jef All three pumps get a start signal from a SI signal, but the "A" pump is locked out to allow the "C" pump to start on its normal (H bus), if its breaker is racked out.

(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

1 l

62 __eLeNI_SygIgdS_DgSIgN2 _CgNIBg62_8ND_INSIBUdENIGIIgN PAGE I .4 OUESYION 6.18 (1.50)

Indicate whether each of the statements below regarding permissive functions associated with the Excore Nuclear Instrumentation is TRUE or FALSE.

a) In order for the P-7 permissive (At Power Trips) to be DISABLED, both reactor power permissive P-10 and turbine power permissive P-13 must clear.

b) The single loop loss-of-flow reactor trip is one of the trips ENABLED by the P-7 permissive.

c) When actuated, the P-10 permissive will automatically DE-ENERGIZE the high voltage to the Source Range Instrument, but it will NOT RE-ENERGIZE the SR Instrument high voltage when P-10 clears.

QUESTION 6.19 (1.25)

Describe how the High Steam Line Flow SI input varies and

.the parameter on which this program is based.

QUESTION 6.20 (1.00)

List the 4 requirements, control manipulations that will make up the logic to manually close the Diesel Generator output breakers (15H2).

QUESTION- 6.21 (1.00)

The Detector Current Comparator receives input from all 4 upper and lower power range detectors. How are these inputs compared, and what conditions are needed to auto bypass circuitry while at power?

(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

l 62__EL6NI_SygIEdS_DE@lGN2_GQNIB063_6NQ_INSIBydENI611gN PAGE -15 QUESTION 6.22 (1.75) a) What consequences could be expected in the Rod Control System's DC Hold Cabinet if 2 or more groups of rod drive mechanisms were placed on hold power (excluding

' control Bank D rods)?

b) _ Why is there both a 125 VDC and a 70 VDC power supply in the DC Hold Cabinet?

(***** END OF CATEGORY 06 *****)

i I

L

Z __PBQCEQUBES_ ,NgBMOL1_GENgBdGL1_EMEEGENCY_8ND PAGE 36 68DIO6901C86_ggNIBg6 QUESTION 7."1 M (1.OC,

( b O O)

Prior to operating Reactor Coolant Pumps in accordance with OP-5.2, Reactor Coolant Pump Operations, the minimum seal flow should be ____ gpm and VCT pressure should be a minimum of ____ psig.

1. O , 10

2. O.2 , 15

3. 2.0 , 30

4. 5.0 , 20 QUESTION 7.02 (1.00)

What operator actions are required upon evacuating the control room if the reactor could not be tripped before exiting the control room?

QUESTION 7.03 (1.00)

Which of the following describes a temporary change which alters the INTENT of a procedure?

a. A change that corrects an incorrect valve lineup.

b. A change that modifies the criteria by which a system's operability is determined.

c. A change that allows partial use of a procedure to test a subtrain without affecting remaining equipment in that train.

d. A change that allows you to change incorrectly specified instruments for data taking.

I

(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

t

I JZs__EBOGEQUBES_ _NQBd86t_8BNQBd@Lt_EdESGENQY_8NQ PAGE Q

88 Dig (QGlG86_QQNIBQL Q'UESTION .7,04 (1.00)

If you are-in a 100 mrad / hour gamma field for 45 minutes, what is your dose in MREM after 45 minutes?

a. 45

b. 75

c. 450

d. 750 QUESTION 7.05 (1.00)

One of the source range channels fails on a reactor startup just above the point where P-6 is actuated. Which one statement below describes the correct action (s) that should be taken by the operator 7

a. Insert the control banks to the fully inserted position - and repair the source range instrument before increasing power above P-6 again,

b. Continue with the reactor startup.

c.- Insert control banks until below P-6 , then repair the malfunctioning source range channel before continueing with the startup.

d. _ Borate the RCS to the shutdown margin requirements of the applicable Technical Specifications section.

(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

r.

Zz__EBOGERUBES_ _NQBd8Ls_GBNQBd8Lt_EdEBGENQY_8ND PAGE 16 88D1969EIG86_GgNIBQL-

-QUESTION 7.06 '(1.00)

A-hydrogen bubble formed in the reactor vessel is eliminated by

a. increasing pressurizer temperature above core thermocouple readings.

b. injecting oxygen into the reactor coolant system via the chemical and volume control system.

c. maximizing. coolant flow by running all reactor coolant pumps, increasing letdown flow to 120 gpm, and placing the cation bed demineralizer in service in parrallel with the mixed

' bed demineralizer.

d. venting the reactor vessel head.

QUESTION 7.07- (1.50)

For each of the following, indicate YES or NO if the conditions violate critical safety function (CSF) red path criteria.

a) Pressurizer level of 5% and RVLIS upper head 80%

b) Total AFW flow 400 gpm with all S/G levels < 6%

c) Containment pressure 65 psig' QUESTION 7.08 (1.50)

Answer the following questions regarding EOP usage TRUE or FALSE:

i a) If a Function Restoration Procedure (FRP) is entered due to an ORANGE Critical Safety Function (CSF) condition, and a HIGHER priority ORANGE condition is encountered, the original FRP must be completed prior to proceeding to the newly identified FRP.

b) Unless specified, a task need not be fully completed before proceeding to a subsequent step as long as that task is progressing satisfactorily l c) If a procedure transition occurs, any tasks still in progress from the procedure which was in effect need not be completed.

(*****-CATEGORY 07 CONTINUED ON NEXT PAGE *****)

Z _ EBgggDUBES_ _NQBM86t_8BNQBM86t_EME8GENGY_8NQ PAGE 19 68D1960 GIG 86 G9 BIB 96 QUESTION 7.09 (2.00)

Prior to a reactor startup, with the RCS at normal operating pressure and temperature, the following RCS leakages exist. For each leak listed below,

' indicate whether.you could STARTUP or would have to remain SHUTDOWN.

(Treat each leak below as an independent event)-

a) A leak from an unknown source of 1.5 GPM.

b) 6.0 GI3 from a manual valve packing gland.

c) 0.4 GPM from one S/G.

d) 0.1 GPH from the reactor vessel head INNER seal.

QUESTION 7.10 ( .50)

'You are releaseing radioactive liquid waste in accordance with 1-OP-22.11, Releasing Radioactive Liquid Waste, when one of the operating circulating water pumps trips. You may continue the release for up to 5 minutes while attempting to restart the pump. TRUE/ FALSE QUESTION 7.11 (2.50)'

Match the terms in column A to the values in column B for the radiation exposure guidelines. Assume whole body dose unless otherwise stated.

CAUTION: Some answers could be used more than once. (0.5 ea)

COLUMN A COLUMN B

a. NRC limits /qtr 1. 0.5 REM

b. Virginia Power limits /qtr 2. 1.25 REM

c. NRC pregnant woman limit / gestation 3. 1.0 REM

d. NRC general public limit / year 4. 0.75 REM

e. NRC quarterly limit with a Form 4 5. 5 REM

6. 3 REM

(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

L

'Zz__EBOCEDUBES_:_NgBd(ks_QBNg8 dolt _EMEBGENQy_6NQ PAGE to 8001060GIC06_GQNIBQL QUESTION 7.12 (1.00)

-List the 4 methods given in the S/G Tube Rupture EOP to identify which S/G is ruptured.

OUESTION 7.13 (1.50)

Following a valid reactor trip and safety injection, what are the Reactor Coolant Pump Trip Criteria? (Assume normal containment conditions)

(l . 2f) '

QUESTION 7.14 -t i . : :"

List the immediate operator actions to initiate emergency bcration if it is required on an Anticipated Transient Without Trip _ondition. Assume Safety Inj ecti on has not accuated and is not desired.

QUESTION 7.15 (1.50)

List the SI termination criteria following a LOCA.(Include all appropriate values)

DUESTION 7.16 (1.00)

List the 4 DISTINCT hazards to which personnel are exposed when an entry into the reactor compartment is made during reactor operations.

QUESTION 7.17 (1.00)

List four of the critical conditions required to be recorded during a startup when 1 X 10E-8 amps is attained.

QUESTION 7.18 (1.00)

List ALL immediate operator actions required by 1-AP-14, Low Condensor Vacuum, if condensor vacuum lowers, but does not increase above 9.5" HO absolute.

(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

i L7- ' PROCEDURES - NORMALt _ADNQBMAL1 _EMEBGENgy_AND PAGE Li B891969 GIG 86_GONIBgL

' QUESTION 7.19 (1.00)

List all of the immediate operator actions if a valid Reactor Coolant Pump Vibration DANGER Annunciator is received while at 30% power?

QUESTION 7.20 (1.75)

List the make up flow paths to the Refueling Cavity, in the order of preference, for a loss of refueling cavity level per AP-52, Loss of -

Refueling Cavity Level During Refueling.

(ie. Containment sump via LHSI pump to refueling cavity)

.36ea, 2P- fcr r_.n c c t m Ja OUESTION 7.21 (2.50)

List FIVE indications of a loss of Component Cooling Water in accordance with AP-15, Loss of Component Cooling.

QUESTION 7.22 ( .75)

What constitutes a Class II reactor trip?

QUESTION 7.23 (1.00)

During normal operations, why is Oxygen concentration in the VCT limited to less than 5% by volume?

QUESTION 7.24 (1.00)

During a natural circulation cooldown, it is desired to cooldown using the steam dumps. Which MODE is the steam dump system operated in and WHY7

(***** END OF CATEGORY 07 *****)

1 92__6DDINISIB8IIVE_PBggEDUBE@2_ggNplIlgNS3 _BND_61dII@IlgbS PAGE 2L QUESTION 8.01 (1.00)

'The Unit i reactor coolant system pressure exceeds 2735 psig when in

-mode 3. According to technical specifications, pressure must be restored within acceptable limits within what time frame given below?

a. 5 minutes.

b. 15 minutes.

c. 30 minutes.

d. one hour.

