ML20154C854

Exam Rept 50-368/OL-88-01 During Wk of 880418.Exam Results: All Six Candidates Passed All Portions of Exam & Will Be Issued Appropriate Licenses

ML20154C854
Person / Time
Site:	Arkansas Nuclear
Issue date:	05/10/1988
From:	Graves D, Pellet J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20154C853	List: ML20154C850 ML20154C854 ML20154C882
References
50-368-OL-88-01, 50-368-OL-88-1, NUDOCS 8805180344
Download: ML20154C854 (41)

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APPENDIX U.S. NUCLEAR REGULATORY COMMISSION REGION IV Operator Licensing Exam Report: 50-368/0L88-01 Operating License: NPF-6 Docket No: 50-368 Licensee: Arkansas Power & Light Company P.O. Box 551 Little Rock, Arkansas 72203 Facility Name: Arkansas Nuclear One. Unit 2 (AN02)

Examination at: AN02 l Chief Examiner: [kY D. N. Graves, Examiner, Date f8b Operator Licensing Section, Division of Reactor Safety i

Approved by: & T[lO!II J.'L. Pellet, Section Chief, Date Operator Licensing Section, Division of Reactor Safety 4 I

1 Sumary NRC Administered Examinations Conducted During the Week of April 18, 1988 (Report 50-368/0L88-01) 1 NRC administered examinations to six (6) candidates. All candidates passed all portions of the examination and will be issued the appropriate license.

8805180344 880511 PDR ADOCK 05000368 V DCD

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DETAILS

1. Persons Examined SR0 License Examinations: Pass - 6 Fafl - 0

2. Examiners D. Graves, Chief Examiner L. Defferding R. Gruel

3. Examination Report Performance results for individual examinees are not included in this report as it will be placed in the NRC Public Document Room and these results are not subject to public disclosure.

a. Examination Review Coment/ Resolution In general, editorial coments or changes made during the examination, or subsequent grading reviews are not addressed by this resolution section. This section reflects resolution of substantive coments made by AN02. Both coments generated by the facility review of the written examination have been incorporated into the master examination key which is included in this report. The full text of the coments is attached.

b. Site Visit Sumary (1) At the end of the written examination administration, the facility licensee was provided a copy of the examination and answer key for the purpose of commenting on the examination content validity. It was explained to the facility licensee that regional policy was to have examination results finalized within 30 days. Thus, a timely response was desired to attain this goal.

(2) At the conclusion of the site visit, the Chief Examiner met with facility representatives to discuss the visit. The following personnel were present:

NRC Facility D. Graves P. Crossland L. Defferding E. Ewing III E. Force D. Johnson W. Perks J. Vandergrift

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Mr. Graves opened the meeting by thanking those present for the cooperation received during the site visit and informing those present that current guidelines do not allow the disclosure of preliminary operating examination results. Other items discussed were as follows:

1. All items that are listed on the Simulation Facility Fidelity Report (included in this report).

2. Mr. Force asked if the Region would consider using facility generated simulator scenarios during the simulator portion of the operating examination.

Mr. Graves responded that the scenarios would be considered for incorporation into the examination process, just as is written examination question bank.

The scenarios would have to meet the examination requirements of NUREG 1021, Operator Licensing Examiner Standards, and would have to be diverse and numerous enough that candidate recognition of a given scenario would not be likely.

3. Mr. Johnson felt that the examiners did not allow enough time to setup out-of-service equipment properly prior to the start of the scenarios. Mr. Graves responded that it was not the examiners' intention to hurry the simulator operator in setting up his initial conditions and would allow the operator the necessary time to place equipment out of service,

c. Generic Comments No areas of knowledge were identified as being generically weak,

d. Master Examination and Answer Key A copy of the final AN02 license examination and answer key is attached. The facility licensee coments have been incorporated into the answer key,

e. Facility Examination Review Coments The facility licensee coments regarding the written examination are attached,

f. Simulation Facility Fidelity Report All items on the attached Fidelity Report have been discussed with the facility and Discrepancy Reports (DR) have been generated.

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U. S. NUCLEAR REGULATORY COMMISSION SENIOR. REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _88K6N181_UUQLE68_QUE:2__

REACTOR TYPE: _EWB:QE__________________

DATE ADMINISTERED: _HQLQ441B________________

EXAMINER: _QB8MElz_Qz______________.

CANDIDATE: _________________________

INSIBUCIIQUS_IQ_Q6NDIQeIEl-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 f.inal grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.

% OF l CATEGORY  %'0F CANDIDATE'S CATEGORY a

__VeLUE- _IQI8L. ___1QQBE___ _V8LUE__ ______________QeIEQQBl_____________ l 251QQ__ _25zQQ ___________ ________ 5. 'i H E O R Y OF NUCLEAR' POWER PLANT OPERATION, FLUIDS; AND THERMODYNAMICS

_2512Q__ _25tQQ ___________ ________ 6. PLANT GYSTEMS DESIGN, CONTROL, l AND INSTRUMENTATION i l

_25zQQ__ _25zQQ ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, I EMERGENCY AND RADIOLOGICAL  !

CONTROL

_25tDQ__ 25tQQ ___________ ______..8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 100zQ0__ ___________ ________t Totals d

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

i Candidate's Signature 4

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

.___~ _o,. .,r., ~. _,_.,.,m_ v - , - . . . .,y - m

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5t__IBEQBl_QE_UUQLEeB_EQWEB_ELeUI_QEEBoI1QUc_ELU1Q1t_eUQ PAGE 2 IBEBdQQ18ed101 QUESTION 5.01 (2.50)

True or False:

a. It will take less time for the reactor to go from 2% to 4% power than from 25% to 50% power if the period is constant.

b. Halving the period will double the Start Up Rate.

c. If reactor power is increased from 1% at a stable SUR of 0.5 DPM, power will be 5%. in one minute,

d. If a positive reactivity addition to a just critical reactor results in a 100 sec. period, the same amount of negative reactivity addition will result in a 100 second negative period.

e. If a shutdown reactor is started up from a 10 CPS count rate, criticality should be expected at a count rate between 2000 and 2500 CPS. (2.5)

ANSWER 5.01 (2.50)

a. Falso

b. True

c. False

d. False

e. False (0.5 ea.]

REFERENCE AN02 PSRT, Pp. 129 - 148 192003K109 2.3/2.3 192003K109 ...(KA'S) l l

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

4 r St__IBEQBl_QE_UURLE6B_EQWEB_ELoul_QEEBoI196t_ELVIRSt_eUR PAGE 3 IBEBdQQ1Ued1RS QUESTION 5.02 (1.00)

Concerning equilibrium Samarium-149 (Sm) reactivity, which of the following statements is correct? 50% equilibrium Sm reactivity is: (1.0)

c. one-quarter of 100% equilibrium Sm reactivity. l l

b. one-half of 100% equilibrium Sm reactivity. l l

c. three-quarters of 100% equilibrium Sm reactivity, j i

d. equal to 100% equilibrium Sm reactivity.

ANSWER 5.02 (1.00) d (1.0)

REFERENCE Unit 2 PSRT, P. 207 192006K115 1.9*/1.9*

192006K115 ...(KA'S)

QUESTION 5.03 (1.00)

Choose the correct response.

