Text

Exam Rept 50-327/OL-85-01 on 850520-23.Exam results:11 of 13 Candidates Passed Both Written & Oral Exams.One Candidate Passed Oral Exam Only.One Candidate Failed Written Exam in All Categories
	ML20128L679
Person / Time
Site:	Sequoyah
Issue date:	06/21/1985
From:	Douglas W, Wilson B; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML19304B354	List: ML20128L659; ML20128L679;
References
	50-327-OL-85-01, 50-327-OL-85-1, NUDOCS 8507250087
	Download: ML20128L679 (150)
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. o' ENCLOSURE 1 EXAMINATION REPORT 327/0L-85-01 Facility Licensee: Tennessee Valley Authority 500A Chestnut Street Chattanooga, TN 37401 Facility Name: Sequoyah Nuclear Plant Facility Docket No.: 50-327 Written examinations were administered at Sequoyah near Soddy Daisy, Tennessee.

Chief Examiner: lh b W. G. Dougla cru.b OblI9l95 Date Signed Approved by: h . .% b bl N Bruc[A. Wilson', Section Chief Da'te Sibned Summary:

Examinations on May 20-23, 1985 Written and oral examinations were administered to 13 candidates; 11 of whom passed. An oral examination was administered to one candidate who passed. A written examination of all categories was administered to one candidate who did not pass.

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. t fEnclosure 11 2 REPORT DETAILS

1. Facility Employees Contacted:

C. S. Benton, Unit Supervisor, Simulator Section

C. O. Brewer, Training Manager

L.4C. Bush, Operations Assistant Group Head

V. E. Keyser, Instructor

B.'C.; Lake, Training Shift Engineer M. J. Lorek, Instructor.

L. M.. Nobles, Superintendent (0&E)

C. H. Noe,. Supervisor,l Operator Training

W. G. Payne, Instructor

L.:H. Sain, NTB

Attended Exit Meeting

2. Examiners:

W.-G.- Douglas, USNRC, Region II F. S. Jagger, EG&G A. J. Vinnola, EG&G

~* Chief Examiner-

3. Examination Review Meeting At ' the conclusion .of the written examinations, the examiners met with V. E. Keyser, B. C. Lake, M. J. Lorek, and W. G. Payne to review the written examination and answer - key. -The following comments were made by the

-facility _ reviewers:

-a. SRO Exam (1) Question 5.02 Facility Comment: Depending upon whether cycle 1 or cycle 3 is assumed, more than one curve is correct.

J NRC Resolution: Agree with facility comment. Based on. cycle 3

-information available at review, answer b or d will be accepted -

for Part 1.

(2) Question 5.07.b Facility Comment: Answer is TRUE for transient, but FALSE if new steady state is achieved.

NRC Resolution: Agree with facility comment. Part b deleted from examination.

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o Enclosure 1 3 (3) Question 5.12 NRC' Resolution: Point value changed from 0.75 to 1.0 to make it

. consistent with other multiple choice questions.

(4) Question 5.26 Facility Comment: Answer is TRUE if cycle 1 is assumed and FALSE if cycle 3 is assumed.

NRC Resolution: Verified comment using Plant Curve Book, TI-28.

Question deleted from examination.

(5) Question 6.02 Facility Comment: There are two possible correct answers for this question.

NRC' Resolution: Agree with facility comment. Answers c and d verified as correct using Systems Manual, Chapter 3 and G0I-3C as reference. Answer key changed to accept c or d as correct.

(6) Question 6.16 NRC Resolution: Answer key was incorrect due to " typo". Answer key changed to accept d as correct answer.

(7) Question 6.19 NRC Resolution: Review of stated reference discovered a third possible answer. Answer key changed to accept any two for full credit.

(8) Question 7.01 Facility Comment: Answer a is also correct.

NRC Resolution: Verified answer a as correct using S0I-68.2.

Answer key changed to accept a or b as correct.

(9) Question 7.11 Facility Comment: FHI-7 -has been recently revised. The correct answer is 4 fuel assemblies which is not one of the four choices.

NRC Resolution: Revision was pointed out by examinee during examination. Instructions were given to disregard choices and put answer for revision to FHI-7. Answer key changed to accept 4 as correct.

i Enclosure 1 4 (10) Question 7.17 Facility Comment: The answer to part a should be FALSE.

NRC Resolution: Agree with facility comment. Using listed reference, verified correct answer as FALSE. Answer key changed accordingly.

(11) Question 8.12 Facility Comment: Containment spray actuation is not a reactor trip signal.

NRC Resolution: Agree with facility comment. Answer key changed to required two answers at 0.5 points each. The total value of question was reduced to 1.0 point.

b. R0 Exam (1) Question 1.02 - See SR0 question 5.07.b.

, (2) Question 1.09 - See SR0 question 5.02.

(3) Question 1.16.c Facility Comment: Depending upon value of beta fraction assumed, answer could either be supercritical or prompt critical.

NRC Resolution: Agree with facility comment. Answer key changed to accept 3 or 4 as correct answer for part c.

(4) Question 2.04 NRC Resolution: Provided reference contained incorrect informa-tion. Using RCS system description, verified answer c as correct.

Answer key changed to accept c as correct.

(5) Question 2.21 Facility Comment: Answer a is correct and answer d is incorrect.

NRC Resolution: Agree with facility comment. Verified, using stated reference and facility supplied handout. Answer key changed to accept answer a as correct.

(6) Question 2.22 Facility Comment: Calculation of AFW load using oasign data

ndicates 5% of full load may be maintained.

Enclosure 1 5 NRC Resolution: Stated references says 3%. However, calculation using data in stated refe' ance yields about 5%. Answer key changed to accept b or c ror full credit.

(7) Question 3.07.b Facility Comment: The answer should be FALSE.

NRC Resolution: Using stated reference, verified the nonlinear gain is controlled by temperature error. The answer key is changed to accept FALSE as the correct answer.

(8) Question 3.12 NRC Resolution: Using stated reference, verified answers c and d as correct. Answer key changed to accept c or d for full credit.

(9) Question 3.21.d NRC Resolution: Delta T does not directly affect the OPAT set-point. However, if it is logically assumed the delta T increased due to power increase which causes a Tavg increase, then the OPAT setpoint would reduce. Since two of the three answers can be correct, this part of the question is deleted from the examina-tion.

(10) Question 3.23 NRC Resolution: Answer key for part c was incorrect. Based on the stated reference, the correct answer is FALSE. Answer key changed accordingly.

(11) Question 4.15 Facility Comment: 501 says 200 degrees, AOI says 225 degrees.

Both answers c and d should be accepted.

NRC Resolution: The RCP must be secured at the most resistive requirements, 200 Jegrees. The answer key is not changed.

(12) Question 4.19 Facility Comment: Parts c and d are not alarms or indications.

NRC Resolution: Agree with facility comment. Parts c and d deleted and parts a and b are worth 0.5 points each.

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Enclosure 1 6

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. Those individuals who clearly passed the oral examination were identified.

There were two generic weaknesses noted during the oral examinations. The first was in the area of radiation protection. The examinees were unable to adequately describe the detectors used for the various survey instruments.

Also, they were unable to explain the sources of radiation (tritium, N-16, secondary activity) in the plant. The other area of generic weakness was in using Tavg and AT indications to figure out whether (and how) T had failed. cold rT hot The cooperation given to the examiners and the effort to ensure an atmo-sphere in the control room conducive to oral examinations was also noted and appreciated.

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

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. o ENCLOSURE 3 U. S. NUCLEAR REOULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION

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FACILITY: _SE2HQ1AE_112 _

REACTOR TYPE: _ EWE-WEC1 DATE ADMINISTERED:_EiL11410 EXAMINER: _1AQQAR. r_

APPLICANT:

Nbk

- -' ' - fDWI v -

3'#L./i=J}h Vj j iU12E22IiQUI_IQ_AEELiCLEIl Use separate' paper for the answers. Write answers on one side only.

Staple question sheet on top ci the answer sheets. Points for each questien 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%. Exa=tnation papers wtil be picked up six (6) hours after the examination starts. .

% OF CATEGOP.Y % OT APPLICANT'S CATECORY

__2AL2;_ _IQILL S22aI _YAL2;__ C.iIIcORY

_22.22__ _11.22 _ 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSTER AND FLUID TLOW

_11.20 _21.22 _

. PLANT DESIGN INCLUDING SATETY AND EMERGENCY SYSTEMS

_12_2E__ _11.22 _

3. INSTRUMENTS AND CONTROLS

_la_2E__ _21.1A 4. PROCEDURES - NORMAL, ABNORMAL, EMERCENCY AND RADIOLOGICAL CONTROL 112.11__ 122.21 __ ____ _

TOTALS FINAL CRADE _ %

l All work done on this er. amination is my own. I have neither l given not received and.

APPLICANT'S SICNATURE l

i L_ '

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1- P R I UCIELES_QE_HECLEAE_EQWEE_ELAMI_QEEEAI1CL. PACE 2 IEEE.4221H&!41CL._HEAI_IEAEEEEE_AEQ_ELEla_ELQW QUESTION 1.01 (1.00)

Which of the following is TRUE concerning RCS operation following the loss of one Reactor Coolant Pump?

a. Core coolant velocity decreases therefore the flowrate in the remaining loops decreases.

b. Flow to the vessel from the remaining three pumps is less than 0/4 of the oragtnal flow.

c. Tiow in the idle loop bypasses the core.

d. Since only thre >

S/C's are providing steam, the steam pressure and temperature in the remaining S/C's is increased.

QUESTION 1.02 (1.00)

TRUE or TALSE?

a. During 100% power operation, Departure from Nucleate Boiling Ratio CDN3E) is greater than the DNBR for ,

20% reactor power.

b. Increasing pressure of the RCS, when operating in the nucleate boiling region of the heat transfer curve'w lj decrea'se the heat _ transfer rate CBTU/hr-square j foot).e,j,.4Ib QUESTION 1.03 (1.50)

TRUE or TALSE?

a. The faster a centrifugal pump rotates, the greater the NPSH required to prevent cavitation.

b. One of the pump laws for centrifugal pumps states that the volume flow rate is inversely proportional to the speed of the pump.

c. Pump runout is the term used to describe the condition of a centrifugal pump running with no volume flow rate.

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. L. 2 ER1HC1 ELE 2_QE_HECLEAE_EQ' DER _Eka3I_QEEEAI1GN. PACE 3

- THEEM2Q13&MIC2 _;iEAI_IRAHEEEl_AU2_ELElR_ELO'd QUESTION 1.04 (2.00)

What is the most significant type of heat transfer (canduction, convectica,-or radtation) taking place under each of the follantng conditions?- Constder each condition separately.

I '. Ilueleate boiling on the cladding surface.

2. Aactdent conditton in wh.ch :solant is botIed and convert =d -!

to' steam.In the reactor vessel.

O. lie a t i t ;ta fis: ten thru a fuel pellet.

4. Oveay heat removal by natural ctreulation.

CU E CT ! C:1 1.03 (0.00)

.. d t : 2 t : hou th. fallcaing ail: iff et "rit eff: t+ncy Cin:rease, d ;raa:c, n2 ;hanga) at a atirady ,,tata pcuer !=ve (Consider each ite :ri:2 tat :/

a. Ab ; ; 1.a t ,s candensar pressure changes f r oaa 1 pst to 1.0S p;t.

t. T;tal 3/G blowdown is changed from 15 gpm to 10 3pm.

c. G a d e n s e r h a t .v e l l t st:p e r a t u r a changes from 123 f to 100 T.

. A : stinte no change tn candenstr p r e .: : u r e )

J. 3 team qua1ity changes f r o m 0 3 . 3 ". to 0 3 . ? . .

QU:'OT IO!! 1.0S ( .50)

TI.UE er FA:0E?

The 3:!R :at n i f l ow valvse ptsvant civitattan of the 2:!2 pump; by apeninJ at $20 dpm and closing at 10!O nm.

1 (tt*** CATECORY 01 CONTIMUED ON I! EXT PACE *****)

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cf' L __EE1U21 ELE 1_QE_HEOLEAE_EQWEE_ELEHI_QEEEAI1CIL. PAGE 4 IEERf'SDYM AMICidEAT. TR A2EEEE_AUQ_ ELE 12_ELQW QUESTION 1.07 (1.00)

When saturated steam is throttled, the down stream fluid:

a. loses energy as it expands.

b. DECREASES in pressure, however, the temperature may INCREASE.

c. may or may not be saturated depending on the upstream conditions.

d. has a higher mass flow rate than the upstream steam.

QUESTION 1.08 (2.00)

a. If the reactor is operating in the power range, how long will it take to ratse power from 20% to 40% with a +0.5 DPM Start-up rate?

1. 12 sec.

2. 21 sec.

3. 36 sec.

4. S4 sec. (1.0)

b. How long will it take to raise power from 40% to 60% with the same +0.5 DPM Startup rate?

1. 12 sec.

2. 21 sec.

3. 3C sec. .

54 sec. (1.0) 4.

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

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PAGE S 1- EE1HC1 ELE 1_QE_NEQLEAE_EQWER PLAUT QEEEA11QN.

IEEEMQQ1EAM1Q1 _HE&I_IEEEEEER_&HD FLUIQ_EEQW QUESTION 1.09 (1.00)

Match the curves, on Tig A.4, with t.he following plant descriptions.

Put your answers on your answer paper. i.e. " Curve 4 - g".

1. Beginning of life (BOL) - 0% power.

2. BOL - 100% power.

3. End of 11fe CEOL) - 0% power.

4. EOL - 100% power.

QUESTION 1.10 (1.00)

Which of the f' ing statements is TRUE?

a. It is NOT pos. ole for the Moderator Temperature Coefficient (MTC) to ever become positive at the Sequoyah plant.

b. It is possible for the MTC to become positive, but ONLY when the reactor is in Mode S.

c. If the MTC is positive, while the reactor is in Mode 2, Technical Specifications must be consulted because there are action statements that must be followed.

d. MTC can be positive in an under moderated core where the moderator to fuel ratio is less than the optimum value.

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

4f t EE1HC1ELEE_QE_EECLEAE_ECEEE_ELAHI_QEEEAI1QM. PAGE 6 IEEEQ21NAMICS EEAI_IEAEEEER_&E2_ ELM 1Q_ELQW QUESTION 1.11 (1.00)

Which of the following best describes the effect on MTC if the RCS temperature is LOWERED?

a. It becomes less negative because boron and water molecules are swept into the core as a result of the outstrge from the pressurtzer, therefere, neutrons spend more time in the resonance legion.

b. It becomes less negative because the rate of change in the density of water per degree temperature change is less at lower temperature which causes a lesser change in rate in resonance escape probability.

c. It becomes more negative because thermal utilization increases and resonance escape probability decreases.

d. It becomes more negative because as temperature is lowered the moderator becomes more dense, this increases the amount of water molecules in the core therefore neutrons have a greater probability of colliding with a water molecule and this is an increased negattve reactivity effect.

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pr: ..

i EE1HC1ELES_QE_EHCLEAE_EQWEE_ELAHI_QEEEAI1QE. PAGE 7 IEEEMQQIGAMICE EEAI_IEMEEER_AEQ_ ELE 1Q_ELQE QUESTION 1.12 (4.00)

Using the attached Xenon worth curve. Tig. 1.1, answer the following.

a. Power at TO was at 70%. What was the power level between T1 and TO?

1. 90%

0. 50%

3. 00%

4. 10% (1.0)

b. What was the length of time between T and T3?

1. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

3. O hours

3. 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months

4. 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months (1.0)

c. What happened at T27

1. Reactor tripped.

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2. Rods were placed in AUTO, and turbine power was raised to 100%.

3. Reactor power was reduced to 10%.

4. Turbine power remained constant, rods were in manual and inserted 50 steps and the steam dump valves failed open (10% of rated power). (1.0)

d. At time T4 ...

1. All Xenon production has stopped.

2. todine decay to Xenon has stopped.

'3 . All Xenon production remains constant, but the burnout increases.

4. Xenon production directly from fission has stopped, but Xenon production from decay lodine continues. (1.0) l l

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

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i PE1HC1 ELE 1_0E HECLEAE_EQWEE_ELAC'I_QEEEAT.1CH. PAGE 8 IHEEM2213 AMICE._HEAI_IEAEEEEE_AUD F LU12_EkQ'd QUESTION 1.13 ( .50)

TRUE or FALSE?

After an extended outage of 1 year, the Secondary Neutron Source is unable to release any neutrons and therefore there are insufficient neutrons available to start a chain reaction in the cote.

QUESTION 1.14 (1.00)

Delayed neutrons play a major role in the operation of the core because they

a. are born at (thermal) slow energy levels (less than i ev) and therefore are more apt to cause a itssion as compared to -

being absorbed by a poison.

b. are considered as epithermal neutrons and therefore they will not travel far enough to leak out cf the core,

c. are born so much later than the prompt neutrons and provide controlability during steady state operations and power transients,

d. provide 70% of the fission neutron inventory and have higher importance factors associated with them as compared to prompt neutrons.

QUESTION 1.15 ( .50)

TRUE or FALSE?

Shortly after a reactor trip from the power range the reactor has a -80 sec. period because the mean life of the longest lived delayed neutron precursor group is 80.7 seconds which corresponds to a SS.9 sec. half life.

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

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i EElHCLEkEE QE_.MU.CLEAE_EQWEE Ek&Hl._QEEEAT.lQlL. PAGE 9 IHEEM2DICAMICi EEAI_IEAESEER AMD FLUfD EEQW QUESTION 1.16 (3.00)

If the Source Range (SR) instruments indicate 50 cps with Keff equal to 0.9, what would the SR instrument indicate if rods were withdrawn to bring Keff equal to 0.95? Assume BOL conditions.

a. 1. 50 cps

2. 75 cps

3. 100 cps

4. 200 cps (1.0)

b. How much reactivity was added?

1. 0.0347

2. 0.0500

3. 0.0506

4. 0.0585 (1.0)

c. If the same amount of reactivity were added again, what would be the state of the reactor?

1. Sub-critical.

2. Critical.

3. Super-critical.

4. Prompt-critical. (1.0)

QUESTION 1.17 (1.50)

Refer to the attached graphs B.6.0 and B.7.1.

a. Which graph represents Total Power Coefficient? C0.5)

b. rt!! in the blanks on the right side of each graph with BOL or EOL. Tear Graphs from the exam and include with your answer sheets. (1.0)

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1- EE1HCIELEE_QE_NE2 LEA 2_EQWER_ELEMI_QEEEAI1QE. PAGE 10 IHEEM221HAC10L._EEAI_IEACEEEE_ACQ_ ELE 12_ELQW QUESTION 1.18 (2.00)

Refer to TI-08. Pig. B.2.c attached to answer the following.

a. The reason the curve changes slope at the point marked "a" is because:

1. this is the point where the Power Range lower detector is no longer able te detect thermal neutrons and the rods have not moved high enough to allow the Power Range upper

, detector to detect thermal neutrons.

2. the rods are moving through the area that has the highest thermal neutron flux.

3. bank "B" rods have Just stopped moving, while bank "C" rods continue to be withdrawn, therefore, only one bank of rods is moving instead of two banks just prior to point "a". -

4. there are two effects here. One, bank "A" rods have just sicpped moving, while bank "B" rode continue to be withdrawn, two, Just prior to point "a", bank "A" rods were moving through an area (top of the core) that has a relatively low concentration of thermal flux. (1.0)

b. The reason for the positive slope on the curve marked "b" is because:

1. the bank "D" rods are moving into an area that has an increasing thermal flux concentration.

3. in addition to bank "C" rods moving through an crea of increasing concentration of thermal flux, now bank "D" is also moving. i.e. overlap.

3. bank "C" has just stopped moving and now bank "D" rods aro moving into an area of the core that has a decreastng Xenon concentration.

4. bank "D" is moving through the uppermost part of the core where the thermal flux is decreasing, therefore, the competition for neutrons is greater and rod worth is higher. (1.0)

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

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.L __EEluc1EEEE_QE_uuCLEAE_EQWEE_EEAMT OPERATION. PAGE 11 IEEEl4021gAlM.CE . .. HEAT TR&giEEE_AND PLUID TEQW QUESTION 1.19 C .70)

Compare the calculated Estimated Critical Position (CCP) for a startup 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months after a trip to the actual Critical Rod Position (ACP) If the following events / conditions occurred. Consider each l

Independently. Limit your answer to:

a. ACP higher than ECP.

b. ACP lower than ECP.

c. ACP would not be significantly different than ECP.

1. One Reactor Coolant Pump is stopped one minute prior to criticality.

2. The steam dump pressure setpoint is increased to a value just below the code safties setpoints. -

3. The startup is delayed 2 more hours.

QUESTION 1.20 ( .75)

Tor each condition in COLUMN A find the correct heat transfer equation in COLUMN B that would be used to calculate the heat transferred.

COLUMN A COLUMN B

a. Across the reactor 1. b=UAdeltaT (cold leg to hot leg) . .

2. Q

M delta T

b. Across S/G U-tubes . .

(primary to secondary) 3. O = MCp delta T

c. Across S/C secondary side 4. h = M delta H Cfeedwater to steam) .