QUESTION 8.02 -(1.00)

If control power is lost to a Unit 2 pressurizer power' operated relief valve while in mode 1, which statement below is correct?

a. Tech specs require no action'provided another PORV is operable and all pressurizer code safety valves are operable.

b. lech specs require the power supply to be removed from the associated block valve after verifying it to be open, if the PORV is not operable within i hour and contnuous operation is desired.

c. Tech specs require the associated block valve to be shut and its power removed if the PORV is not made operable within one hour and continuous operation is desirable.

d. Tech specs require action to be initiated within one hour to place the plant in at least hot standby within the following hour if the PORV is not made operable.

(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)

8- eod1NISIBellyE_BBQCEDQBES2_QQUDlIlgNS _6ND_L1011811QUS 1 PAGE 13

~

OUESTION 8.03 (1.00)-

A Unit 2' control rod is determined to be INOPERABLE in mode 2 as a result of excessive friction. Tech Specs' require which action below in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ?

a.- -]Be in hot standby.

b. Restore the rod to operable status.

c. Position the remainder of'the rods in that group to within

"+" or "" 12 steps of the inoperable rod.

d. Determine that the tech spec shutdown margin requirement is satisfied.

QUESTION 8.04 (1.00)

According to Tech Specs, which of the following is the correct action to be taken if the Radwaste Effluent Monitoring Line Process Monitor is out of service?

a. Effluent releases cannot be performed until the Monitor is back in service,

b. Effluent releases may be performed if Grab Samples are analyzed every twelve hours during the release.

c. Effluent releases may be performed provided two samples taken prior to the release are analyzed and do not exceed 10CFR2O limits and two qualified staff members verify the release rate calculations and the discharge valve lineup,

d. The effluent release may be performed provided a sample prior to the release indicates that the Lower Limit of Detection (LLD) is not exceeded for all the analyses required and subsequent hourly samples during the release confirm this condition continues to exist.

f

! (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)

L

ei__8Dd1NigIB8Ilyg_BBggEQUBES 2 _CQNp111gNS 2 _8Np_L10118IlgNS PAGE L4 OUESTION 8.05 (1.00)

Which of the following require activation of both the TSC and OSC7

a. Either an unusual event, alert, site area emergency or general emergency.

b. Only an alert, site area emergency, or general emergency.

c. Only a site area emergency or general emergency.

d. Only a general emergency.

QUESTION B.06 (1.00)

Which one of the following statements is correct regarding the control and issuance of Special Order Tags (Blue tags)?

a. These tags may be used by all departments except Health Physics

b. These tags may be used in lieu of a mechanical danger tag.

c. The Control Room Operator may authori=e tag removal.

d. The tag indicates who must be contacted to operate the equipment.

QUESTION 8.07 (1.00)

Answer TRUE or FALSE to the following:

a) IF a component's emergency power supply is INOPERABLE but all other supporting equipment for that component is OPERABLE, than surveillance requirements on that component must still be performed within the proper time frame.

b) If it is required by an LCO Action Statement to be in HOT STANDBY in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and then HnT SHUTDOWN in the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , it is permissable to be in HOT STANDuY in 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months then use the next 9 to be in HOT SHUTDOWN.

(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)

E 8 2 __6QMINISIB611VE_BBQCEQQBES _CQNQlllQNS 2 _6NQ_

1 LIM 1I611QNg PAGE 15 OUESTION 8.08 (2.50)

Use the attached Technical Specifications to determine the correct response to the questions below regarding Nuclear Instrumentation.

a) What is the MAXIMUM # of each NI that can be out,of service at any time without requiring action to reduce the plant operating mode?

Assume you are in Mode 1 at 75% power.

b) You are at the minimum # of operable Power Range NIs, when an IC tech requests permission to put an operable PR NI in test for a channel functional test. Can this be done? Refer to applicable TS in answer. (

QUESTION 8.09 (1.50)

List 3 additional administrative precautions that must be met to enter a Locked High Radiation Area (> 1r/hr) that are not required for entry into [

a High Radiation Area (< 1000 mr/hr).

QUESTION 8.10 (1.00) a) During a non-emergency situation, who must authorize a temporary change to a operating procedure which does not change the procedural intent?

b) What 2 forms are temporary changes and permanent changes to procedures documented on?

OUESTION 8.11 (1.00)

List 3 of the 4 pieces of information that the. Shift Supervisor must obtain for transmittal to the appropriate medical facility prior to transporting a contaminated, injured worker uff-site, IAW EPIP 5.01.

QUESTION 8.12 (1.00)

While in a ref ueling mode of operation, with A RHR pump in operation circulating reactor coolant, and the normal power supply to the J Dus out of service due to maintenance, it is discovered that the EDG supply to the H bus has a malfunctioning air distributor making it INOPERABLE and that the water level above the reactor vessel flange has dropped below 23 feet. What action, if any, is required? Use the attached Technical Specifications and identify those which are applicable.

(***** CATEGORY 08 CONTINUED ON NEXT PAGE ****+-)

0 __6Dd1NISIB8IlyE_BBQQEQUBEgi_QQNDillgNg1_8ND_LidlI8IlgNS PAGE. 16 QUESTION 8.13 (1.50)

List the support equipment in TS 3.8.1, "AC Sources", required for a Diesel Generator to be considered' operable (there are 3 different-criteria that must be met).

QUESTION 8.14 (1.50)

List five hard copy sources of information that are referred to when performing a post trip review, following an unplanned reactor trip.

QUESTION 8.15 '(1.00)

List the four conditions, as stated in ADM-20.9, " Containment Egress and Ingress", that will initiate containment' evacuation.

QUESTION 8.16 (2.00)

What are the 4 conditions listed in the EPIPs that dictate when updates should be given to offsite authorities regarding an emergency, subsequent to the initial 15 minute notification?

QUESTION 8.17 (1.50) a) What are the two emergency exposure limits addressed in EPIP 5.06,

" Emergency Radiation Exposure Authorization", for damage control considerations. (0,5) b) List 4 criteria that should be considered by the Station Emergency Manager in selecting personnel for emergency radiation exposure (1.0)

QUESTION 8.18 (1.00)

What is meant by the statement in Technical Specifications that says "The provisions of Specification 3.0.4 are not applicable"?

(***** CATEGORY 00 CONTINUED ON NEXT PAGE *****)

u_. . .. . . . .

o "Dz__6Dd1NISIB6IIME_EBgCEgyBES2_GQNplIlgNS 2 _SUp_LIMIISIlgdg PAGE Z7 OUESTION 8.19 (1.00)

Why, in the attached Tech Spec 3/4.4.8, is it a requirement to cooldown to less than 500 degrees as an action if the RCS activity limits are exceeded?

QUESTION 8.20 (1.00)

As stated in 10CFR50.54, under what conditions may actions be taken that depart from a license condition or a technical specification, and who, as a minimum, must approve such action?

QUESTION 8.21 (2.00) a) State one of the two accidents specified in the bases for TS 3.4.9.3

" Overpressure Protection Systems" that having~two operable PORVs when less than 340 deg F in a cold leg is supposed to sufficiently protect against.

b) Explain why the TS for the OPMS require that Pressurizer level.be no more than 457 cubic feet when Tavg is between 320 and 340 degrees F.

QUESTION 8.22 (1.50) a) What are the three criteria which define an Excluded Radiation Worker? Include any applicable dose limits.

b) Assuming an Excluded Radiation Worker is a station employee, what entry conditions, if any, are required to enter a Radiation Area or a Restricted Control Area.

QUESTION 8.23 (1.00)

What is different in the forms and procedures that are required when performing a simple jumper installation (e.g. lifted lead) as opposed to a jumper installation requiring multiple steps? Assume that no approved procedure exists in either case initially.

i i

f (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)

83 __8pMINISIB811ME_669CEgyBEg2_CONDIIIQNg3_8ND_LIMIIBIIONS PAGE 18 OUESTION 8.24 (1.00)

If is is determined, OUTSIDE of normal working hours, that an Emergency Work Order is necessary, what two personnel shall be notified by the Shift Supervisor and what two forms need to be processed / initiated prior to commencement of work?

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

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

2-1-85 O

3/4 LIMITING CONDITIO!!S FOR OPERATION AND SURVEILLANCE REQUIREMENTS

-3/4.0 APPLICA8ILITY LIMITING CONDITION FOR OPERATION

! 3.0.1 Limiting Conditlons for Operation and ACTION requirements shall be l i

applicable during the OPERATIONAL MODES or other conditions specified for each l Specification.

3.0.2 Adherence to the requirements of the Limiting Condition for Operation and/or associated ACTION within the specified time interval shall constitute compliance with the Specification. In the event the Limiting Condition for Operation is restored prior to expiration of the specified time interval,

completion of the ACTION statement is not required,.

3.0.3 When a Limiting Condition for Operation i's not met, except as provided in the associated ACTION requirements, within one hour ACTION shall be initiated to place the unit in a MODE in which the Specifica-tion does not apply by placing it, as applicable, in

i 4 1. At least HOT STAND 8Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

2. At least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and g 3. At least COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Where corrective measures are completed that permit operation under the ACTION l requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual l

Specifications. This specif.ication is not applicable in MODES 5 or 6.

l l

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l 11-26-77 I

3/4.3 INSTRUMENTATION

-3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION

(-

LIMITING CONDITION FOR OPERATION s

3.3.1.1 As a minimum, the reactor trip system insbrusentation channels and interlocks of Table .3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2.

APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLANCE REQUIREMENTS

~

4.3.1.1.1 Each reactor trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-1.

4.3.1.1.2 The logic for the interlocks shall be demonstrated OPERABLE prior to each reactor startup unless performed during the preceeding 92 days. The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall.be demonstrated to be within its limit at least once per 18 months.

Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once' every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the " Total No. of Channels" column of Table 3.3.1.

e NORTH ANNA - UNIT 1 3/4 3-1

TABLE 3.3-1 15 3

=

REACTOR TRIP SYSTEM INSTRUMENTATION

!5 MINIMUM l: TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES A_CTION f

1. Manual Reactor Trip 2 1 2 1, 2 and

12 i p .

l 2. Power Range, Neutron Flux 4 2 3 1, 2 2 1

) 3. Power Range, Ne'utron Flux 4 2 3 1, 2 2 j High Positive Rate i

! 4. Power Range, Neutron Flux 4 2 3 1, 2 2 i

High Negative Rate j

d 5. Intermediate Range, Neutron Flux 2 1 2 1, 2 and

3 i"

!Y 6. Source Range, Neutron Flux gg j" A. Startup 2 1 2 2 4 B. Shutdown 2 1 2 3*, 4* and 5* 15 C. Shutdown 2 0 1 3, 4 and 5 5

]

i

7. Overtemperature AT Three Loop Operation 3 2 2 1, 2 7#

, bro Loop Operation 3 1** 2 1, 2 9

) ,

a Ib Q

a

?

13 ,

t i

6-9-86 TABLE 3.3-1 (Continued)

TABLE NOTATION With the reactor trip system breakers in the closed position and the control rod drive system capable of rod withdrawal.

l' The channel (s) associated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped condition.

With the Reactor Trip Breaker open for surveillance testing in accordance i j with Specification Table 4.3-1 (item 21A). l The provisions of Specification 3.0.4 are not applicable.

High voltage to detector may'be de-energized above P-6.

ACTION STATEMENTS i'

ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1.1 l provided the other channel is operable. .

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and POWER OPERATION may proceed provided the following conditions are satisfied:

i a. The inoperable channel is placed in the tripped condition within I hour.

b. The Minimum Channels OPERABLE requirement is met; however,

the inoperable channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months f

for surveillance testing of the redundant channel (s) per i Specification 4.3.1.1.1.

^

c. Either, THERMAL POWER is restricted to 5 75% of RATED THERMAL POWER and the Power Range, Neutron Flux trip

~ set reduced to s 85% of RATED THERMAL POWER within Jg ,

,4 the QUADRANT POWER TILT RATIO is monitored at

'leasF6 per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

d. The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75 percent of RATED THERMAL l POWER with one Power Range Channel inoperable by using the moveable incore detectors to confirm that the normalized symmetric power distribution, obtained from 2 sets of 4 symmetric thimble locations or a full-core flux map, is consistent with the indicated QUADRANT POWER TILT RATIO at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 3 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level NORTH ANNA - UNIT 1 3/4 3-5 Amendment No. 81

11-26-77 TABLE 3.3-1 (Continued)

a. Below P-6, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint.

. =-

b. Above P-6 but below 5% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 5% of RATED THERMAL POWER.

c. Above 5% of RATED THERMAL POWER, POWER OPERATION any continue.

ACTION 4 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL 4

POWER level:

a. Below P-6, restore the inoperable cb nnel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint. -

l b. Above P-6, operatiog may continue. ,

I ACTION 5 - With the number of channels OPERABLE one less than required by

.' the Miniaua Channels OPERABLE requirement, verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per i 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

! ACTION 6 - Not applicable.

2 i'

ACTION 7 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and POWER OPERATION may proceed until performance of the next required CHANNEL FUNCTIONAL TEST provided the inoperable channel is placed in the tripped

. condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 8 - Not applicable

(

e 4

NORTH ANNA - UNIT 1 3/4 3-6 .

1 1

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6-2-81 REFUELING OPERATIONS-RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION 28 ALL WATER LEVELS L

. z-LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal loop shall be in operation. l 28 APPLICABILITY: MODE 6. .

ACTION:

a. With less than one residual heat removal loop in operation, except as provided in b. below, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,

b. The residual heat removal loop may be removed from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs.

~

c. The provisions of Specification 3.0.3 are not applicable.

SURVFILLANCE REQUIREMENTS 4.9.8.1 A residual heat removal loop shall be determined to be in operation and circulating reactor coolant at a flow rate of greater than or equal to 28 3000 gpm at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

NORTH ANNA - UNIT 1 3/4 9-8 Amendment No. 32 L

6-2-81 REFUELING OPERA 1 IONS LOW WATER LEVEL LIMITINGCONDITIONFOROPERITION 3.9.8.2 Two independent Residual Heat Removal (RHR) loops shall be OPERA 8LE.*

APPLICA8ILITY: MODE 6 when the water level above the top of the reactor pressure vessel flange is less than 23 feet.

ACTION:

a. With less than the required RHR loops OPERA 8LE, immediately initiate 28 corrective action to return the required RHR loops to OPERA 8LE status as soon as possible.

b. The provisions of Specification 3.0.3 are not applicable.

l I

SURVEILLANCE REQUIREMENTS 4.0.8.2 The required Residual Heat Removal loops shall be determined OPERABLE per Specification 4.0.5.

e "The normal or emergency power source may be inoperable for each RHR loop.

NORTH ANNA - UNIT 1 3/4 9-8a Amendment No. 32

'I - -

a l

11-26-77 l l

REACTOR COOLANT SYSTEM

~

3/4.4.8 SPECIFIC ACTIVITY ,

! l LIMITING CONDITION FOR OPERATION 4

3.4.8 The specific activity of the primary coolant shall be limited to:

a. 1 1 0 vC1/ gram DOSE EQUIVALENT I-131, and

b. 1 00/T 1 vC1/ gram.

APPLICABILITY: MODES 1, 2, 3, 4 and 5 ACTION: ,

J MODES 1, 2 and 3* ,

a. With the specific activity of the primary coolant > 1.0 vCt/ gram DOSE EQUIVALENT I-131 but within the allowable limit (below and to the left of the line) shewn on Figure 3.4-1, operation

. may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months provided that operation under these circumstances shall not exceed 10 percent of the unit's total yearly operating time. The provisions of Specification 3.0.4 are not applicable.

4

b. With the specific activity of the primary coolant > 1.0 uCi/ gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one con-tinuous time interval or exceeding the limit line shown.on l Figure 3.4-1, be in at least HOT STANDBY with T"V9 < 500'F within i 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With' the specific activity of the primary coolant > 100/T

vCi/ gram,' be in at least HOT STANDBY with T,yg < 500*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

MODES 1, 2, 3, 4 and 5

(

a. With the spei:ific activity of the primary coolant > 1.0 uCf/ gram DOSE EQUIVALENT I-131 or > 100/E uCt/ gram, perfom the sampling and analysis requirements of item 4a of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits. A special report shall be prepared and submitted to the Comission pursuant to Specification 6.9.2.

This report shall contain the results of the specific activity analyses together with the following infomation:

With T,yg >,500"F.

NORTH ANNA - UNIT 1 3/4 4-22 e

j

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

i 1-15-66 REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING' CONDITION FOR OPERATION -

3.4.9.3 Atleastoneofthefollowingovefpressureprotectionsystemsshall be OPERABLE:

a. Two power operated relief valves (PORVs) with a lift setting of:

1) less than or equal to 420 psig whenever any RCS cold leg temperature is less than or equal to 375'F, and 2) less than or equal to 350 psig whenever any RCS cold leg temperature is less than 185'F, or

b. A reactor coolant system vent of greater than or equal to 2.07 square inches, or

c. A maximum pressurizer water volume of 457 cubic feet with all RCS cold leg temperatures greater than or equal to 320*F.

APPLICA8ILITY: When the temperature of one or more of the RCS cold legs is less than or equal to 375'F except when the reactor vessel head is removed.

ACTION:

a. With one PORY inoperable, either restore the inoperable PORY to OPERABLE status within 7 days or depressurize and vent the RCS through 2.07 square inch vent (s) within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; maintain the RCS in a vented condition until both PORVs have been restored to OPERABLE status,

b. With both PORVs inoperable, depressurize and vent the RCS i

through a 2.07 square inch vent (s) within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; maintain the RCS in a vented condition until both PORVs have been restored to OPERABLE status. . . . .

c.

In the event either the PORVs or the RCS vent (s) are used to mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circtaistances initiating the transient, the effect of the PORVs or vent (s) on the transient and any corrective action necessary to prevent recurrence.

d. The provisions of Specification 3.0.4 are not applicable.

