Text

Exam Rept 50-206/OL-86-01 on 860603-05.Exam Results:All Five Candidates Passed Written & Oral Exams.Two Senior Reactor Operator Candidates Withdrawn by Facility.Related Info Encl
	ML20203L055
Person / Time
Site:	San Onofre
Issue date:	07/31/1986
From:	Elin J, Meadows T; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:	;
Shared Package
ML20203L054	List: ML20203L050; ML20203L055;
References
	50-206-OL-86-01, 50-206-OL-86-1, NUDOCS 8608220353
	Download: ML20203L055 (96)
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Enclosure (1)'

U. S.-NUCLEAR REGULATORY COMMISSION REGION V

, EXAMINATION REPORT.

Examination; Report =No.: 50-206/0L-86-01 ,

s Facility: San Onofre Nuclear Generating Station, Unit 1 Docket No.: 50-206. .

Facility License No.: DPR-13 Examinations administered at San Onofre NCS, San Clemente, California.

' Chief Examiner: s _ p g r/2 y/n .

Thomas R. Meadows i Date Signed n

-Approved by: 4 h / 6- _

s Nh Ddte Signed Johp.Elin,Chier,OperationsSection

. Summary:

Examinations on June 3-5, 1986.

Written and oral examinations were administered to four SRO candidates and one R0 candidate. Two SRO candidates were withdrawn by the facility. All of the remaining candidates passed.

8608220333 860806

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PDR ADOCK 05000206 PDR

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- ' REPORT DETAILS

^ 1 '. Examiners: -

.

Thomas'R. Meadows,'RV: i '

' John O. Elin, RV- t

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Chief Exa$iner ,

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, . 2. Persons' Attending the-Exit Meeting:

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s' T. R. Meadows,t RV - ' ' "^

R. Pate, RV- '

_ M. Kirby, SCE ,, .

R. Mette. SCE J. Wambold, SCE , , . ;

M. Barr,.SCE ,

+

3. Written Examination and Facility Review: ~4- ,

Writ' ten examinations were administere'd to four SRO candidates and one RO candidate on Jene 3, 1986 in the facilities training building.

The facility staff reviewed the exams immediately upon their conclusion.

The comments made.by the staff are attached. All comments were resolved.

Appropriate changes were made to the applicable SRO and RO exam keys prior to the grading of the exams.

4. Operating Examinations:

A general training weakness was noted during oral examinations conducted June 4-5, 1986. The candidates knowledge level was very weak in the operation and application of portable radiation detectors, and in the identification of types of ionizing _ radiation.

5. Exit Meeting:

On. June 5, 1986, the Chief Examiner met with the licensee representatives s 11sted in paragraph 2. The weakness noted in paragraph 4 was discussed.

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-ATTACHMENT 1 SONGS 1 RO EXAM (JUNE 3, 1986)

FACILITY COMMENTS /NRC RESOLUTIONS NOTE: Facility comments were submitted on a copy of an exam key,provided for this use. This key-is attached.

1.4 Comment

Part 1.4.b is a double jeopardy question; Resolution: Comment not' accepted. Answer key eliminates possibility-of double jeopardy application in grading.

1.7 Comment

Correct TYPO error. Change." takes always takes" to "always takes".

Resolution: . Comment accepted. Answer key corrected.

1.8 Comment

Part 1.8.C is not applicable to SONGS'I since the facility does not have variable speed pumps.

Resolution: ' Comment not accepted. Facility equipment has no' bearing.

on this academic question. Candidates should associate pump speed with system flow and effects on NPSH.

1.10 Comment: Correct TYPO error. Change point value from 0.5 to 1.0 for applicable Sections 1.10a and 1.10b.

Resolution: Comment accepted. Answer key corrected.

2.6 Comment

The trip setpoint (1500 PSIG) is not specifically asked for by the question, therefore, should not. be required for full credit by the answer key.

Resolution: Comment accepted.- The answer key was modified such that it is clear that a specific number is not necessary for full credit, as long as the answer provided establishes' the knowledge level required.

2.7 Comment

The facility does not consider a responding alarm to be an " automatic'. function". Also, alarms given in answer key don't'ref1'ct e the information in actual training references.

Resolution: Comment accepted. The answer key was modified to reflect'the specific reference requested by the facility.

The facility was also informed of general inconsistencies throughout the references provided for this exam.

2 l

3.1 Comment

(1) We are concerned.that the candidates will not

. consider " Total Instrumentation" when addressing the subcooling monitor, due to logic channel separation.

(2) The reference given in the answer key is not appropriate for this question.