5. O = UA delta H

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i P E1HC1ELEE_2E_HUCLEAE_EQ'dEE_EL&NI_QEEEAIlQth. PAGE 12 IEEEMQO1HAMICL* dEAI_IEAMEEEE_&E2_ ELE 1D ILOW QUESTION 1.21 (1.00)

TRUE or TALSE?

a. During a RCS heatup, as temperature gets higher, it will take a smaller letdown flow rate to maintain a constant pressuriser level (0.S)

b. Increastng condensate depression (subcooling) will cause BOTH a decrease in plant efficiency AND an increase in condensate (hotwell) pump available NPSH. (0.5)

QUESTION 1.22 (1.00)

Steam extting the HP turbine is at 785 psig, 90% quality. Steam entering the LP turbine is superheated to 100 T. What is the enthalpy change of the steam?

a. 85 BTU /lbm

b. 140 ETU/lbm

c. 154 BTU /lbm

d. 705 BTU /lbm

(***** END OF CATECORY 01 *****)

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EL&gT DESIGN INCLUDINQ__ SAFETY &EQ_EMEEQEECY SYSTEM 1 PAGE 13 QUESTION ~2.01 (1.00) l The order of transfer of power sources to the 6.9 kV Shutdown Board 1B is

a. From 6.9 kv Unit Board 1D to 6.9 kV Unit Board 1C to Emergency Diesel Generator,

b. From 6.9 kv Unit Board 1B to 6.9 kV Unit Board 1D to Emegency Diesel Generator.

c. Trom 6.9 kV Unit Board ID to 6.3 kV Unit Board 1B to Emergency Diesel Generator.

d. From 6.9 kV Unit Board 1C to 6.9 kv Unit Board 1D to Emergency Diesel Generator.

QUESTION 2.02 ( .50)

TRUE or TALSE?

Following the receipt of an emergency start signal, diesel generator speed and voltage may be manually adjusted from the control room.

QUESTION 2.03 (1.00)

The 125 vde vital batteries are sised to supply the de power required to maintain the plant in a safe shutdown condition for hours,

a. 0.5

b. 1.0

c. 1.5

d. 2.0

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

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, d 9 twt Encu_innunuG_nEEIl_aNR_EMEEGEEC1_111IEME PAGE 14 QUESTION 2.04 (1.00) l Which of the following best describes the Reactor Coolant Pump Thermal Barrier and Heat Exchanger?

a. Allows a contrciled amount of relatively cool water to enter the seal section of the pump.

l

b. Prevents RCS water entering the lower radial bearing and the seal section of the pump.

c. Component Cooling Water flows through the cooling coils.

d. During a loss of seal injection, minimi:es the amount of hot l

RCS water into the lower radial bearing.

QUESTION 2.05 (1.00) .

Match the following parameters in COLUMN A to their value's in COLUMN B.

COLUMN A COLUMN B

a. Safety injectten pump normal 1. 600 discharge pressure (psig).

2. 925

b. RHR pump normal discharge pressure (psig). 3. 1 50

c. Minimum volume of cold leg 4. 1500 accumulator (cu. ft.)

S. 1800 d.-Volume of UH1 accumulator (cu. ft.)

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', 1 PLAGI_QE11GE_lNOLEQ1HG_i&EETY AHQ_EMEEQLUCY S Y ST EME PAGE 15 QUESTION 2.06 (1.00)

Which of the following statements is TRUE concerning the use of the Residual Heat Removal system?

a. Used during plant startup in conjunction with the CVCS to equalt:e boron concentration between the RCS and pressurtzer.

b. Used during plant cooldown in conjunction with the CVCS to equalt:e boron concentratton between the RCS and pressuri:er.

c. Used during plant cooldown as an alternate letdown flow path,

d. Used during plant heatup as an alternate letdown flow path.

CUESTION 2,07 (1.00)

Wr.ich of the following is the preferential order of valve operation for Emergency Boration?

a. fCV-62-929 (Emergency bore.te manual valve), TCV-62-138 (Emergency borate MOV),62-135 & 6 (RWST suction valves), BIT injection valves.

u

b.62-135 & 6. BIT injection valves, TCV-62-929, TCV-62-138.

A

c. FCV-62-138, YCV-62-929. BIT injection valves,62-135 & C.

4

d. FCV-62-138, KCV-62-929,62-135 & 6. BIT injection valves.

QUESTION 2.08 ( .50)

TRUE or FALSE?

An interlock of the Manipulator Crane hoist drive circuit in the up direction permits the hoist to be operated only when the open indicating switch on the gripper is actuated.

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

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PLALLDE11CH_1COLEQ1NG 12.EETY ANQ_.EMEEGIUCL11EIEMS PAGE 16

, E-QUESTION 2.G9 (1.00)

Whten of the following is a component supplied by a Component Cooling Water System Safeguards Train pump?

a. Non-regenerative heat exchanger,

b. RHR heat exchanger,

c. RCP thermal barrier heat exchanger

d. Reciprocating charging pump.

QUESTION 2.10 (1.00)

Which of the following will cause a trip of a running Main Feed-water Pump? ,

a. Low feedwater temperature,

b. Low Main reed Pump turbine speed.

c..-Recirculation valve open.

d. Safety injection.

QUESTION 2.11 (2.00)

State whether the following statements are TRUE or FALSE concerning the Upper Head Injection System (UHI).

a. The membrane in the Nitrogen pipe between the water and nitrogen UH1 accumulator prevents nitrogen absorbtion in ANY of the water in the UH! system.

b. The membrane prevents the injection of nitrogen into the RCS following UH1 actuation.

c. The UH1 system operates during a LOOA which is greater than the capacity of the charging or Si pumps,

d. The UH! system can be isolated from the RCS by closure of their motor operated valves.

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

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bkANT DE11GE_lEChilD1HQ_1AEST.L.&ER_EMEECEROY SYSTEME PAGE 17 QUESTION 2.12 (1.00)

Which of the following is a purpose of the atmospheric relief valve on the main steam line?

a. Will actuate during a trip from full load while preventing the lifting of the safety valves.

b. Used as a dummy load during plant startup.

c. Used as a steam dump when the condenser is unavailable.

d. Used to control cooldown rate during normal plant cooldown.

QUESTION 2.13 (1.00)

What signal inputs are used to control the position of the governor ~

on the Turbine Driven Auxiliary Peed Pump?

a. Turbine speed and pump discharge flow.

b. Turbine speed and pump discharge pressure.

c. Steam pressure and pump discharge pressure.

d. Steam flow and pump discharge flow.

QUESTION 2.14 (1.00)

What is the normal' operating speed of the Emergency Diesel Generator?

a. 800 RPM

b. 850 RPM

c. 900 RPM

d. 950 RPM

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

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E- Ek1HI QE112H_1HCLE2122 E EEII AND. EMEECEEC1_11EIEME PAGE 18 QUESTION 2.15 (1.00)

Match the following Emergency Diesel Generator speeds to the event that occurs as the generator is started.

a. 40 RPM 1. Engine running alarm.

b. 200 RPM 2. Field flash.

c. 550 RPM 3. Opens ERCW valve to Jacket Water heat exch.

d. 850 RPM 4. Diesel muffler room exhaust fan starts.

S. Diesel engine tube oil pump starts.

QUESTION 2.16 (1.00)

When operating the RHR Heat Exchanger outlet flow control valves, FCV-74-16 & FCV-74-28 from the Unit-1 and Unit-2 control rooms, state if the fo!!owing are'TRUE or False.

a. Reset FCV-74-16 counter-clockwise (to the left) on-Unit-1.

b. Reset FCV-74-28 counter-clockwise (to the left) on Unit-2.

QUESTION 2.17 (1.00)

The purpose of the Reactor Coolant Pump (RCP) Seal standpipe is to provide a:

a. final collection point for the #3 seal leakoff.

b. lubricating water supply for #3 seal.

c. final collection point for the #2 seal leakoff.

d. pressure head for #3 seal.

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

r 2 PLANI _QE11CN I NCMQ1gq_1&EET,Y AND EMEEQENCY _SYSID41 PAGE 19 QUESTION 2.18 (1.00)

The purpose of the interlock that prevents the letdown isolation valves from opening or shutting unless all three orifice isolation valves are shut is to prevent:

a. exceeding design flow rates of the demineralizers.

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

c. flashing of water on the shell of the regen. heat' exchanger.

d. unnecessary lifting of relief valves downstream of orifices.

QUESTION 2.19 (1.00)

What are TWO reasons for maintaining a minimum spray bypass flow '

to the pressurtzer? Choose only one of the following combinations.

a. 1. Prevent excessive cocling to the surge line.

2. Reduce the delta pressure across the spray valves.

b. 1. Reduce thermal shock to the spray nozzle.

2. Ensure that the backup heaters cycle on.

c. 1. Prevent excessive cooling to the spray line.

2. Equalize boron between pressurizer and the RCS.

d. 1 Minimize stress to the surge line thermal sleeve.

2. Remove gases from the RCS.

QUESTION 2.20 (1.00)

The Condensate Storage Tank minimum water volume of 190,000 gallons is sufficient to maintain the plant in hot standby for how many hours?

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

b. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

c. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 1

d. 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months

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2. PLAMT_QEstCM IEOCE21HC.5ACETY AMD EMEEGEHCY SYSTEME PACE 20 CUESTION 2.21 (1.00)

Which action below is taken to protect the Motor Driven Auxiliary Teedwater Pumps from a runout condition?

a. Flow ortfices in the discharge line of each pump.

b. Trip of motor breaker.

c. Automatic closure of the flow control valves.

d. Rectreulation of the excess flow to CST.

CUESTION 2.22 (1.00)

When ALL the Auxiliary Feedwater Pumps are used for normal plant startup, what appro:.. percentage of full load can be maintained?

s. 1

b. 3

c. 5

d. 7 QUESTION 2.23 (1.00)

Whtch of the following provides a direct input to cause the feedwater isolation valves to close automatically?

a. Reactor trip from 100% power.

b. Steam Generator Hi Hi level.

c. Steam Generator Hi level

d. Both Main reed pumps trip.

(***** CATECORY 02 CONTINUED ON NEXT PACE *****)

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, ' PLANT DE11GE_1NCLUDING EATETY AND . EMERCERCL_iHTEMS PAGE 21 QUESTION 2.24 (1.00)

The Component Cooling Water system in conjunction with the RHR system ts designed to reduce the RCS temperature to _r within hours after shutdown.

a. 350, 16

b. 120, 24

c. 250, 18

d. 140, 16 QUESTION 2.25 (2.00)

Por the following components, indicate whether they will receive .

an OPEN, CLOSE, or No signal as a result of a safety injection initiation signal (with Phase "A").

a. Control room supply ducts

b. Main feed bypass valves

! c. S! accumulator discharge isolation valves

d. Normal charging header isolation valves l e. Llain steam isolation valves i f. RWST to Si pump suction valves l g. Seal water return isolation valve l h. Component cooling isolation valve from RHR system i 1. Component cooling isolation from letdown heat exchanger J. Steam supply valves to turbine-driven feed pump I

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P I AM"' DEE1CLIECLLU11HQ_11EEll_&MD EMERCEMC1_111IEM1 PACE 22 QUESTION 2.26 (1.00)

Reactor Coolant Pump #3 is started with the plant in Mode 5, after its seal has been replaced. After operating for approximately 20 minutes the following is observed.

1. #1 seal detta P >390#

2. Standpipe low level

3. #1 seal leakoff has increased ASSUME:

1. Plant pressure is at 400#

2. Seal injection at 6 gpm.

Which of the fo!!owing is a probable cause for the abnormal indications?

a. #3 seat failure,

b. VCT pressure is low. .

c. #1 seat bypass ts closed.

d. RCDT pressure has increased.

QUESTION 2.27 ( .50)

TRUE or TALSE The automatic switchover of the charging pump's suction from the VCT to the RWST is designed to maintain proper seat injection flow to the RCP's.

GJESTION 2.28 (1.00)

Which of the following components served by the Component Cooling Water System (CCW) would NOT cause an increase in both CCW surge tank level and CCW radiation monitor indication if a failure of the component were to occur?

a. Reactor Coolant Pump bearing coolers.

b. RHR pump seal coolers,

c. Seal Water Heat Exchanger.

d. Spent Tuel Pit Heat Exchanger.

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

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, L ELANT DEElGM_1 HCL 11Q1gG J &[,ETY AMD EMERCggq1_SYSTIME PACE 23 QUESTION 2.29 ( .50)

TRUE or TALSE7 The Reheat Stop Valves close on a turbine trtp to control turbine overspeed due to expansion of entrained steam in the MSR's through the low pressure turbine.

CUESTIO!! 2.30 (1.00)

If an unsaturated bed of H-OH resin is placed in service, what will be the result? ,

a. RCS Oxygen concentration will increase,

t. No ion exchange will occur for the first 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of operation.

. RCS Bcron concentration will decrease,

d. RCS Soron concentration will increase.

(***** EllD Or CATECORY 02 *****)

c :- .

9 i gETR'U1rMT S AMD COMTROL1 PAGE 24 QUESTION 3.01 (1.00)

Which of the following ts true concerning the operation of the Reactor Trip (RT) and Reactor Trip Bypass (BY) Breakers?

a. The Train B trip signal trips RTB and BYB.

b. To allow testing of the RT's. BOTH BY's may be closed while the teactor is at power.

c. Tripping is accomplished by an undervoltage relay, normally held open by 46 volt de power frcm the logic panels,

d. The Train A trtp signal trips RTA and BYB.

CUESTION 3.0* (1.00)

Which of the following Reactor Protection signals is actuated by a 2/4 coincidence?

a. Pressurtser Low Pressure.

b. Pressuriser High Level.

c. Low RCS riow,

d. RCP Low Voltage.

(***** CATE00RY 03 CONTINUED ON NEXT PACE *****)

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. 1___LEETruerMIg At:D cogIggk1 pAgg g5 QUESTION 3.03 (1.00)

Match the following permissives in COLUMN A with their function in COLUMN B. ( t.e. e. 6 )

COLUMN A COLUMN B

a. P-6 1. Below setpoint allows operation with one RCP off,

b. P-9

2. Celow setpoint blocks reactor trip due

c. P-10 to turbine trip.

d. P-12 3. Allows manual block of high steam flow Si and blocks steam dumps below setpoint.

4. Allows manual block of SRM trip above setpoint. .

S. At setpoint allows manual block of IRM and P.".M l o w setpoint trips and rod stops and removes P-7.

l QUESTION 3.04 (1.00)

Whtch of the following to a correct statement concerning CCCS equipment operation?

a. A Safety injection signal starts the diesel generators and aligns them to the shutdown boards,

b. A Safety injection signal starts the charging pumps. SI pumps.

! RHR pumps and Containment Spray pumps.

c. A Safety injection signal opens the' valves in the boron injection tank rectreulation lines,

d. A Safety injection signal will initiate a reactor trip. Isolate the feedwater system, and start the auxiliary feedwater pumps.

, QUESTION 3.05 (1.50)

! On the attached pressuriser level control diagram. Figure 3.1. fill in the setpoints or actions indicated a. through f. (Put your answers on your answer paper.)

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(***** CATCOORY 03 CONTINUED ON NEXT PACE *****)

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1 1E2IEEMEKI1_Ada COMTEQL2 PAGE 06 i

QUESTION 3.06 (1.00)

Match the following total temperature error signals (Tavg -Tref) in l

COLUMN A to their corresponding automatic rod speed signals in COLUMN E.

COLUMN A COLUMM B

s. -0.5 T 1. O steps / min.

b. -0.0 T 2. 8 steps / min.

c. -4.0 T 3. 40 steps / min.

d. -6.0 T 4. 48 steps / min.

5. 72 steps / min.

QUESTION 3.07 (1.50)

Tor the following statements concerning the rod centrol system, indicate whether each is TRUE or TALSE.

a. The programmed Tavg is increased linearly with power.

b. The nonlinear gain unit adjusts circuit gain depending on turbine power.

c. The rod drive mechanisms receive their power from two parallel motor generator sets through two series Reactor Trip Breakers.

QUESTION 3.08 (1.80)

TRUE or TALSE7 VA=W4

a. The AAep and reheat stop valves are closed by the Overspeed Protection Controller,

b. The governor and intercept valves are the ONLY valves closed by the Overspeed Trip Mechanism.

n.ellIt,

c. The A+ep reheat stop, and intercept valves are opened when the turbine is latched.

,V Valves opened when the turb y s- 6: latched.

(***** CATECORY 03 CONTINUED ON NEXT PAOg maans)

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-3 L MSIEtNFMTR AMD CQ'IgQ(1 d

QUC5 TION 3.09 (1.00)

What signals are sent to the reactor protection system to indicate a turbine trip?

a. Covernor valves closed and Auto-stop-ott pressure low.

4

b. J45p valves closed and Auto-stop-ott pressure low,

c. Covernor valves closed and EHC fluid pressure low.

A ,.wVa

d. Shop valves closed and E!!C fluid pressure low.

QUESTION 3.10 ( .75)

Unit I is operating at 45% power with all systems in automatic control. For each of the following conditions, give the direction

of inttal rod motion. (In. Out, or None)

a. A steam generator Atmospherte Rettef Valve falls open,

b. A feedwater heater string becomes isolated,

c. The lower detector of the power range channel N-44 falls high.

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! (***** CATCOORY 03 CONTINUED ON NCY.T Pact *****)

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F .

. .___....3-,..-.

. , , , , . . . . ...._a._..r.,.,

P .,w,, C , ,J

. . ,s . .. .

.. . . . . I ,

s. 1 (...n0)w

: t:t tht saswer whi ch aa s t les-ly de:: rites the normal :tatu: of th O t v a:u C u::.p : y s t s ta . A:.su::.e untt i .; at 1 0 0 */. p o w e r

a. "4 . /Itin ts dtsarmed in the steam pressure mode with load i sjection controller selected to effect steam dump valve

.. ; e r a t i o n upon raceipt of an atror signal between Tav3 and Tav load of >/: 5 T.

v.eesa N

'J . '" h e s '/ 0 1.s ta la da;armis) in the T a V J rao d e wtth t h e We ittp

't,';er :' ;t+d t .. ff ' :t 3 ; ' ;.:; vitve .; 4 : ati:n a p e. n J - - t i A i. S . j .. . . a,.e n . .a / J s fi . r at a a ;3 'J

64. '/- .i . J .a

. 6 avg

, a t W. .

. T'. .j.t a .. : . :. a r a. s l ta !!. ; Ta > J mu !.s .v.th the ! ad r e 3 e : t t . :.

. . . . . ..: .;.e:tsi t: +!fe-t . '. : ; ra 1. ; val"e . ;. - t a t : ;n 1; . n t v e + t ,:. t v an trrst J a g e. a 6 b e t .v .s -s n T a s y and Ti f <. C f if

.1 ;.ai -;e:t.>n a r m . r. ; .: ;na! ; 3 s p .. ! t e d .

.. T;. , . i ; L . r.. s. i . . a rt:.; J .a t r. 2 T.v; .:,,, J e n i t ?. : .. ;s J 4 .s j e .; t t a n

. t . . '. ; r . 1 '. s . 1 3 r. : ?s- ; > r. ;

. -- - - 3:.11 '. , t : : ~ ~. T:t 3 2n1 T a. . .. ! ,.....a'. .r 1 . . ! ' . : . a .. ; t. ....;t ; .r s:a .i u:ap va1vs

- .. .: .. at ;/ ';... t >t;t

1 ! aid r .' s st s a, 3 : m. t r. ;

l

'J ! . . . ; .! th4 f si ; ra t n j :tatements :or.carning th> " t + va Ot.mp Ce n t r e !

Cy,t .;. I$ eerr90t' 3 Tho 3 team dump v.1 v94 !3i! opan on loss of 31r

b. In vrles tu ecv!Jvwn beloa 010 T. tha atuam J.2:ap rav d e avlactor

.a.t:h must Lt ra n.; > n t i r t l y t 1L rn to " it s t a t " and returnef to

' J '. ).a Pressute"

4. Wi. i n s ai t !. e Tav; raole, thi s t o a:a J a.a p . ars 4 r .a er d by the ts.etst trip b r .r t'. o r s . .a r i n t n j .

d. Ia thJ lJ3d .' 1 j s . I 1 s .1 . :. s 1 J . ! !. J 4 t . A i!. d '. ... e ' a ' / 4 e aL J / iJeJl J 3, in 2: t 3.jnst t) trip '.?n !"; a n ! t 4; x . 6 ; . 3 '. t, . . ,, i t. >'.

. ,, s n t ;. e .aagnituda cf the error ;;J.tal.

(aeaie cA;;ger,y o ; n: 7 g :19,f ce c;j ;-v,7 p Act eeeie)

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., 1 t urTEMENTE AMD CCMTECLE PAGE 29 QUESTION 3.13 (1.00)

What signal " ARM 3" the Steam Dump system when in the Tavg mode?

a. 5% / min.

b. 10% / 2 min.

c. 10 T Tavg/Trei error,

d. 13 T Tavg/ Tref error.

QUESTION 3.14 (1.00)

If the Intermediate Range Compensating Voltage fatted low, would the indicated flux level be H!CHER. LOWER, or the SAME AS the actual flux at ,

a. 100% power indicated on the power range.

-10

h. 10 amps indicated on the intermediate range.

QUESTION 3.15 (1.00)

Which of the fctlowing statements describes the signal path from the Source Range detector to the Source Range level meter on the MC87

a. Detector. Pre Amp. Discriminator. Log integrater. Meter

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

c. Detector. Pre Amp. Log Integrator. Discriannator Meter

d. Detector. Log Amp. Meter l

(sesse CATE00RY 03 CONTINUED ON NEXT PA0g samas) l _ _ _ _ _ - _ _ - - _ _ _ _ _ _ -

. a IMshauvrM*1_Ag2 CQKICQki PACE 30 QUESTION 3.16 (1.00)

What is the signal used to develop the Control Rod Insertion Limits?

a. Auctioneered high Tavg and Auctioneered high detta T.

b. De;ta T and T&vg.

c. Auctioneered high detta T.

d. Auctioneered high Tavg.

QUESTION 3.17 (1.00)

Which of the following is a function of the Power Range Instruments?

a. Provides 20% power rod withdrawal stop.

b. Provides input to the Rod Insertton Limit computer.

c. Provides indt stion of Startup Rate on M-4.

d. Provides input to OTdT protection circuit. ;

QUESTION 3.18 (1.00)

T!te Tl!REE input signals to the Steam Generator Water Level Control ares

a. Tavg. compensated feed flow, uncompensated steam flow.

b. Teod flow, compensated steam flow. Teod pressure.

c. Compensated feed flow, water level, compensated steam flow.

d. Uncompensated feed flow, compensated steam flow, water level.