NORTH ANNA - UNIT 1 3/4 4-31 Amen $nent No. 74

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

~~

52__INEgBy_gE_NgC6E88_EgWEB_E68NI_gEgB6IIgN1 _ELUIgS2 _8N9 PAGE 8 ISEBdggyN6dICS ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMANg J ANSWER 5.01 (1.50)

a. true

b. false

c. false REFERENCE surry LP-ND-86.3; NA TS 2.1 bases 193009; K1.05(3.5)

ANSWER 5.02 (1.00) d REFERENCE Surry l esson pl an ND-86. ;:-LP-7, p7.28-7.34; NA NCRODP-86.1 192003; K1.01(2.8)

ANSWER 5.03 (1.00)

C REFERENCE Surry lesson plan ND-83-LP-8, Rev 1; NA NCRODP-83 191004; K1.14(2.5)

ANSWER 5.04 (1.00) b REFERENCE Surry lesson plan ND-86.2-LP-1, p1.16; NA NCRODP-86.1 192004; K1.05(2.4)

ANSWER 5.05 (1.00)

A T* O L

. h,___91EQ8V WMt.,E@$_@RDi@_,@h@s2_92@H@ldBsi_fWEDSa_@l@ ~ ~MT IHEBdQQYNAMICS ANSWERS - NORTH ANNA-itd2: -87/02/09-MOORMANg J i REFERENCE Surry lesson plan ND-86.2-LP-4, p4.11; NA NCRODP-86.1 192006; K1.02(3.1)

ANSWER 5.06 (1.50)

a. more negative

b. more negative

c. more negative i

REFERENCE Surry lesson plan ND-86.2-LP-2, p2.11-2.17; NA NCRODP-86.1 192004; K1.06(3.1)

ANSWER 5.07 (1.50)

a. DECREASES

b. -OCCOC^CEO- 2N46 case 5

c. DOES NOT CHANGE REFERENCE Surr.y lesson plan ND-86.2-LP-1, p1.4, 1.11; NA NCRODP-86.1 192004; K1.07(2.9)

ANSWER 5.08 (1.00)

a. Lower

b. Higher REFERENCE Sorry' lesson plan ND-83-LP-8, pB.9/10; NA NCRODP-83 191004tK1.04(3.4) l l

L

5 t__IUEQBy_QE_NtjGLEAR POWER PLANT OPERATIONi_Ebu1QS,_ANQ ~ PAGE 10 IUEBdQQYU601CS ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J ANSWER 5.09 '(1.50)

(acceptable answers, 3 of any_4)

a. fuel burnup

b. fission product poison buildup

c. power def ec:t

d. heat up REFERENCE Surry lesson plan ND-86.2-LP-5, p5.5; NA NCRODP-86.1 192002; K1.09(2.7)

ANSWER 5.10 (2.00) '

1. reactor. power .

2. coolant flow rate ,_ ,

10Q or fri

3. RCS eeFd temperature I.Tc1 4 .RCS pressure.

REFERENCE Surry lesson plan ND-86.3-LP-2, p2.10; NA TS 2.1 193008; K1.05(3.6)

ANSWER 5.11 (1.50)

Start up rate is positive and constant, reactor power is increasing, and there is no outward rod motion.

REFERENCE Surry lesson plan ND-86.2-LP-7, p7.51; NA NCRODP-06.2 192OOB;K1.11(3.8) 1 r

Q,.__IBEQBY_QE_UURLE88_EQWE8_.ELOUI_QEEBOIl0U._ELulDS _8UQ PAGE 11 IMEBdQQYueblGS ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAMg J ANSWER 5.12 (1.50)

1. Gradual warm _up of steam lines

2. Proper venting of tanks and components during warm up and operation.

3. Steam traps

4. Lines kept full (others as appropriate)

REFERENCE Surry lesson plan ND-83-LP-8, p8-36; NA NCRODP-83 193006; K1.04/1.10(3.6/3.4)

ANSWER 5.13 (1.00) i The presence of adjacent control rods may cause a significant change in an individual control rod worth.

REFERENCE Surry lesson plan ND-86.2-LP-6, p6.19; NA NCRODP-86.2 001/000; K5.05(3.9)

ANSWER 5.14 (1.50) a) > DNB, have partial film boiling, where the fuel rod is alternately covered with steam and water (+.25). Steam has poor thermal conductiv-i ty -- cap ab i l i t i es (+.25), so heat transfer rate drops and Delta T rises

(+.25) b) As fuel surface temperatures rise, stable steam layer forms (+.25) causing a further increase in fuel rod temperatures (+.25). Eventually, significar.t radiative heat transfer occurs causing heat xfer rate to incarease (+.25)

REFERENCE Westinghouse Thermal / Hydraulic Principles II, pp 13-18/20 EPE-074; EK1.02(4.6/4.8) u _

5 t__IBEQBY_QE_UQQ6E88_EQWEB_E68NI_QEEB811QN1 _E6Q1QSx_8ND PAGE 12 IHEBdQDYN8 digs ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN 9 J ANSWER 5.15 (2.00)

A.(AstheU-235/U-238isotopesaredepletedtheir fraction of fissionsdecreasesthus)Deff decreases.

B. Production of plutonium isotopes with smaller delayed neutron fractions decreases the average delayed neutron fraction over core life.

C. As the delayed neutron fraction decreases, one is likely to see a quicker response to change in power (i . e. more of a prompt jump / prompt drop)

REFERENCE Surry lesson plan ND-86.1-LP-7, p7.10; NA NCRODP-86.1 192003; K1.07/K1.08(3.0/2.9)

ANSWER 5.16 (2.00)

A. It doubles (or increases) the head for a given mass flow rate.

B. It will double (or increases) the mass flow rate capacity for a given head.

REFERENCE Surry lesson plan ND-83-LP-8, Rev 1, p8.18; NA NCRODP-83 191004; K1.09/1.10(2.5/2.4)

ANSWER 5.17 (1.00)

With a DNBR of 1.3, during normal operation and anticipated operational occurrences, there is (a 95'/.) confidence that DNB does not occur. When > 1.3, the likelyhood of DNB occurring decreases.

REFERENCE Surry lesson plan ND-86.3-LP-2, p2.10; NA NCRODP-86.1 193008; K1.10(3.1) l

[

I I

u

Gi__IUEQBY_QE_UUGLEGB EQWEB_ELGUI_REEBBI1QU _ELUIDSa ONQ PAGE 13]

IUEBdQQXN801CS

' ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMANg J ANSWER 5.18 (2.00)

1. Density difference (or DELTA T) created by heat addition by the heat source and heat removal by the heat sink.

2. The_ heat sink'must be elevated physically above the heat source.

REFERENCE Surry lesson plan ND-86.3-LP-4, p4.5; NA NCRODP-83, ARR-12 193008; K1.21(4.2)

ANSWER 5.19 (1.00)

~

Neutrons at or near the edge of the core have a higher probability of leaking out than the ones at the center which have a higher probability of causing fission. (Hence: DRW at center i s > than at edge) .

' REFERENCE Surry lesson plan ND-86.2-LP-6, p6.12; NA NCRODP-86.1 192005; K1.14(3.5)

ANSWER 5.20 (1.00)

(any 2-of the the 3 )

1. Neutron production is relatively high, so' power is constant when the reactor is critical.

2. Below 10 exp -8 amps the output of the intermediate range may not be directly proportional to the neutron population.

3. Reactivity has not yet been changed by the moderator or

, fuel temperature.

REFERENCE Surry lesson plan ND-86.2-LP-7, p7.57; NA NCRODP-86.2 192008; K1.12(3.6).

i i

L

~ ~

T

. 5z__'HEORY OF NUCLEAR POWER' PLANT OPERATION1 _FLUIQS2_ANQ PAGE 14 IUE80QQYN8dlCS ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J ANSWER 5.21 (2.50)

a. Maintain DNBR < 1.3 and core exit enthalpy <

saturated (+1.0)

b. Prevent-bulk boiling _during normal operations (+0.5)

c. Ensure fuel clad temperature < 2200 deg F during a LOCA(+1.0)

REFERENCE

-Surry lesson plan ND-86.3-LP-3, p3.12; NA NCRODP-86.3 193OO9;K1.07(3.3) 1 m

6 t__ELONI_SYSIEMS_ DESIGN1 _CQNIBQL1_8ND_INSIBUMENIBIlgy PAGE 38 ANSWERS -- NORTH ANNA 18<2 -87/02/09-MOORMAN, J ANSWER 6.01 (2.00)

(any 4 of 5 at 0.5 ea)

1) Max. Fuel Element Cladding Temp. < 2200 Deg. F

2) Cladding Oxidation < 17% thickness

3) Hydrogen generated by Zirc-Water reaction <1% of max.

possible.

4) Core remains in a coolable geometry

5) Provides for long term decay heat removal REFERENCE 10CFR50.46 NA NCRODP 91.9 ESF p.2.6/2.7 006/050; PWG 4 (4.2/4.3)

ANSWER 6.02 (1.00) a (1.0)

REFERENCE NA NCRODP 93.8 PZR Press. Control and Protect.

010/000; A4.01 (3.7/3.5)

ANSWER 6.03 (1.50)

To provide a path to keep the RHR system full (0.50) and to allow for expansion of the system during heat up of the RCS .

(0.50) (0.50). or

}o cu j ^ d 'd **" Q " f

@4 C H f- Sp h L f ) o randrrthus ov ambiently' { heating up RHRerp r(5WA f(o lec REFERENCE NA NCRODP 88.2 RHR 004/000; K1.01 (3.4/3.9)

ANSWER 6.04 (1.50)

1) Prevents shocking the regenerative heat exchanger and the orifices (0.75)

2) Keeps the regenerative heat exchanger and associated piping pressurized to prevent flashing (0.75)

REFERENCE )

NA NCRODP 88.3 CVCS p. 6 004/020; K4.03 (3.0/3.4)

-6 __E68NI_SYSIEMS_ DESIGN2 _CQNIBQ62_8ND_lNSIBUMENI8IlgN PAGE 39 ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J 004/02O; K6.12 (2.9/3.1)

ANSWER 6.05 (1.Ov) d (1.0)

REFERENCE NA NCROPD-77 RPS p 39 015/000; K4.01 (3.1/3.3)

ANSWER 6.06 ( .50)

TRUE (0.5)

REFERENCE NA NCRODP 93.5 Rod Control 001/000; K4.03 (3.5/3.8)

.3S ANSWER 6.07 N

,sd' (3) motor generator set b) 2 (4) reactor trip breaker 2 (2) power cabinet 4 (7) logic cabinet 1 (6) rod position indication cabinet 4 (5) automatic rod control unit 1 (1) DC hold cabinet 1

-(0.75 for 23 rull, c er rc b, 0.75 for b) fully correct

--- 0 . 1 Ic- cach ewitch nonded to plume a cen>punent in p op- crdcri REFERENCE NA NCRODP 93.5 Rod Control System 001/000 K4.01 (3.5/3.8)

ANSWER 6.08 (1.50)

Sufficient water available to maintain the RCS at Hot Standby for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (1.0) with a steam discharge to the atmosphere (0.25) with a total loss of offsite power (0.25).