Resolution: (1) Comment not accepted. The question clearly states " Total Instrumentation" and is consistent with the training references provided.

(2) Comment accepted. TYPO in answer key corrected.

. 3.2 Commenti Reference stated in answer key is not appropriate for this question.

Resolution: Comment accepted. Same as 2.7.

3.3 Commest

Terminology used for specific values are not clear. The

- meaning of "immediate" RCP trip is confusing.

' Resolution: Comment not accepted. The question accurately reflects the reference provided.

3.5 Comment

The' point break down is not consistent between question parts.

Resolution: Comment not accepted. Points assigned are appropriate for R0 knowledge level. The answer key is designed to reflect the entire scope of knowledge required.

Verbatum compliance to the answer key is not necessary as icng as the correct intent of the subject area is expreesed.

3.6 Comment

The references provided are not appropriate for this question.

Resolution: Comment not accepted. References were appropriate for distinguishing exactly what kind of instruments used at the facility; however, reference'to " Westinghouse-Thermal-Hydraulic Principles and Applications to Pressurized Water Reactors" was added to the key to accommodate theory aspects of this question.

3.9. Comment

Clarify terminology concerning rod position indication by inserting (digital).

Resolution: Comment accepted. Answer key modified accordingly.

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4.2 Comment

. Facility procedures and system descriptions do not

, provide consistent information addressing this subject, 7

therefore, candidates could provide various answers.

Resolution: Comment not accepted. Same as 3.3.

,4.3' Comment: Part'4.3.b is not appropriate for an RO since at SONGS 1, he would not make this decision.

Resolution: Comment accepted. Answer key was adjusted to delete ~

part 4.3.b, however, over-all question point value will remain the same.

4.4 Comment

Correct part 4.4.b TYPO in answer key to read "200*F" vice 200*F per hour.

Resolution: Comment accepted. Answer key modified accordingly.

4.5 Comment

Question is not performanced based. The operator does -

not have to know RCP restart limits verbatum.

Resolution: Comment not accepted. The operator should be familiar with RCP starting duty requirements. Again, verbatum answer key compliance is not required for

, full credit, as long as intent of answer is met.

4.7 Comment

Part 4.7.a is an open-ended question in that the candidate is required to specify 8 items for full credit' from memory.

Resolution: Comment not accepted. Full credit will be given if intent of answer is met.

4.10 Comment: Point assignment not appropriate for question.

Resolution: Comment not accepted. Point value within acceptable range for question.

General Comment:

The RO exams over-all point structure was too high.

The facility would prefer more questions at a lower per question point structure.

Resolution: Comment not accepted. Point structure will remain as is for this exam. The point structure is appropriate, considering R0 knowledge level and time to respond to associated questions.

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ATTACHMENT 2'

. SONGS 1 SRO EXAM (JUNE 3, 1986) -

FACILITY COMMENTS /NRC RESOLUTIONS.

e l NOTE: ' Facility comments were submitted on a copy.of an exam key provided for this use. This key is attached.

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5.5 Comment

Possible double jeopardy.

I Resolution: Comment not accepted; answer key ~ eliminates possible-double jeopardy problems in grading process.

6.3 Comment

Various references have different answers for this question. ' Request you use-S01-2.3-1 for consolidating i

answer key.

Resolution: Comment accepted. Answer key modified to accommodate

- specific reference requested.- (Facility informed of

general inconsistencies within references provided).

. 6.5 Comment: . Part of 6.5.b answer should be optional. Additional _

information not required for full credit.

T Resolution: Comment accepted.' Answer key modified to clarify this intent.

3 6.12 Comment: Part 6.12.b answer should not include, " 1500 pounds

, load", for full credit since. question does not specifically ask for setpoints.

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Resolution: Comment accepted. Same as 6.5.

7.3 Comment Wrong reference stated for question.

I Resolution: Comment accepted. Typo in answer key corrected.

j 7.4 Comment: 7.4.b answer should read 200*F vice "200*F per hour."

Resolution: Comment accepted. Typo corrected in answer key.

7.7 Comment

The definition of "E" was not asked for and, therefore, should only be considered as extra information in answer.

Resolution: Comt nt accepted. Answer key modified to clarify intent, i-

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8.1 Comment

Answer key.is.not correct. Answers should be (b)' and -

(c_) vice (b) and (d_) . Question was' confusing due to last phrase "...in which the safety limit is exceeded."

i Resolution:. Comment accepted. ,~iiiiswei key changed, incorporating (b) and (c). '?Confusitiig" statement was deleted.