The preferred method for dampening a Xenon oscillation is to initially:

a. Insert control rods into a low flux area.

b. Insert control rods into a high flux area.

c. Withdraw control rods from a low flux area,

d. Withdraw control rods from a high flux area. (1.0)

ANSWER 5.03 (1.00) b (1.0)

REFERENCE l Unit 2 PSRT, P. 206 192006K114 3.2/3.2

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't o' U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _68E68Se$_UMQLE68_QUE:2__

REACTOR TYPE: _EWB:QE__________________

DATE ADMINISTERED: _SH4Qel10.._______________

EXAMINER *. _QB6VE$t_Qt______________

CANDIDATE: _________________________

INSIBUQI1QUS_IQ_QeUQ10eIE1 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

__VeLUE_ _IQIeL ___SQQBE___ _VeLUE__ ______________QeIEQQBX_____________

_251QQ__ _25tQQ ___________ ________ S. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS

_251QQ__ _251QQ ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION

_2510Q__ 25tQQ ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL

_25tQQ__ _25tQQ ___________ ________ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 1001Q0__ ___________ ________% Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.

Candidate's Signature

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5t__IHEQBl_QE_UUQLEeB_EQWEB_ELeUI_QEEBoIIQUt_ELU10St_8UQ PAGE 2 .

IBEBdQQlded10S QUESTION 5.01 (2.50)

True or False:

a. It will take less time for the reactor to go from 2% to 4% power than from 25% to 50% power if the period is constant,

b. Halving the period will double the Start Up Rate.

c. It reactor power is increased from 1% at a stable SUR of 0.5 DPM, power will be St. in one minute.

d. If a positive reactivity addition to a just critical reactor results in a 100 sec. period, the same amount of negative reactivity addition will result in s 100 second negative period.

e. If a shutdown reactor is started up from a 10 CPS count rate, i criticality should be expected at a count rate between 2000 and 2500 j CPS. (2.5) l l

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ANSWER 5.01 (2.50) )

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a. False I

b. True

c. False  !

d. False  !

e. False (0.5 ea.)

l REFERENCE AN02 PSRT, Pp. 129 - 148 I 192003K109 2.3/2.3 I 192003K109 ...(KA'S) )

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St__IBEQBl_QE_UUGLEeB_EQWEB_ELoul_QEEBoIIQUt_ELVIDSt_oNQ PAGE 3 IBEBUQQYU6 DIGS QUESTION 5.02 (1.00)

Concerning equilibrium Samarium-149 (Sm) reactivity, which of the following statements is correct? 50% equilibrium Sm reactivity is: (1.0)

a. one-quarte- of 100% equilibrium Sm reactivity.

b. one-half of 100% equilibrium Sm reactivity,

c. three-quarters of 100t equilibrium Sm reactivity,

d. equal to 100% equilibrium Sm reactivity.

ANSWER 5.02 (1.00) d (1.0)

REFERENCE Unit 2 PSRT, P. 207 192006K115 1.9*/1.9*

192006K115 ...(KA's)

QUESTION 5.03 (1.00)

Choose the correct response.

The preferred method for dampening a Xenon oscillation is to initially:

a. Insert control rods into a low f ' u.c area.

b. Insert control rods into a high flux area.

c. Withdraw control rods from a low flux area.

d. Withdraw control rods from a high flux area. (1.0)  !

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ANSWER 5.03 (1.00) '

b (1.0)

REFERENCE Unit 2 PSRT, P. 206 192006K114 3.2/3.2

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< .e l 5t__IBEQBY_QE_UUQLEeB_EQWEB_ELoNI_QEEBoIIQUt_ELVIQ1t_eUQ PAGE 4 IBEBdQQYUedIQS l

l 192006K114 ...(KA'S)

QUESTION 5.04 (1.00)

Why will a positive reactivity insertion at End of Core Life (EOL) cause a greater reactor Startup Rate (SUR) response than the same positive reactivity insertion at Beginning of Core Life (BOL)? (1.0)

ANSWER 5.04 (1.00)

At EOL the delayed neutron fraction (Beta) is smaller (1.0).

REFERENCE Unit 2 Plant Specific Reactor Theory, pg 138 192003K107 3.0/3.0 192003K107 ...(KA's)

QUESTION 5.05 (2.00)

You have just completed a reactor startup and power level is above the point of adding heat (P0AH). In the following cituations, EXPLAIN in terms of Reactivity Coefficients, why the Reactor Power changes? (2.0)

(Assume the core is at mid-life, no other operator action, and treat each situation separately).

a. The Steam Dump pressure setting is raised by 20 psig resulting in a lower final Reactor power.

b. A 1% steam leak develops outside of containment resulting in a higher final Reactor power.

ANSWER 5.05 (2.00)

a. The steam dump pressure setting increase causes an RCS temperature increase (0,5). MTC (0.25) and FTC (Doppler) (0.25) both add negative reactivity to lower reactor power.

b. The increased flow will result in a lower.RCS temperature (0.5).

MTC (0.5) will add positive reactivity and power will rise.

REFERENCE Unit Plant Specific Reactor Theory, chapter 17 192008K117 3.3/3.4

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5t__IHEQBl_QE_UUCLEaB_EQWEB EL6HI_QEEBoI198t_ELU101t_66Q PAGE 5 !

I8E80001060101 192008K121 3.6/3.8 192008K121- 192008K117 ...(KA'S)

QUESTION 5.06 (2.00)

For each of the factors listed below, does the magnitude of the negative Power Coefficient increase or decrease as the factor is changed in the direction described? (2.0)

a. Moderator temperature decreases

b. Core age increases

c. Boron concentration increases

d. Control rods inserted ANSWER 5.06 (2.00)

a. Decrease (0.5)

b. Increase (0.5)

c. Decrease (0,5)

d. Increase (0.5)

REFERENCE Unit 2 Plant Specific Reactor Theory, pg 194-197 192004K113 2.9/2.9 192004K113 ...(KA'S)

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

5t__IBEQBl_QE_UVQLEaB_EQWE8_ELeUI_QEEBoIl00t_ELU101t_80Q PAGE 6 IBEBdQQ1860101 QUESTION 5.07 (1.00)

As a suberitical reactor nears criticality, the length of time to reach equilibrium count rate after an insertion of a given fixed amount of positive reactivity... (1.0)

(SELECT THE ONE CORRECT ANSWER)

a. increases because of a larger number of neutron life cycles required to reach equilibrium.

b. increases primarily because of the increased population of delayed neutrons in the core.

c. decreases primarily because of the increased population of delayed neutrons in the core.

d. decreases because the source neutrons are becoming less important in relation to total neutron population.

ANSWER 5.07 (1.00) o (1.0)

REFERENCE Unit 1 Plant Specific Reactor Theory, pg 157 193008K104 3.8/3.8 192008K104 ...(KA'S)

QUESTION 5.08 (1.00)

Why does Boron Worth decrease with increasing coolant temperature? (1.0)

ANSWER 5.08 (1.00)

Boron worth decreases with increasing moderator temperature because of the decrease in moderator density (1.0) (displaces boron out of the core - at a given ppm there will be more pounds of boron in the core when the system is cold than when hot).