REFERENCE NA Technical Specifications Bases 3/4 7.1.3 026/000; PWG-5 (3.3/4.1)

6:__eL8NI_SY@IEUS_ DESIGN2_CQNIBQ62_6ND_1NSIBQUENI@IlgN PAGE 40 ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J ANSWER 6.09 (1.00) a (1.0)

REFERENCE NA NCRODP 93.11 041/020; K4.17 (3.7/3.9); K4.18 (3.4/3.6) l,25 ANSWER 6.10 ( 1. 50 F a) 3,43 5 (0.25 ea) dddd b)c) 6-d) 1,6 e) 2 REFERENCE NA NCRODP 88.1 RCS 002/000; K1.09 ( 4.1/ 4.1 )' , K1.06 (3.7/4.0)

ANSWER 6.11 (2.00) a) The C charging pump must be put on the alternate bus.

(1.0) b) A jumper must be installed (to provide a signal that a J bus (C) charging pumps are running). (1.0)

REFERENCE NA NCRODP 88.3 004/000 K2.02 (3.3/3.5) 004/020 PWG-1 (3.6/4.1)

ANSWER 6.12 :(2.00)

1) PZR Hi Press. Trip (0.2) 2385 psig(0.1), 2/3(0.1)

2) PZR Lc Press. Trip (0.2) 1870 psig(0.1), 2/3(0.1)

3) PZR Lo-Lo Press. SI(0.2) 1765 psig and not blocked (0.1),

2/3(0.1)

4) P-11(0.2) <2000 psig(0.1) on 2/3 (0.1)

5) Press. input to the OT Delta T (0.4) i'

62__E66NI_SYSIEMS_ DESIGN3 _CQNISQ63_8ND_INSIBydENI611gN PAGE 41 ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J REFERENCE NA NCRODP 93.8

-010/000; K1.01 (3.9/4.1)

ANSWER 6.13 (1.00)

The design change resulted because of experiences where the undervoltage trip signal alone was not sufficient to trip the breaker. (1.0)

REFERENCE NA NCRODP 77 RPS p.35 012/000; K6.03 (3.1/3.5) r '"

ANSWER 6.14 (1.50) 4g , mininie'^^ 6hn'I S 8" ^ "

1) Compensate f or power def ect er dD '"# (0.5 ea)

2) To minimize the amount of positive reactivity inserted during a rod ejection accident, and

3) To minimize radial flux tilt (peaking) er jo m aiw}nig MdPM E fe #

REFERENCE dM b d W n I' NA-NCRODP-77 RPS 001/000; K5.04 (4.3/4.7)

ANSWER 6.15 (1.50) a) STSP decreases (0.5 EA) b) STSP decreases c) STSP decreases REFERENCE NA NCRODP 77 RPS p 25 012/000; A1.01 (2.9/3.4)

ANSWER 6.16 (1.00) a (1.0)

REFERENCE NA NCRODP 93.9 Main Generator Control & Protection 045/010; K1.11 (3.6/3.7)

6c__E68NI_SYSIEMS_QESIGN2 _CQNIBQ61_6NQ_INSIBUMENI8IlgN PAGE 42 ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J I.O O ANSWER 6.17 A4-rStW a) TRUE (0.5 EA) b) FALSE g' c) -TPJ IC -

REFERENCE NA NCRODP 91.1 p.2.18 013/000; K1.11 (3.3/3.8)

ANSWER 6.18 (1.50) a) TRUE (0.5 ea) b) FALSE c) TRUE REFERENCE NA NCRODP 93.2 Excore Instrumentation Sys.

015/000; K4.07 (3.7/3.8)

I

! ANSWER 6.19 (1.25) l 40% setpoint from 0-20% (0.5) Turbine power (0.25) and linearly from 40-110% as Turbine Power goes from 20-100%

(0.5)

REFERENCE NA NCRODP 91.1 "ESF-SI of ECCS" 013/000; K1.01 (4.2/4.4)

ANSWER 6.20 (1.00)

1) Control switch to close (0.25)

2) Synchronizing selector switch is ON (0.25)

3) DG terminal voltage is 95% (0.25)

4) Efreaker 86 and 87 protective relay's are reset (0.25)

( 0(s - 6 t entu r outrcarred ,91 phase ddhere. din l REFERENCE NA NCRODP 90.4 EDG p2.57 064/000; A4.01 (4.0/4.3)

I

.6t __E68NI_SYSIEd@_DESIGU1_CQNIBQ62_8ND_IN@~[BydENI8Ilgd PAGE 43 ANSWERS -- NORTH ANNA 1&2 -87/ )2/09-MOORMAN , J ANSWER 6.21' (1.00)

~

The highest reading upper / lower detector is compared to the -

average cf the upper / lower detectors (0.5). The circuit auto def eats below 50% power on ALL charinels (0.5).

REFERENCE NA NCRODP 93.2 Excore' Instrumentation Eys.

015/000; K6.04 (3.1/3.2) & A1.04 (3. 5 /:3. 7)

ANSWER 6.22 (1.75) a) Cabinet has the capacity to support up to 6 stationary gripper coils simultaneously (0 5). So with 2 groups or mere, would overload / heat the c abinet (0.5) .

b) 125 'DC-Latching Rods 70 VDC-Holding Rods (0.5 for reasons, 0.25 for correctly associating voltaqes).

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

_Zt__BBQGEDUBES_ _NQBM86t_8DNQBM86t_EMEBGENGY_6NQ PAGE 8 8891969EIGGL_GQNIBQL ANSWERS -- NORTH ANNA 18<2 -87/02/09-MOORMAN, J ANSWER M Ms !J -f 1. 00 )

2M 90 CcaucE u w ,600)Qp A /Y > k' f b 6 pn 3 REFERENCE VCS, SOP-101 p1 NA OP-5.2 p 4 SUR OP-5.2 p 2,4 ANSWER 7.02 (1.00) 9 0.5 points each:

1 .- Trip turbine locally.

2. Manually oprin reactor trip breakers or the rod drive MG output breakers.

REFERENCE

-NAPS 1-AP-20, p.3.

SUR 1-AP-20, p5 ANSWER 7.03 (1.00) b REFERENCE NA ADM 5.8, pp 2/3 Sur SUADM-ADM-21 p 21 PWG-23: Plant Staffing and Activities (2.8/3.5)

ANSWER 7.04 (1.00) b

. OF=1 for gamma l 100(45/60)(1)=75 l REFERENCE 10 CFR 20.

PWG-15: Radcon Knowledge (3.4/3.9) 1 I

L

!ZA__EBQgEQUBES - NQBd86t_8BNQBd86t_EMEB@ENQY ANQ- .PAGE 9.

,88DI96991G86_GQNIBQL.

TANSWERSE--. NORTH ANNA 1&2 -87/02/09-MOORMAN,-J

-ANSWER '7.05 (1.00) 1 b

REFERENCE

.VCSg T/S p 3/4.3-2, 3/4 3-6

'NA T/S Table 3.3-1

~SUR T/S Table 3.7-1 ANSWER 7.06 (1.00) l d

> REFERENCE

'MNS.EP/2/A/5000/16.3 CNS-EP/1//A/5000/2F3,'p.7.

NAPS 1-FRP-I.3A, p.3.

__SUR. FRP-I.3, p 9' ANSWER -7.07 (1.50)

's)- 'No (+.5 ea)'

-b). No c) Yes

' REFERENCE

NA CSF.F-0.4, F-0.5, F-0.6

.Sur CSF F-3,-F-5, F-6 PWG-10: Recognize abnormal indications for EOPs (4.1/4.5)

-ANSWER 7.08 (1.50)

a) False -(+.5 ea) b) True

<>t) False

. REFERENCE Westinghousc User's Guide for EOPs, pp'5-12 PWG-22(4.3/4.3)

I . .. .

Za__EBgggpUBgS_ _NQBdQLt_G@NQBd@Lt_EdEBgEUGY_6NQ PAGE 10 BGQ1969GIGOL_GOUIB06 ANSWERS -- NORTH ANNA 162 -87/02/09-MOORMAN, J ANSWER 7.09 (2.00) a) Shutdown (+.5 ea) b) Startup c) Shutdown d) Sh u t d a. 61arI4p REFERENCE SON TS 3.4.6.2 NA TS 3.4.6.2 SUR TS 3.1-13 002/020; PWG-8 (3.5/4.4)

ANSWER 7.10 ( .50)

FALSE REFERENCE NA 1-OP-22.11, p4 ANSWER 7.11 (2.50) a 2 b 4 c 1 d 1 e 6 REFERENCE Virginai Power GET pp 14-17 PWG-15 3.4/3.9 ANSWER 7.12 (1.00)

-Unexpected rise in S/G level (+.25 ea)

-High radiation on a S/G blowdown line

-High radiation on an MS line monitor

-High radiation as determined by sampling and analysis REFERENCE Surry EP-4.OO, pp 2 NA 2-EP-3, pp 2

Zn__EBOGEDUBES_:_NQBd86t_OBNQBMGLt_EMEBGENQY_6NQ PAGE 11 BGQ1060 GIG 86_QQNIBQL L ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J EPE-038; EA2.03 (4.4/4.6)

ANSWER 7.13 (1.50)

1) Verify Charging /SI flow

2) RCS Pressure < 1230 psig

(+.5 ea) [l,2 or 3)

3) If component cooling water to any pump is lost REFERENCE SONP Foldout Page NA Foldout page for 2-EP-O Surry Foldout page for EP-1.OO 003/000; PWG-10 (4.1/4.4)