' 8'. 4 Comment: "Minimumcrew"..-(norma 1Iyreferstothecontrolroom

.l ~ area. Therefore,goly this requirement should be requested in the' answer key. Also, answer key should

be corrected to two(2) licensed Senior Reactor Operators vice one(1).

Resolution: Comment accepted. Answer key modified to reflect.

comment end reference (Table 6.2-1 does specify crew composition as, " control room area").

8.10 Comment: Question is not performanced based.

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Answer: Comment not accepted.' Question is applicable and, "need to know" for safe operation.

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f A U.S. NUCLEAR REGULATORY COMMISSION j s

REACTOR OPERATOR LICENSE EXAMINATION Facility: SONGS UNIT 1 Reactor Type: PWR WESTINGHOUSE 3 LOOP Date Administered: JUNE 3, 1986 Examiner: THOMAS R. MEADOWS, RV Candidate:

INSTRUCTIONS TO CANDIDATE Use separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at i least 70% in each category and a final grade of at least 80%. Examination ,

papers will be picked up six (6) hours after the examination starts. -

Category % of Candidate's % of Value Total Score Cat. Value Category 25 25 1. Principles of Nuclear Power Plant Operation.

Thermodynamics, Heat Transfer and Fluid Flow 25 25 2. Plant Design Including

~~

Safety and Emergency Systems 25 25 3. Instruments and Controls 25 25 4. Procedures - Normal, Abnormal. Emergency, and Radiological Control 2

! 100 TOTALS 4

Final Grade %

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

Candidate's Signature

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s' EQUATION SHEET f = ma v = s/t w = mg a=v,t+ at 2 Cycle efficiency =

E E = mC a = (v - y )/t f

KE = my A = AN U^

vg = v, + a t A = A,e ,

1 PE = agh m = 0/t A = in 2/tg = 0.693/tg W = vaP (t, )(ts)

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AE = 931Am h '

(t +t) b Q = jnC AT 7 ,-Ex

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Q = UAAT y ,7 ,-px Pwr = Wg m I=I o 10

M P=P 10 SR(t) TVL = 1.3/p P=P 0 et /T HVL = 0.693/p SUR = 26.06/T i

T = 1.44 DT SCR = S/(1 - K,gg)

(A*gro)

SUR = 26 i g_, CR x = S/(1 - K,ff )

T = (1*/p ) + [(f ;p)/x 1(1 - K,gg)1 = CR2 (1 - K,gg)2 p]

T = 1*/ (p ) M " I/(1 - Kegg) = CR /CR g 0 T = (3 - p)/ Aeffp M = (1 - K,gg)0 (1 - K,gg)g

" ( eff- )! eff = AK,fg/Keff SDM = (1 - K,ff)/K,gg p= [1*/TKygg] + [E/(1 + A,ggT )] 1* = 1 x 10- seconds P = E$V/(3 x 10 ) A,gg = 0.1 seconds E = No Id gy =Id 22 WATER PARAMETERS Idg =Id 2 i 1 gal. = 8.345 lbm R/hr = (0.5 CE)/d (meters)

I gal. = 3.78 liters R/hr = 6 CE/d (feet) 1 ft3 = 7.48 gal. MISCELLANEOUS CONVERSIONS .

3 10

( Density = 62.4 lbm/ft 1 Curie = 3.7 x 10 dps 3

Density = 1 gm/cm 1 kg = 2.21 lba 3

Heat of var orizations = 970 reu/lbm I hp = 2.54 x 10 BTU /hr 6

, Heat of fusica = 144 Btu /lbm 1 Mw = 3.41 x 10 Btu /hr k 1 Atm = 14,7 psi = 29.9 in,l'g. 1 Btu = 778 ft-lbf 1 ft. H 2O = 0.4333 lbf/in 1 inch = 2.54 cm F = 9/5 C + 32 "C = 5/9 ( F - 32)

SECTION 1

. Principles of Nuclear Power Plant Operation,

, Thermodynamics, Heat Transfer and Fluid Flow 1.1 (1.0)

SELECT THE BEST ANSWER. As a subtritical reactor nears criticality, the length of time to reach equalibrium count rate after an insertion of given fixed amount of positive reactivity, (a) decreases primarily because of the increased population of delayed neutrons in the core.

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

(c) decreases because the source neutrons are becoming less inportant in relation to total neutron population.

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

(d)

Ref: I Westinghouse fundamentals of nuclear physics, (Chapters 6 and 7) 1

1.2 (3.0)

Is Keff DEPENDENT or INDEPENDENT of initial source range counts?

EXPLAIN YOUR ANSWER.

INDEPENDENT- Keff(neutron life cycle) only conciders fission neutrons in the self-sustaining reaction.