REFERENCE Unit 2 Plant Specific Reactor Theory, pg 198 192004K110 2.9/2.9

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St__IBEQBX_QE_UuQLE68_EQWEB_EL6HI_QEEBoIIQUt_ELulRSt_680 PAGE 7 IBEBdQQXUedIQS 192004K110 ...(KA'S)

QUESTION 5.09 (1.00)

The Xenon peak that occurs after a reactor trip from'100% equilibrium Xenon condition is greater than the peak for a trip from 50% power because:

(Complete the statement by selecting the correct response from the choices listed below) (1.0)

a. The fission yield for Xenon is higher at 100% power.

b. There are more thermal neutrons in the core at 100% power.

c. There is more Iodine in the core at the time of the trip frein 100% power.

d. There are more delayed neutrons in the core at 100% power.

ANSWER 5.09 (1.00) c (1.0)

REFERENCE Unit 2 Plant Specific Reactor Theory, pg 20S 192006K102 3.0/3.1 192006K102 ...(KA'S)

QUESTION 5.10 (2.00)

Given two reactor startups with identical plant conditions. One is performed using a continuous CEA withdrawal and the other is performed using a pull and wait method. l

a. Which would go critical first AND why? (1.0) 1 l

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b. Which would have the highest count rate at criticality and why? (1.0) l 1

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Ez__INEQBl_QE_UURLEaB_EQWEB_ELoNL QEEBoI120t_EL'JIQEt_68Q PAGE 8 IBE800R1U8b1Rd ANSWER 5.10 (2.00)

a. The continuous rod withdrawal startup (0.5) because the required reactivity to take the reactor critical will be inserted first (0.5).

b. The Pull and Wait startup will have the highest count rate (0.5) at criticality due to suberitical multiplication accounting for a higher equilibrium value (0.5) which the continuous pull startup does not have an opportunity to do.

REFERENCE Unit 2 Plant Specific Reactor Theory, Chapter 15 192003K101 2.7/2.8 192003K101 ...(KA'S)

QUESTION 5.11 (1.50)

Wnat steam generator pressure is required to maintain 200 deg's F subcool-ing margin in the RCS when RCS pressure is 595 psig. Show all work. (1.S) l ANSWER 5.11 (1.50)  ;

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1. Add 15 psi to 595 psig = 610 psia (0.25) I l

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2. Using steam tables, 8 610 psia, Tsat = 488 +/- 2 deg's F (0.5) i 1

3. Tsat in S/G = Tres - Tsubcooling = 488 - 200 = 288 +/- 2 deg's F(0.25) l

4. Using steam tables. Psat 8 Tsat = 288 deg's F= 56 +/- 4 psia (0.5) l l

I REFERENCE Steam Tables 193003K125 3.3/3.4 193003K125 ...(KA's)

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

,5t__IBEQBl_QE_UVQLEeB_EQWE8_EL6UI_QEEBoIl08t_ELVIQSt_88Q PAGE 9 IBE8dQQ16ed102 QUESTION 5.12 (1.00)

The equation Q=m (h2 - hl) is used to perform the calorimetric calculation of Reactor Thermal Power. Based on this calculation the power range nuclear instruments are adj usted to read in percent rated power. If the heat input from the Reactor Coolant pumps were neglected in the calculation, how would indicated power from the power range nuclear instruments differ from actual thermal power? (1.0)

ANSWER 5.12 (1.00)

The nuclear instrumentation would indicate higher than the actual thermal power. (1.0)

REFERENCE ANO HT, Thermodynamics, Fluid Handbook, pg 193 - 195 193007K108 3.1/3.4 193007K108 ...(KA'S)

QUESTION 5.13 (2.00) l Regarding Nucleate Boiling:

o. What are the FOUR Reactor Coolant System parameters that the DNB Heat Flux (CHF) is dependent upon? (1.0)

b. Where in the core (exially) is the DNBR the largest and why? Assume normal operating conditions. (1.0)

ANSWER 5.13 (2.00) 1

a. Flow l Temperature i Pressure Power [ 0.25 each) l I

b. Toward the bottom of the core (0.5) because this is where the temperature is the lowest and pressure the highest (0.5).

1 REFERENCE I ANO HT, Thermodynamics, Fluids handbook, chapter 8 I I

193008K105 3.4/3.6

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Ez__IBEQBl_QE_UUQLEoB_EQWEB_ELeUI_QEEBaI10Ni_ELulQft_6UQ PAGE 10 IBEBdQQ16ed10S 193008K105 ...(KA'S)

QUESTION 5.14 (1.50)

What Reactor Coolant System chemistry parameters are controlled by the following additives? (1.5)

a. Hydrazine

b. Hydrogen

c. Lithium Hydroxide ANSWER 5.14 (1.50)

a. Hydrazine controls Oxygen concentration (0.5)

b. Hydrogen controls Oxygen concentration (0.5)

c. Lithium Hydroxide is used to control RCS pH. (0.5)

REFERENCE AA-52002-022 pg 6, 7 194001A114 2.E*/2.9 194001A114 ...(KA'S)

QUESTION 5.15 (3.00)

The reactor is operating at 100% power. STATE the INITIAL effect that each of the following will have (INCREASE, DECREASE, NO EFFECT) on fuel center-line temperature and BRIEFLY JUSTIFY your answer.

a. Decrease in reactor coolant flow (1.0)

b. Increase in RCS pressure (1.0)

c. Decreasing fuel / cladding gap thickness (1.0)

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St__IBEQBX_QE_UUQLEeB_EQWEB_ELeUI_QEEBeI1QUt_ELul0St_8UR PAGE 11 IBEBdQQXUedIQS ANSWER 5.15 (3.00)

a. Increase (0.5). Q is proportional to the mass flowrate, so as flow decreases moderator and therefore fuel must heat up (0.5].

b. No effect (0.5). Subcooled heat transfer is not affected by pressure changes (0.5).

c. Decrease (0,5). Q is proportional to delta-T/L (gap width) -->

delta-T is proportional to QL, so as L decreases, delta-T (and fuel center-line temperature) decreases [0.5).

REFERENCE ANO HT, Thermodynamics, Fluids Handbook, chapter 8 193008K116 2.4/2.6 193008K116 ...(KA'S)

QUESTION 5.16 (1.50)

For each of the following concerning centrifugal pump operation, STATE whether available NPSH INCREASES, DECREASES or REMAINS THE SAME, Consider eacn item separately,

a. Suction temperature is reduced. (0.5)

b. Discharge valve is closed slightly. (0.5)

c. Sr . ion valve is closed slightly. (0.5)

ANSWER 5.16 (1.50)

a. increases (0.5)

b. increases (0.5)

c. decreases (0.5)

REFERENCE ANO HT, Thermodynamics, Fluids Handbook, pg 144 191004K106 3.2/3.3 191004K106 ...(KA'S)

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

ht__ELaNI_SYSIEUS_QES196t_QQUIBQL&_88Q_1USIBudENIeIIQU PAGE 12 QUESTION 6.01 (3.50)

a. List three (3) suction sources for the Emergency Feedwater System, including when each would be utilized. (1.5)

b. When starting the EFW pumps manually (non-auto starts), why should the turbine driven pump be allowed to accelerate to full speed prior to starting the motor driven pump? (1.0)

c. How is pump minimum flow verified with the EFW pumps running even though the steam generators are not being fed? (1.0)

ANSWER 6.01 (3.50)

a. The condensate storage tank (0.25) in normally aligned as the suction source when above 10% power (0.25). When below 10% power (0.25), the CST and the Startup and Blowdown Demineralizer effluent (0.25) are both lined up. The Service Water System (0.25) is utilized if a EFAS signal is present (0.125) and low suction pressure exists (0.125).

b. To prever.t drawing suction pressure excessively low (1.0) which could cause the turbine driven pump to overspeed and trip.

c. EFP racirculation flows are included in the indicated flows to the steam generators so those flow indicators should be utilized (1.0).