(1,25 )

ANSWER 7.14 +1-(tt++-

Surry (+.25 ea) North Anna (+.25 ea)

Verify SI/CHG pumps running / flow 1. Verify 2 SI/CHG pumps running / flow

. Check RCS pressure <2335 psig 2. Switch BATP to fast speed Switch BATP to fast speed 3. Open MOV 2350 e- In; mt tF : "I7 Open MOV-( )350 4. Check p r press <2335 REFERENCE Surry FRP-S.1 p3 NA FRP-S.1 p 4 EPE-029; PWG-11 (4.5/4.7)

ANSWER 7.15 (1.50) l NORTH ANNA (+.3 ea) l-RCS Press >.2000 psig & increasi-ng

-RCS Subcooling > 50 Deg F l-PZR Level > 50%

l-SG Level > 10% or > 30% Advrs Cntm

OR l-AFW Flow > 730 GPM PEFERENCE NA EP-O Foldout page l

7

,Zs__EB9GEQUBES - NQBd@Lt_G@NQBMGLt_EdEBQENQY_GMQ PAGE 12 8891969 GIG 86_GQNIBQL

' ANSWERS -- NORTH ANNA it<2

-87/02/09-MOORMAN, J

_ :OOo/050; PWG-7-(3.8/4.2)

ANSWER 7.16. (1.00)

'ioinizing radiation; heat stress; differential pressure; O2 deficiency

(+.25 ea)

REFERENCE Surry SUADMO-19 p 3

.NA:ADM 20.9, pp 1

-PWG-18: Knowledge of Safety Procedures (3.0/3.1)

ANSWER 7.17. (1.00)

Any'4 & O.25 points.each:

North Anna Surry.

1.cBank-C position. 1. Date Critical

2. Bank D position. 2. Time Critical

-3. Auct. .High Tavg. 3. Average RCS temp.

4. IR N35. 4. RSC Boron concentration

5. IR N36. _

5. Bank C Control rod position

6. RCS baron concentration. 6. Bank D Control rod position

7. Actual critical position within admin. requirements

. REFERENCE-NAPS 1.OP-1.5, p.12.

SUR7 1-OP-1C App. A p 10 of 10 001.010; K5.08 (2.9/3.3)

ANSWER 7.18 (1.00)

'1.; Reduce generator load until vacuum stabilizes

2. Check. vacuum breaker (MOV-AS-100) closed (.25) and a-water seal present (.25) iREFERENCE-NA '1- AP-14, p3 _

~ . - - - -

Z:__EBOGEDUBEE_:_NQBd86t_8ENQBM86t_EMEBQgNCY_6NQ PAGE 13 88DIRLOGIG86_GQNIBg6 ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J

. ANSWER 7.19 (1.00)

Manually trip the reacter (.50) and the affected RCP (.50)

REFERENCE NA AP-29, p2 000/015 PWG-10 4.2/4.5 ANSWER ~ 7.20 (1.75)

(i 3E en)

1. RWST to hot legs via LHSI . Z. m s , .25 #cr corr ct crdcr

2. RWST to cold legs via LHSI

3. RWST via HHSI to cold leg

4. RWST via HHSI to hot leg

5. RWST via RP system to reactor cavity REFERENCE NA AP-52, p 4,5 034/000 PWG-11 2.8/4.1 ANSWER 7.21 (2.50)

CC surge tank low level alarm CC-pump auto trip alsrm CCW low flow discharge header alarm CCW low pressure discharge header al arm Reactor coolant pump low flow /high temp alarm Excess letdown HX low flow /high temp Non-regenerative HX high temp REFERENCE NA AP-15, p2 008/030 ~PWG-10 3.8/4.2 l

I

,y

.Zc__EBQCEQUBE@_ _NQBd6Lt_GBUQBd8(z_EdEBgEUGy_GNQ PAGE 14 80QIQLQQICOL_CgNIgg6 ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J I

. ANSWER 7.22 ( .75)

Cause not clearly understcod ~ (+.25) or safety related/important equipment operated in an abnormal ar degraded manner (+.5)

REFERENCE

-Surry SUADM-O-02 p3 NA ADM 19.18, pp 1 PWG-10: Recognizing abnormal indications (4.1/4.5)

ANSWER 7.23 (1.00)

An oxygen concentration of <5% by vclume must be maintained in the VCT to avoid explosive mixtures in the gas space REFERENCE VCS, SOP-102 p 1 SUR OP-8.6 p2 NA- OP-8.6 p4

'004/000 K5.04 2.8/3.2

' ANSWER 7.24 (1.00)

Steam pressure mode EO.25]

Tavg input to the steam dump control is not valid without f orced flow in the loops. E.75] or Tog mode coeo4 be mie d to ecol Jew n Loo low 5 0 F (o lonj Toug )

glpeid REFERENCE NA 1-AP-10, Att. 2, p2

'SUR AP-39, p4 4

i_-

T:

y

8m__eQdlNISIB611ME_BBQGEQUBES2_GQNQlIlgNS1_8NQ_LldlI@IlgNS PAGE 9 ANSWERS -- NORTH ANNA 1842 -87/02/09-MOORMAN, J ANSWER G.01 (1.00)

(a)

REFERENCE NA U1 TS.2.1.2 010/000; PWG-5 (2.9/4.1)

ANSWER 8.02 (1.00)

(c)

REFERENCE NA U2 TS 3.4.3.2

.. TPT TS.3.1-la Surry TS 3.1-5.6 010/000; A2.03 (4.1/4.2)

ANSWER -8.03 (1.00)

(d)

REFERENCE

, NA U2 TS 3.1.3.1 001/050; PWG-5 (2.9/4.3)

. ANSWER 8.04 (1.00)

'b ,

REFERENCE.

.TPT TS 3.9 NA TS 3.3-12 Surry TS table 3.7-5a

.073/000; PWG-5(3.0/3.8) q.

8 t __eQululgIBOIlyg_eggggQugggi_GQNQlllQNg2_8NQ_LldlIGIlQNg PAGE 10 ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J ANSWER. 8.05 (1.00)

'(b)

REFERENCE EPIP 3.02 and 3.03.

PWG-36: E-Plan (2.9/4.7)

ANSWER 8.06 (1.00) .

d REFERENCE NA ADM 14.0, pp 6/7 SU-ADM-O-13 PWG-14(3.6/4.0)

ANSWER 6.07 (1.00) a) TRUE-(+.5 ea) b) TRUE REFERENCE TPT TS B3.0.1, B3.0.5 NA TS B3.0.3/B3.0.5/D4.0.3 PWG-5: TS Knowledge (2.9/3.9)

ANSWER 8.08 (2.50) a) SR-2 (+.5 ea)

IR-1 PR-1 b) Yes (+.5) TS 3.3.1 action 2.b applies (+.5)

REFERENCE NA TS 3.3.1 015/020;PWG-5(2.8/3.9)

e s__8QdlNI@IB811ME PRQQEQURE@i_GQNQIIIQN@2_ANQ_6IdITATIQNS PAGE 11 ANSWERS -- NORTH ANNA'i&2 -87/02/09-MOORMAN, J ANSWER 8.09 (1.50)

1) Use of buddy system is required (two people in constant contact or communi cati on ) (+. 5 ea)

2) The entrance is guarded while area is occupied

3) -T v.O pcrcanaal~must sign for key REFERENCE NA HP Manual PWG-15(3.4/3.9)

ANSWER 8.10 (1.00) a) 2 SROs (+.25)of which one must be the shift supvsr. or Ops Supt.(+.25) b) Temp: Procedure Deviation (+.25 ea)

Perm: Request to Change Procedure REFERENCE Surry SDM-60, pp 19/21 NA ADM 5.8, pp 4,5 PWG-23(2.8/3.5) Change Procedure ANSWER 8.11 (1.00)

1) Time of accident (+.33 ea for any 3)

2) Severity of injuries

3) Dose received by victim

4) Is victim neutron irradiated REFERENCE NA EPIP 5.01, pp 2/3 Surry EPIP 5.01, pp 3/4 PWG-36(2.9/4.7)

ANSWER 8.12 (1.00)

No action is required

8 2 __8Dd1NISIB811ME_BBQCEQQBEQ2_CQNQlligNS 2 _8NQ_LldlI611gNQ PAGE. 12 ANSWERS --: NORTH ANNA 1&2 -87/02/09-MOORMAN, J REFERENCE

'NALTS 3.9.8.1/3.9.8.2 064/050; PWG-5(3.1/4.1)

ANSWER 8.13 (1.50)

1) Day tank (+.3) level of at least 750 gal (+.2)

2) On-site supply of fuel (+.3) of greater than 45,000 gallons (+.2)

3) Separate operable fuel transfer pump (+.5)

REFERENCE NA TS 3/4.8.1 1064/050; PWG-5(3.1/4.1)

ANSWER 8.14 (1.50) 1)' Sequence of' events recorder (+.3 ea for any 5)

, 2) . P-250 Alarm Typewriter

3) Strip Charts 4)- Logs 5)- Completed Procedures 6). Post trip review printout REFERENCE SU ADM O-02, Attachment B NA ADM-19.18, Attachment B PWG-28(2.9/3.5)

ANSWER 8.15 (1.00)

1) Loss of source range audible counts (+.25 ea)

2) High flux at shutdown alarm

3) Station evacuation alarm

4) Announcement of containment evacuation REFERENCE NA ADM-20.9, pp 9 103/000; A2.04(3.5/3.6) i

-n .

e z__8QMINISIB811ME_BBQCEQQBESz_QQNQlIlgNS 2 _8NQ_(IMlI8IlgNS PAGE 13 ANSWERS -- NORTH ANNA 1842- -87/02/09-MOORMAN, J

\

ANSWER 8.16 (2.00)