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f 2'- 13&IguurnTs_&gq_qqgIggig pAgg 3g l

QUESTION 3.19 (1.00) l Considering only the Steam Generator Level Control System, what

would be the response of the INITIAL feedwater flow to the S/C 1f the controlling S/C pressure transmitter fatted LOW during 50% power

operations?

a. The flow would decresse due to the loss of the steam pressure l

input to the steam flow signal.

b. The flew would remain the same due to the steam pressure not affecting the steam flow,

c. The flow would increase due to the steam pressure input to the feed control valve posttlon centroller,

d. The flow would increase due to the loss of steam pressure input to the steam flow signal.

CUESTION 3.00 C .50)

TP.UE or PALSE?

The indicated steam generator level would INCREASE If the diflerential pressure transmitter reference leg temperature decreased by 10 degrees. ( ASSUME actual level does NOT change.)

QUESTION 3.21 (2.00:

How would a significant increase in the following affect the OTdT setpoint? ( Increase. Decrease. or No effect ).

a. Tavg.

b. RCS pressure.

c. Delta flux.

+ w..., -... T w\o (sames CATE00RY 03 CONTINUED ON NEXT PACE asams)

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l INg;UfMEMTE AMD CQEIEQL1 PAGE 32 l

QUESTION 3.22 (1.00) f Which of the following Radiation Monitors has an automatic isolation i function?

a. Tuol Pool Radiation Monitor (RE-90-102).

b. Condenser Vaccuum Pump High Range Air Exhaust Monitor tRE-90-99).

c. ERCW Liquid Effluent Monitor (RE-90-133).

d. Shield Building Vent Monitor (RE-90-100).

QUESTION 3.03 (0.00)

TRUE or PALSE

a. The Source Range instrument uses a itssion chamber for detecting neutrons.

b. Compensating current is used in the Intermediate Range detector to eliminate the gamma signal contributton.

c. The lower Range High Flux Setpoint Deviation alarm ts artomatically defeated below 50% power.

d. If control rods are in automatic and reactor power is at 50%

when Power Range channel N!-42 falls low, the rods will automatically drive out.

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11 fusTEUME"T1_&E2_C2HIEQL1 PAGE 33 l -

i QUESTION 3.34 (1.00)

Which of the following statements about temperature detectors is correct?

a. The thermocouple is connected to one leg of a bridge circuit and as the temperature changes the output voltage across the bridge changes,

b. When a thermocouple falls open it will respond in the same manner as an T.TC and will indicate a full scale reading on the meter,

c. When a faster responding temperature signal is needed a direct immerston (wet bulb type) RTD is used instead of the thermowell mounted RTD.

d. A RTD is comprised of two wires of dasssmilar metals in contact with each other and generates a voltage that is -

proportional to the temperature difference between the cpan ends of the wires.

CUESTION 3.05 (1.25)

Select the Emergency Diesel Generator (EDO) PARAMETER that will be adjusted by the METHOD and conditions listed. (i.e. f. S.)

METHOD EDO PARAMETER

a. Voltage adjust switch WITH EDO 1. KW output breaker open

2. VARS

b. Voltage adjust switch WITH EDO output breaker closed and normal 3. Voltage 6.9 kv shutdown board supply breaker open. 4. Frequency

c. Voltage adjust switch WITH EDO output breaker closed and normal 6.9 kv shutdown board supply breaker closed.

d. Speed switch WITH EDO output breaker open.

e. Speed switch WITH EDO output breaker and normal 6.9 kv supply breaker closed.

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

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PACE 34

, b1METEt"'MI1_MID CQ!!IE2L1 I

l l QUESTION 3.26 (1.00)

Which of the following, by itself, will cause an automatic start of the Turbine Driven Auxiliary reed Pump?

a. Actual Low-Low level in Loop e3 Steam Generator.

b. Loss of either fiain Teod Pump G 50% reactor power.

c. 35 seconds after a loss of offsite power.

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d. Immediately after a loss of offstte power.

QUESTION 3.37 (1.00)

Which of the following t s !!OT a purpose of the time delay in

' r after a reactor trip?

trtpping t h e ma i n 1" *

G 'onake-9 tur-a Reduce arcing of the generator output breakers,

b. Prevent overspeeding of the main turbine.

c. Prevent lifting Steam Generator safeties following the trip,

d. Remove excess steam from the motsture/seperator reheaters.

(8888 END OT CATE00RY 03 *****)

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1 PEOCEDUEEE_ _E2' MAL. AEEQEMAL. EMEEGEECY AMD PACE 35 EA21QLQG1 CAL _C2E'EEQL QUESTION 4.01 (1.00)

It is necessary to dilute 200 ppm to get the critical boron concentration prior to pulling the control banks. Prior to the dilution the source range instruments read 30 and 37 cps. After diluting 100 ppm the same instruments read 62 and 75 cps. Which of the folicwin3 is the proper operator action in accordance with C01-07

a. Stop the dilution and borate back to the original count rate,

b. Stop the dilution and evaluate the situation.

c. Continue the dilutton and continuously monitor the count rate,

d. Continue the dilution and recalculate the CCC.

CUESTION 4.00 (1.00)

I According to a precaution in 001-g during solid plant operatton, what is the minimum Reactor Coolant System (RCS) , temperature +,t- 'Jk *k

.4 .1,5 s ai v a Reactor Coolant Pump (RCP)?

epests4

a. 100 F

b. 160 F

c. 190 r

d. 250 t QUESTION 4.03 (1.00) rtil in the blanks in accordance with the Precautions Section of the " Col's",

s. A load change rate of +/ _____%/ min. or a step change of should not be exceeded,

b. The boron concentration difference between the pressur1:er and the RCS must not exceed _____ ppm.

c. The maximum allowable heatup rate for the RCS is _____deg /hr.

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4* ECOOEQUEEi ECEL!1WEEQEL'.1LJEEQEECL.&EQ PACE 36 EAQ1QLQG1 CAL _CCEIECL l '

QUESTION 4.04 (1.00)

COI-2 " Plant Startup from Hot Standby to Minimum Load" states that the shutdown banks must be at the fully withdrawn position whenever posittve reactivity is being inserted, except when . .

a. the Shutdown Margin has been calculated to be 900 pcm.

b. the RCS is borated to the cold shutdown concentration with plant l cooldown in progress.

c. the reactor is an the source range with the High Flux at Shutdown alarm operable,

d. the actual boron concentration is greater than the predicted crstical boron concentration.

CUE TION 4.05 (1.00)

TT.UE or TALGE7 During solid plant operations with pressure being mainta'ined by the low pressure letdown valve. TCV-62-81 in automatic ...

a. If the RHR system pressure exceeds 700 psig, the RHR suction from Loop 4 hot leg valves will close.

b. with RCS pressure at 300 psig and no steam bubble in the pressurtzer, it is permissible to isolate the RHR suction line from the RCS.

QUESTION 4.06 (1.00)

During normal CVCS operatton, which of the following is an abnormal condition and would require operator action to correct?

a. VCT pressure is 15 psig.

b. The temperature of the fluid leaving the letdown heat exchangers is 137 T.

c. The RCP seal injection water tempeature is 100 F and flow to the seals is 8 gpm/ pump.

d. RCP seal differential pressure is 300 psid.

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

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4. PRQOEQQEE1___EQE'&L._AggQEg&

d L EMERGESCY,_&gQ pact 37 EA21QLQGlC&k_COEIEQL QUESTION 4.07 (1.00)

At what point during plant cooldown from hot standby using CO!-3B is the remaining Turbine-driven Main reed Water Pump removed from service?

a. Tavg is between 540 T and $25 T.

b. Main steam pressure drops to 1000 psig.

c. Tavg is between 440 F and 425 T.

d. RCS pressure is between 450 psig and 425 psig.

QUESTION 4.08 (1.00)

When an RCP is stopped, either Component Cooling Water through the thermal barrier, or :eal water to the ROP must be continued until the ROS temperature ts reduced below: (Choose the most correct answer)

a. 150 T.

b. 160 T.

c. 170 T.

d. 175 T.

(***** CATECORY 04 CONTINUED ON NEXT PACE *****)

r t .

1 EEQ2E2EEEE - EQZUEL. AEUQEMbk _EMEEGEEC1_&MQ PAGE 38 E121.QLQ21. CAL _CQEIEQL QUESTION 4.09 (1.00)

Ustng the following 5 actions, which of the below sequences is correct for STARTING the first Control Rod Drive M.G. set?

1. Plash the field.

2. Close the Aux. 150 VAC supply breaker to rod drives.

3. Adjust generator voltage.

4. Close the motor circuit breaker.

5. C. lose the generator circuit breaker,

s. 1. 2, 3. 4. 3.

b. 5. 3, 4, 1. O.

c. 4, 1, 3. 5, 2. ,

d. 4. O. 1. 3. 5.

QUESTION 4.10 (1.00)

Whtch of the following statements concerning the procedure for a dropped RCCA is correct?

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

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

c. If two or more RCCA's have dropped, manually trip the reactor and proceed in accordance with EP-1.00.

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

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

r

. i.__EECCEQUEEE EC'"AL. ARECEMAL. E"EEGENCY AN2 PACE 39 RAOLQLCG1CAE CQE"ZE2L QUESTION 4.11 t!.00)

During an inadvertent dilution acetdent while at 100% power, which of the following will be the most probable cause of a reactor trip?

a. Pressurt:er low pressure.

b. Over-temperature dT.

c. Over-power dT

d. Power range monitor positive rate.

OUESTION 4.10 (1.00)

Which of the following would be a symptom of a Power Range instrument.

failing HIGH?

a. Rods stepping out.

b. Tavg increase.

c. OPdT reactor trip,

d. Rods stepping in.

QUESTION 4.13 C1.00)

Prter to a reactor startup with normal operating temperature and pressure the following RCS leakages exist. Por each leak rate below, indicate ti you gg3)) STARTUP or REMAIN SHUTDOWN.

1. C.5 GPH from a cracked weld on a narrow range temperature instrument manifold.

2. 1.0 GPM from a manual valve packing gland.

3. 0.4 CPM tube leakage en one Steam generator.

4. Leak from an unknown source of 1.2 CPM.

(***** CATECORY 04 CONTINUED ON NEXT PACE *****)

s _. .

. 4 .* J ALJEEGENCY ANQ PC2CECHEE1_ ._HQE!/iALmaaM2* ,PAGE 40

! EA21CLCG1 CAL _C2ET.a2L QUESTION 4.14 (1.00)

If the minimum temperature for criticality Technical Specification is violated, what option is available?

a. Restore to within its limits within 15 minutes or be in Hot Standby within the next 15 minutes,

b. Restore to within its Itmits within 15 minutes or be in Hot Standby withtn the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. Restore to within its limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in Hot Standby within the next 15 minutes.

d. Restore to within its limits within 15 minutes or be in Hot Standby within the next 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

CUESTICN 4.15 (1.00)

If CCW flow to the RCP motor is lost, at what Upper or Lower bearing tempeature must the effected RCP be stopped?

l a. 180 T.

b. 185 T.

I

c. 300 T.

d. 02S T.

QUESTION 4.16 (1.00)

Which one of the following is a symptom of " Rods fail to insert following a decrease in turbine load"?

a. Low pressurizer pressure.

b. " Pressurizer level high backup heaters on" alarm.

I

c. " Reactor coolant loops Tref-Tauct, high-tow" alarm and Tre!-Tavg * +5 T.

d. Rod insertion low limit alarm, i

j

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

i L_ _

r .

. 1 __EEQCE2MLEE - SCE'&L l _&Eu2EM&k._EMEEGEEC1_&E2 PAGE 41 i EL21QLQGLCik_.C2EIEOk QUESTION 4.1? (1.00)

An operator receives the following whole body exposures of radiation:

30 RAD of Gamma 4 RAD of rast Neutrons 3 REM of Thermal Neutrons What is the total dose the operator received? Show all work.

QUESTION 4.18 (1.00)

Which of the following would be a result of a loss of a RCP with reactor power at 30%? Assume No operator action.

s. Tavg increase.

' " effected S/G level increase.

c. Reactor trip,

d. Unoffected loop flow decrease.

l QUESTION 4.19 (1.00) l Match the following RCP seat alarms / indications in COLUMN A to the causes in COLUMN B. (i.e. f. 6.)

COLUMN A COLUMN B

a. #1 Seal high flow alarm 1. Increased flow to RCDT.

[ b. et Seal low flow alarm 2. Loss of injection water.

c. #2 Seal high flow 3. #1 Seal dP too low.

l

d. e3 Seal high flow 6 Ek. 4. Increased filling of seal g standpipe.

ss O

Y C A'toC\n I) ko\"

s

$)d #

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

1__E2QCE2'J.1ES_. EQEMLL._AMQEMAL._EME1QMCY AND PAGE 42 EA21QLQG1 CAL _CQEIEQL QUESTION 4.20 (1.00)

Which of the following would NOT require Emergency Boration?

a. Fa11ure of a control rod to fully insert following a reactor trip.

b. Excesstve control rod withdrawal when at power.

c. Failure of Boric Acid Flow Controller TC-62-139 to function properly.

d. Uncontrolled reactor heatup following a reactor trip.

QUESTION 4.21 (1.00)

Arrange the following Critical Safety Function Status Tree colors .

by nu= bering in crder of priorsty. Place answers on answer paper,

a. Creen

b. Yellow

c. Red

d. Orange .

QUESTION 4.22 (1.00)

FR-S.1 " Response to Nuclear Power Generation /ATWS" is entered from what procedure?

a. ES-0.1 Reactor Trip Response.

b. E-1 Loss of Reactor or Secondary Coolant.

I

c. FR-C.1 Response to inadequate Core Cooling.

I d. E-0 Reactor Trip or Safety Injection.

C***** CATECORY 04 CONTINUED ON NEXT PAGE *****)

- L ._u.po.c..e k . _ A a no n..n .n .e .- _ .n.e .e.s. d n. c.i.e 1. , _ e J.. r. n .,- d c..v _ i.d a PACE 4.,

ne,

. .n s 2.1w 3.s 2e.r. e p . ,._c.e.sd*tu..

QUESTION 4.33. (1.00)

Tc!!owing a raactor trip, how much Doron (gallons) must be a d d e :1 for each control rod not fully anserted?

a. 350

b. 100-

-... s.C 4

f.

1.

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

.i s ] n, )

!h . : h s i - t ha fcllcwing 1: MOT an immedtsie operator a:* ion for a '

. :: a f e t y I n j e s t i o n as stated in C-0?

a. *1 e r : f y C c n't a i n m e n t II:lat:on.

. C.. : : :. ' T a v ; .

, '.*: r i l y AT':!: status d .. Verafy Steam Cump; actuated.

,w,..--.,..

a _ a . s s.. . ..a ( 1 . 0 ns,3 T2U2 or FALS"?

s

/

a. Ths transfer of ECCS suction to the containment cump t:

! a c t omp l i s h e d whe n R'r/CT - l e v e l is ( C M'i .

.. 'ilh e nf 2*.13T level raaehe 0 * '. . the Centa:nment 3; ray Pumps are chtf!ed.to the'.Contataceat Camp.

OUCGTION l.26 ( .503-

......-,,I

._ . .A,-r, Araa: 'whare[do;e:rato: are 500 mr/hr a:e required to be locked and access controlled.

(*225* CATECOP.Y 04 COMTINUED ON NEXT PACC *****)

m

PAGE 44 LEEQCE2MEEi XQ2!JEkdEHQE!JAk T.l4EEGEMCY._AED.

EAQ1QLQ21 CAL CCET.E2L QUESTION 4.27 (1.00)

Which of the following conditions would require re-initiation of safety injection according to ES-0.2 " Safety injection Termination"?

RCS PRESSURE SUBCOOLING PZR LEVEL %

a. stable 25 15

b. increasing 45 IS

c. stable 45 35

d. stable 45 30 QUESTION 4.28 (1.00)

Prom the choices below pick one that best completes the following statement A trip after a long period of reactor shutdown leaves little decay heat to be removed thus causing the possibility of excessice cooling of the reactor coolant if too much feedwater is being added. The operator should NEVER restore the steam generator water level, after a plant trip, at the cost of a reduction of the .

a. plant pressure

b. shutdown margin

c. CST level

d. steam generator pressure.

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

9

4* paQCE22EE1 - UQCZAL _AlHQEM&ku_EMEEGEEOl_&M2 PAGE 45 E&21QLCGiC&L_CQuItQL QUESTION 4.29 (1.50)

For the power levels in Column A, find the one associated conditions or actions in Column B, as stated in GOl-SA.

COLUMN A COLUMN B

a. 20 % 1. p-8 light goes out.

b. 30 % 2. P-9 light goes out.

c. 35 % 3. Observe turbine startup drains closed.

d. 50 % 4. Open HP drains to the No. 1 heater shells.

5. Start two condensate demin pumps.

6. Prtor to exceeding % power, steam generator chemtstry must be below the libits for exceeding thts specific power level QUESTION 4.30 (1.00) in accordance with the " Loss of Reactor or Secondary Coolant" procedure, E-1, if one charging pump is operating and RCS pressure is uncontrollably decreasing, at what RCS pressure must the Reactor Coolant Pumps be stopped?

a. 1450 psag.

b. 1350 psig.

c. 1250 psig.

d. 1150 psig.

C***** END OF CATEGORY 04 *****)

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

EQUATION SHEET f = .ua v = s/t Cycle afficiency = (.'iet werx cut)/(Energy in) w=q s = V ,:

1/2 at-7 .

g x-a = (Vf - V,)/t A = 1:1 A=Ae3 KE = 1/2 mv PE = mgn vf= V, + at * = e/t 1 = tn2/t1/2 = 0.593/t1/2 1/2" " M

n0 2 y,y j ; + (; )j A= 4 [(t cE = 931 m -Dt m = V,yAo ,

O Q = mCoa:

6 = UA A.T I*I Q Pwe = W,2 I = I,10** E T7L = 1.3/u HVL = -0.593/u .

P = o,- P 10,sur(:)

P=Pe*' o SUR = 25.05/T SCR = 5/(1 - K,g)

CR x = 5/(1 - K ,ffx)

SUR = 25e/t* + (a - o)T CR;(1 - K ,g)) = CR2 II ~ *eM 2)

. T = (1=/a) + [(s - s /Io] i M = 1/(1 - K,g) = CR)/CR 3 i u(, - a) M = (1 - K ,go)/(1 - K ,g3)

T = (3 - s)/(Ia) SCM = ( - K,g)/K,g a = (K,ff-l) A,ff = d,fdK,g t' = 10 seconcs

I I = 0.1 seconds

=[(t*/(TK,g))+[I,g/(1+IT)]

Ijd j = I k 2 ,2 ge 2 P = (:av)/(3 x 1010) Id j qq 2

= sti R/hr = (0.5 CE)/0 (meten)

R/hr = 5 C2/d2 (f,,g) .

Watae Parneters Mises11aneous C:nve.siens 1 curie = 3.7 x 1010:3, 1 gal. = 8.345 lem.

1 gal. = 3.7B litars 1kg=2.2110m$Stu/nr 1 np = 2.54 x 10 1 f:4 = 7.48 cal. 1 :::w = 3.41 x 100 3:u/hr Density = 52.5 lbrp/ft3 Density = 1 gm/c9 lin = 2.54 :::s Heat of vacorization = 970 Stu/lem *F = 9/5'c + 32 Heat of fusion = 1 4 Stu/lem 'C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. Hg. 1 STU = 778 ft-1bf 1 ft. H 2O = 0.4335 luf/in.

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i REACTOR CCOLANT SYSTEM

~

3/4.4.6 REACTOR COOLANT SYSTEM LEAXAGE LEAKAGE DETECTION SYSTEMS LIMITING CON 0! TION FOR OPERATION 3.4.6.1 The following Rea:: tor Coolant System leakage ostection systems sna11 be OPERABLE:

a. The lower containment atmospnere particulata racioactivity monitoring ;

system.

b. The containment pocket sump level monitoring system, and

c. The lower containment atmosphere gaseous radioactivity monitoring

. system.

APPLICASILITY: MCDES 1, 2, 3 and 4.

. ACTION:

With only two of the above required leakage detection systems OPERABLE, operation may continue for up to 30 days croviced grab samples of the containment atmosphere are obtained and analyted at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the required gaseous R16 or particulate radioactive monitoring system is inoperable; otherwise, ce in at least HOT STAND 8Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REOUIREMENTS 4.4.6.1 The leakage detection systems shall be demonstrated OPERA 8tE by:

a. The lower containment atmosphere gaseous and particulate monitoring systems-performance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3, and .

b. Containment pocket sumo level monitoring system-perfonsance of CHANNEL CALIBRATION at least once per 18 months. .

0

-~

MAk 251982 3/4 4-13 Amendment No. 12 SEQUOYAH - UNIT 1 i

i k

4

-- - . - - - - , ._ , . - - .--,_ ,n, _ _ , _ , , , . , ._,-_,._-_..,_-n ,_ - - ---_m,,,._ _ - _ _ _ - ,

~

/

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CON 0! TION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE. BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE,

c. 1 GPM total primary-to-secondary leakage through all steam generators and 500 gallons per day through any one steam generator,

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and

e. 40 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 + 20 psig.

f. 1 GPM leakage at a Reactor Coolane System pressure of 2235 1 20 psig K16 from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.