Refs West i nghouse Fundamen tal s of Nucl ear Physi cs, (Chapters 7 and 8) 1 2

1.3 (3.0) \

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Concerning the Neutron Life Cycle; Match each factor of the six factor formula with it's applicable definition listed below (SELECT the correct letter, 0.5 pts each):

l (1) fast fission factor ( 6 >= (a) # of neutrons from !

neutron induced fission 1 (2) fast nonleakage factor (7)= that are thermalized in l Core.

(3) resonance escape probability (P)= # of fast neutrons from neutron induced fission (4) thermal nonleakage factor (7,,)= remaining in the core.

(5) thermal utilization factor (b) # of thermal i z ed (f)= neutrons from neutron induced fission absorbed (6) reproduction factor Os)= in the core.

of neutrons from neutron induced fission ,

thermal ized in the core.

(c) # of fast neutrons produced by thermal -

neutron induced fission.

thermalized neutrons from neutron induceo fission absorbed in the fuel.

(d) # thermalized neutrons from neutron induced fission absorbed in fuel.

thermalized neutrons from neutron inauced fission absorbed in core.

'( e ) # of fast neutrons from neutron induced fission remaining in the core.

of fast neutrons produced by all neutron induced fission.

(f) # of fast neutrons produced by all neutron induced fissions.

of fast ' neutrons produced by thermal induced fissions.

(1) f. (2) e. (3) a, (4) 6. (5)d. (6)c Ref: Westinhouse Fundamentals of Nuc Reactor Physics 3

1.4 (4.0)

A reactor is ini tially cri tical at a power level of 1.0 kw.

At time =0, the Keff of the reactor core was changed to 1.002 by rod withdraw.

(a) Cal cul ate the reactivity insertion in percent (1.0) millicho (pcm).

(b) Cal cul a te the resul tan t .reac tor per i od (T) . -- (1.0)

(Assume Beff=0.0072 and.1 =0.1) cAF (c) Cal cul ate the resultant startup rate (SUR). (1.0)

(d) At what time will the reactor power equal 100 kw? (1.0) 5 5 (a) (pcm)= Keff-1)(10 )/Keff= (1.002-1.0)(10 )/1.002

=200 pcm (b) T= Beff-@ / h for 4Beff T= (0.0072-0.002)/(0.002)(0.1) = 26 sec (c) SUR= 26/T = 1.0 dpm (Whatever T used from above, acceptable)

(SUR)(t)

(d) P=Po10 t

100 kw =(1 kw)10 t= log (100)= 2 minutes Ref:

Westinghouse Fundamentals of Nuclear Reactor Physics, (Chapter 7, Neutron Kinetics)

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4

1.5 (3.0)

During a reactor startup near the end of core life,will the actual critical position (ACP) be HIGHER, LOWER, or the SAME as the estimated critical position (ECP) if the f oll owing condi ti ons occur ? EXPLAINATION IS NOT REQUIRED

[ Concider each condition separately 3 (a) Actual boron concentration was 50 ppm higher than (1.0) the value used for the ECP.

(b) The steam dump control pressure setpoint is raised (1.0) by 50 psi about 15 minutes prior to criticality.

( Assume hot standby conditions)

(c) The startup was delayed four(4) hours beyond the (1.0)

ECP time estimate of sixteen (16) hours since normal shutdown from 100% equilibrium power.

(a) ACP higher (boron contributing negative reactivity)

(b) ACP higher (S/G pressure higher raises Tave; MTC

& FTC add negative reactivity)

(c) ACP lower ( Xenon concentration lower; adds positive reactivity)

Ref: Westinghouse Reactor Core Control, Integrated Knowledge l

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l 1.6 (1.0)

The Fuel Temperature Coefficient (FTC) increases (becomes MORE l negative) from BOL to EOL primarily due to:

(a) the reduction of fuel to clad gap distance.

1 (b) the reduction in the modarator's boron !

concentration. l (c) the increase in Pu-240 in the core.

(d) the increase in thermal neutron flux.

(c)

Ref: Westinghouse Fundamentals of Nuclear Reactor Physics J

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. 1.7 (3.0) y f t" 0 Answer True or False:

(a) Heat transfer te%#s[alwaystakesplacefrom high (1.0) to low temperatures.

(b) As water flows around a bend in a pipe, the (1.0) velocity of the water is uniform throughout the diameter of the pipe .

(c) Pressure measured in a static fluid (e.g. reference (1.0)

. leg of a level indicator) at different levels decreases with increasing elevation of measurement.