REFERENCE AA-42002-021, Emergency Feedwater, pg 2, 17 061000K105 2.6*/2.8* 061000K107 3.6/3.8 061000K401 3.9/4.2 061000A301 4.2/4.2 061000K408 2.7/2.9 061000A204 3.4/3.8 061000K105 061000A204 061000K401 061000K107 061000A301 061000K408 ...(KA'S)

QUESTION 6.02 (2.50)

List the automatic actions that occur as a result of a RAS. (2.5)

ht__ELaul_111IEd1_QE11 Gut _GQUIBQLt_eUQ_18118MdEUI6IlQU PAGE 13 ANSWER 6.02 (2.50)

- LPSI pumps trip

- Containment sump recirc line isolation valves open

- Minimum flow line isolation valves close

- RWT isolation valves close

- Service water supply valve to containment spray cooler (Shutdown Cooling Heat Exchanger) opens (5 at 0.5 etch)

REFERENCE AA-52002-004, Emergency Core Cooling, pg 10 006020A402 3.9/3.8 006020A402 ...(KA'S)

QUESTION 6.03 (1.50)

During normal power operation, the two temoerature sensing elements on the outlet of the letdown heat exchanger fail HIGH. What control function (s) occur (s) as a result. (1.5)

ANSWER 6.03 (1.50)

- CCW flow controller valve from the letdown heat exchanger goes full open.

- Letdown flow to the CVCS rad monitors and boronometer is isolated.

- Bypasses letdown flow around the demineralizers to the VCT.

(3 at 0.5 each)

REFERENCE AA-42002-003, CVCS, pgs 4, 5 004020K404 2.6/3.0 008000K102 3.3/3.4 004020K404 008000K102 ...(KA'S)

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

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kt__ELeUI_SISIEUS_QESIQut_QQUIBQLt_eUQ_INSIBudEUIeIIQU PAGE 14 r

QUESTION 6.04 (3.00)

a. An Emergency Diesel Generator is running (manuel start) and the "governor" handswitch and the "voltage" control handswitch are taken to RAISE or INCREASE. What effect does this have on the machine if:

(2.0)

1. The EDG is paralleled with another source on the grid (output breaker shut)

2. The EDG output breaker is open

b. If the EDG has automatically started, what are two (2) ways in which the machine may be shut down? (1.0)

ANSWER 6.04 (3.00)

a. 1. Kw increases (0.5) and reactive load increases (0.5)

2. EDG speed, or frequency, increases (0.5) and output voltage increuses (0.5)

b. - Placing the local AUTO-LOCK 0UT-START switch to LOCK 0UT Pushing the Emergency Manual Stop PB on the engine

- Tripping the fuel racks (2 at 0.5 each)

REFERENCE ANO Procedure 2104.36 L 064000A401 4.0/4.3 064000A402 3.3/3.4 064000A401 064000A402 ...(KA'S)

QUESTION 6.05 (1.50)

"A" CCW pump is serving Loop 1 CCW, "C" CCW pump is serving Loop 2, and "B" CCW pump is NORMAL AFTER START. "C" CCW pump fails (shaft breaks).

Describe the response of the CCW system. Start the response with discharge pressure on "C" pump decreasing. (1.5) l

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6t__ELoul_111IEd1_DE1106t_CQUIBQLt_6UQ_1811BudEUI611QN PAGE 15 ANSWER 6.05 (1.50)

When C pump discharge is <88 psig (0,5), B CCW pump's crossover valves open (0.5), B CCW pump starts (0.5), supplying Loop 2.

REFERENCE AA-52002-030, Cooling Water, Table 30.1 AN02 Question Bank 202-AA-52002-030-3G 008000A201 3.3/3.6 008300A201 ...(KA'S)

QUESTION 6.06 (1.50)

For the following situations, indicate whether a reactor. trip will occur:

a. High Linear Power trip in channel A and low DNBR trip in channel D

b. "A" CEDM MG set trips with the synchronizing circuit breaker open

c. High LPD trip in channel B and CPC channel C fails (loses power)

ANSWER 6.06 (1.50)

a. no

b. no 4 c. yes (0.5 each)

REFERENCE AN02 Question Bank AA-52002-006-70 012000K401 3.7/4.0 012000K403 2.3/2.7* 012000K103 3.7/3.8 012000K103 012000K401 012000K403 ...(KA'S)

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ht__PL6UI_1111EUS_QE11QUt_QQUIBQLt_8UQ_INSIBQUEUI611QU PAGE 16 QUESTION 6.07 (1.00)

In which area listed below (a- d) will void formation have the greatest effect on neutron leakage as detected by the Source Range Nuclear Instru-mentation during an accident? (1.0)

a. core

b. vessel downcomer

c. vessel head

d. RCS loops ANSWER 6.07 (1.00) b (1.0)

REFERENCE AA-62012-003 015020A202 3.3/3.8 015020A202 ...(KA'S)

QUESTION 6.08 (2.00) ,

l What are the four (4) functions / interlocks provided in the CPCs and PPS by 4 the 1.0E-44 bistable in the log safety channel? (2.0) !

l ANSWER 6.08 (2.00)

When above 1.0E-4% (0.2):

- allows the operator to bypass the high log power trip (0.4)

- inserts (removes the bypass) the LPD/DNBR trips (0.4) ,

1 l

When below 1.0E-4% (0.2)

- Allows the operator to bypass the CPC (LPD/DNBR) trips (0.4)

- inserts the high log power trip (0.4)

REFERENCE AA-52002-014, Nuclear Instrumentation 015000K406 3.9/4.2 015000K407 3.7/3.8 015000K406 015000K407 ...(KA'S)

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i .-

kt__ELeUI_SISIEdS_QESIQUt_GQUIBQLt_eUQ_INSIBudEUIol10U PAGE 17 QUESTION 6.09 (1.50)

Match the plant area (a - f) with the type of fire water system available in that area (1 - 3). (1.5)

c. Containment 1. Wet pipe sprinkler

b. Cable spreading rooms 2. Pre-action sprinkler

c. Chemistry lab 3. Deluge

d. Switch gear transformers

e. Diesel generator rooms

f. Secondary sample room ANSWER 6.09 (1.50)

a. 2

b. 3 l

c. 1 l

d. 3 l

o. 2  !

b.25each)

REFERENCE AA-52002-021, pg 4, 5 l 086000G004 3.1/3.3  !