APPLICABILITY: MODES 1, 2, 3 and 4 -

ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STAN08Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE SOUNOARY LEAKAGE, and leakage frem Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANOSY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the aeove limit, isolate the hign pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least two closed manual or coactivated automatic valves, or be in at least HOT STAN08Y within tne next 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REOUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within R16 eacn of the aoove limits by:

hiAR 251982 SECUOYAH - UNIT 1 3/4 4-14 Amenoment No.1:

REA; TOR CCCLANT SYSTEM SURVEILLANCE REOUIREMENT5 (Cor.tinued)

a. Monitoring ne lower containment atmosobere particulate racioac-ivity monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D. Monitoring the containment pocket sump inventory and disenarge at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c. Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump seals wnen sne Reactor Coolant System pressure is 2235 ; 20 psig at least once per 31 cays with the modulating valve fully open. Tne provisions of Scecification 4.0.4 are not applicaole for entry into Mode 3 or 4. R16

d. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

e. Monitoring the reactor head flange leakoff systes at least once per

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

,e-4.4.6.2.2 Each Reactor Coolant System Pressure Isolation valve specified in

- Table 3.4-1 shall be demonstrated OPERABLE pursuant to Specification 4.0.5, except that in lieu of any leakage testing requirements required by Specification 4.0.5, each valve snail be demonstrated OPERA 8LE by verifying leakage to be within its limit:

a. At least once per 18 months.

t. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN K16 for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or more and if leakage testing has not Deen performed in the previous 9 months. .

c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve.

d. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow tnrough the valve.

The provisions of Specification 4.0.4 are not applicaele for entry into M00E 3 g, or 4.

t ft.An 2 o.1982 SEQUOYAH - UNIT 1 3/4 4-15 Asenoment No.12

TABLE 3.A-1 REACTOR COOLANT SYSTEw PRES 5URE ISOLATION VALVES VALVE NUMSER FUNCTION 63-560 Accumulator Discharge 63-561 Accumulator Disenarge R16 63-562 Accumulator Discharge 63-563 Accumulator Disenarge .63-622 Accumulator Olscharge 63-623 Accumulator Disenarge 63-624 Accumulator Discharge 63-625 Accumulator Discharge 63-551 Safety Injection (Cold Leg)63-553 Safety Injection (Cold Leg)63-557 Safety Injection (Cold Leg)63-555 Safety Injection (Cold Leg) .

Residual Heat Removal (Cold Leg)63-632 Residual Heat Removal (Cold Leg)

R16 63-633 63-634 Residual Heat Removal (Cold Leg)

Residual Heat Removal (Cold Leg)63-635 Residual Heat Removal / Safety ^

63-641 Injection (Hot Leg)

Residual Heat Removal / Safety 63-644 *'

Injection (Hot Leg)63-558 Safety Injection (Hot Leg)63-559 Safety Injection (Hot Leg)63-543 Safety Injection (Hot Leg)63-545 Safety Injection (Hot Leg)63-547 Safety Injection (Hot Leg)

Safety Injection (Hot Leg)63-549 Resiaual Heat Removal (Hot Leg)63-640 Resioual Heat Removal (Hot Leg)63-643 87-554 Uoper Head Injection 87-599 Upper Head Injection 87-560 Uoper Head Injection 87-561 Uoner Head Injection 87-562 Upoer Heac Injection 87-563 Uccer Head Injection FCV-74-l' Residual Heat Removal Residual Heat Removal FCV-74-2" l Al*

I .

"inese valves co net have to De leak tested follcwing manual er automatic actuation or flew througn tne valve.'

t e

biAk 2 6 ICS2 3/4 4-15a Amendmene No. 12 SEQUQYAH - UNIT 1

r PAGE 46 1__EE1301E' EE_QE_t2CEEAE_ECEEE_EEEI_QEEEAI1QM.

ur,m: 2Eum1CE._EEA;_IE&tEEEE_at2_EEE1D._EEQW ANEWERS --

S EQUOY All 1&2 -85/05/20-JACCAR. T.

ANSWER 1.01 (1.00) f/.

o g j Q) M j j C.

RETERENCE WNTC, HTTT. Chap. 12, p. 15 ANSWER 1.02 (1.00)

a. False . , a t .L us , D L %.h ~ n

b. Trus (d ML~ k*' a i - (0.5 ea.1 RETERENCE SONP, HTTT text p. 200 ANSWER 1.03 (1.50)

~

a. True

b. Talso

c. False (0.5 eachl RETERENCE WNTC Thermal Hydraulic Principles and Applications to the Pressurized Water Reactor Chapter 10, pp 32,38,49. Chapter 11, p 27 ANSWER 1.04 (2.00)

1. Convection

2. Radiation / convection (large Delta T)

3. Conductnon

4. Convec' tion (natural) (0.5 ea.1 RETERENCE SQNP HTTT text, pp. 191-206

i ' EELEC1ELEi_QE_EECLEAE_EQWEE_ELANT._QEEEAT.1QL PAGE 47 IEEEM22EEAMIC L _EEAI_IEANEEER_AER_ELEla_ELCW ANSWERS --

SEQUOYAH 1&O -85/05/60-JAGGAR, T.

ANSWER 1.05 (2.00)

a. Decrease

b. Decrease

c. Increase

d. Decrease 10.5 ea.)

RETERENCE SONPs HTrr, page 15 ANSWER 1.06 ( .50) .

TRUE RETERENCE System Descriptions, Chapter 3, pp 6 & 7 ANSWER 1.07 (1.00) c.

REFERENCE SQNP, HTTF, p. 11 & 12 ANSWER 1.08 (2.00)

a. 3

b. 2 (1.0 es.1 RETERENCE SQNP, Q & A Bank , see 1-11

1___EEINCIELEE_QE_EECLEAE_EQWEE_ELAMI_QEEEAI1QE. PAGE 48 IEERIA021MMICE. HEAI_Ig&ggEEE_MQ ELQ1Q_EEQW ANSWERS --

SEQUOYAH 1&3 -85/05/30-JACCAR, T.

ANSWER 1.09 (1.00)

1. D o* D pv h.14 equd Mtvdadmg o- ads ea '*96'*

3. C

3. D

4. A RETERENCE WNT. Chap. 3, pp. 3-44 to 3-53 ANSWER 1.10 (1.00) ,

c.

RETERENCE SQNP. T.S. 3.1.1.3: Revtew of Reactivity Coefficients p. 4 ANSWER 1.11 (1.00) b.

RETERENCE SONP, 6 Tactor Tormula lesson, pp. 3 - 5 ANSWER 1.13 (4.00)

a. I

b. 3

c. 3

d. 4 RETERENCE SQNP. Revtew of core poisons lesson, p. 6

1. P R I HC1 ELE 1_QE_EECLE&E_EQWEE_EL&E'L_QF,EE&T,1QEu pAGE 4g T,gERMODYM AMlQ1. HEAT 'IgiggEER_&gQ_ELQ1Q_ELQg AN3WERS -- SEQUOYAH 1&2 -85/05/20-JACCAR, T.

ANSWER 1.13 ( .50)

TALSE.

RETERENCE 3CNP. Review of Neutron Phystes Iesson, p. 3 ANSWER 1.14 (1.00) c.

RETERENCE 3CNP, Review of Neutron Kinetics, p. 5 ANSWER 1.15 ( .50)

TRUE.

RETERENCE SONP, Review of Neutron Kinettes, p. 5 ANSWER 1.16 (3.00)

a. 3

b. 4

[Cy e43)

C. 3 e R 4 3. , l. - hwelo~ p,,34 u t- 4 4# g RETERENCE GQNP. Suberttical Multiplication lesson, p. 5: Revtew of Kinettes, p. 3 ANSWER 1.17 (1.50)

a. D.7.1 (0.5)

b. See graphs attached. (0.5 es.) (1.0)

. 50h*P

, TI-28 , . , , . -., .

Figure B.7.1 7. J , . ,' 't 'e 3 ,

Page 1 of 1 CTCLE 3 J t; t C [

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e : : :

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1 2 2 3 2 2 d. d.g 0

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( W3 Mod Ul3::IMId l'.','dd ) u:31:::ggios w3 mod 1yloz

. SQNP TI-28 Figure B.6.2 ' * * "?' S Page 1 of 1 l8}i 8l' ['

Rev. 61 U, CYCII 3 EE sesas IIItit:

...... t 1a a aa aaaaa a OOoOO o O. O. O. 22W2W W .

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- +' 7 -

1 _EE1HC1 ELE 1_QE_EECLEAE_ECWEE_ELAEI_QEEEAI1QL PAGE 50 IEEEMOOINAM10E._EEAI_IEAEEEEE_AEt_ELEla_ELQW ANSWERS --

SEQUOYAH 110 -85/05/20-JACCAR. F.

REFERENCE SONP. Curve book. Fig's B.6.2. B.7.1 ANSWER 1.18 (2.00)

a. 3

b. 1 (1.0 ea.)

REFERENCE SONP, Review of Core Poison lesson. pp. 4 -

5 ANSWER 1.19 ( .75) 1 e

2. a

3. b (0.25 ea.)

REFERENCE SONP. Review of Core Poisons. pp. 4 -

7 ANSWER 1.20 ( .75)

a. 3 or 4

b. I

c. 4 REFERENCE SQNP. HTrr text, pp. 174 -185 ANSWER 1.21 (1.00)

a. TALSE

b. TRUE (0.5 ea.1

PAGE 51 L_EE1MC1ELEE_QE_EECLEALEQWELELAULQEEEAT.1QEu IEEEM2DluaMLCS. HEALIE&ggEEE_&gLELELLELQW

-85/05/20-JAGGAR, T.

ANSWERS -- SEQUOYAH 1&O RETERENCE General Physics. HT&TT, pp. 155 and 320 and Subcooled Liquid Density Tables ANSWER 1. (1.00)

C RETERENCE SQNP, HTTT text, pp. 23 -

24 4.

2- P L AMI_QEEj,GN_1NGLQQLNG_E&EElll3Q_EMEEGE'LT *._1111CM1 PACE 52 ANSWERS -- SEQUOYAH 1&2 -85/05/20-JACCAR. F.

ANSWCR 2.01 (1.00) d.

RETERENCE 3CNP System Description. Electrical Distribution, p. 5.7-13 ANSWER 2.02 ( .50)

! TALSE.

RETERENCE SONP System Description, Electrical Distribution, p. 5.2-13 ANSWER 2.03 (1.00) d.

RETERENCE SQNP System Description. Electrical Distribution, p. 5.3-3 ANSWER 2.04 (1.00)

)( . C.. I2t hben t e i f. 16 Ca c e t a Y~

RETERENCE SQNP System Description. RCS, pp. 12 & 13 of 38 ANSWER 2.05 (1.00)

a. 4

b. 1 4 c. 2

d. 5 RETERENCE SQNP System Descriptions, ECCS. p. 4.2-27 -

29: RHR. p. 4.1-17

L _gt. ANT D E S I GE_1.ECLEal.E2_i&EEnlED._EMERGEECY S Y STDJE PACE 53

-85/05/20-JACOAR, T.

ANSWERS -- SEQUOYAH 1&2 ANSWER 2.06 (1.00) d.

RETERENCE SQNP System Descriptions. RHR, pp. 4 & 9 of 11 ANSWER 2.07 (1.00) d.

RETERENCE SCNP System Description, CVCS, p. 14 of 40 Rol-34 .

ANSWER 2.08 C .50)

TALSE.

REFERENCE SQNP System Description, Tuel Handling, p. 9-6 ANSWER 2.09 (1.00) b.

REFERENCE SONP System Description, Component Cooling System, p. 7.1-17 ANSWER 2.10 (1.00) d.

RETERENCE SQNP System Descriptions, Condensate and Teodwater, pp. 9& 10 of 11

9

' ' 2- ' P L AMT QEilGE_lEOLU.21EG_1AEEU._AED E"ECGEECY SYSTEM 1 PACE S4 ANSWERS -- SEQUOYAH 1&2 -85/05/20-JACCAR, T.

ANSWER 2.11 (2.00)

a. TALSE

b. TALSE

c. TRUE

d. TALSE RETERENCE SQNP System Description, ECCS, pp. 4.2-7 & 8 ANSWER 2.12 (1.00)

C.

RETERENCE SQNP System Description, MainSteam, p. 7 of 10 ANSWER 2.13 (1.00) a.

RETERENCE SQNP System Description. ATW, p. 6 of 8 ANSWER 2.14 (1.00)

C.

RETERENCE SQNP Diesel Generator Handout

Ek&MT 2E11GH lECLU.D.lliG._1&EETY AND EMEEGEECY SY STIME PAGE 55

. 0-ANSWERS -- SEQUOYAH 1&2 -85/05/20-JAGGAR, F.

ANSWER 2.15 (1.00)

a. 3

b. 4

c. 2

d. 1 RETERENCE SONP Diesel Generator Handout ANSWER 2.16 (1.00)

a. TRUE B. TALSE RETERENCE 501 74-1Ai pp 5 & 7.

ANSWER 2.17 (1.00) d.

RETERENCE SONP System Description, RCS, p. 13 of 38 ANSWER 2.18 (1.00)

C.

RETERENCE SONP System Description. CVCS, p. 9 of 40 ANSWER 2.19 (1.00) i l c.

I

2' ELagI_QE11.qu_1ECLUDING_EAEET.LAMIL EMEEGENCLEY.EIEEE PACE 56 ANSWERS -- SEQUOYAH 1&O -85/05/20-JACGAR, T.

RETERENCE SONP System Description, RCS, p. 28 of 38 ANSWER 2.00 (1.00) a.

REFERENCE SONP System Desertption, p. 4 of 8 ANSWER 2.21 (1.00)

[

'RETERENCE ^

SONF System Desertption, ATW, p. S of 8 ANSWER 2.22 (1.00)

b. Il'* s O YC 34 4 C- * "I

C #dC ^ - '" (d 1 ~ I'3 RETERENCE SONP System Description, ATW, p. 3 of 8 ANSWER 2.23 (1.00) b.

REFERENCE SONP Condensate and reedwater, p. 10 of 11 o

ANSWER 2.24 (1.00) d.

RETERENCE CONP System Description, RHR, p. 4.1-3

p .

. 1_ PLA M E11CH 1NCLQQlEQ_1&ELIY'AND FMFRGMC1_111121 PAGE S7 I

ANSWERS -- SEQUOYAH 1&2 -85/05/20-JACGAR. T.

ANSWER 2.25 (2,00)

a. CLOSE

b. CLOSE

c. OPEN

d. CLOSE

e. No

f. OPEN

g. CLOSE

h. NO

1. NO

j. NO [0.2 ea.1 RETERENCE SQNP System Description. ECCS, CVCS, MNSTM, CCW ANSWER 2.25 (1.00) b.

REFERENCE SONP Sol 68.2 p. 2 of 11 ANSWER 2.27 C .50)

TALSE REFERENCE SONP System Descriptions, CVCS, p. 3-28 ANSWER 2.28 (1.00) 3.

RETERENCE SQNP System Desertption. CCWS, p. 7.1-4 ANSWER 2.29 ( .50)

TRUE.

. L _ _EL&gt DEsicN It!qggnit!c saggTY AMD gggggggcY SYSTEMS PAGE 58

-85/05/20-JAGCAR, T.

ANSWERS.-- SEQUOYAH 1&2 RETERENCE SONP System Description, Steam Systems, p. 10.1-11 ANSWER 2.30 (1.00)

C.

RETERENCE SQNP Sol 62.1B, p. 2 of 8 L

~~

t PAcg Sg 1.__1EEIRUMENTS AND CONIEQL1 ANSWERS -- SEQUOYAH 1&2 -85/05/20-JAGGAR. F.

ANSWER 3.0 (1.00) d.

REFERENCE 3QNr System Description, Reactor Protection, pp. 6 of 13 ANSWER 3.02 (1.00) 4.

RETERENCE SQNP System Description, Reactor Protection, p. 9 of 13 ANSWER 3.03 (1.00)

4. 4

b. 2 C. 5

d. 3 (0.25 ea.)

RETERENCE SQNP System Description, Reactor Protection, pp. 10 & 11 of 13 ANSWER 3.04 (1.00)

d. ,

RETERENCE SQNP System Description. Protection System. p. 11.10-32: ECC5. p. 4.2-10.11 l

i i

I l

l i

i 1- 1ESIEUMEMTK AE22CQEIEQL1 PACE so ANSWERS -- SEQUOYAH 110 -85/05/20-JACCAR, T.

ANSWER 3.05 (1.50)

a. 90%

b. High level alarm

c. 60%

d. 24.7%

e. Letdown line isolation or low-low level heater cutout

f. 578.2 T (0.25 ea.]

RETERENCE SONP PLS. .pp. 16. 28, 34 ANSWER 3.06 (1.00)

a. 1

b. 2

c. 3

d. S RETERENCE SONP System Description. Rod Control, pp. 6 & 7 of 11 ANSWER 3.07 (1.50)

a. TRUE

b. SA UG G \& 4.

c. TRUE REFERENCE SQNP System Description. Rod Control. pp. 5 - 9 of 11

. 2- IN11guMEMTS AND CQgigQki PAGE 61

' ANSWERS -- SEQUOYAH 1&2 -85/05/20-JAGGAR, T.

ANSWER 3.08 (1.50)

a. TALSE

b. TALSE

c. TRUE (0.5 ea.)

REFERENCE SONP System Description. Turbine Control, p. 5: E-H Control, p. 10.3-12, 10.3-20 ANSWER 3.09 (1.00) b.

REFERENCE SONP System Description, Protection System, p. 11.10-131 N

ANSWER 3.10 ( .75)

a. Rods out,

b. Rods out.

c. Rods in.

REFERENCE SONP System Description, Rod Control, p. 5 & 6 of 11 ANSWER 3.11 (1.00)

C.

REFERENCE SONT System Desertption, Steam Dump Control, p. 7 of 8 3

l

.- f MET 9_rvrffTE AMD CQEIEQL1 PACE 60 1___

ANSWERS -- SEQUOYAH 1a2 -85/05/20-JAGGAR. T.

ANSWER 3.12 (1.00)

d. o f C REPERENCE SQNP, System Description. Steam Dump Control System, pp. 2 - 8 ANSWER 3.13 (1.00) b.

RETERENCE SONP System Descriptions. Steam Dump Control, p. 7 of 8 ANSWER 3.14 (1.00)

a. Same

b. higher (0.5 ea.! ,

s '

RETERENCE SQNP System Descriptions. Excore Instrumentation, pp. 11 & 12 of 18 ANSWER 3.15 (1.00) a.

RETERENCE SQNP System Description. Excore Nuclear Instrumentation, p. 7 of 18 l

ANSWER 3.16 (1.00) ,

C.

RETERENCE GQ:!P System Desertption, Rod Control, p. 11 of 11 l

.. 11 iM2TguuruTs Ann eggIEQki PAGE 63 ANSWERS -- SEQUOYAli 1&2 -85/05/20-JACCAR. T.

ANSWER 3.17 (1.00) d.

RETERENCE SQNT System Description. Excore Instrumentation, p. 13 of 18 A!1SWER 3.18 (1.00) d.

RETERENCE SQ!!P System Description. SCLCS. p. 11.7-1 ANSWER 3.19 (1.00) a.

RETERENCE SQNP System Desertption. SOLCS. p. 11.7-8 thru 11.7.10 ANSWER 3.00 ( .50)

TALSE.

RETERENCE SQNP System Description. SCLCS. 11.7 ANSWER 3,21 (2.00)

a. Decrease.

l

b. Increase.

l

c. Decrease,

d. S :!!::P.oc beev..u. h e?o < 4. (0.5 ea.)

9. d k wu % d a o \r b '

I REFERENCE SQNP System Desertption. Reactor Protection, p. 11.10-24, 25 l

I l

l l

k.

I ngT2tn3rMT3 AMD CONTROLS PAGE 64

'. 2

-85/05/20-JAGGAR, T.

ANSWERS -- SECUOYAH 1&2 ANSWER 3.00 (1.00) a.

RETERENCE SCNP System Desertption. RMS, pp. 41 - 49 of 53 ANSWER' 3.33 (2.00)

a. TALSE

b. TRUE c . .':JHff IM A & 8*=4 8 O.

d. TALSE (0.25 ea.1 RETERENCE ^

SCNP System Descrtption. Excore Instrumentation, p. 6, 17, 10 of 18 ANSWER 3.24 (1.003 c.

RETERENCE Nuclear Power Plant Instrumentation Systems Manual. Ch. 4 ANSWER 3,05 (1.25)

a. 3 .

b. 3 l c. C

d. 4

e. 1 10.25 es.1

)

I RETERENCE SQNP Sol-82 s

l

o

3. IMETRg3gggIs _ AND qQgIggkg PACE 65 ANSWERS -- SEQUOYAH 110 -85/05/20-JACCAR, T.

AN3WER 3.26 (1.00) d.

REFERENCE SONP System Description. Auxiliary Teodwater System, p.3 of 8 ANSWER 3.27 (1.00) a.

RETERENCE GCNP System Desertption. Electrical Distribution, p. 5.2-6 i

s

4- PROCEBgREE - M C't" A L. ADNo m a L EL1M G M EL.MlE PAGE 66 oAB1oLocicAL CQgIEltk ANGWERS -- SEQUOYAH 1&2 -85/05/20-JAGCAR. T.

i ANSWER 4.01 (1.00) b.

RETERENCE SQNP 001-2. pg. 1 ANSWER 4.02 (1.00) b.

RETERENCE GCNP C01-1, pg. 3 ANSWER 4.03 (1.00)

a. 5, 10

b. 50

c. 100 RETERENCE SQNP CO!-1, p. 2: 001-5. p. 2 ANSWER 4.04 (1.00) b.

REFERENCE l SQNP 001-2. p. 2 ANCWER 4.05 (1.00)

a. TRUE b, FALSE

_ + _ - - _ - - - - - - - - . - - - - _ - . - - - - - - - - - - - - - - - . - - - - - - - - - - _ - - - - - - - - - - - - - - - - - - _ - - - - - _ - - - - - - . - - . - - - - - -

o j , [__ _ ERcc ED L'E E g . McPMAL.'ARNQ&g&L._LME&QENcY AND PACE 67 RiniStocicJ_L COMTEQL l ANSWERS -- SEQUOYAH 110 -85/05/20-JACCAR, T.

l l

i I RETERENCE l

SQNP 501-74.1. pp. 3. 4 l

l ANSWER 4.06 (1.00) a.

P.CTERENCE SQMP 301-C2.10, pp. 8, 9 ANSWER 4.07 (1.00) l t

1 c.