(a) True (b) Fal se (c) True Reft Westinghouse Thermal-Hydraulic Principles and Applications to PWR's (Chapters 1-5) 7

1.8 (2.0)

Of the following operations,which one will have a negative effect on available Net Positive Suction Head (NPSH) of a given centrifugal pump:

(a) Trottling open the pump's suction valve.

(b) Throttl ing open the pump's discharge value.

(c) Decreasing the pump's speed.

(d) Decreasing the temperature of the fluid (water) being pumped.

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Ref:

West i nghouse Thermal-Hydr aul i c Pr i nc i pl es and Appl i cat i ons to the PWR, (Chapter 10, Fluid Mechanics in Pumps and Pipes) 8

. 1.9 (3.03 Refer to FIGURE 1.9 Sketch the HEAT GENERATION PROFILE for heat produced by the fission process from fuel pin centerline to the adjacent bulk cool an t .

attached FigJre 1.9 KEY Ref: Westinghouse Thermal-Hydraulic Principles and Applications, Westinghouse Reactor Core Control for Large PWR's

FIGURE 1.9

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. 1.10 (2.0)

Listed below are two(2) factors that can affect Departure from Nucleate Boiling Ratio (DNBR). Given initial equilibrium power conditions, What will be the INITIAL effect on DNBR (increase or decrease) if the f actors l isted bel ow are INCREASED ?

(a) Reau to.- Power JIk57 [/ #)

(b) Reactor Coolant System flow Mg Sj (a) DNBR decreases.

(b) DNBR increases.

Ref: Westinghouse Reactor Core Control for I.arge BWR's, SONGS 1 Technical Specifications END OF SECTION 1.0

+++++++++++++++++++++++++++++***++++++++++++++++++++++++++++++++

l 10

, SECTION 2.0 Plant Design Including Safety and Emergency Sys.t ems 2.1 (3.0)

State the valve and pump lineups necessary for Main Feedwater Pump minimum suction requirements for each of the conditions belows (a) One Main Feedwater Pump in operation. (1.5)

(b) Two Main Feedwater Pumps in operation. (1.5)

(a) Feedwater Pump Suction Valve open and two Condensate Pumps running for the operation of one Main Feedwater Pump.

(b) Feedwater Pumps Suction Valves open and three Condensate Pumps running for operation of two Main Feedwater Pumps.

Ref: SONG 1 Operating Instruction S01-7-2 1

2.2 (4.0)

(a) What are two sources of actuating power for opening (1.0) the PORV's (CV-545& CV-546)?

(b) With the reactor coolant system solid at 400 psig .(1.5) and the OMS armed, what is the action of the OMS if a 200 psig pressure increase occurs?

(c) What are the three control room al arms directly (1.5) associated with the event described above?

(a) (0.5 each) (1.0)

Instrument air header Nitrogen Bottles (b) At approximately 500 psig CV-545 and CV-546 (1.5)

(PORV's) open[ relieving pressure to the P2R relief tank),

(c) The following annunciators alarm in the control room: g OMS HIGH PRESSURE (480 psig) (0.5) !

PRESSURE TRANSIENT in PROGRESS (500 psig) (0.5)

PORV Open(Valve limit switch) (0.5)

Ref:

SONG 1 Operating Instruction 501-2.1-11, Reactor Coolant System Instrumentation Study Guide, Technical Specifications 2

d _ . . _. _.

2.3 (2.0)

Concerning the emergency Diesel Generating System, answer the f oll owing TRUE or FALSE:

(a) The D/G speed droop is set to protect the diesel (0.5) ,

from over loading upon bus syncronization.

(b) The D/G Regulator Mode Selector switch is located (0.5) on the D/G Iocal panel.

(c) D/G excitation can only be controlled from the (0.5)

Standby Power Generation Panel during emergency operations.

(d) The MANUAL Voltage Regulator is normally used when (0.5) paralleling with anouther generator.

(a) TRUE (b) TRUE (c) FALSE (d) FALSE Ref:

SONG 1 System Description SD-SO1-600, section IX, S01 Diesel Generator and Fuel Oil operating instructions.

l l

2.4 (2.0)

FILL IN THE BLN4K:

(a) When a D/G is to be paralleled to the 4kV bus, (1.0) real load (kW) is adjusted by operation of the (b) When a D/G is to be paralleled to the 4kV bus, (1.0) the reactive load (KVAR) is controlled by operation of the automatic .

(a) Governor Speed Control Switch (b) Vol tage Adjuster (regulator)

Ref:

SONG 1 System Description SD-SO1-600, section X ,

4

2.5 (1.0)

What is the minimum water volume, specified by SONGS 1 Technical Specifications, for the auxiliary feedwater system Storage Tank ?