086000G004 ...(KA'S) i l

l l

l l

l l

l 1

l l

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kt__ELeUI_1111EU1_QE11QUt_GQUIBQLt_eUQ_1U11BudEUI6IIQU PAGE 18 QUESTION 6.10 (3.0.0)

For each of the following conditions (a- c) LIST those actuation signals and trips (1 - 8) that should have automatically occurred. CONSIDER each condition separately. Each condition may have more than one answer from the list of actuation signals or trips. (3.0)

1. SIAS

2. CSAS

3. RAS

4. MSIS

5. CIAS

6. EFAS

7. CCAS

8. none

a. A stea.n line break has occurred in containment and containment pressure : 7 osig containment radiation = normal background S/G 1evels = A: 60% B: 50%

Pzr pressure = 1800 psig S/G oressures = A: 700 psig B: 650 psig RWST level = 93%

b. A LOCA has occurred and containment pressure = 25 psig containment radiation =6 R/hr S/G 1evels = A: 38% B: 38%

Pzr pressure = 1300 psig S/G pressures = A: 380 psig B: 450 psig RWST level = 5%

c. A feedwater problem has occurred containment pressure = 1 psig containment radiation = normal background  !

S/G 1evels = A: 63% B: 39% l Pzr pressure = 2180 psia l S/G pressures = A: 900 psig B: 910 psig RWST level = 85% l 1

Zz__EBQQEQUBE1_:_UQBdekt_oBUQBdelt_EUEBGEUQ1_6NQ PAGE 21 86010LQQ1GoL_GQUIBQL QUESTION 7.01 ( .50)

TRUE or FALSE. Activation of the Emergency Plan supersedes the use of other plant procedures. (0.5)

ANSWER 7.01 ( .50)

False (0.5).

REFERENCE ANO 1903.10 194001A116 3.1/4.4*

194001A116 ...(KA'S)

QUESTION 7.02 (1.00)

According to the E0P/ Tech Guide, during an Inadequate Core Cooling (ICC) event. there is a preference for which systems / components should be used for restoration of feedwater flow. List three (3) systems or components in their preferred order. (1.0)

ANSWER 7.02 [1.00) l

1. Emergency feedwater

2. Main feedwater pump

3. Condensate pump l (0.25 for each component, 0.25 for order) l REFERENCE AN02 2202.01, E0P Tech Guide  ;

000074G012 4.3*/4.4* l 000074G012 ...(KA*S) l QUESTION 7.03 (2.00)

List the four (4) basic entry conditions into 2202.01, Emergency Operating Procedure. Values are NOT required. (2.0) ;

l l

l i

I

Zt__EBQQEQUBE1_:_UQBd6Lt_6BUQBdekt_EdEBQEUQ1_6UQ PAGE 22 BeQ10LQQ1Q6L_QQUIBQL ANSWER 7.03 (2.00)

- Any automatic reactor trip

- A manual reactor trip due to a failure of the RPS to function upon reaching the appropriate setpoint.

- A manual reactor trip due to the occurrence of any malfunction or event i which, in the opinion of the operator, is necessary to protect the plant equipment or personnel.

- If the reactor has not tripped and indications of a S/G tube rupture greater than Tech. Spec. limits are present.

(4 at 0.5 each)

REFERENCE ANO2 EOPs, 2202.01, Rev 3, pg 1 000007G011 4.1*/4.3*

000007G011 ...(KA*S)

QUESTION 7.04 (3.50)

List the seven (7) safety functions having immediate action verifications in the E0Ps. (3.5) l ANSWER 7.04 (3.50)

, - Reactivity Control Electrical Power RCS Pressure Control

- RCS Inventory Control

- RCS Heat Removal

- Core Heat Removal

- Containment Integrity (7 at 0.5 each)

REFERENCE AN02 E0Ps, 2202.01, Rev 3, pg 2 194001A102 4.1*/3.9 194001A102 ...(KA'S) j l

l l

QUESTION 7.05 (1.00) i 1

Where should one look to find a list of SIAS actuated components for post actuation verification? (1.0) l i

- _ _ - . . - - - _ ... _o

o. a Z1__EBQQEQVBES_:_dQBdaL&_oEUQBdelt_EUEBQEUQY_8UQ PAGE 23 BeQIQLQQ108L_QQNIBQL ANSWER 7.05 (1.00)

Appendix to E0P 2202.01 REFERENCE AN02 E0Ps 006020A301 4.2/4.3 006020A302 3.9/4.2 006020A501 006020A302 ...(KA'S)

QUESTION 7.06 (1.00)

During rapid power changes (ramp rates in excess of 30% per hour), which indications should be used for monitoring ASI and WHY not the normal steady state monitoring method? (1.0)

ANSWER 7.06 (1.00)

The CPC indications should be used for monitoring ASI (0.5) due to the slow response time of the Rhodium incore detectors (0.5).

REFERENCE ,

AN02 2102.04, Power Operation, Rev 11, pg 4 01200CV.607 2.9*/3.2*

012000K608 3.6*/3.7*

012000K607 012000K608 ...(KA'S)

QUESTION 7.07 (2.00)

a. What three (3) conditions s11ow the uwe of the Reactor Trip Recovery procedure, 2102.067 (1.5)

b. Whose authorization, as a minimum, is required for conducting the restart? (0.5)

j ZA__88QQEQV8El UQBd8Lt_6EUQBd8kt_'idEBQEUQ1_eUQ PAGE 24 BeQ10LQQ1Q6L_QQUIBQL ANSWER 7.07 (2.00)

o. 1. The cause of the trip is known and is or will be corrected (0.5).

2. The trip did not result from failures of any safety related systems such that a related reportable occurrence occurred (0,5).

3. A cooldown is not required (0,5).

b. Operations Superintendent (0,5) or designee REFERENCE t ANO2 2102.06, Reactor Trip Recovery, Rev 9, pg 1, 3 194001A102 4.1*/3.9 194001A102 ...(KA*S)

QUESTION 7.08 (3.00)

a. Section I of Abnormal Procedure 2203.14, Alternate Shutdown, deals with actions taken during a control room evacuation. What is the minimum number of operators required to perform these actions, AND what function does each have with regard to general responsibility during this event? (2.0)

• . If a fire in the Unit 1 control room causes them to implement an alternate shutdown, what action (s) should be taken in Unit 2 (assume the Unit 2 control room remains habitable)? (1.0)

ANSWER 7.08 (3.00)

a. 6 minimum (0,5)

SS to the TSC (0.5)

SRO and 2 R0s out in the plant (0.5)

Aux operator and waste control operator to the fire brigade (0.5)

b. Begin an immediate plant shutdown (0.5) at the maximum safe rate (0.5).

REFERENCE AN02 AOP 2203.14, Alternate Shutdown, Rev 24, pgs 2, 6 000068G010 4.1*/4.2*

000068G010 ...(KA*S)

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a

?t__E80GEQUBCS_ _UQBd8Lt_eBUQBdokt_EUEBGEUQ1_eUQ PAGE 25 <

BBQ10LQQJC6L_GQUIBQL QUESTION 7.09 (3.00)

a. Explain the two major concerns involving an inadvertent actuation of SIAS and RAS simultaneously? (2.0)

b. If the actuations are determined to be invalid, what two (2) verif ications should be made prior to securing the SIAS? (1.0)

ANSWER 7.09 (3.00)

a. 1. HPSI pumps' recirculation valves shut (0.5) and the pumps will eventually overheat and may.be permanently damaged (0.5).