1) .approximately 30 minute intervals (+.5 ea) 2)' Significant change to meteorlogical data

3) plant status

4) . radiological data REFERENCE Surry/NA EPIP Note following State / County Notification Step

_ PWG-36(2.9/4.7)

ANSWER 8.17 (1.50) a) Whole Body- 25 Rem (+.25-ea) Thyroid- 125 Rem b) --volunteers: (+.25 ea for any 4)

--professional rescue personnel

--good physical health

---above the age of 45

--should not be a~ woman capable of reproduction

--familiar with consequences of exposure REFERENCE Surry/NA EPIP 5.06, pp 3/4 PWG-36(2.9/4.7)

ANSWER 8.18 (1.00)

Entry into an Operational Mode may be made (+.5) even if the conditions for an LCO are not met (+.5)

REFERENCE TS 3.0.4 PWG-5: T/S Knowledge (2.9/3.9)

ez__8Dd1NISIB811VE_BBQCEQQBES 2 _QQNQlliQNS 2 _8NQ_61dlI8IlgNQ PAGE 14 ANSWERS -- NORTH ANNA 1&2 -87/02/09-MOORMAN, J

~ ' ANSWER 8.19 (1.00)

With. Temperature < 500~ degrees, the release of activity to the environment

-due'to a SGTR is.. precluded (+.7) since Psat is < SG PORV lift setpoint(+.3)

REFERENCE-TPT TS.B3.1-6 NA TS B 3/4.4.8 Surry TS 3.1-17 002/020; PWG-5(2.9/4.1)

ANSWER 8.20 (1.00)

In an emergency when the action is needed to protect the health and safety of the public (+.75), approved by at least a licensed SRO (+.25).

REFERENCE 10CFR50.54 PWG-36(2.9/4.7)

ANSWER 8.21 (2.00) a) CHG pump starts and injects into a solid RCS (+1.0 for either) or' start of idle RCP with secondary temp within 50 deg of RCS cold leg b) Gives sufficient time for an operator (approx-10 minutes) to respond in case a malfunction resulting in max charging flow from one Chg pump. (+1.0)

REFERENCE NA TS Bases B 3/4-4-16 010/000; PWG-5(2.9/4.1)

az__8Dd1NISIB811ME_EBQCEDQBES2_CQND1I1QNS1_8NQ_LldlI@IlgNS PAGE 15 ANSWERS --- NORTH ANNA ~1&2 -87/02/09-MOORMAN, J' ANSWER 8.22 (1250) a) Radiation worker with an accumulated whole body quarterly dose of 2750 mrem, calender year dose.cf.4800 mrem -

(+.5) or a Limited Radiation worker with a similar dose exceeding 1000 mrem. (+.25) b) Cannot' enter.a radiation area (+.375 ea)

Need HP_ approval and instructions to enter a_ Restricted Control Area l.

REFERENCE NA HP Manual 2.3-7/8 PWG-15(3.4/3.9)

ANSWER 8.23 (1.00)

Simple jumpers require a Jumper Log Form (+.5) whereas a_more complex jumper operation would also require a controlling procedure (+.5)

REFERENCE NA ADM-14.1, pp 5 SU-ADM-O-11, pp 14 PWG-14(3.6/4.0)

ANSWER 8.24 (1.00)

Notify: _SRO on Call (+.25 ea)

Craft foreman on shift Forms: Emergency Work Order

- Equipment History / Failure Analysis REFERENCE

~

.NA ADM 16.5, _ pp 5/6 PWG-23(2.8/3.5) l

o L*NCLOSURE 3 Attachm:nt Page 1 02/12/87 NORTH ANNA POWER STATION COMMENTS ON WRITTEN NRC EXAMINATIONS ADMINISTERED ON FEBRUARY 9,1987 A. Reactor Operator Examination

1. Question 1.03:

Comment: According to ES-202 Section E, General Guidance p. 3 of 6 " Technical Specification questions for reactor operators should be conceptual in nature (e.g., recognition of limiting conditions for operation and Technical Specifications that exist for a given area)." The 95% confidence level is a detail from the Technical Specifications Bases 2.1.1 Reactor Core that is beyond the required knowledge of a reactor operator.

Recommendation: Do not remove any credit for not mentioning 95% confidence.

Reference:

ES-202 Section E, General Guidance p. 3 of 6 and North Anna Technical Specifications 2.1 Safety Limits Bases 2.1.1 Reactor Core p. B2-1. Refer to Attachment 1.

2. Question 1.04 (3):

Comment: Tech Specs Figure 2.1.1 specifies Tavg, not T.

The negative slope of these curves is based on the actual thermal limit being the most restrictive on T ' is also H c acceptable because of the programming relationship between T and T .

ave Recommendation: Accept either RCS temperature, T ,, TH rT as a correct answer.

Reference:

North Anna Tech Specs 2.1 Figure 2.1.1 Refer to Attachment 2.

3. Question 1.05:

Comment: Our students may also mention that no superheating exists. Superheating would create vapor spaces (i.e. voids) that would interrupt natural circulation flow.

Recommendation: Accept above answer as correct in addition to reasons given on answer key. Suggest 2/3 grading criteria.

Reference:

North Anna Lesson Plan: Reactor Energy Removal, NCRODP-86.3, Section IV: Natural Circulation, p. 4.8 and 4.12.

Refer to Attachment 3. .

Attcchment P gs 2 02/12/87

4. Question 1.08b:

Comment: FTC magnitude increases with core age due to lower fuel temperatures and buildup of Pu-240.

Recommendation: Change aitswer key to INCREASE.

Reference:

North Anna Lesson Plan: Reactor Operating Principle: , NCRODP-86.2, Section I: Fuel Temperature Coefficient and Defect and Attachment 4.

5. Question 1.11:

Comment: Another acceptable answer would be that an inserted rod at the center of the core would cause a greater curvature of the neutron flux, than a rod inserted at the edge (increased buckling). This effect would increase neutron leakage and make the center rod worth more.

Recommendation: Accept above answer in addition to reason given on answer key.

Referenco: Westinghouse Nuclear Training Operations: Station Nuclear Engineer's Text: Section 5: PWR Core Physics p.

I-5.36 to 5.38. Refer to Attachment 5.

6. Question 1.19:

Comment: In order to solve problem, students needed formula for Right Circular Cylinder lateral surface area. It was not provided. Our students are not required to have this geometric formula memorized, but only to solve problems once given formula.

Recommendation: No credit be removed for incorrectly calculating U-tube surface area.

Reference:

North Anna Lesson Plan: Mathematics, NCRODP-79, Section VII: Geometry. Refer to Attachment 6.

7. Question 1.23a:

Comment: Unit A could have a higher source range count. Unit A's source strength could be greater. Unit A's secondary source (Sb-Be) and intrinsic source (T 'H) will be stronger if 1

Unit A has not been shutdown as long as the refueled unit.

Recommendation: Accept either A higher or both the same as correct.

Reference:

North Anna Lesson Plan: Reactor Operating Principles, NCRODP-86.2, Section VII: Reactor Startup, p.

7.19. Refer to Attachment 7.

,4 - .

Attachment Pagt 3 02/12/87

8. Question 1.23b:

- Comment: The amount of reactivity added by the control rods will be the same for both units since they are shutdown by the

, same amount of reactivity. It is true . that rod worth is greater at. E0L, and as a consequence Unit B at BOL will have a higher critical rod height. However, North Anna lesson plans

do: not require control room operators to know how rod worth

. . varies with core age . without referring . to the North Anna

" Station Curve Books. In other . words, the question is beyond the knowledge and skills of a CRO, unless the curve books are supplied. Consequently, CRO's may very well answer the critical rod heights are. the same, since the amount of rod

+

reactivity added would be the same. In addition, the question

says that "all systems and parameters are identical at the commencement of the startup." From this the student could infer that the rod worths are assumed to be the same.

The overall problem with this question is that sutdents do not know what " parameters" are assumed to be the same and what

" parameters" are assumed not to be the same.

i Recommendation: Accept either B higher or both units the same as correct.

Reference:

North Anna lesson plans: Reactor Operating

. Principles, NCRODP-86.2, Section VII: Reactor Startup p. 7.3.

Refer to Attachment 8.

1

, 9. Question 2.06:

Comment: Total point value does not match sum t of individual point values. Total is 1.50. Sum is 1~25.

</

An RO is responsible for which heater banks are powered off the 4'

emergency bus and which are off the station service buses. ,//

Strict memorization of which MCC implies RO should have all MCC loads memorized which would be a vast undertaking.

L Recommendation: < Accept identification of which banks are [ '

powered off < emergency bus MCC's and whih banks powered off station service MCC's.

Reference:

Candidates have been trained to use load lists for

MCC loads. Only.the power supplies for major components are to be memorized (i.e, MG sets, major pumps).

10. Question 2.10e:

. Comment: Emergency borate valve is an MOV which will be f unaffected by loss of IA. Answer states fails as is, agree

, remains as is but does not fail.

Recommendation: Accept failed as is or recognition valve is an MOV not affected by IA.

{~

Reference:

ESK 6EA (attached)

-_:, _z , -- - _

Attcchment Pigs 4 02/12/87

11. Question 2.11a:

Comment: Lesson plan states floating ring seals limit leakage to 50 gpm. Further analysis from W has indicated floating ring seals will limit leakage to i 100 gpm. The report of this analysis came from RCP Seal group in Pittsburgh via phone conversation thus lesson plan has not been updated until official documentation is received.

Recommendation: Since candidates have been exposed to the 100 gpm analysis as well as the 50 gpm, delete question.