RETERENCE SQtlP Kol-3D. p. 15 l

0 i ANSWER 4.08 (1.00) l a.

l RETERENCE SQNP Sci 88.2, p.3 ANSWER 4.09 (1.00) c.

REFERENCE SQNP SO!-85.1A, p. 3 of 4 ANSWER 4.10 (1.00) a.

I RETERENCE SQNP A01- D. pp. 10 - 12 l

I i

l l

l

r-

&_ .PRCCEDURES - Monuit. AREORMAk._ EMERGENCY Af1D PAGE 68 o

gAQtaLoc1 CAL CQMT,1QL ANSWERS -- SEQUOYAH 1&2 -85/05/20-JAGGAR, T.

ANSWER 4.11 (1.00) b.

I RETERENCE l

SQNP A01-3D, p. 1 of 2 l

l ANSWER 4.10 (1.00) d.

RETERENCE SQHP AOI-4D. p. 1 of :

AN3WER 4.13 (1.00)

1. Shutdown

2. Startup

3. Shutdown

4. Shutdown (0.25 eachl RETEAu!!CE SQNP Technical Spectfications 3/4.4.6.2 l

ANSWER 4.14 (1.00) a.

RETERENCE SONP Tech. Spec. 3/4.3.1.1.4 ANSWER 4.15 (1.00) c.o=$~ 302 . <54 &co , Q gay $ h f' REFERENCE SONP A01-tSA, p. : of 3 l

__ ~

I 0 A. P ROCEDL'EES - MQEttAL..ARMQRMAL. E!.fERGENCY AND PAGE 69 l

EA21Ck221 CAL QQEEQL l ANSWERS -- SEQUOYAH 1&2 -85/05/20-JACCAR. T.

l l

l ANSWER 4.16 (1.00) t f b.

REFERENCE CONP AOI-2, p.

i l ANSWER 4.17 (1.00) 30 RAD camma x ! Rem / Rad a 30 Rom 4 RAD rast Neutrons x 10 Rem / Rad a 40 Rem

! 3 REM Thermal Neutrons s 3 Rom 73 Rem REFERENCE 10 CTR 20.1 and 10 CTR 55.21 ANSWER 4.14 (1.00) a.

RETERENCE SQNP A01 5, pp. 2, 3 ANSWER 4.19 (1.00) a.

b. 36f.le=Il+dl*0Id' A k' 4 #* { J Tl7Il m -

h g \. C' e 'fAbkd%== l**b*'"

d. 4 g, ggi &,, Q M g s, A Q ( , G w g h d / d (s M6/

REFERENCE SONP A0l=23A, pp. I - 3 of 4 l

l

s e 0 1- PtocGMFJ1 ECEMAL ARMQRMAL. r.ArnCGCL.AME PAGE 70 PAniotoG1CA M TdQL f ANSWERS -- SEQUOYAH 1&O -85/05/20-JA00AR. F.

l ANSWER' 4.20 (1.00) d.

RErCRENCE SQNP AOI-34A. p. 1 Of 3 ANSWER 4.21 (1.00)

! a. 4

b. 3

c. t

d. 2 RETERENCE ,

SQNP TRO's ANSWER 4.32 (1.00) d.

REFERENCE SQNP TR-5.1. p. 1 of 5 ANSWER 4.23 (1.00) ,

s.

REFERENCE SQNP ES-0.1. p. 2 of 13 ANSWER 4.24 (1.00) d.

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5 A 10W::R 4.05 (1.00) a T3UC L PALOC.

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00:l? CO-1.3 p. I of la CO 1.0 p. 1 of 3. App. A

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-e o o ,A_ e m o c E Stf R E E - NORMAL. ABNORMAL. EMERCENCY AND PAGE 72 '

m otocicAL ccM m k ANSWERS -- SEQUOYAH 1&2 -85/05/20-JA00AR. T.

I. . i I I RcrencNCE j s

SQNP 001-5A, pp 5-8 l

ANSWER 4.30 (1.00) c, ;

REFERENCE SC!lP E- 1. p. 2 of 11 f

I I

i E

l i

f, 9

l I

i

o ENCLOSURE 3 U. S. NUCLE AR REGULATORY COMMISS ION SENIOR REACTCR OPERATOR LICENSE EXAPINATION Sewswed Dy FACILITY 8 _SESuQ2AM.lLZ_ .______

l. IA kayter REACTOR TYPE 8 _EhE-h!Ch................

J. C.O.1E<ewee DATE ADMINISTERE08.122Q212Q...... ........

3. W d' hyna iXAMINER: .kibbOLAt_A4...___ . ...

4. 1.C.L k* APPLICANTI ..@ %13.Tr3....___.......

5. m . 7. L e v e. w IU115UEIIMDA.1W.htrbl6&G11 Uso separate paper for the answers. Jrite answer s on one side only.

Sttple question sneet on top of the answer sheets. Points for each question are Indicated in parentheses af ter the question. The passing gr ade requires at lesst 707, in each category ano a final grade of at l e a s t 6 0 /. . Examination papers will be picked up s ix (61 hours after the exaninstion starts. .

/. O F CATEGORY ? 0F APPLICANT'S CATEGORY

. 2ALuf. .IQIAL .. 300EC... _MALUA._ ..............CAIEGQEI... ... ...

.3QcQQ.. 22400 ........... ....... 5. THEORY OF NUCLEAR POWER PLANT OPEMATICNs FLUIDS, Atl0 THERNUQYNAMICS

.22aQQ.. 21499 ........... ....... 6. PLANT SYS TEPS DESIGNS CONTROL, AND IflS TR LMENTAi!ON

.22cQQ.. 224QQ ........... ....... 7. PROC EDUR E S - NORM ALs ABNORMALS EMERGENCY AhD RA010LCGICAL CONTROL 22400 ADMINIST1ATIVE P R OC E D LR E S ,

JD4QQ.. ........... ........ d.

CON 0!TIONS, AND LIMITATIONS

........ TOTALS 120ccQ. 19Q400 ...........

FINAL GRADE .................?.

All corn done on this stamination is my own. I h 3ve f el ther olvcn nor received sid.

ApPLIC ANT'S SIGb ATURE l

' 1.__IMEDEX.CE.MUCLEA8.1DM11.ELAMI.DEERAllDHa.ELUIQ34.ANQ PAGE 2 IW1850015A21G3 QUESTION 5 01 (2 00) so If tne reactor is operating in the power r anges tow long hill it take to raise power f rom 204 to 404 with a +0 5 OPP Startup rate?

1. 12 secones i

2. 21 seconos

3. 36 seconos

4. 54 seconds t be How long will it take to raise power f rom 404 to 60% with the same

+0 5 DPM Startup ritet

1. 12 seconds ,

2. 21 seconds

3. 36 seconds

4. 54 seconds OVES TION 5.02 (1 00)

Match the curvess on Figure A-4 on the next pages with the (allowing plant descriptions. Put your snswers on your answer papers e.g. "C ur w e A -6. ".

lo teJinning of life (BCL) - 04 powe r .

Zo 80L - 1004 power.

3. End of life (EOL) - 04 power.

4. EOL - 1001 power.

( ***** C ATEGCRY 05 CONTINUED ON NEXT P AGE *****)

4 100 W - - - - --

0 0 */ . . __

h g

Curve A. Curve 8.

PCAKS I SOW W 50% - *-- P E AX.

l

'AxlAL taxing

ioog ._ _ ___

ioow. __ __

4 h lhPEAK I So%

Curve C. I So% - Curve D.

1 W \

5 u l

0 0 I AXIAL 0 AXIAL

4 Figure A - 4

\

l j

8

l t

l t l'3 IMEQ83.01.5MGl.EAA 2GM11.2LedI 0211 ail 084_ELW1014 AMS PAGE 3 (:

l h1850015451G1 :

I l \

QUESTION 5.03 (1 00) j t

chleh of the following statements is TRUET i

s. It is NOT possible f or the Moderator Temperature Coef fielent (MTC) to ever besene positive at the Seguoyah plant.

l be It is possible f or the MTC to become positivea but ONLY when the I reseter is in Mode S. t o, If the MTC is positives while tne reactor is in Pode 2, Technical l Speelfloations must be consulted because ther e are action statements j l that must be followed. l l i

d. HTC can be positive in an under-moderated core where the moderator to fuel r ette is less then the optimum value. i i

Q UE S TION 5 04 (1 00) t I

Wnion of the following best describes the ef f ect en PTC If the AC5 !

teoper a ture is LCWERE01 ae It becomes less neJative because boron and water molecules are swept !

into the core as a result of the outsurge from the pressurlaers l tnererore, neutrons s,end more ti.e in ene resonance ,e.ien.

be It becomes less neJative boosuse the rate of change in the density of I water per degree temperature change is less at lower temperature l wnlon causes a lesser chanje in rate in resonance escape probability. i

e. It becomes more neJative boosuse thermal utillastion incrosses and resonance escape probacility deeresses.

de It becomes more negative because as temperature is lowered the moderator becomes more denses this increases the amount of water molecules in the sore therefore neutrons have a greater probability of colliding with a water molecule and this is an increased negative reactivity effect.' ,

l P

i l

i i

t es*** C ATE 0 CRY 09 CCNT!NUED ON NEXT P AGE e***el j i

r f

is IMEDEX QE.SUCLEAE.EQWER.ELAMI EEEEAIIDHA.ELUIQia_ASQ PAGE 4 IMEREDEXbe51G3 QUESTION 5.05 (4.00)

Use the attacned Xenon worth curves (on the next page) to answer the following four questions.

a. Power at To was at 704. What was the power l evel between T1 and T27

1. 90%

2. $0%

3. 20%

4. 10%

be wnat was the length of time between T2 and T37 -

1. I hour

2. 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months

3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

c. Wnat happened at T27

1. Reactor tripped.

2. Rods were placed in Auf0s and turoine power was raised to 100%.

3. Aeactor power was reduced to 104.

4. Turbine power remained constants rods wer e ir manual and inserted 50 steps and the steam dump valves failed open(10% of rated power).

d. At time T4, what happened?

1. All Xenon production has stopped.

2. todine decay to Xenon has stopped.

3. All Xenon production remains constante but tre burnout increases.

4. Xenon production directly from fission has stoppeds but Xenon production from deca odine continues.

~5 l ***** C ATE 0 CRY 05 CONT!NUE0 DN NEX T P AGE **** *)

;; :f

' I' '

Xe Worth

's e

8 g

(pcm) 8 o

I n. ,

i

-5500 _ ! .! !

t0 t1 t2t3 t4 XENON vs. TIME CURVE FIGURE 1.1 I

f

\

i PAGE 5 i ' 2a__ISECRX Of SUCLEAR.EIMER ELASI EEELAIIGSA.ELUIDSA.ANQ l IH18500%BA51C3 l

QUESTION S.06 (1 00) l l

Which of the following is TRUE concerning PCS operation f ollowing the loss of one Reactor Coolant Pumpf l

so Core coolant velocity decreases therefore the flowrate in the remaining loops decreases.

be Flow to the vessel f rom tha remaining three ptsps is less than 3/4 of the original flow.

ce Flow in the idle loop bypasses the core.

1

d. $1nce only three S/G's are pr oviding steams the steam pressure and temperature in the renalning $/G's is inc reased.

l Q UES TION 5.07 11 001 TRUE or FALSE 7

a. During 1 C 0 ?. power operations Departure from Nucleate Bolling Ratio 10NORI is or ester than tne ONBR f or 20% reacter power.

be Increasing pressure of the RC$s when oper ating ir the nucleate bolling region of the heat transf er curve will decrease the heat transf er rate in (3fU/hr-square foot).

QUEST!0li 5.08 (1.00)

TRUE or FALSE 7

s. During a RCS heatups as temperature Jets nishers it will take a smaller letdown flow rate to talntain a constant pressuriser level.

b. Increasing condensate depression (subcoolingl will cause 80TH a decrease in plant ef ficiency AN0 en increase in condensate (hotwelli pump avillable NPSH.

( * * **

C A TE00RY 05 C ONTINUE D Dil N9 X T P AGE * * ** * )

i

1.__IMED81.DE_SUCLEAR EQWES ELANI DEELAIIDMA_ELUID34_AdQ PAGE 6 IWEtn0QISA51C3 QUESTION 5.09 (1 00)

The resctor is operating at 100 4 powers all-rods-cut, near end of cycle prior to a scheduled power reduction to 50 % power f or a surveillance. The Unit Operator observes AFD (delta Il to be in the doghouse and decides to loner power and temperature by borating, while le aving rods f ully althdrawn. Actual Tav; f ollows the programmeo Tavg.

Which of the following best describes the AFD Ini tial change?

ao AFD will change in the POSITIVE direction; Decause relatively more POSITIVE reactivity is added to ths top nalf of tne core.

be AFD will change in the NEGATIVE direction; because relatively more NEGATIVE reactivity is aaded to the top half cf the core.

ce AFD will not change because rods have not moved. ,

de AFD will change in the NEGATIVE direction; because relatively more NiGATIVE reactivity is added to the bottoe half of the core.

QUESTION 5 10 (1 00)

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

a. The reactor can be made suberltical from all cperating conditions.

b. The reactor will not be made critical with th e p eactor Coolant System everage temperature less than 541 degrees F.

c. The reactivity tr snslents associated witn pos tulated accident conditions are controllaole witnin acceptacle llelts.

d. The reactor will be raintained sufficiently s6ccritical to preclude inadvertent criticality in the shutdown condition.

l l

{

(***** C ATEGC#Y 09 CONTINUE 0 ON 4GX f P AGE *****)

PAGE 7

'As._IMEDEX.QE_SUGLEAR.EQWER_ELadI_CEEEAIIQua_ELUIQ5a_AHQ IH18500ZhaM CS

(

QUESTION 5 11 ( .75)

Fcr each condition in COLUMN As find the correct heat transfer equation in COLUMN 8 that would be used to calcul ate the heat t r ans f er red.

COLUMN A COLUMN B so Across the reactor 1. 0 = UA delta T (cold leg to hot leg) . .

2. Q = M delta T be Across S/G U-tubes . .

(primary to secondary) 3. 0 = MCp asits T

c. Across S/G seconoary side 4 0 = M delta H Ifeedwater to steam) .

5. 0 = UA delta H .

Q UE S T10fl 5.12 ( .75) thich of the following is a "tell-tale" sign that the Point of Adding Heat has seen reached?

ao Decrease in the aver a0e temperature.

b. Increase in the Pressurizer level.

c. Increase in the Start-up rate.

de Decrease in the Reactor CoJiant System pressure.

QUkSTION 5 13 ( .50)

TRUE or FALSE 7 Convective heat transf er capability is increased by cecreasing coolant flow velocity.

l ** *** C A TE0 CRY US CONT!NVE0 ON NEX T P ACE *****)

1

14 _IMEDRI_QE SUCLEA1.EQME8_ELadI CEERAIIDut_ELUIDSA AdQ PAGE e IdEidDDIDA51CS l

QUESTION 5.14 (1.00)

Heat Flux Hot Channel Factor, RCS Flowr ate and Nuclear Enthalpy Rise Hot Channel Factor are power distribution limits and are determined pcriodically. This periodic surveillance is suf f icient to ensure the lloits are maintainea provided four conditions are ret. One condition to be met is the control rod inser tion limi ts are ma intained.

Which of the followinJ choices is NOT one of the other three conditions?

se The reactor shall not be made critical with a positive MTC.

be Con tro l rods in a single group are to move togetter with no single rod differing by more than 13 steps f rom the group comand position.

-> c. Control rods ire to be sequenced th proper overlapping groups.

de Axi al power distribution (AFO) is to be maintained within limits.

QUcSTION 5.15 ( .50)

TRUE or FALSE 7 Increasing the RCS boron concentr ation causes the thermal utlllzation f setor to increase, tnerefor e the MTC becomes less negative.

QUESTION 5 16 ( .50)

TRUE or FALSE 7 Thermal neutron flux is higher in the f uel rod than in the moderator.

QUESTION 5.17 ( .50)

TRUE or FALSET fna use of a slidin) Tavo program provides plant operation witn a higher tnermodynamic efficiency than does oper ation with a constant Tav2 program.

(***** C ATEGORY 05 CCNTINUED Oft NEX T P AGE *****)

1a_.IMED&1.QE.BUCLEA2..EDWER_ELA3I_DEERAIIDHa ELUIQSA AtiQ PAGE 9 IDEREQQIBASICS QUESTION 5.18 (1 50)

TRUE or FALSE 7 ,

ao One of the pump laws for centrifugal pumps states that the volume flow rate is proportional to the speed of the pump.

be As VCT temperature decreases, volume flow rate from the positive displacement (PD) pump increases.

ce Pump runout is the term used to describe the conoition of a centrifugal pump running witn no volume flow rate.

QUESTION 5.19 (1 00)

Which of the below describes the Inverse multiplication plot (graph)?

The Vertical Axis The Horizcntal Axis l a. ini tial count rate final count rate be initial count rate divided control red reactivity by final count rate (No. o f asserblies) ce control rod reactivity final count rate divided (No. of assemblies) by the initial count rate L

! de f i n al count rate divided control red reactivity by initial count rate (No. of asserblies) i QUESTION 5 20 (1.00) thl ch o f the below i tems will cause the control r ed alf ferential rod worth to increase 7

a. An increase in the boron concentr ation.

be An increase in the Moder ator temper ature.

c. A decrease in the neutron f as t flux.

de An increase in the fission product concentration.

j ( ***** C ATEGCRY 05 CONTINUED ON NEXT P AGE *****)

i - - - . _ - _ _ - _ _ . - - _ _ , . . , , - . - - , . . _ _ . _ . _ - _ - _ _ _ _ _ _ . _ . _ _ _ , - _ _ _ _ _ _ - _ _ _ _ -____

- 1a__IMEDSI_DE_SUCLEAE_EDMEE_ELAMI_DEEEAIIDMt_ELU1DSa_AUQ PAGE 10 IMEddQDIDA51CS I

QUESTION 5.21 (1.00)

Which of the below choices correctly completes the fcilowing statement?

Assuming all other factors are identicals the MASS flow rate of fluid through a 10 inch diameter pipe will be a p p r o x i m a t e 4 y _____________ times as great as the MASS flow rate through a2 inch disseter pipe.

a. 2.5

b. S.O

c. 12.5

d. 25.0 QUESTION 5 22 (1 00) unich of the following nuclides is NOT a fissile nuclide, i.e. fissionaole by a thermal neutron, curing powe r op er ati ons ?

a. Ur['anium235

c. Or/anium238

c. Plutonium 239

d. Plutonium 241 QUESTION 5.23 (1.00)

To what approximate pressure must tne steam gener ator pressure De reduced to maintain a'200 degree F subcooling margin in the RCS when reducing RCS pressure to 1600 psig?

a. d45 psig.

D. 645 p sig.

c. 445 psig.

de 245 pslo.

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

S a__ IB EDEI_ DE_B U C LE AE_ ED M Et_E LA BI_ DEEE AIldu t_ ELUIDSz_ AdQ PAGE 11 IMERdQDISABICS QUESTION 5.24 (1.00)

From which of the below choices does the majority of Tritium originate?

a. Directly from fi ss ion as a fission f ragment.

b. Activation of Deuterium by a neutron.

c. Activation of Lithium by a neutron.

d. Activation of Boron by a neutron.

l QUESTION 5.25 (1.00)

If RCP's are tripped following a LCCAs and the break has been isolateds -

which of the following situations would be MOST desirable?

PZR HOT LEG COLD LEG Press. Temp. Temp.

l a. 600 500 480

b. 800 530 520

c. 1000 540 530

d. 1200 575 565 QUESTION 5.26 ( .50)

TRUE or FALSE?

The Doppler only power coefficient ( PC M / ?.p o w e r ) at hot full power be'comes more negative during the life of the core (BOL to ECL).

QUESTION 5 27 (1.00)

TRUE or FALSE?

a. As Keff approaches unitys a smaller change in neutron l ev el will result for identical changes in Keff.

b. As Keff approacnes unitys a longer period of time is required to reach the equilibrium neutron level for i dentic a l changes in Keff.

(***** C ATEGGRY 05 CONTINUED ON NEXT PAGE *****)

' la__IMEQRI_DE_duCLEAR_EDMEE_ELAUI DEEEAIIDut_ELUIQ3A AdQ PAGE 12 IdERdDDINADICS QUESTION 5.28 (1 00)

TRUE or FALSE?

a. The di f f erential temperature necessary to transfer heat is inversely proportional to heat flux.

b. The latent heat of vaporization is another term for the l atent heat of condensation.

QUESTION 5.29 ( .50)

TRUE or FALSE?

High energy neutron exposure incr eases the possibili ty of bri ttle f racture of the reactor vessel by i ncr easing the compressi ve stress on the react 6r vessel, d

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

' ha__ELABI_11SIEd1_DE11GUA.GQUIRQkt_Ada_IB11EudESIAIIGd PAGE 13 J

l 1

QUESTION 6.01 (1.00)

The purpose of the CVCS deminer' al izers is tot

a. Remove all chemicals f rom the RCS fluids.

b. Remove soluable and insoluable material from the RCS.

c. Replace inso luable materi al with soluable ions.

d. Provide a method for boron control during reactor operations.

QUESTION 6 02 (1 00)

Near the end of a plant cooldenn how is the Pressur i zer (PZR) volume cooled after the RCPs are securec? .

a. Ambient losses deterrine the cooling of tne PZR.

b. RCPs are to be run f or 5 minutess every 15 min. to provide spray flow.

c. PZR level is raised and vapor is venteo to tne PRT.

c. Aux spray path is used from the regen heat exchanger outlet.

QUESTION 6.03 (1.00) t!hi ch of the below f eatures enhances the operation of the ice condenser and containment spray for heat removal? ,

a. Containment designs such that the delta P between upper and lower con ta i nmen t drives the air circulation.

b. Ventilation coolers and recirculation fans are Lseo to mix the air ano provide additional cooling.

c. Air return fans provide flow to return the ai r from the upper containment to the lower containment.

d. Pressure operated doors open to all ow upper containment air to flow through to tne lower con tainment.