2. The RWT will gravity daain to the containment sump (0,5) during the time the suctions are shifting. Concerned about loes of RWT water (0.25) and adequate pump suction pressure (0.25).

b. - >30 degree F margin to saturation (0.5)

- secondary decay heat removal (0,5)

REFERENCE AN02 AOP 2203.18, Inadvertent Safety Injection Actuation, Rev 0, pg 1, 3 006050A201 3.9/4.2 006050A201 ...(KA*S)

QUESTION 7.10 (3.00)

a. What three (3) conditions require emergency boration per AOP 2203.32, Emergency Boration? (1.5)

b. What are three (3) methods of aligning high concentration boron to '

the charging pump suction in the above procedure? (1.5) i

)

l 1

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

Zi__EBQQEDVBES_:_UQBdekt_6BUQB06Lt_EdEBGEUQ1_eUQ PAGE 26 BeQ10LQQ1CeL_CQUIBQL ANSWER 7.10 (3.00)

a. - CEAs inserted below the Transient Insertion limit

- Shutdown Margin in Modes 3, 4, 5, 6 less than required

- Boron concentration during refueling operations <1731 ppm (0.5 each)

b. - BAMT gravity feed (0.5) 2CV-4923-1 or 2CV-4921-1 RWT (0.5) 2CV-4950-2

- Boric Acid Makeup Pumps (0.5)

Valve or pump numbers not required REFERENCE AN02 AOP 2203.32, Emergency Boration, Rev 2, pg 1 000024K301 4.1/4.4 000024K302 4.2/4.4 000024K302 000024K301 ...(KA'S)

QUESTION 7.11 (1.00)

TRUE or FALSE

o. During refueling (core alterations), all shutdown cooling may be stopped for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period. (0.5)

b. Fuel handling operations may continue if continuous voice communications between the reactor fueling area and the fuel storage area are lost provided restoration of the lost communication is being pursued. (0,5)

ANSWER 7.11 (1.00)

a. true (0.5)

b. false (0.5)

REFERENCE AN02 2502.01, Refueling Shuffle, Rev 16, pg 16, 17 ,

034000K001 2.3/2.9 l 034000K001 ...(KA's)

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I

_ . - .I

l Zi__EBQQEQUBES_=_UQBdelt_eaN0Bdelt_EdEBGEUQY_eUQ PAGE 27 BeQIQLQQIQaL_QQUIBQL QUESTION 7.12 (1.50)

s. L i s t - t. h r e e (3) different modes of tritium (1H3) production at a commercial nuclear power plant. (0.75)

b. Other than the fact that it is radioactive, list three (3) reasons tritium is a hazard to personnel. (0.75)

ANSWER 7.12 (1.50)

a. 1. Ternary fission

2. Boron activation

3. Lithium activation

4. Deuterium reaction (3 at 0.25 each) ,

b. 1. Long half life

2. Easily absorbed through skin or clothing

3. Reacts chemically the same as hydrogen (0.25 each)

REFERENCE AA-52009-001 ANO2 Question Bank 194001K103 2.8/3.4 194001K103 ...(KA'S) l 4

l

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

I i

1

- - . - -m ~ . , - - - - - - - - -------.,-=--------,-n.,,- ,-- e , - . - .n . ,, ,,e, -,-....,---,,-----n ,,

^

u .s r

Zs__EBQQEQUBES_:_80Bdelt_eBUQBdelt_EdEBGEUQ1_eUD PAGE 28 BeQ10LQQ1G6L_GQUIBQL QUESTION 7.13 (2.50)

A condition arises which requires entry into a high radiation area. The operator entering the area will receive a whole body dose of 40 mrem. The personnel listed below, with their related personal information, are available to do the work. Each candidate is technically competent and physically capable of performing the task. Emergency limits do not apply and time constraints do not permit obtaining authorization for an exposure limit increase. Which candidate (s) have acceptable exposure margins to perform the task? Indicate the reason (s) for rejecting a candidate for the job, if applicable. (2.5)

NOTE: Each exposure below (qtr, yr, life, etc.) includes the exposure above it. Assume the current quarter is the fourth calendar quarter. All exposures are in mrem.

Candidate 1 2 3 4 Sex male male female male Age 27 38 24 20 Today's exposure 50 10 10 20 Wkly exposure 90 150 90 250 Qtr exposure 1220 600 90 1230 Yr exposure 2200 2995 210 2810 Life exposure -

54730 5200 9770  ;

Remarks history -

4 months - '

unavailable pregnant  ;

ANSWER 7.13 (2.50) )

Candidate #1: Rejected (0.25). Will exceed 100 mrem /wk (0.25) and will i exceed 1250 mrem /qtr (0.25).

Candidate #2: Acceptable (0.5)

Candidate #3: Rejected (0.25). Will exceed 125 mrem /qtr (0.5).

Candidate #4: Acceptable (0.5)

} REFERENCE ANO 1000.31, Radiation Protection Manual, Rev 5, pg 37, 38 J 194001K103 2.8/3.4 194001K103 ...(KA'S)

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

Dt__eDd1GISIBoIIVE_EBOCEQUBES&_CQ001110USt_860.L10116I1003 PAGE 29 QUESTION 8.01 (3.00) 1 j a. What action must be taken, and in what time frame, if the unit is

! operating with the minimum shift crew composition and one of the Reactor Operators becomes ill? (0.5) l

b. How long may the shift crew remain below the minimum composition and what action must be taken if this time is exceeded? (1.5)

! c. Fifteen minutes before the scheduled arrival of the on-coming shift, l one of the three on-coming R0s calls in sick and says he will not be

! coming in. The Shift Supervisor decides to call in another operator due to the overtime status of his own shift. He also decides that since the relief operator should arrive shortly after shift change (approximately 30 minutes) that his shift can go home and let the on-coming shift start with two R0s. Were his decisions correct?

JUSTIFY your answer. Assume the plant is operating in Mode 1. (1.0)

ANSWER 8.01 (3.00)

a. Immediately (0.25) attempt to restore minimum crew composition (0.25) such as call in replacement operators.

b. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (0,5). Place the plant in a mode where the minimum crew composition is met (1.0).

c. Decisions were correct (0,5). Minimum crew composition was met without calling in the additional operator (0.5).

REFERENCE ANO2 Technical Specification, Table 6.2-1 194001A103 2.5/3.4 l 194001A103 ...(KA'S) i

(

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

at__eQU1U111BoIIVE_2800EQUBES&_QQUQ1Il0 Ult _eUQ_ Lid 1IeIl00$ PAGE 30 QUESTION 8.02 (2.00)

The plant is in Mode 3 when the following information is turned over to the on-coming Shift Supervisor:

1.8 gpm - leakage past check valves from RCS to SI Tanks (0.9 gpm each to SI Tanks A and B) 1.2 spm - Primary to secondary leakage (total) 4.8 gpm - Total RCS leakage Indicate whether any RCS leakage limits are exceeded, including the limit (s) that was(were) exceeded. (2.0)

ANSWER 8.02 (2.00)

The leakage limits for total SG 1eakage (0.5) of 1 gpm (0.5) and unidentified leakage (0.5) of 1 gpm (0.5) was exceeded.