150,000 gallons (minimum)

Ref:

SONG 1 Technical Specifications, 3.4.4 5

2.6 (3.0)

During refueling operations:

(a) What systems are required to be continuously (2.0) operable to monitor for inadvertant criticality ?

(b) What interlock on the fuel lifting hoist prevents (1.0) lifting more than one fuel assembly at a time?

(a) Core subcritical neutron flux shall be continuously monitored during the entire refueling period by not less than two neutron monitors (source range)( each with continuous visual indication and one with continuous audible indication).

(b) Lifting hoist circuitopens(at >1500 pounds LBS load}andstopsoutmotion.

Ref:

swwA-A pd !

Technical Specifications 3.8 SONG Operating Instruction S01-3-7, Feul Handling and Refueling Operations Study Guide 6

2.7 (4.0)

In accordance with procedure S01-3-1," Plant S/U from Col d Shutdown to Hot-Standby":

(a) With the pressurizer temperature at 430 degrees (2.0)

F,how is the CVCS aligned to form a steam bubble in the Pressurizer ?

(b) How may additional letdown be provided (per (1.0) procedure) when forming a Pressurizer steam bubble and the reactor coolant pressure is below normal ?

(c) What d automatic functions occur on a letdown (1.0) line high temperature condition ? (include both the alarm and functional condition responces- 0.5 pts.

each)

(a) Reduce the charging flow by taking manual control of the letdown flow control valve ( FCV-1112 ).

Establish letdown flow via east RHR pump bypass &

RHR loop inle ts (or ifice valves al so open) .

Verify or Set the letdown pressure control valve (PC-1105) to control at Psat for 430 degrees F.

(b) The excess letdown heat exchanger may be placed in service to increase the letdown flow rate.

(c) The demineralizers are bypassed by the demineralizer temperature control valve ( TCV-1105).

(J ? *r 'r)

C,,u f Letdown Line Hi Temp" al arm ? ? SC

&vieen F '/T ca 47a -~s n e n s i s a s A Z& c cf //ics fr**

c14- [Y Ye 'f)

Ref: S01-3-1 Plant Startup From Cold Shutdown to Hot Standby S01-4-6 Charging and Letdown System S01-4-11 Letdown Demineralizer System S01-4-36 CVCS Alingment 7

2.8 (4.0)

The plant is operating at 50% power when a rapid load reduction to 10% power occurs. The STEAM DWP OPERATION MODE SELECTION SWITCH is in the " AUTO" position.

(a) What TWO(2) input signals are re.cuired to actuate -(1.0) the steam dump system in the " AUTO" mode ?

(b) Explain the Otaam Dumn talve(both the condenser (2.0) dumps and atmospherse dumps) operation that would occur from the above transient ?

(c) What is the purpose of the Condenser Low Vacuum (1.0) interlock associated with the condenser dump valves?

(a) Two signals required

-An error signal from r ap i dl y decreasing Tref (function of ist stg. Turbine pressure-auto adjusted by the Reactor Control and Protection System) , measured against Tave.

-a rapid " loss of MWe" load (10% load instantaneously).

(b) Two condenser dump valves (CV 3&4) and four atmospheric dump valves (CV 76,77,78,&79) fully open. The atmospheric dumps are subsequently modulated completely closed (controlled via Tref control, via Reactor Control and Protection System /TC-417c ) as the " loss of load" signal decays. Steam dump control (TC-417c) continues with the Condenser Dump valves (CV 3&4 ) until the new load average is restored, at which time, these ,

valves also close completely.

(c) Prevents opening the condenser dump valves (CV 3&4) when condenser vacuum is lost (2 in. Hg).

Reft SD-S01-190 Main Steam System S01-7-15 Main and Extraction Steam System 8

2.9 (2.0)

Concerning the Containment Spray & Actuation System (CSAS):

What TWO(2) conditions (setpoints and logic) au t omat i cal l y ,

initiate the CSAS ? (1.0 each)

(1) Safety Injection (2/2) (1 sequencer per train [1/13 )

(2) High sphere pressure-10 psig (2/3)

Reft SD-S01-580 Safety Injection, Spray, and Recirculation systems SD-S01-590 Sequencer System SD-501-630 Containment and Containment Isolation Systems END OF SECTION 2.0 9

SECTION 3

. Instruments and Controls 3.1 (3.0)

The RCS SUBC00 LING MARGIN MONITOR channels are required to have two Pressurizer Pressure inputs and one of three(3) possible combinations of loop temperature inputs (That RTD's,and/or Tcore exit Thermal Couples) to be considered operable (Total instumentation).