(***** CATEGGRY 06 CCNTINUED ON NEXT PAGE *****)

' h u_ELASI_SISIEBI_DiSIGUz_LQUIROLA_AbD_IUSIEudESIAIICE PAGE 14 QUESTION 6.04 ( .50)

TRUE or FALSE?

Increasing RCS pressure while sol i d on RHF. Is most effectively and and quickly accomplished by increasing the temper ature control setting of the RHR Heat Exchanger outlet temperature controller.

QUESTION 6.05 (1 00)

Uny are the protection Distacles associated with a f ailed ins trument placed in a tripped condition?

a. Ensures that the required protection coincidence will be met.

b. Prevents inadvertant tripping of the plant curing repair work.

c. Ensures that a failed channel will not cause a trip when another valid signal is present.

d. Ensures adequate plant reli ablity is maintained by ensuring that two additional signals are needed to trip.

QUESTION 6.06 ( .50)

TRUE or FALSE?

After tripping a bistable in a 2/4 logic systems one of three remaining signals reaching the bistable setpoint will cause a trips even though the logic SYSTEM remains as a 2/4 system.

QUESTION 6 07 (1.00)

Whi ch o f the fol lowing protects the Upper Head In jection Accumulator f rom

, overpressurization, if inleakage or temperature changes occur red during normal operations?

a. Relief valves

b. Surge tank

c. Rup tur e di sk

d. Alarms for operator action i

(***** C ATEGCRY 06 C ONTINUED ON NEXT P AGE *****)

e ha__ELASI_111IEES_DESIGUA_CDHIR2La_AllD_INSIEUBEHIAIIDH PAGE 15 QUESTION 6.08 (1 00) ~

What interlock must be satisfied prior to opening the containment sump suction valves to the RHR system?

a. RHR pump must be off.

b. Low alarm on RWST.

c. RHR discharge valves shut.

d. RWST valves shut.

QUESTION 6.09 (1.50)

Refer to the Log ic Di agr am on F igure 6-1s on the next p ages to answer the following. -

a. hhat are the setpoint values for the items labeled "A"?

o. What is the coincidence of the bistables l abe lec "B"?

c. What is the P-ll permissive setpoint labeled "C"?

QUESTION 6.10 (1.00)

The plant is at 1007, power and stable. A maintenance person inadvertently trips the turbine but the reactor does not trip. After 30 seconds the AUO ce-energizes tne rod drive MG sets. Assume no further operator action.

Which o f the below is the response of the plant?

a. Steam dump arms, all valves trip open and Tavg is reduced to Tref no-load setpoint.

b. Steam dumps opens turbine trip controller reduces Tave to Tref plus the '2 degrees F dead band (549 degrees).

c. Steam dump arms and valves will open as a resul t of a signal from the load rejection controller and reduces Tavg to Tref no-load plus the 2 degrees dead band (549 degr ees) .

de Steam dump arms and valves will open as a result of a signal from the load rejection controller until the rods drops at whicn time the reactor tr ip controller reduces Tavg to Tref.

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

t

LOW PRESSURIZER PRESSURE-ESF ACTUATION LOGIC PRESSURIZER A

AUTO UNBLOCK 4

i e s

b e

( ,PM s

PRESSURIZER I e

yy ; Si BLOCK CONTROL UNBLOCK

-+ .

BLOCK $

7

. s @ g (P-11 L3N '

]:

r 0THER o LOGIC TRAIN

-1 ESF ACTUATION e

r hs__ELABI_SISIEUS DESIGUA_CCBIR0tt_Ah2_ISSIEudEHIAIILb PAGE 16 QUESTION' 6.11 (1 00)

Whicn statement is correct concerning the level c en t ro l valves on the Motor Driven Auxiliary Feedwater Pump discharge?

a. They are hydraulic operated by the pump and r eservoir mounted locally.

b. They fait closed.

c. Wnen pressure downstream of the valves drops to 500 psigs they will Close.

d. Can De operated from the local control panel, Just outside the Terry TurDine Room.

QUESTION 6.12 (1 00)

Wnich o f the below choices correctly completes the f ollowing statement?

The Tur bine Dr iven Auxili ary Feedwater Punp _____________________________.

a. will trip on a tnermal over load of its trip / throttle valve.

b. can operate properly on 50 psig steam.

~

c. will automaticly start if one main feedpump trips at 607. power.

d. suction valves will transfer to ERCW when suction oressure drops to 23 psig. for 2 seconcs.

QUESTION 6.13 (1.00)

The controlling Pressurizer (PZR) level channel (459) fails high during 100% power operation. Assuming NO operator action is takens which of the following best describes the response o f the p l an t?

a. Charging flow goes to minimums PZR level decr easess l e t do wn isolates and the plant continues to operate a t the same power.

b. Cnarging flow goes to minimums PZR level decr eas es, letdown isolates and the pl ant eventually trips on high PZR level.

c. Cnarging flow goes the maximums PZR level increases and the plant trips on high PZR level.

d. Charging flow remains the s ames PZR level increases due to letdown isolating and the plant trips on nigh PZR level.

(***** CATEGCRY 06 CONTINUE 0 ON NEXT PAGE *****)

ha__ELAMI.11SIEMS_DEllGUA CLUIR2LA_Ad2_INSIRudsdIAIIDS PAGE 17 QUESTION 6.14 (1.00)

Besides the overspeed shutdowns which of the following diesel engine /

scner ator shutdowns is enabled during an emergency start of the diesel?

a. Voltage restraint overcurrent rel ays (51V).

. b. Gener ator di f f eren ti al relays (87).

c. Pnase balance relays (46 ) .

d. Low lube oil pressure.

QUESTION 6 15 tl.00)

Which statement describes the signal path from the Source Range detector to tne Source Range level meter on the main control bo aro?

a. Detectors Pre Amps Discriminators Log Integrator, Meter D. Detectors Log Integrator, Pulse Shapers Pulse Counters Mater

c. Detectors Pr e Amos Log In te gr ators Discriminators Meter de Detectors Log Amp, Meter Q UES TION 6 16 (1 00) kni ch statement concerning the Power Range Nuclear Instrument detectors is CORRECT 7

a. They are uncompensated ion chambers that operate in the proportional region of the gas amplifi cation (detector character isti cs) curve.

b. Uses compensation inner chamber current to cancel out gamma current measured in the auter chamber.

c. Uses Boron Trifluoride gas to make it neutron s ensi ti v e.

d. Its detector current is cal ibrated using the cata from a secondary heat balance.

(***** CATEGORY 06 CONTINUE 0 ON NEXT PAGE *****)

I

6a_.ELASI_SISIE53_DE11Gda_CQUIEDLA_ASD_lHEIEudEBI&lILS PAGE 18 QUESTION 6.17 (1.00)

Using the following 7 actions, which of the below sequences is the correct sequence for a rod withdrawl?

Seven Rod Control Actions

1. Movabl e gr ipp er coil energizac.

2. Stationary coil ene rg i zed.

3. Lift coil energized.

4. Stationary gripper energized.

5. Novaole coil DEENERGIZED.

6. Stationary coil OEENERGIZED.

7. Lift call DEENERGIZED.

S equ enc es to Choose From

a. As 3r is 6s 2, 5, 7.

b. As is 6s 3, 2, 5, 7.

c. 2, 4s is 6, 3s 7, 5.

d. 3, is 7s 5s 6s 4, 2.

Q UES TIO!! 6 18 (1.00)

Which statement concerning the Rod Control System is CORRECT 7

a. The power cabinet provices AC power pulses to drive the control rod drive mechanism.

b. The reactor control unit generates a rod speec arc direction signal in response to three ERROR signals.

c. Tur bi ne impu l se pr ess'ure pr ov ides s i gn al s to the rate comparators summing unit and the vari able gain unit in the rod control circuits.

d. Rod power is supplied by two motor generator se ts with a 260VDC cutput through an isolation transformer.

QUESTION 6.19 (1.00)

What are TWO interlocks that would prevent the au tomati c tr ansf er of a 6.9 KV Shutdown Board from its normal supply to alternate supply?

(***** C ATEGORY 06 CONTINUE 0 ON NEXT PAGE *****)

6.__ELauI_111IE31_DE11GSz_COBIRDLA_AhD_lS11RudEulAllEh PAGE 19 QUESTION 6.20 (1.00)

On a sustained (greater than 1.9 seconds) loss of voltage to a 6.9 kV Shutdown Boards what is the sequence of events?

a. The diesel is starteds and in 2.0 seconds the bus i s strippeds then the diesel breaker shuts upon reaching 800 RPM an o 6.9 kV.

b. The bus is strippeds except the 480v shutdown tr ans f o rmer s, and in an additional 3.5 secondss the diesel is started and i ts breaker shuts upon reaching 6.9 kV and 900 RPM.

c. 3 5 seconds later the diesel is started with the bus being completely s tr ipped and then the diesel breaker shuts wi thin the following 10 seconds.

d. Tne diesel is starteds then after an add i t i on al 3.5 seconds, the bus is stripped except the 480 y shutdown board transformerss and the diesel breaker snuts upon reaching 900 RPM and 6.9 kV.

QUESTION 6.21 (1 00)

Witn normal power unavailable and CNE vital battery cut of services how long wi ll the remaining THREE Datteri es Do capaole of supplying all loads required for safe shutdown of 30TH units?

a. 30 minutes.

b. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

c. 1 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months de 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months QUESTION 6.22 (1.00)

TRUE or FALSE 7

a. A Diesel Gener ator should never be isolated on a Shutdown Board during surveillances.

b. If a Olesel Generator was loaded at 40 f. for four nourss it is then permissable to snutdown the Olesel without f u rther loading until the next surveillance.

(***** CATEGCRY 06 CONTINUE 0 GH NEXT P AGE *****)

6.__ELANI_SISIEUS_QE11GUA_IDUIEDLa_AdQ_IUSIEudENIAIILS PAGE 20 QUESTION 6 23 (2 00)

List the automatic actions that occur if one Main Feed Pump trips when the plant is operating above 80 7. power.

QUESTION 6.24 (1.00)

What action will occur if the #3 heater drain byp ass to the condenser v al ve leaves its fully closed position with plant at 957 power?

a. Standby dr ain pump starts.

b. Main Turbine runback to 857..

c. Main Turbine runoack to 60r..

d. Main Feed Regulating (Con tr o l ) BYP ASS val ves open f ul ly.

Q UES TION 6 25 (2.00)

For the f ollowing components, indicate whether they will receive an OPENS CLOSES or NO signal as a result of a safety injection (with Phase 'A')

initiation signal.

a. -Control room supply ducts D. Main feed bypass valves

c. SI accumul ator disch arge isol ation valves

d. Normal charging header isolation valves

e. Main steam isolation valvas

f. RwST to SI pump suction valves 0 Seal water return isolation v alve

h. Component cooling isolation valve f rom RHR system

1. Component cooling isolation from letdown heat exchanger J. Steam supply val ves to turbi ne-dr i ven f eed purp l

l l

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

1

ha__ELAHI 1XSIEBS.DE11GUA CDUIEGLA AdQ_Id1IEMBESIAIILh PAGE 21 QUESTION 6.26 (1.00)

Ccnsidering only the Steam Generator Level Contro l System, what would be the response of the IllITIAL f eedwater flow to the S/G if the controlling S/G pressure transmitter f ailed LOW during 507. power operations?

a. Tne flow would decrease due to the loss of the steam pressure input to the steam flow siJnal.

b. The flow would remain the same due to tne steam pressure not affecting tne steam flow.

c. The flow would increase due to th e steam pressure input to tne feeo control valve position controller.

c. Tne flow would increase due to the l oss o f s t ear. p r essur e input to the steam flow signal. ,

QUESTION 6.27 ( .50)

TRUE or FALSE?

The Pressurizer PORV arming temocrature setpoint f or Unit 2 was recently changed to that of Unit is i.e. 380 degrees F.

QUESTION 6.28 (1.00)

Which of the following will cause a trip of a running Main Feedwater Pump?

a. Low f eedwater temper ature.

b. Low Main Feed Pump turbine speed.

c. Recirculation valve open.

d. Safety Injection.

(***** C ATEGORY 06 CONTINUED ON NEXT PAGE *****) )

l

- . _. . . - , ._ . - . - . _ - _ . _ _ - . ._. .\

ha__ELASI_S11IEds_QE11Sua_CDMIBDLA_AUQ IBSIHudENIAIIES PAGE 22 i

QUESTION 6.29 (1.00)

If an unsaturateo bed o f H-OH r es in i s p l aced in ser vice, what will be the result?

a. RCS 0xygen concentration will i ncr e ase.

b. No lon exchange will occur for the first 12 h our s o f oper at i on.

c. RCS Boron concentration will decrease.

d. RCS doron concentration will increase.

4 h

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

y

Ia__EEQCEQUEE1_=_dDadAL4_A3dQ&5&La_EMERGENC1_AdQ PAGE 23 BAQIDLDGICAL_CDNIEQL QUESTION 7.01 (1.00)

G01-1 gives a precaution limiting the opening of the RCP seal bypass return volves 62-53. It must be open if any #1 seal return flow is less than Igpm and the RCS temperature is hign enough to cause the lower radial bearing teaperature to exceed the al arm point.

What is the other limi tation required prior to opening 62-537

a. CVCS injection to the seals are at least 6 sps.

c. RCS pressure is at least 100 psig.

c. Tne RCP thermal barrier flow is normal .

d. Tne RCP #1 seal delta Pressur e is less tnan 4C0 psid.

! QUESTION 7.02 (1.00)

If tne reactor trip breakers are closed and the steam generators are under nitrogan pressures the nitrogen pressure must be vented off the steam gener ators prior to opening the MSIV's.

uny must this De done?

a. To prevent SIS actuation on steam generator high celta Pressure.

b. To prevent aamaja to the MSIV seats.

c. To prevent ESFAS actuation on high steamline fl ow.

d. To prevent EST actuation on steam generator 10-10 level.

(***** C%TEGORY 07 CONTINUED ON NEXT PAGE *****)

Zs__ERQCEQUEES_=_UDRBALa_ABBQadALa_EMERGENC1_AdQ PAGE 24 RADIDLDGICAL GQUISQL QUESTION 7.03 (1.00)

When the RCS pressure is below 500 psig, why is a centrifugal charging pump roquired to be used instead of the posi tive di spl acesent pump?

a. The smoother flow fror the centrifugal pump is desired instead of the pulsation flow of the positive displacement pump.

D. The flow from the centrifugal pump can be regulated to a lower value than the flow from the posi tive displ acement pump.

c. The seal injection flow is easier to control wi th the centrif ugal pump than the positive displacement pump.

d. One centrifugal pump must be tagged out f or overpressurization reasons in case of an SI and the secono is oesired to be running.

QUESTION 7.04 (1.00)

The con trol rods wer e wi thdr awn 5 steps to preven t "the rmal lock-up" during RCS hea tup. If the control rods were NOT fully inserted using bank-select prior to withdrawing rods using manual, what woul c ce the result accord i ng to GOI-2, " Plant Startup from Hot Standby to Minimum Loao"?

a. Error in rod height for the ECP cri tical data.

b. Rod bottom lights malfunction.

c. Rod bank overlap malfunction.

d. Rod upper limit stop malfunction.

QUESTION 7.05 (1.00)

According to a note in G01-2s what condition must be met prior to exceeding 600 RPM on the main turbine?

a. Main Feedwater Regulating valves are to be in automatic.

b. Tavg is to be at the no-load value.

c. The low pressure turoine inlet metal temperature must be greater than 400 degrees F.

d. Steam dumps must be in Tava mode.

(***** C ATEGCRY 07 CONTINUED ON NEXT P AGE *****)

Z$__EEDCEQUEEi_=_SDE3ALa_AadQESAL4_EMEEGEUC1_ASQ PAGE 25 BAQ1DLDGICAL.GQhIRDL Q UE S TI0ft 7.06 (2 50)

TRUE or FALSE 7

a. When the axial flux difference monitor is inoperables the AFD must be logged once a shif t by perf orming SI-44

b. If the loop boron concentration is changed by 10 ppm or greaters pressurizer sprays will be actuateo by manual operation of sprays.

c. Any off-frequency turbine operation is to be reported to tne results sec ti on for record keeping.

a. If the " Rod Control Banks Lim i t Low" alarm comes in when cr itical, commence boration to clear the alarm.

e. dnen the quadrant power til t ratio alarm is i nop er ables the OpTA must '

be calculatec every 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s by perf orming SI-133.

QUESTION 7.07 (2.00)

For the power levels in Column As find the one associated conditions or actions in Column Bs as stated in G01-5A.

COLUNil A -

COLUMN B

a. 20,7 1. P-8 light goes out.

b. 30 % 2. P-9 light goes out.

c. 35 4 3. Observe turbine s tar tup dr ains closed.

de 50 7 4. Open HP dr ains to the Nc. I heater shells.

5. Star t two condensate derin pumps.

6. P r i o r to e x c e e d i n g _________ powers steam generator chemi str y must ce below the limits for exceeding this specific power level.

(***** C ATEGORY 07 CCNTINUED ON NEXT P AGE *****)

s Za__E80CEDuBES_= UD3dakt_AEUDE3ALA_EEEEGEUC1_AUQ PAGE 26 BAQ1DLOGICAL_CDSI3QL QUESTION 7.08 (1.00)

In accordance wi th ES-1.2s "Transf er to RHR Conta inrent Sump"s what is the guidance for performing all actions?

a. S lowl y and deliberately with time taken to an al y ze the proceeding steps.

b. Complete each step as rapidly as possible unl ess action does not take places then complete corrective measures.

c, Stop all running pumpss then tran s f er the valwe lineup per the procedure.

de Quickly in a precibes orderly sequences without interruption of changeover operation until all acti ons ar e corpl eted.

Q UES T I0ti 7.09 (1 00)

After going into the recircul ation modes at wnat RWST level shoul d the Centainment Spray Punps suction be realigned to t he containment sump?

a. 0;

b. 84

c. 154

d. 257 QUESTION 7 10 (2 00)

LIST tne THIRTEEN immediate actions to be taken for a Saf tey Injections in accordance with Emergency procedures E-0.

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

Ic._E10CEDURES_= UDadALa_AESDadAla_EBERGESC1_AdQ PAGE 27 BAQ10LDGICAL_CDdIs0L QUESTION 7.11 (1.00)

What is the maximum quantity of fuel (assemblies) that shall be allowed out of approveo stor ace locations at any one time dur ing f u e l-h an dl i n g operations?

a. 1 oisy.g.,J L ?J y o- %wu 5 4* *\y v e o s5*S 7'* *h t *-

c. 6

d. 8)

QUESTION 7 12 (2.00) ,

Maten the whole cody radiation exposure terms in Column A to their limit in Column B.

CAUTION: Some answers could De used more than once.

CCLUMN A COLUMN B -

a. 10CFA2C limit /qtr wi thout 'IRC Form 4 1. 0.3 REM l

b. TVA limits /qtr f or TVA (occupational) 2. 1.25 REM workers

3. 0.75 REN

c. TVA limit /qtr for Non-TVA personnel l without thei r history 4. 5.0 REM

d. 10CFR20 limi t/qtr with an NRC Form 4 5. 3.0 REM CO.5 each)

(***** C ATEGCRY 07 CONTINUED ON NEXT P AGE *****)

G Ia__EaQC100RE1_=_dQad&La_ABUQESALA EdE1QESCI AMQ PAGE 28 BAQIDLOGICAL_COMIRQL QUESTION 7.13 (1.00) that group of THREE indications below are used to monitor RCS cooldown during natural c ircul ation acco rd i ng to ES-0. 3s " Natural Circulation Cooloown"?

a. Tcold (NR), Pzr levels Core exit TC's.

b. That (NR)s RCS pressures RCS subcoo l ing.

c. Thot (WR), Core exit TC's, RCS suocooling.

d. Tcold (WP)s Pzr pressures Thot (HR).

QUESTION 7.14 (1.00)

G01-2s " Plant St1rtup f rom Hot S tandby to Minimum Lo ad" states that the shutdown banks must be at the fully althdrawn pos ition whenever positive resctivity is being inserted.

When can exceptions to this rule be applieo?

a. Anen the Shutcown Margin has been calculated to be 900 pcm.

D. when the RCS nas been borated to the cold shutoonn concentration and tne plant is being cooled down.

c. When the reactor is in the source r ange wi th the High Flux at Shutdown alarm operable.

d. When the actual coron concentration is greate r than the predicteo critical boron concentration.

i f

(***** CATEGORY 07 CONTINUED ON I4 EXT PAGE *****)

Ia__EEQCEQUEf1_=_bC33AL4_ABUQadAL4_EEEEGEUC1_AdQ PAGE 29 dADICLGGICAL_CDMIs0L QUESTION 7.15 (1.50)

Aro the three statements below TRUE or FALSE?

ASSUhE the plant (ACS) is soliJ with press,ure being maintained by the low prOssure letdown valve, FCV-62-81, in automatic.

a. Tne stopping of an RHR pump will cause a decr eas e in RCS pressure.

b. If the RHR system pressure exceeds 700 psig, the RHR suction valves from Loop 4 hot leg will close.

c. With RCS pressure at 300 psig (no steam bubble in the pressurizer), it is permissible to isolate the P.HR suction line from the RCS.

QUESTION 7 16 (1 00)

Quring normal CVCS operation, which of the following is an abnormal condition ano would require optrator action to correct?

a. VCT pressure is 15 psig.

b. The temperature of the fluid leaving the letdcwn heat exchangers is 127 F.

c. Tne RCp seal injection water tempeature is 120 F anc flow to the seals is 8 gpm/ pump.

d. RCP seal olfferential pressure is 250 psid.

QUESTION 7 17 (1.50)

TRUE or FALSE 7

a. Tha tr ansf er of ECCS suction to the containmert sump is accomplished when pWST level is greater th an 29 4.