REFERENCE AN02 Technical Specification 3.4.6.2 002020G0ll 3.3/4.0 002020G011 ...(KA'S)

QUESTION 8.03 (2.00)

The ANO2 Technical Specifications require that the four SI Tanks be operable in Modes 1, 2, and 3. What four (4) conditions are required to be verified to satisfy this requirement? VALUES ARE NOT REQUIRED, only the parameter need be listed. (2.0)

ANSWER 8.03 (2.00) I

- Isolation valve open

- Water volume  ;

- Boron concentration 1

- Cover pressure I (4 at 0.5 each)

REFERENCE ANO Technical Specification 3.5.1 006020G005 3.5/4.2* l 006020G005 ...(KA'S)

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

1

=

u '

at__600lulSIB6I1VE_EBQQEQuBESt_QQURIIl00St_oNQ LIU1IeIl0NS PAGE 31 QUESTION 8.04 (1.00)

Per 10 CFR 55. "Operators' Licenses", what must be done by a licensed operator to maintain his/her license in on "active" status?" (1,0)

ANSWER 8.04 (1.00)

The operator shall actively perform the functions of the appropriately licensed operator (0.25) on a minimum of seven 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shifts (0.25) or five 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts (0.25) per calendar quarter (0.25).

REFERENCE 10 CFR 55.53(e) 19400iA103 2.5/3.4 194001n103 ...(KA'S)

QUESTION 8.05 (2.50)

TRUE or FALSE: (2.5)

a. A 10 CFR 50.59, "Changes, Tests, and Experiments", review is required for changes to non-safety related procedures as well as safety related procedures.

b. A Senior Reactor Operator is a "certified 10 CFR 50.59 reviewer" due to having an active SRO license,

c. A procedure change that expands the acceptance criteria of a surveillance does not constitute a change of intent of the procedure,

d. The Interim Procedute Approval Process may be used to implement a procedure change that changes the intent of a procedure as long as it does not involve an unresolved safety question.

e. Following interim approval of a procedure change applicable to BOTH units, the change form will be placed in the procedure change update manual binder in the Unit 1 Control Room.

's .'

at__eQUIBISIBellVE_EBQQEQUBEft_QQUQlIIQUSt_oGR_L1011eI1QU$ PAGE 32 ANSWER 8.05 (2.50)

a. true

b. false

c. false

d. false

e. true (0.5 each)

REFERENCE ANO 1000.06, Procedure Review, Approval, and Revision Control, Rev 27, pg 3, 9, 15 194001A101 3.3/3.4 194001A101 ...(KA*S)

/ QUESTION 8.06 (2.00)

a. List, in order of proference, the three (3) means of communication to be used for IMMEDIATE NOTIFICATIONS to the NRC. (1.5)

b. WHO, by title, is normally designated to make this notification? (0.5)

ANSWER 8.06 (2.00)

n. 1. ENS

2. Commercial

3. HPN (0.4 for each line, 0.1 for each preference)

b. Shift Administrative Assistant (0.5)

REFERENCE ANO 1000.08, NRC Reporting and Communications, Rev 24, pg 10 - 12 '

194001A104 3.0/3.2 194001A104 ...(KA*S)

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at__oQd1UISIBeIIVE_EBQQEQUBEnt_QQUQ1Il0 Ult _oUQ_LIdII6Il0US PAGE 33 QUESTION 8.07 (3.00)

a. What two (2) persons, by title, may sign for authorizing the placement of Hold Cards? (0.5)

b. What two persons, by title, are required to review the tagout for adequate boundary isolation? (0.5)

c. What two (2) persons, by title, may authorize intentionally entering an Action Statement of Tech Specs due to tagging? (1.0)

d. List the order of placing and removing contractor tags and ANO Hold Cards when being used on the same component (which tags should be hung first, removed first, etc.). (1.0)

ANSWER 8.07 (3.00)

a. SS or CRS (0.25 each) on the affected unit

b. Licensed Operator and Lead Craftsman (0.25 each)

c. Affected Unit's Operations Superintendent or Operations Manager (0.5 each)

d. Install ANO Hold Card, then contractor tag. Remove contractor tag, then ANO Hold Card (1.0).

REFERENCE ANO 1000.27, Hold and Caution Card Control, Rev 10, pg 12-14, 23 194001K102 3.7/4.1 194001K102 ...(KA's)

QUESTION 8.08 (3.00)

a. List five (5) examples of temporary modifications. (2.0)

b. Arrange the following list of personnel in order of preference for performing the independent verification of a temporary modification installation. STATE WHICH ORDER THE LIST IS IN, 1. e. most to least preferred. (1.0)

1. Cognizant engineer

2. Another lead craftsman of the same discipline

3. Responsible supervisor

4. Cognizant SRO on the affected unit

Si__eQd1UISIBoIIVE_EBQGEQVBESt_GQUQ1110NSt_eUQ_LidIIeIIQU$ PAGE 34 ANSWER 8.08 (3.00)

a. - Lifted leads for the purpose of altering a function

- Electrical j umpers Pulled circuit cards

- Intentionally disablod annunciator alarms

- Mechanical j umpers

- Blank flanges

- Disabled relief or safety valves

- Temporary power supplies

- Rotation of spectacle flange (5 required at 0.4 each)

b. 3, 2, 1, 4 (0.33 for each manipulation to put in the correct order)

(most to least preferred)

REFERENCE ANO 1000.28, Temporary Modification Control, Rev 8, pgs 4, 19 194001A103 2.5/3.4 194001A103 ...(KA'S)

QUESTION 8.09 (2.50)

a. When en event occurs that is common to both units such as a security aler . fire in a common building, etc., which Shift Supervisor is responsible for responding to the event? (0,5) l l

b. Key shift personnel shall not enter areas from which they cannot  ;

respond to the control room within 10 minutes. Name TWO (2) positions i which fall under "key shift personnel." (1.0)  !

I

c. List two (2) areas from which returning to the control room may take j more than 10 minutes. (1.0) i I

_ _ _ _ _ , . _ _ . . , . _m . . - . . _ . . _ . . _, _ . _ . _ _ , _ . . . , . _ _ _ _ . , . . , . , , . , , _ _ . _ __

at__eQdlu1SIBoIIVE_EBQQEQUBESt_QQU01110U1t_e00_L1d11611081 PAGE 35 ANSWER 8.09 (2.50)

e. The SS receiving the notification (0.5) is responsible for incident response,

b. Shift Supervisor and STA (0.5 each)

c. - cooling tower area containment

- emergency cooling pond area

( 2 at 0.5 each)

REFERENCE ANO 1015.01, Conduct of Operation. Rev 32, pg 10, 25 194001A103 2.5/3.4 194001A103 ...(KA's)

QUESTION 8.10 (1.00)

TRUE or FALSE:

a. Non-licensed operators may manipulate the reactor reactivity controls if directly supervised by a licensed operator and the trainee is enrolled in a license training program. (0.5)

b. Non-licensed operators may operate indirect controls of reactivity such as steam generator pressure or feed flow with the knowledge and consent of a licensed control room operator (direct supervision is not required). (0,5)

ANSWER 8.10 (1.00)

a. true (0.5)

b. true (0.5)

REFERENCE ANO 1015.01, Conduct of Operation, Rev 32, pg 17, 18  !