Reference:

N/A

12. Question 2.15:

Comment: Discharge canal is a correct source of water to fire main system. Water from discharge canal is pumped into a bladder tank which the warehouse 5 fire pumps take suction from.

Recommendation: Accept either the bladder or - discharge canal as third source of water.

Reference:

Fire Protection lesson plan, NCRODP-92.1

13. Question 2.17:

Comment: Answer key states range of associated parameters, question asked for minimum values.

Recommendation: Delete upper limit from answer key.

Reference:

N/A

14. Question 2.23:

Comment: Question asked to place components in proper flow path order, starting with normal power supply ending at CRDM's.

Question was written in reference to attached section of the lesson plan. This step, step 4, is strictly the overview stating components of the Rod Control system. Looking at the attached block diagram, step 4 is not stating components in any order by power supply or flow' path, but merely in what order the components are to be taught in the lesson plan.

Recommendation: Use attached block diagram of Rod Control system to evaluate candidates knowledge of system. If candidate tried to answer the question without adding drawing / explanation, etc., request the poor quality of the question will be taken into consideration /or delete the question.

Reference:

Rod Control lesson plan

Atts.chment Piga 5 02/12/87

15. Question 3.06:

Comment: Question addresses pressurizer control. The answer key addresses pressurizer protection. North Anna has the following:

2 pressurizer pressure control channels 3 pressurizer pressure protection channels 3 pressurizer level / control-protection channels.

Pzr. level control does not use an isolation amplifier per se but a card which serves the same purpose. To be qualified to distinguish between the two, candidate would require Inst. tech knowledge which is not required as per ES 202 B.3. Thus "C" is a correct answer. Also, clarification during exam defining

" pressurizer control" as the pzr. Instrumentation as a whole makes "A" a correct answer.

Recommendation: Accept "A" or "C" as correct answers.

Reference:

Attached

16. Question 3.12c:

Comment: Question as written with the clarification provided during the exam does not have an answer. A charging pump that is racked out cannot be " locked out" because the 86 relay is deenergized. Question was written in reference to attached lesson plan which refers to 1-CH-P-1C breaker when it states "its" breaker, not 1-CH-P-1A as clarified during the exam.

Recommendation: Delete question.

Reference:

ESF Lesson Plan, p. 2.18.

17. Question 3.16:

o Comment: Answer key states: Compensate for power defect and minimize radial peaking as 2 of 3 purposes.

Recommendation: Accept as stated in answer key or

1. Acceptable power distribution limits

2. Maintain minimum shutdown margin

Reference:

Tech Spec Bases 3/41.4 attached. Page B 3/4 1-4.

18. Question 3.18:

Comment: Underfrequency on 4KV buses reactor trip is blocked by P-7.

Recommendation: Include underfrequency as a correct answer.

Reference:

Westinghouse logic 5655D33 Sheet 5 attached.

Attachm;nt Psg3 6 02/12/87

19. Question 3.24a:

Comument : Inputs to main feedwater bypass valve controllers included N-44.

Recomumendation: Include N-44 as an acceptable answer.

Reference:

SGWLC Lesson Plan attached.

20. Question 4.01:

Conunent : Question asked for minimum seal flow. Answer key correct answer is for minimum #1 seal leakoff flow. Term seal flow can imply:

seal injection flow, or #1 seal leakoff flow Recomunendation: Delete question

Reference:

OP 5.2 attached

21. Question 4.13:

Comment: Clarification of answer key. Answers 1 and 2 must be present or 3.

Recommendation: Correct answer key to reflect I and 2 as trip criteria with both present, not independent of one another.

Reference:

Foldout page EP-0 attached.

E

Attachment Pega 7 02/12/87 B. Senior Reactor Operator Exam

1. Question 5.05:

Comment: The [X,] dip is not independent of the magnitude of the power increase.

Recommendation: Change correct answer to a.

Reference:

North Anna lesson plan: Reactor Operating Principles, NCR0DP-86.2, Section IV: Fission Product Poisons

p. 4.11. Refer to Attachment 9.

/2. Question 5.07b:

Comment: FTC magnitude increases with core age due to lower fuel temperatures and buildup of Pu-240.

Recommendation: Change answer key to INCREASE.

Reference:

North Anna Lesson Plan: Reactor Operating Principles, NCRODP-86.2, Section I: Fuel Temperature Coefficient and Defect and Attachment 4.

3. Question 5.10b:

Comment: Tech Specs Figure 2.1.1 specifies Tavg, not T.

The negative slope of these curves is based on the actual thermal limit being the most restrictive on T . T is also H c acceptable because of the programming relationship between T and T ,y ,.

Recommendation: Accept either RCS temperature, T,y,, T #

H c as a correct answer.

Reference:

North Anna Tech Specs 2.1 Figure 2.1.1 Refer to Attachment 2.

4. Question 5.14b:

Comment: It is not only radiative heat transfer that causes heat transfer rate to increase. Conduction also increases heat transfer rate. Q = UAAT. Since U is approximately constant after the stable film is established, the increased AT implies greater conductive heat transfer.

Recommendation: Accept either radir '.ive or conductive heat transfer as reason for increased heat transfer in Region IV.

Reference:

North Anna lesson plan NCRODP-83: Thermodynamics, Fluid Flow and Heat Trensfer, Section X: Heat Exchangers pp.

10.10 and 10.11. Refer to Attachment 10.

Atttchm nt P gs 8 02/12/87

5. Question 5.17:

Comment: According to ES-202 Section E, General Guidance p. 3 of 6 " Technical Specification questions for reactor operators should be conceptual in nature (e.g., recognition of limiting conditions for operation and Technical Specifications that exist for a given area)." The 95% confidence level is a detail from the Technical Specifications Bases 2.1.1 Reactor Core that is beyond the required knowledge of a reactor operator.

Recommendation: Do not remove any credit for not mentioning 95% confidence.

Reference:

ES-202 Section E. General Guidance p. 3 of 6 and North Anna Technical Specifications 2.1 Safety Limits Bases 2.1.1 Reactor Core p. B2- 1. Refer to Attachment 1.

/.

6 Question 5.18:

Comment: Our' students may also mention that no superheating exists. Superheating would create vapor spaces (i.e. voids) that would interrupt natural circulation flow.

Recommendation: Accept above answer as correct in addition to reasons given on answer key. Suggest 2/3 grading criteria.

Reference:

North Anna Lesson Plan: Reactor Energy Removal, NCRODP-86.3, Section IV: Natural Circulation, p. 4.8 and 4.12.

Refer to Attachment 3.

37. Question 5.19:

Comment: Another acceptable answer would be that an inserted rod at the center of the core would cause a greater curvature of the neutron flux, than a rod inserted at the edge (increased buckling). This effect would increase neutron leakage and make the center rod worth more.

Recommendation: Accept above answer in addition to reason given on answer key.

Reference:

Westinghouse Nuclear Training Operations: Station Nuclear Engineer's Text: Section 5: PWR Core Physics p.

I-5.36 to 5.38. Refer to Attachment 5.

Attcchment Pega 9 02/12/87 Y.8 Question 6.07:

Comment: Question asked to place components in proper flow path order, starting with normal power supply ending at CRDM's.

Question was written in reference to attached section of the lesson plan. This step, step 4, is strictly the overview stating components in any order by power supply or flow path but merely in what order the components are to be taught in the lesson plan.

Recommendation: Use attached block diagram of Rod Control system to evaluate candidates knowledge of system. If candidate tried to answer the question without adding drawing / explanation, etc., request the poor quality of the question will be taken into consideration /or delete the question.

Reference:

Rod Control lesson plan

9. Question 6.10A:

Comment: Excess letdown can be aligned to any of the three intermediate legs.

Recommendation: Change answer key to reficct correct answer.

A. 3,4,7

Reference:

FM 93A attached J10. Question 6.14:

Comment: Answer key states: Compensate for power defect and minimize radial peaking as 2 of 3 purposes.

Recommendation: Accepted as stated in answer key or:

1 Acceptable power distribution limits

2. Maintain minimum shutdown margin

Reference:

Tech Spec Bases 3/41.3 attached. Page B 3/41-4

/ 11. Question 7.01:

Comment: Question asked for minimum seal flow. Answer key correct answer is for minimum #1 seal leakoff flow. Term seal flow can imply:

seal injection flow or #1 seal leakoff flow Recoe:mendatio.t: Delete question

Reference:

OP-5.2 attached

, e. -

Attrchment Psgs 10 02/12/87

12. Question 7.09D:

Comment: Refer to RO 4.07 answer key. Answer is "startup" .1 gph inner seal leakage is collected in PDTT which is identified.

Recommendation: Change answer D to startup

Reference:

RO 4.07 answer key and attached Annunciator

Response

13. Question 7.13:

Comment: Clarification of answer key. Answers 1 and 2 must be present or 3.

Recommendation: Correct answer key to reflect 1 and 2 as trip criteria with both present, not independent of one another.

Reference:

Foldout page EP-0 attached to RO 4.13

14. Question 7.14:

Comment: ' Inject the BIT' is a separate means of emergency boration thus a separate step not combined with P.0V 2350 (MOV1350 Unit 1) as stated in answer key.

Recommendation: Accept ' Inject the BIT' as a separate step.

Reference:

1-FRP-S.1 attached

15. Question 7.20:

Comment: In order to obtain full credit, examinee must memorize longterm actions. Response as well as RN0's. Step 5.6 states use one or more of following methods, thus does not specify a specific order.

Recommendation: Accept answers in any order.

Reference:

AP 5 2 attached

16. Question 8.09:

Comment: Answer key states "two personnel must sign for key".

This applies to Surry not North Anna.

Recommendation: Accept "must sign for key".

Reference:

HP Manual pp. 2.3 - 10, step 2.3.5 attached References are attached as appropriate.

ML20206F466

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