-> b . When the RWST level reaches 0 7., JAf a l l equipment taking suction on

,, tne RWST 2et, stopped.

is e k s W Q s % oeb.

c. Transfer to hot leg r eci r cu l a t i on is accomplisheo 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months after initiation of Safety Injection.

(***** C ATEGCRY 07 CONTINUE 0 ON NEXT P AGE *****)

Za__ESQCEQu!E1_=_h0!dALa_A3hCadALA EdEEGENC1_AND PAGE 30

-EAd1GLLGICAL CCMIADL QUESTION 7.18 ( .50)

TRUE or FALSE 7 Areas where dose rates are 500 mr/ hrs are requirac to be locked and access controlled.

QUESTION 7.19 (1.00) t hich of tne f ollowing statenents concerning the procecure for a dropped RCCA is correct? .

a. Upon starting recovery of the dropped pCCA, a n U AGENT F AILURE alarm will occur because the lift colls f or the other rods in the group have been cisconnec ted.

b. The delta flux target band is not applicable curing a dropped l RCC A mal function and recovury.

c. If two or more RCC A's have droppeds manually trip the reactor i and pr eceec in iccoraance wi th Ep-1.00.

t l d. Aecovery fror a dropped RCCA will be facilitatec if Tavo is higher than Tref prior to commencing withdrawal of the croppec RCCA.

QUESTION 7.20 (1 00)

Followin) a reactor trips how many callons must b e emergency barated for oach control rod not fully insertect

a. 350 l 0. 400

c. 450

d. 500

(***** C ATEGORY 07 CONTINUED ON NEXT P AGE *****)

Ia__EuQCiDUREl_=_h0a3AL4_AabOEMAL4_EEEEEEUC1_AMQ PAGE 31 RAQ1CLEGICAL_CQ3IdQL QUESTION 7.21 ( .50)

TRUE or FALSE Fuol Handling equipment interlocks shall only De byp assed by approval of, and under the direct supervision ofs the fuel hanollrg SRU.

Q UES T ION 7 22 (1.00)

At least ___(how many).__ steam generator (s) must be maintained available for RCS cooldown in accordance with procedure E-2, " Faulted Steam Generator Isolation". s QUESTION 7.23 (1 00)

To terninate SI 4ith adverse containment conditions, the pressurizer level I

cus t be Jreater than ______,1 in accor d ance wi th E S-0.2.

a. 20

b. 30

c. 40 i

de 50 QUESTION 7 24 ( .50)

TRUE or FALSE 7 During a steam senerator tube rupture and subsequent controlled RCS cepressurizations the RCP Trip Cr iteri a can be al sr egar ded.

-> a.yo ca l

L

(***** C ATEGCRY 07 CCNTINUFO CN NEXT P AGE *****)

Ia__ERQGE2WRE1_=_dQE3ALa_ABEQ35ALA EtERGENC1 AdQ PAGE 32 8AQ10LUSIGAL_GuuldDL QUESTION 7.25 (1.00)

In accordance with the " Loss of Reactor or Secondary Coolant" procedures E-1s I f one char ging pump is operating and RCS pressure is uncontrollably d0 creasing, at what ACS pressure must the Reactor Coolant Pumps be stopped?

a. 1450 psig.

b. 1350 psig.

c. 1250 pslo.

de 1150 psig.

QUESTION 7.26 (1.00)

Wnich of the following conditions would REQUIRE f uel shuf fle operations be lcceolately stoppeo? Assume the Initial nucleus of ten assemblies are loadeo, AND exclude AflTICIPATE3 change in count r ates due to detector and/or source movement.

a. An increase in count rate by a factor of 2 on ANY nuclear channel or by a f actor of 1.5 on ALL nuclear channels curing any single loading step.

D. An increase in count rate by a factor of 3 on ANY nuclear channel or by a f actor of 15 on ALL nuclear enannels during any single loading step.

c. An increase in count rate by a factor of 4 on ANY nuclear channel or by a f actor of 2 on ALL nuclear channels duri ng any single loading step.

de An increase in count rate Dy a factor of 5 on ANY nuclear channel or by a factor of 2 on ALL nuclear channels curing any single loading step. -

( *** ** END OF C ATEGORY 07 *****)

Ra__AQUIDIIIEAIIME_EEQCfDUEfia_CDUQIIIQ31t_AUQ.LI311AIIQN1 PAGE 33 QUESTION 8.01 (1.00)

Which of the below cisssifications of drawings !s permi tted to be used in the Nain Control Room?

a. Information Only.

b. As-Designed.

c. Workplan Drawin2 Copy.

de Safeguard Information.

QUESTION 8 02 (1 00)

Concerning AI-30 Procedures "Nuc l ear P l ant Me thoo o f Op er at i on"s which one of tne below statements is correct?

a. Under emer gency conditions, a licensed reactor operator may APPROVE reasonable action that departs from a license condi tion or Technical Speci f i cati ons.

b. The operator is required to have Category "B" instructions present dnen performing work.

c. Thsre is no specific written culcance concerning the repeat back of verbal communication 3s however it is a 3000 operating practice.

c. Permission shall be received from the Lead Unit Operator prior to the performance of any maintenances tests or mool ficati on act ivity one or that may affects p l ant equi pment.

l l

l

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

Ra__AD5151SIEAIIME_EEQCEQUB21t_CDUDlIIQuit AUQ.LI51IAIIQd3 PAGE 34 QUES TION 8.03 (1.00)

Whicn statement is correct about Independent verifications?

a. Independent veri fication shall be provided for return to service of equipment th at is placed in an "off normal" configuration for the operating mode to allow surveillance testing to be done.

b. When clearances tr e hung independent veri fica tion is required f or the placing of tags and not for the removing of tags.

c. Incependent checks on remote, hard to reach equipment can be verified by the second person by verbal report from the first person.

de S yp as s i ng inoependent verifications can be cone by the Shift Engineer i f an "af ter-tne-f act" review is suomitted by a licenseo SRO.

QUESTION 8.04 (2 00)

List FIVE circumstances which require direct noti fication within one hour to the NAC Operations Center via the ENS.

QUESTION 8.05 (3.00)

For tne items in Column As identi fy the proper co lor code from Column 8 for plant instruments and common equipment.

COLUMN A COLUMN B se Unit 0 1. Red

b. Unit 1 2. Blue

'c. Unit 2 3. Green

d. A l a r.3 point (not trip) 4. 3 lack

e. Trip point 5. Yellow

f. Operiting Band 6. Oranse

7. White

8. 3rown

(***** CATEGCRY 08 CONTINUED ON NEXT P AGE *****)

Ac_.ADelSISIBAIIMf_2EQCEDUBEla_CDSDIIIQUSA_AUQ_LI3lIAIIQdi PAGE 35 Q UES TI0tl 8.06 (1 00)

Uhich of the following is a direct responsibility of the Assistant Shift Engineer (Shift Supervisor) as stated in the AI-2 Procedure?

a. Serves as site emergency director until relievec as specified in the Raalological Emergency Plan.

b. Can exercise control over any action which could affect reactivity of the reactor for which he is responsible.

c. It is his responsibilty to analyze tne cause ano determine if operation can continue safely before returning to power.

d. Responsible for the safe and correct operation of all electrical boards and maintains a constant source of power to the equipment ano controls of the uni t that he/she is assioned to.

QUESTION 8.07 (1.00) defore a person may be placed on the official cl e ar ance list he/she must complete wnicn of the below i t em s ?

a. An oral checkout with the shi f t engineer.

b. A written exam with the grade of at least 70%.

c. A 3 0 hour0 days 0 hours 0 weeks 0 months training course on clearances.

de feo complicated clearances under supervision of an approved person.

QUES TI0f4 8.08 (1.00)

When utilizing the Emergency Clearance procedure, who is responsible fcr all work performed and the safety of all workers involved in the clearance? .

4. The Shif t Engineer.

De The person clearing the equipment.

c. Tne Plant Manager.

d. The Emergency Dir ector.

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

Aa__AD51HISIEAII2f_E1QCsDUS11t_CDHDlIIQMat_ASQ_LI52IAIIQal PAGE 36 QUESTION 8.09 (1.0C)

Who should review a Temporary Alteration (TA), to determine the need for spacific training or for formal information for the licensed operators to assure their awareness of the TA and its implications?

a. NKCs Region II.

D. The Unit ASE.

c. The Shift Technical Advisor.

d. Tne PORC.

Q U ES TION 8 10 (1.00)

From the choices below pick tne term that best corpletes the following-statement.

On l y _________________ c h ang e s s h o u l d oe installea using plant instructions or Maintenance Requests (MA's). ,

a. long-term

b. PORC-approved

c. SQA-approved

d. short-term Q UE S TI0tl 8.11 (1.00)

From the choices below pick one that Dest completes the following statement A trip after a long period of reactor shutdown leaves little decay heat'to be removed thus causing the possibility of excessice cooling of the reactor coolant if too much feedwater is being added. The oper ator should NEVER rostore the steam generator water levels after a plant trips at the cost of a r eduction of the _________________________________.

a. plant pressure be shutdown margin

c. CST level de steam generator pressure.

(***** C ATEGORY 08 CONTINUED ON NEXT P AGE *****)

d a__AD $1311IS AIIEf _EEDC f DU Belt _CDSDIIIONit_A3Q_LI511 AI1QS1 PAGE 37 QUESTION 8.12 (1.50)

List the THREE reactor trlP CIRCUITS that may be ad m ini st r at i ve l y BYPASSED for maintenance on a single channel.

Q U E S TI0tl 8.13 (1.00) uhat is the minimum Technical Specification quadr ant power tilt ratio (QPTR) which requires actions to be taken when op er a tin g at 757 power?

a. 1.02.

o. 1 03.

c. 1.05.

~

d. 1.09.

QUESTION 8.14 (1.00)

Durin] cold shutcown, Technical Specifications recuire the two RHR loops to bo OPEAABLEs with a note modifying tni s requi r ement to allow substitution of certain equipment for one of the RHR loops.

Which of the below Jroups of equipment can be sub st i tut ed for the one RHR loop?

a. Four filled RCS l oops wi th one Saf ety Injecti on Pump and RWST level > 507..

b. Four filled RCS l oops wi th one OPER ABLE RCPs anc one Safety injection pump.

c. Four filled RCS loops with two GPER A3LE RCPs ano one OPERABLE euxiliary feed pump with CST level > 5 0 7. .

,, d. Four filled RCS loops with at least two steam generators having levels >= 107 WR.

(***** CATEGCRY 09 CONTINUED ON NEXT PAGE *****)

Sa__AQuidISIBAIIMS_EEDCEDUEEIt_CQHDIIIQuiz_AdQ_LIdIIAI1QdS PAGE 38

(

)

QUESTION 8.15 (3.00)

Match the condition in Column A to the maximum system leakage as stated in Tcchnical Specifications from Column 8.

COLUMN A COLUMN B (spm)

a. controlled leakage 1. Zero

b. RCS pressure isolation valves 2. 0.5

c. Primsry-to-secondary 3. 1.0

d. Pressure Doundary leakage 4. 5.0

e. Identified leakage 5. 10.0

f. Unidentified leakage 6. 20.0

7. 40.0 QUESTION 8 16 (1.00)

During refueling operations i t is discovered that the Kef f is greater than 0.95 but the boron concentration is at 2050 ppm.

What action must be takens in. addition to suspending all core alterations?

a. Sorate at greater than or equal to 10 gpm with a solution greater than or equal to 20,000 ppm boron, until Keff is less than 0.95.

b. No action required since the boron concentrat ion is gr eater than 2000 ppm.

c. Borate at 10 gpm with 20,000 ppm boron until the boron concen-tration is 100 ppm higher than previous concentration.

d. Notify tne reactar engineers and immediately recalculate the Keff v al ue.

(***** CATEGORY 08 C CNTINUED ON NEXT P AGE *****)

14__AQdid1SIRAIIME_EEQCEDURESt CDSDlIIDSSt_ANQ_LillIAllQHS PAGE 39 QUESTION 8.17 (1 00)

How are loads in excess of 2000 pounds prohibi ted from tr av e l over fuel assemolles in the stor sge pool when suspended from the Spent Fuel Area Crane?

a. All loads are less than limit except fuel and exceptions are made for the fuel movement.

b. The area is restricted by adminstrative control.

c. The crane is prevented f rom moving in tne area by interlocks and stops.

d. The crane is not rateo for loacs in excess of 2000 pounds.

QUESTION 8.18 (1 00) ,

What actions shall take place in the event a Saf e ty Limit is violated?

a. Place unit in HOT SHUTDCWN in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and notify NRC in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. Place unit in COLD SHUTDOWil within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months and notify NRC in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. Place unit i n HOT SHUTDOWN wi thin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and CCLD SHUTDOWN within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months and notify the NRC within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d. Place unit in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ano noti fy the NRC within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

QUESTION 8.19 (1.00)

What is the MINIMUM crew composition (SSs SRO, R0, A C, STA) as defineo in the Technical Specification when n o th uni ts are oper ating in mode 17

a. 7

c. 9

c. 10

d. 12 4

(***** CATEGCRY 09 CONTINUED ON NEXT PAGE *****)

. Sa__AQ5151SIEAIIME_EEDCEDU8ESz_CDBQlIlQBSz_ASQ_LidlIAIIQB1 PAGE 40 QUESTION 8.20 (1.00)

Each time the main control room drawing files ar e up cated to provide new "as-constructed" drawingss wi th which o f the below labels are tne new drawings identified?

a. CONTROL COPY

b. S AF EGU ARDS INFORM ATICN

c. FIELD CONTROL COPY

d. THIS ORAkING EXPIRES AFTER QUESTION 8.21 (1.00)

Whose approval is required prior to tne implementation of Handwr itten Instructions?

a. SE-SRO and Operations Supervi sor

b. ASE-SRO and 00

c. ASE-SRO ana SE-SRO

d. SE-SRO and the preparer QUESTION 8.22 (1.00)

Whenever a condition requires a temporary (one day) ceviation from normal system alignements tnese deviations are to be traCKeC by entr ies in which of the below logs?

a. Status Log.

b. Deviation Log.

c. Configuration Log.

d. Surveillance Log.

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

, aa__ADdIUISIBallMf_2dDCEDUSESt_CDUDIIIDHSz_AUQ_LI311AI1QUS PAGE 41 QUESTION 8.23 (1.00)

Wno nas the responsibility for initiation of a hold order clear ance on the incore flux drive motor control p ower s prior to personnel entering into the lower containment or annulus?

e. Health Physics Representative

o. Shift Engineer

c. Public Safety Officer

d. ASE-SRO QUESTION 8.24 ( .50)

TRUE or FALSE?

The inoperability of vital inverter 2-II is cause f or the Uni t 1 Technical Speci fication LCO cetion statement to be entereas when Lnit 1 is in moce 1.

QUESTION 8 25 (1.00)

If a reactor trip was caused by personnel errors who (by job ti tl e) is allowed to grant approval for the subsquent starttp of the rehctor?

LIST ALL APPLICABLE J03 TITLES.

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

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

YMASTS.R Sa__IBEORX_DE_duCLEA3_EDWEE_ELA3I_DEEEAIIDut_ELUI2SA_Ah2 PAGE 42 IbsR5DQ15A51CS ANSWERS -- SEQUOYAH 162 -85/05/20-VIbNCLAs A.

A NS tJ ER 5.01 (2.00)

a. 3.

b. 2.

REFER 6HCE SONPs Q& A Bank s sec l-11.

S Qr4 P s Review of ?!eutron Kinetics Lesson Plans p d.

ANSWER 5.02 (1.00)

Curve 1. - 8.< o ,

Curve 2. - C Curve 3. - D Curve 4. -A REFERENCE e,WNTs Chapter 3, pp 3-44 to 3-53.

2. Cye.it 3 h kov m s ksen hAbdsk ctA- v evsb.

ANSWER 5 03 (1 00) c.

REFERENCE SQNPs T.S. 3.1.1.3; AND Review of Reactivity Coef ficients Lesson Plans p 4.

ANSWER 5.04 (1 00) b.

REFERENCE S Qi4 P s 6 Factor Formula Lesson Plans pp 3 - 5.

-1 __IBEDSY_DE_SUCLEAS_EDMEB_ELAdI_DEEEAIIDHA_ELUIDSA_AUD PAGE 43 IufR50DISABICS ANShERS -- SEQUOYAH 162 -85/05/20-VIbHOLAs A.

ANSdER 5.05 (4.00)

a. 1.

b. 3.

c. 3.

d. 4.

REFERENCE SQNPs Review of core poisons lessons p. 6 ANSWER 5.06 (1 00) c.

REFERENCE WNTCs HTFFs Chapter 12s p 15.

o. s-ANSWER 5.07 _L1,40r"

a. False.

b. True. bh<. Le 5%sa h,fo <( no4 3 <ader)

REFERENCE SONPs HTFF text p. 202 A N Si!E A 5.08 (1.00)

a. FALSE

b. TRUE CO.5 ea.]

REFERENCE GCneral Physicss HT6FF, pp. 155 and 320 and Subcocied Liquid Density Tables

.' 2t__IMEDEX_DE_SUCLEAE_EDMEB_ELAMI_DEEEAIIDUt_ ELD 123A_AMQ PAGE 44 IdEEEQDINA51CS ANSWERS -- SEQUOYAH 1E2 -85/05/20-VINNOLAs A.

ANSWER 5.09 (1.00) 4.

REFERENCE UNTCs Reactor Controls Chapter 8. pp 19-22.

ANSdER 5.10 (1.00) o.

REFERENCE Tec nn i c al Specifications, B 3/4 1-1.

A NS W ER 5.11 ( .75)

a. 3. or 4.

b. 1.

c. 4.

R EF ER EtiCE S QtlPs HTFF texts pp. 174 -185 ANSWER 5.12 ( .75) b.

REFERENCE Wilt C s Reactor Control, Chapter 9s p 17.

ANSWER 5.13 ( .50)

False.

REFERENCE S QtlPs HTFF texts pp 169-170.

6

. la__IMEQBI_QE_uUCLEAR EQWEE_ELASI DEEEAIIQut_ELU1HSA_ANQ PAGE 45 IMEEBDQ1SABICS ANSWERS -- SEQUOYAH 152 -85/05/20-VIhNCLAs A.

ANSWER 5.14 (1 00) a.

REFERENCE SQNP, Technical Specificationss p B3/4 2-2.

A t45 W ER 5 15 ( .50)

False.

REFERENCE SON, Review of Reactivity Coef ficients Lesson Plans pp 4- 10.

ANSWER 5 16 ( .50)

False.

REFERENCE SQNs Review of MNeutron Kinetics Lesson Plan, pp 6 - 8.

ANSWER 5 17 ( .50)

TRUE.

REFERENCE SONS HTsFFsTHERMCs Lesson Plans pp 12 - 15.

AtlSWER 5 18 (1.50)

a. True.

b. False.

c. False.

REFERENCE 1 SQNP) HTFF, pp 16, 17 and 19.

2.__ IM EQEI_ D E_B u C LE A E _E Q'JE R_E L A dl_D E12AIIDS4_ EL U 1Gla_ AB Q PAGE 46

, IMEE5001dAMICS ANSWERS -- SEQUOYAH 162 -85/05/20-VINNGLAs A.

ANSWER 5.19 (1.00) b.

REFERENCE SQNs Neutron Sources and S ubc r i t i ca l Multiplicaticn Lesson Pl an, pp 4 - 6.

ANSWER 5 20 (1.00) o.

REFERENCE SQNs Review of Core Poisons Lesson Plans pp 3 - 5.

ANSWER 5.21 (1.00) a.

REFERENCE SQNs HisFFsTHERMCs po 15 - 17.

ANSWER 5 22 (1 00) 1 b.

REFERENCE SQNs Review cf B asic Nuclear Concepts Lesson Plans p 4.

ANSWER 5.23 (1 00) d.

REFERENCE Steam Taules.

. la__IBEQal_DE_duCLEAa_EDMEB_ELAdI EEEEAI1QUA_ELu1DSA_4UQ PAGE 47 IMEadQDIDA51CS ANSWERS -- SEQUOYAH 1&2 - 85/05/20-VIhNGLA, A.

ANSWER 5.24 (1.00) de REFERENCE WNTOs Radiations Chemistry and Corrosion... for Nuclear Power Plants, pp 7- 13.I 14.

2 ANSWER 5.25 (1.00) c.

REFERENCE S team T ab l es s

ANSWER 5.26 . ( .50)

False. (T

b eyeia O REFERENCE SQNs Revi ew of Reactivi ty Coef f icients Lesson Plan.

- ANSdER 5 27 (1 00)

a. False.

b. True.

REFEnENCE WNTO Station Nuc Eng Text; 1-3 13 through 19 ANSWER 5 28 (1.00)

< a. False.

b. True.

REFERENCE WostinJhouse Thermal Sciences C ha pter s 3s 5 & 10.

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

14__IMEQ81_DE_duCLEAE_EDMEB_ELAUI_DEERAIIQut_ELUIQSA_AHQ PAGE 48 IME85JDIDA51CS ANSWERS -- SEQUOYAH IC2 -85/05/20-VIhNOLAs A.

ANSWER 5.29 ( .50)

False.

REFEAENCE WNTCs Tnermal-Hydraulic Principles and Applications, pp 13 - 18.

f 4

0

-c- . , - . , . , , _ . . ,----n. -,y n,..-. - - p , , - - . - . , . - , - - -

ha__ELAMI_SISIEd3_DE11GNt_EQUIRQLt_AUQ_1U1IEudENIAIIGN PAGE 49 ANSWERS -- SEQUOYAH 1&2 -8 5/ 0 5/ 20-VI hNC LAs A.

ANSWER 6.01 (1.00) b.

REFERENCE System Manuals Chapter 3, pp 4s 10 SQNP System Descriptionss CVCS ANSWER 6.02 (1.00) de av C REFERENCE System Manuals Chapter 3, pp 3-5 C OT -14, pp M -11. .