194001A111 2.8/4.1*

194001A111 ...(KA'S)

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az__oQd1U111BoI1YE_E80GEQuBEnt_GQUQlIl0NSt_oNQ_ Lid 1IoIIQU$ PAGE 36 QUESTION 8.11 (3.00)

a. Define a Category "E" valve? (1.0)

b. What physically ensures that their position remains as intended? (0,5)

c. Who may waive the independent verification on Category "E" valve operations (non-emergency)? (0.5) d.- TRUE or FALSE. When an independent verification is performed on a Category "E" valve, it is not necessary to actually reposition the valve. (0.5)

e. TRUE or FALSE. Independent verification of throttle valve position may be accomplished by observing the first operator throttling the valve. (0.5)

ANSWER 8.11 (3.00)

a. Valves in the flow path of a safety related system (0.33) required to be in a specified position for the system to perform its safety function (0.33) and whose mispositioning could go undetected frem the control room (0.34).

b. All category "E" valves are locked (0.5).

c. Operations Superintendent (0.5)

d. True (0,5)

e. True (0.5)

REFERENCE ANO 1015.01, Conduct of Operation, Rev 32, pgs 40-41 194001K101 3.6/3.7 194001K101 ...(KA'S)

n  ;' .

SIMULATION FACILITY FIDELITY REPORT-Facility Licensee: Arkansas Power & Light Company Facility Licensee Docket No.: 50-368 Facility License No.: NPF-6 Operating Tests Administered At: Arkansas Nuclear One, Unit 2 Operating Tests Given On: April 19-20, 1988 During the conduct of the simulator portion of the operating tests identified above, the following apparent performance and/or human factors discrepancies were observed:

1. CCW Pump A (2P33A) red, run light is off even when the pump is running.

The green light stays on constantly. Candidates noticed this during their board walkdowns prior to starting the exams. This had been previously)

(DR88-036 had identified by the facility been previously and a Discrepancy Report generated.

2. We received two inadvertent reactor trips during the exams. One was a low reactor pressure of approximately 1640 psig. No evolutions were in progress that should have caused a spike pressure drop, and return to approximately 2200 psig. The cause of the trip, low pressure, was determined later by simulator personnel. The cause of the pressure indication is unknown. The second trip was preceded by abnormal S/G pressure on the "A" S/G. Again, no evolutions were in progress that would have caused a S/G pressure transient. One of the candidates observed the RCS pressure indication on the CPCs drop from approximately 2200 psig to 14 psig, then return to 2200 psig. Cause of the pressure transients is unknown.

3. The annunciator horn occasionally sounds with no visual annunciator illuminated. Lamp test does not indicate any burned out lamps. This malfunction and the inadvertent trips are being investigated as they l occur.

4. Bypassing Nuclear Instrumentation inputs to the protective system is very l cumbersome and' time consuming and cannot faithfully reproduce plant  !

perfonnance in a timely manner. Instead of having a bypass switch feature like in the plant, each light and CPC/PPS input or light must be individually manipulated by the simulator operator, or the indicating lights on the panels must be individually operated in override to get the desired result. Facility is presently working on obtaining and installing bypass switch panels.

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5. When a Linear Power Channel fails low, a CPC Sensor Failure light on the CPCs was anticipated, as well as a CPC Trouble annunciator. Neither was received. Subsequent investigation by the facility revealed that the malfunction model does not insert the failure into the circuitry where the indications would function. The simulator responds faithfully for where the failure occurs, but was not what was anticipated. The

facility personnel stated that a new malfunction would be created to model the desired failure properly (DR 88-057).

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6. During an inadvertent SIAS actuation, the SIAS indicating lights on the CPC/PPS panels never went out. It was determined that where the malfunction inserts the failure into the circuitry is downstream of the '

indicating lights, which was not anticipated. The failure insertion point will be remodeled (DR 88-061).

7. With EDG "A" out of service, EDG B was given a tail-to-start malfunction, and SIAS "B" was inadvertently initiated. The A4 vital bus could not be reenergized, from anywhere, and it should have been able to be manually reenergized by the operators. Investigation by the simulator staff after the exam found that the transfer logic from SIAS interfered with the normal bus transfer logic and would not allow the operators to reenergize the bus. That is incorrect. The problem was identified in the model and will be corrected (DR 88-054).

8. During a pressurizer safety valve failing open malfunction, subcooling and RCS temperature indications on the SPDS fluctuated drastically. RCS temperature varied from 440*F to 15'F to 440*F instantaneously with similar changes in subcooling. The SPDS indication in the plant has a lower range  ;

l of approximately 400 F and would not show these type changes. The drastic changes were explained as being due to the heat transfer model showing slugs of voids /superheat moving through the system in a linear fashion with the large changes occurring instantaneously as these slugs reach the various temperature detectors. DR 88-058 was generated to remodel

, the range of the simulator SPDS to match the range of the actual SPDS in the plant, i 9. It appears that too many area radiation monitors alarm in containment on a leaking pressurizer safety valve. The monitor at the personnel hatch gl alarmed which seems unrealistic. The leak was 75 gpm into the quench J tank, which was still intact. This is being investigated under DR 88-059.

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ht._ELeUI_11SIEdS_QE1100&_QQUIBQLt_8UQ 1USIBWdEUIel100 PAGE 19 ANSWER 6.10 (3.00)

a. CIAS (5)

SIAS (1)

CCAS (7)

MSIS (4)

(0.25 each)

b. SIAS (1)

CSAS (2)

CIAS (5)

CCAS (7)

MSIS (4)

RAS (3)

(0.166 each)

c. none (8)

(1.0)

REFERENCE AA-52002-013 ESFAS STM-2-70 PPS/ESF 013000K101 4.2/4.4 013000K101 ...(KA'S)

QUESTION 6.11 (2.00)

The operability of the main steam isolation valves ensures that no more than one steam generator will blow down in the event of a steam line rupture. What are two reasons for limiting the effect of a steam line  !

rupture to the blow down of one steam generator? (2.0) i l

! l ANSWER 6.11 (2.00)

1. Minimize the positive reactivity effects of the RCS cooldown associated with the blowdown. (1.0)

2. And limit the pressure rise within the containment in the event that the rupture occurs inside the containment. (1.0)

REFERENCE Tochnical Specifications 3.7.1.5 Bases 039000G006 2.2/3.1 039000G006 ...(KA'S) i (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

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ht__EL6UI_111IEd1_QE1100t_CQUIBQLt_6UQ_IUSIBudEUIeIIQU PAGE 20 QUESTION 6.12 (1.00)

Fouling of the surface of the throat of a Feedwater Venturi has occurred during extended operation. Will indicated feedwater flow be GREATER, LESS, or the SAME as actual flow? JUSTIFY YOUR ANSWER. (1.0)

ANSWER 6.12 (1.00)

The effective inside diameter (from increased friction and actusi narrowing) of the throat will be decreased, this will raise the differential pressure being measured (0.5), and thereby give a GREATER than actual flowrate indication (0.5)

REFERENCE ANO HT, Thermodynamics, Fluids, pg 126 191002K101 2.2*/2.4 191002K101 ...[KA'S)

QUESTION 6.13 (1.00)

Indicate whether each of the following is TRUE or FALSE concerning 480 volt Motor Control Center Operation. (1.0)

a. If the control power fuses are blown, the load may still be operated by depressing the contactors,

b. If the control power fuses are blown, the load may still be operated by using the local handswitch.

c. Depressing the open/close contactors does not bypass the valve position / torque contacts in the control circuit.

ANSWER 6.13 (1.00)

a. true (0.33)

b. false (0.33)

c. fe'se (0.33)

REFERENCE AA-52002-007 062000G009 3.2/3.3 062000G009 ...(KA'S)

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