What is one possible combination of loop temperature inputs needed to consider the Subcooling Margin Monitor operable?

Any one of the following for full credit:

(1) 2 RTD's per HOT leg (2) 1 RTD per HOT 1eg and 4 Core exi t Thermal Couples (3) 8 Core Exit Thermal Couples L/ "

Refa SONG 1 S01 M. ?e ac t c.- D ' :e t-I .. . i unm .. t & t ; c.. Og e . & ; i m.

/'

1

3.2 (3.0)

The Power Range nuclear instrumentation (comparator) compares the four Power Range channels and provides three protective features in order to protect the fuel in the event of a dropped rod. List these three protective features.

(a) Nuclear Dropped Rod Stop (1.0) 61p0 Auto Rod Withdraw Prohibit (1.0)

(7- 7 m c)

(c) Turbine 3 Runback to 70V.( i f : .. . t ! &l t y abe"a 7W (1.0) p e"'e a ? n d-tii e de i e a i- =; ; t c t. .ct ia " defeat"Y-LTwe b i n e Ruoved iv 7C% accept 2 e .' v r C.5 #

Ref:

SONGS 1 System Description SD-SO1-360 SG/ -z.3-3 2

l

.. = - . - - . . - = -. _ .- -

- 3.3 (3.0) i Consider the LOAD SEQUENCING SYSTEM upon a combined Safety Injection Signal / Loss of Power (SIS / LOP) event at time =0 .

Three of the below listed sequencer responses DO NOT occur.

IDENTIFY the three responses that do not occur. [ 1.0 pts each]

(1) Trip Reactor Trip breakers (2) Closes the PZR PORV's (Interlocked closed)

(3) Deenergize reactor trip breaker UV's

. (4) Energize L/0 relays for switchgear 1,2,and 3 (5) Energize L/0 relays for MCC's 1,1C,2,2A,and 3 (d) Trip 4160 VAC Bus 1C and 2C Tie Breakers (7) Trip D/G 1 and 2 output bkers

] (8) Isolates the Steam Generator stemn dumps (bypasses interlocked closed)

(9) Block D/G 1 and 2 excitation S/D ckts (10) Reset D/G field

! (11) Trip lighting transformer (12) Lockout Motor Heater Panels (13) Start D/G's 1&2 (2 cRts per D/G)

(14) Close Feedwater Bypass / control valves (15) Open SI header ~ Iso. valves (16) Close HP & LP Feedwater Hdr. Iso. valves (17) Trip Reactor Coolant Pumps ( immediately )

(18) First out annunciators, Auto alert systems +TSC

! (19) Trip Heater Drain Pumps 1

(20) Trip condensate pumps

! (21) Initiate event recorder

] (22) Trip Feedwater Pumps (23) Initiate Containment Isolation System (24) Open SI loop Iso valves (25) Close letdown orifice Iso-valves j (26) Close 480 VAC bus 2&3 Tie bkrs l (27) Signals Feedwater control system to start Feedwater pump j after 11 see time-delay.

(28) Trip Turbine Plant Cooling Water pumps 1

i I

(2)

(8) l (17)

I Refs SONGS 1 System Discription SD-S01-590 i

i 3

3

~

3.4 (2.0)

What is the only Diesel Generator engine trip enabled when the D/G is auto started by the sequencer?

Engine Overspeed Ref:

SONG 1 SYSTEM DESCRIPTION SD-501-600, section X 4

3.5 C4.0)

(a) What is the principle of operation of the (2.0)

Source Range Nuclear Instrumentation detector?

(b) How is Gamma flux compensated for in the source (1.0) range instrument?

(c) What inputs to the Reactor Protection and Control (1.0)

System are provided by the Source Range Nuclear Instrument Channels?

(a) The SRM detector is a BF3 gas filled chamber.

Thermal neutrons are absorbed by boron-10 atoms in the gas . These atoms subsequently disintegrate into Li7, emitting a high energy alpha partical This Alpha partical causes ionization within the gas that is collected by the chamber cathode / anode circuitry via applied voltage .

The chamber voltage is in the PROPORTIONAL range,therefore,the output event pulse AMPLITUDE is proportional to the magnitude of the ionizing event. The number of pulses is proportional to the number of neutrons reacting with the BF3 gas.

(b) Since the amolitude of the pulses from the detector are proportional to the magnitude of the ionizing event, the pulses from neutron interactions (alpha ionizations) are larger than pulses from gamma ionizations. A pulse height discriminator (gate circuit) in the log level amplifier rejects pulses below a set range (approximately 3.5 Mev) in magnitude, eliminating the gamma pulses from the remaining circuitry.