ANSWER 6 03 (1.00) c.

REFERENCE System Manuals Chapter 4s p. 4.0-2 ANSWER 6.04 ( .50)

False.

REFERENCE System Manuals Chapter 4s p. 4.1-8 ANSWER 6 05 (1 00) a.

REFERENCE Reactor Protection Lessons p. 8 o f 13.

. 44__ EL AUI_111IE ES _ DE11G U A _ C Q UIED L a _ A UQ_INSIEU M E D IAIIL B PAGE 50 ANSWERS -- SEQUOYAH 162 -85/05/20-VINNOLA, A.

ANSWER 6.06 ( .50)

True.

REFERENCE Roactor Protection Lesson, p. 8 o f 13 s item d.

ANSWER 6 07 (1 00) b REFERENCE System Manuals Chapter 4.2s p. 4. 2- 20 ANSWER 6 08 (1.00) d REFERENCE System Manuals Chapter 4 2, pp 4.2-13 and 4.2-45 ANSWER 6.09 (1.50)

a. 1870, 1970

c. 2/3

c. 1970 [0.5 each3 REFERENCE Reactor Protection Lessons pp 11-13.

ANSWER 6.10 (1.00) c.

REFERENCE Sys te m De s c r i p't i on, Chapter 7s pp 768 of 8 l

l

I h a__ EL AMI_111IE51_ Q11 igd a_CCUIED La _ AdQ_Iu1IEud E MI AIIE h PAGE 51 ,

1

ANSWERS -- SEQUOYAH IE2 -85/05/20-VINNCLAs A. j l

ANSWER 6.11 (1.00)

C.

REFERENCE Aux Feed Lesson Plans p. 5 of 8.

ANSdER 6.12 (1.00) a REF ER ENCE Aux. Feed Lesson Plans pp 4-6.

i l

! ANSWER 6.13 (1.00) b DEFERENCE System Manuals Chapter 11.9s pp 11.9-1 thru 5.

SQNP RCS Lesson Plans pp 24-31.

ANSWER o.14 (1.00) be REFERENCE SONP Olesels handouts p. 6.

l A NS rJER 6 15 (1 00) l a

> REFERENCE Sys tem Manuals Chapt e r 11.5s Fig. 11 5-1, p. 11.5-39

. ha._ELAMI_11SIE53_EE11Gua_CQUIEDLa_AUQ_ISSIEMBENIAIILh PAGE 52 ANSWERS -- SEQUOYAH 162 -85/05/20-VIhNCLA, A.

ANSWER 6 16 (1 00) a or d REF EREt4CE SQNP Excore NI Lesson . Plan, p 14. ,

inm.de c s pa.L q 6,4 MM E3 sw.3 <s y r %4. vw* A . d wsn$ sih 0is&.

ANSWER 6 17 (1.00) be REFERENCE Rod Control Lesson, p. 4 of 11.

ANSWER 6.18 (1 00) c REFERENCE SQNP nod Control Lesson, pp 5&6 o f 11.

Sys tems Manual, Chapter 11.1, p. 11.1-63.

ANSWER 6 19 (1.00)

1. Lack of normal voltage on alternate feeder (bus).

2. Overcurrent on shutdown board.

CO.5 each)

2. Tvder w d.h ook tw a &wti e.

REFERENCE Review of Elec. Distribution, p. 4.

ANSWER 6 20 (1 00) de R EF ER EilCE Review of Elec. Distribution, p 4.

ha__ELa3I_111IE51_QE11GUA_CDHIRDLA_ABQ IUSIEUBENIAIIDH PAGE 53 ANSWERS -- SEQUOYAH 1&2 -85/05/20-VINNGLAs A. E' ANSWER 6.21 (1.00) a REFERENCE Review of Elec. Olstribution Lesson, p 10.

ANSWER 6.22 (1.00)

a. True.

b. False.

REFERENCE SQNP Olesel handout, p. 7. .

ANSWER 6 23 (2 00)

1. Auto start of all aux feed pumps, CO.53.

2. Operating NFPT goes to max. speeds (0 33.

3. Isolation of the affected MFPT condenser, CO.53.

4. Main Turbine runoack to 757 Power, Co.53

5. S t e am generator clowdown isolates, Co.23.

REFERENCE System Manual, Chapter 10.2, p. 10.2-7 ANSWER 6 24 (1 00) l l b.

REFERENCE System Manuals Chapter 10 2s p. 10.2-7 l

l

4.__ELABI_SISIE21_DillGSA_IQUIEDLA Ah0_1SSIEU5ENIAIIDb PAGE 54 ANSWERS -- SEQUOYAH 162 -8 5 / 0 5 / 2 0-V I NN G L A, A.

ANSWER 6 25 (2.00)

a. CLOSE

b. CLOSE

c. OPEN

d. CLOSE

e. N0

f. OPdN

g. CLOSE

n. NO

i. NO J. NO [0 2 ea. ]

REFERENCE SQNP System Descriptions ECCS, CVCS, MNSTMs CCW ANSWER o.26 (1.00) 3.

REFERENCE SQNP System Descriptions SGLCSs p. 11 7-9 thru 11 7.10 AN3WER 6 27 ( .50)

False.

REFERENCE SQNs G 01- 1s p 4.

ANSWER 6 28 (1 00) de R EF ER ENC E SQNP System Descriptions, Condensate and Feedwate rs pp. 9 6 10 of 11 ANSWER 6.29 (1 00) ,

c.

, 64._ELASI.SISIE53_DESIGSA_LDUIRDLa_Aha_1HSIZUSEMIAIIED PAGE 55 ANSWERS -- SEQUOYAH 1C2 -85/05/20-VIhNGLA, A. I REFERENCE SQNP SOI c2.18s p. 2 of 8 e

r

. Za__EEQCEQUEE1_=_hQadALa_ASUQEEALA.E5ERGEBCX_AMQ PAGE 50 RADIDLOGICAL_COMI30L ANSWERS -- SEQUOYAH 162 -dS/05/20-VIhNGLA, A.

ANSWER 7 01 (1 00) b of a REFERENCE SQNP GOI-1s p. 4

sot 69, t , p 5.

ANSWER 7.02 (1.00) d PEFERENCE SQNP G01-1, p. 47 precaution T.

ANSUER 7 03 (1 00) b REFERdNCE SQNP G01-1s p. 4 ANSWER 7.04 (1.00) c REFERE4CE SQNP G01-2s p. 9 ANSWER 7,e 0 5 (1 00) b REFERENCE SQi4P G01-2, p. 16

r

< Za. 2AGCEQu1ES.=.dQEdALa.ABSDRBALA_EBERGEBC1 AMD PAGE $7

, SA210 LOGICAL.CQSIRQL ANSWERS -- SEQUOYAH 1&2 -8 5 / 0 5/ 20-VI NNO L As A.

l l

ANSWER 7 06 (2.50)

, so False.

b. True.

c. True.

de True.

o. True. CO.5 each)

REFER 6NCE SQNP GOI-5 As pp 2&3 ANSWER 7 07 (2 00)

a. 3 .

b. 6 l

c. 1 Os 2 l REFER 6HCE SQNP G01-5As pp 5-8 ANSWER 7.08 (1.00) d REFERENCE SQNP ES-1.2s p. 1 i

l ANSWER 7.09 (1.00)

O.

REFERENCE l SQNP ES-1.2s p. 7.

L

r-

. Za__EEQCEQUdES =_hQldALA_ADUQEd&La_EfEEGEBCI_ASQ PAGE 58 14GIDLDEICAL_CONIEQL ANSWERS -- SEQUOYAH 162 -85/05/20-VIhNOLAs A.

ANSWER 7.10 (2.00)

1. Veri f y Reactor Tr ip.

2. Verify Turbine Trip.

3. Verify Shutcown Boards Energized.

4. Cneck If SI Actuateo.

5. Verify ECCS status.

6. Verify Cntmt. Isolation.

7. Verify PFW Isolation.

8 Verify AFw statas.

9. Verify CCS Pumps Running.

10. Verify ERCW Pumps Running.

11. Verify EGTS and ASGTS Running.

12. Check Cntmt. press less tnan 2 81 pslo.

13. Check Tavg. .

R EF ER E *1C E SQNP E-Os pp 2-5 ANSWER 7 11 (1.00) t' 4 REFERENCE SQHP FHI-7s p. 1, w v4 vision ANSWER 7.12 (2 00)

^

a. 2 .

b. 5

c. I

d. 5 CO.5 each]

REFERENCE SQNP RCI-1s p.7 A NSW ER 7.13 (1.00) 0

Za..E1QCEDURES.:.hD&5&LA ABNQEd&LA.EMERGENC1_AMQ PAGE 59 RAQ1DLDGICAL CDdI3DL ANSWERS -- SEQUOYAH 152 -85/05/20-VIhNCLAs A.

REFERENCE SQNP ES-0.3, p. 6 ANSWER 7.14 (1.00) b.

REFERENCE SQNP G01-2s p. 2 ANSWER 7.15 (1 50)

a. False

b. True

c. False REFERENCE SQNP 501-74.1, pp. 3, 4 ANScER 7.16 (1 00) a.

REFERENCE SQNP S01-62.1Bs pp. 8s 9 ANSWER 7.17 (1 50)

a. F+44. F he

b. True.

c. False.

R EF EP 6NCE

.SQNP ES-1.3 p. 1 of 4; ES-1.2 p. 1 of 3s App. A

. Ia _EEQCEQURE1_=_bORBAtt_AAUQRBALA_E5EREESCI A3Q PAGE 60 RADIGLDGICAL_CDMIRQL ANSWERS -- SEQUOYAH 1&2 -85/05/20-VIhNCLA, A.

ANSdER 7 18 ( .50)

False REFERENCE SQNP RCI-1s p. 4 A NS d ER 7.19 (1.00) ao REFERENCE SQNP A01-ZDs pp. 10 - 12 ,

ANSWER 7.20 (1 00) a.

REFEREt4CE SQNP ES-0.ls p. 2 of 13 ANSWER 7.21 ( 50)

True REFERENCE SQNP FHI-7s p. 2 ANSWER 7.22 (1 001 ONE (1)

REFERENCE SQNP E-2, p. 3 t __

o Za__tt0CEQutE1_=. bat 3&La_AADQi3ALA.ECERGENC1.AMQ PAGE 61 RA01DLDEIGAL GQNIdQL ,

l ANSWERS -- SEQUOYAH 162 -85/05/20-VIhNCLAs A.

I ANSWER 7 23 (1.00) d REFERnNCE SQNP ES-0 2s p. 4 ANSWER 7.24 ( .50)

True.

REFERENCE SQNPs E-3s P 4. .

ANSWER 7.25 (1 00) c.

REFERENCE S QN P E- 1, p. 2 of 11 ANSWER 7.26 (1.00) d REFERENCE SQNPs FHI-7s p 4.

i l

o a.. 4GaluisIt&IIVi_EEDGEDu!Est CDUQ1110 Usa _Asa_LidlIAI1 Qui PAGE 62 ANSWERS -- SEQUOYAH 162 -95/05/20-VINNCLA, A.

ANSF.ER 8.01 (1.00)

C REFERENCE SQNP AI-25s P. 5 ANSWER 8.02 (1.00) d

\ REFERENCS SGNP Al-30, pp 1-4 ANSWER d.03 (1 00) a t

REFERENCE SQNP AI-37, pp 1-4 l

l l

l t

L

4 ta__AQuidliIRAIIME.EEQCEQUAElt CQHD1I1 QUIA _AHQ.LinlIAIIQSI PAGE 63 ANSWERS -- SECUOYAH 1E2 -85/05/20-VIbNGLAs A.

ANSWER d.04 (2.00)

SEE NEXT THREE PAGES, WHICH ARE EXCERPTS FRO l1 AI-18.

ANY ITEM LISTED IN 10CFR$0.72 WILL BE ACCEPTED INCLUDING ANY OF THE BELOW1

1. Initiation of AEP.

2. Exceeding a Tech. Spec safety limit.

3. Any event that places the unit in an unexpected or uncontrolled condition.

4. Loss of pnysical security ef rectivenesss sabo tage or acts of sabotage.

Se Snutdown due to a Tech. Spec. LCO.

6e Personnel error or procedural insdequacy which prevents or could prevent the fulfi llment of saf ety functions Irportant to safety.

7. SIS actuation.

8. Accidentals unplanneds or uncontrolled radioactive release.

9 Any fatality or serious injury requiring transport to offsite redical facility.

10. Personnel contanination requiring extensive ensite decontamination or outsice assistance.

11. Any event meetin) the 10CFR20.403 cr i ter i a.

12. Strikes by oper atino employees or secur i ty g uar ds, or honoring of picket lines by these employees.

REFERENCE 10CFR50.72 AI - las pp 64 - 67s 73 - 74.

ANSWEp 8.05 (3.00)

a. 1-Red be 5-Yellow

c. 6-Orange de 5-Yellow

e. 1-Red

f. 3-Green (0.5 esch3 REFEAENCE SQNP AI-2s p. 9 SONP Question and Answer Hanns sec. 9a p. 6 ANSWER d.06 (1 00) b

8.o %

o SQNP AI-18 Appendix A Page 1 of 11 Rev.1s3)

FII.E PACKAGE NO.18 FILE C0\T.R PAGE NOTIFICATION AND LICENSEE E\T.NT REPORT (LER)

Responsible Sect n: Operations / Compliance Report Initiation uditions: Conditional References - Program rocedures 1200R03 and 1200R06 10 CFR .72 10 CFR 50.'

10 CFR 20.4 10 CFR 20.405 10 CFR 50.54 (x)

Sequoyah Technical pecifications NUREG 1022 Standard Practice SQA-NOTE: Section 6.9.1.12 and .1.13 o e technical specifications are no longer applicable aft J ry'1, 1984, and should not be used for determining repor ility. .

Type of Notification / Report I. Immediate Notification - NRC A. The Immediate Notif ation Criteria of 10 CFR 50. ' is divided into I hour and 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months phone alls. Notify the NRC Operations ' Center (red phone) within the applicabl ne or four hour time limit for any it which is identified in the Imm ate Notification Criteria.

NOTE: e checklist (pages 79 and 80) as a guideline for the vpe of information which may be requested by the NRC Operations enter.

Complete a PRO form (SQA-84). -

B. Immediate Notification Criteria -

1. The'following criteria require I hour notification (50.72.a.1.1) a. ,The declaration of any of the Emergency Classes specified in the licensee's approved Emergency Plan.

(50.72.b.1.1.A) b. The initiation of any nuclear plant shutdown required by the plant's technical specifications. -

(50.72.b.1.1.B) c. Any deviation from the plant's technical speicifications authorized pursuant tog 10 CFR 50.54(x). ,

(50.72.b.1.ii) d. Any event or condition durina operation that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded; or results in the nuclear power plant being:

%.o4 SQNP 3

AI-18 Appendix A Page 2 of 11 Rev. 33 File Package No. 18 (Continued)

d. (Continued) (i) In an unanaly:ed condition that significantly compromises plant safety; (ii) In a condition that is outside the design basis of the plant; or

. (iii) In a condition not covered by the plant's operating and emergency procedures.

(50.72.b.1.iii) e. Any natural phenomenon or other external condition that poses

/ an actual threat to the safety of the nuclear power plant or

, significantly hampers site personnel in the performance of duties necessary for the safe operation of the plant.

( '

, (50.72.b.l.iv) f. Any event that results or should have resulted in emergency core cooling system (ECCS) discharge into the reactor coolant system as a result of a valid signal.

, (50.72.b.l.v) 3 Any event that results in a major loss of emergency assessment capability, offsite response capability, or communications capability (e.g. , significant portion of control room indication,

. emergency notification system, or offsite notification system).

, (50.72.b.l.vi) h. Any event that poses an actual threat to the safety of the nuclear power plant or significantly hampers site personnel in the performance of duties necessary for the safe operation of the nuclear power plan.:. including fires, toxic gas releases,

[ or radioactive releases.

(20.403.a) 1. Any event meeting the criteria of 10 CTR 20.403 for notification, p (see following list for 10 CTR 20.403 reporting requirements, also see additional reporting requirements in file package 19).

(i) Any incident involving byproduct, source, or special nuclear

! material possessed by the licensee that may have caused or 2

' threatens to cause:

i I (a) Exposure of the whole body of an individual to 25 rems j or more of radiation; exposure of the skin of the whole body of any individual of 150 rems or more of ,

,' radiation; or exposure of the feet, ankles, hands, or forearms of any individual to 375 rems or more of radiation;,or

\

, %. 04 p SQNP

. AI-18 I Appendix A l

Page 3 cf 11 Rev. 33 File Package No. 18 (Continued)

(b) The release of radioactive material in concentrations wcich, if averaged over a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , would exceed 5,000 times the limits specified for such l

materials in 10 CFR 20, Appendix B Table II, or (c) A loss of one working week or more of the operation of

. any facilities affected, or (d) Damage to property in exesss of $200,000.

i b. The following criteria require 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months notification:

(

(50.72.b.2.1) Any event, found while the reactor is shut down, that, h I l . it been found while the reactor was in operation, woul have esulted in the nuclear power plant, including its p ncipal l s ety barriers, being seriously degraded or being an una lyzed condition that significantly comproni s plant ,

safet

! (50.72.b.2.ii) 2. Any event r condition that results in man or automatic actuation o any engineered safety featu (EST), including l the reactor p tection system (RPS). pwever,actuationof an ESF, includt the RPS, that resul s from and is part of the preplanned se uence during test' g or reactor operation

! need not be reporte .

(50.72.b.2.iii) 3. Any event or condition at a ne could have prevented the I fulfillment of the safet f etion of structures or systems

( that are needed to:

.c (i) Shut down the re tor and aintain it in a safe shutdown l

'. condition; r

(ii) Remove res ual heat; i -

(iii) Contro he release of radioactive terial; or (iv) Mit gate the consequences of an acciden j

(50. 72. b . 2. iv. A) 4. Any rborne radioactive release that exceeds 2 es the .

i ap icable concentrations of the limits specified l pendix B, Table II of 10 CFR 20 in unrestricted a s, when averaged over a time period of one hour.

\

G6

3 Ra__Ad51811IEaIIVE_Ea2CEDUEElt_CDSDIIIQuit_ANQ LIMIIAIIQNX PAGE 64 ANSWERS -- SEQUOYAH 162 -85/05/20-VINNOLAs A.

REFERENCE SQNP AI-2s pp 2-5 l

ANSUER 8.07 (1 00) c

, REFERENCE SQtlP AI-3, p. 8

(

l l

ANSWER d.08 (1 00) l D

l R E F ER dtiC E SQitP AI-3s p. 20 ANSWER- 3.09 (1 00)

c. (STA)

REFERENCE SQNP AI-9s p. 4 ANSWER 8.10 (1 00) d REFERENCE SQMP AI-9s p. 5 ANSWER 8.11 (1 00)

D REFERENCE SQt4P PLS, p. 6 E _--- - - - - - - - -

a o d c__ AQdIBISIRAIIM E_REDC EDURElt_CDH DIIIDH3t_ AHQ _ LIdlI AIIDH1 PAGE 65 ANSWERS -- SEQUOYAH IC2 -8 5 / 0 5/ 2 0-VI hN G L As A.

ANSbER 8.12 (1 50)

1. Source range high neutron flux trip

2. Intermediate ranJe high neutron flux trip Cert;;..n.snt hl-h-h!;5 ;.:::ure s. a rey ?ctu:t;;n ^.- s e,;h :

E (M Aw,La.ma w Tey)

REFERENCE SONP PLSs p. 7 ,

s At45W ER 8 13 (1 00) a REFERENCE -

SQNP Technical Specifications 3.2.4s p. 3/4 2-15 ANSdER 8.14 (1.00) d REF ER Ef4CE SQNP Tecnnical Specification 3. 4.1. 4s p. 3/4 4-2b A t4 SU ER 8 15 (3.00)

a. 7. - 40 gpm D. 3. - 1 gpm

c. 3. -1 gpm

d. 1. - Zero apm

e. 5. - 10 gpm

f. 3. -1 gpm REFERENCE SQNP Technical Specification 3.4.6.2, p. 3/4 4-14 ANSWER 8 16 (1 0C) a

r O

a d a__ AD 51 SISIE AIIE E_EE D C E D UB El t_CD U QlIIQd St_AdQ_ LIdII A IIQB1 PAGE 66 ANSWERS -- SEQUOYAH IC2 -85/05/20-VINNCLAs A.

REFERENCE SQNP Tecnnical Specification 3.9.ls p. 3/4 9-1 ANSWER 8.17 (1.00) c REFERENCE SONP Tecnnical Specification 3.9.7s p. 3/4 9-7 ANSbER 8.18 (1.00) d REFERENCE SQNP Tecninecal Specification 6.7s p. 6-14 ANSWER 8.19 (1.00)

-b REFERENCE SQNP Tecnnical Speci fications 6.2 2 and Table 6 2-1 ANSWER 8 20 (1.00) a REFERENCE S Q:4 P AI-2 5 s p. 6 ANSWER 8.21 (1 00) c REFERENCE SQNP OSLA58s p. 11 and Appendix F

,d as__A2did1SIEA111E_E30CEDUREla_CDUD1110dSz_ASQ_LidlIAI1Qdi PAGE 67 D

ANSWERS -- SEQUOYAH 162 -85/05/20-VIhNCLAs A.

3.

ANSWER 8 22 (1 00) c REFERENCE SQNP OSLA58s p. I r

ANSUER 8.23 (1.00) b REFERENCE SQNP AI-3s p. 2 ANSWER 8.24 L .50)

True.

REFERENCE Tech. Spec. 3/4.8 E 12-20-83 occurrence.

ANSWER 8.25 (1.00)

P l ant Man a2er. U'issbSw++d Plant Superintendent (GEE). (4s.4 Em, - T Operations Superintendent.

9c$ . .- 4r-19 , p S1.T 3 AI-2. , p W .

ML20128L679

Text

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