(c) Rod stop signal (at2DPM .)

Refs SONG 1 System Description SD-501-380 Technical Specifications, Reactor Plant Instumentation 501-4-34 5

3.6 (2.0)

For the following transmitters, Explain how the specified failure would affect the transmitter output Indications (State all assumptions)

(a) The failure of a Delta-P diaphragm on a (1.0) pressurizer level detection system transmitter.

(b) A rupture of the reference leg for a Steam (1.0)

Generator level detection system transmitter.

(a) The f ailure of a diaohragm would eventually result in no D/P across the instrument. This would resul t in a high indicated PZR level on the associated indicator.

(b) A ruptured reference leg would result in a low (negative ) D/P across the affected S/G transmitter. This would result in a corresponding erroneous high level indication.

Ref: S01-1.3-2 (Response to S/G High Level)

S01-1.6-1 (Response to PZR High Level)

S01-4-34 (Reactor Plant Instrumentation Ops.)

- piacis<rr no W4 'f f / W0 iW S$'- , /ptAMJ/ . g y psugsg , jp,,, ,

6

3.7 (3.0)

Two Runbacks are provided to limit Turbine load:

(a) What is the protective runback initiated by an (1.5) under frequency condition and associated power?

(Include all SETPOINTS,and source of Power sensing point)

(b) What is the protective runback initiated by a (1.5) control rod drop and associated power condition?

(Include all SETPOINTS, and source of Power sensing point)

(a) When Turbine load is >380 MW ( S01-6-4 requires a manual runback to 385 MW 3 >385 MW), as indicated by Turbine first stage pressure,AND frequency drops to 58 Hz, the Load Limit Motor runs back the Turbine to (380 MW (To about 85% load).

(b) When the Turbine load is >315 MW . as indicated by Turbine first stage pressure,AND a Control Rod drops, the load limit motor runs the Turbine Load back to 315 MW (To about 70% load)

Ref: SD-S01-270 (Turbine Control System)

S01-6-4 (Load Limit Operation)

S01-13-9 (Turbine Generator Annunciator) 7

. 3.8 (3.0)

(a) What is the purpose of the Reheater Steam Dump (1.0)

System ?

(b) Under what conditions will the Reheater Steam Dumps (1.0) operate ?

(c) How does the Control Room Operator know that the (1.0)

Reheater Steam Dump system is armed ? ( Control room indication only)

(a) To reduce the Turbine Generator Overspeed by dumping reheated low pressure turbine inlet steam to atmosphere in event of sudden loss of load.

(b) The reheater steam dump system operates when the load is >70% (350 MW)( as sensed by pressure in the crossover line), AND the generator breakers open before or simul taneously wi th stop valve closing.

(c) "

MSR STM DUMP RELIEF VALVE ARMED" light illuminated at the control panel.

Ref S01-13-9 S01-6-1 Reheater Steam Dump System SONGS 1 Study Guide 54 i

E

3.9 (2.0)

Concerning Control Rod Position Indication:

(pic,, rs i)

(a) How is Demand Rod Position and Indication,and (1.0) associated dire,ction of rod motion sensed ?

S (b) How is Actual Rod Position Indication sented ? (1.0)

(a) The digital detection system consists of step counters that track the number of times the lift coil of the rod drive mechar' ism is energized. The direction of rod motion is obtained from auxiliary relays (sensed from) the Reactor Control and Protection system.

(b) The Anal og de tec tion system consists of a LVDT with its primary and secondary coils mounted around the pressure housing of the control rod drive shaft.

Output LVDT AC voltage is rectified to DC, which is propor t i on al to rod position.

Ref SD-S01-400 Rod Control System i

END OF SECTION 3.0

- - - - ++++++++++++++++wess+++++++++++++++n*+++++++++++++++.

9

SECTION 4

- Procedures - Normal. Abnormal, Emergency, and Radiological Control 4.1 (3.0)

In accordance with 10 CFR 20, " Standards for Protection against Radiation":

(a) What are the Radiation Dose Standards for (1.5) individuals in restricted areas per Calender

-Quarter?

(b) What are the three requirements that must be met if (1.5) the Whole Body limits for a Calendar Quarter are to be exceeded?

(a) (0.5 each) 1.25 Rem - Whole Body; head and trunktactive blood forming organs; lens of eyes; gonads.

18.75 Rem - Hands and forearms; feet and ankles 7.5 Rem - Skin of whole body.

(b) (0.5 each) 3.0 Rem per calender quarter-maximum 5*(N-18) total accumulated dose to the whole body where N is the individuals age in years at his last birthday.

Form NRC-4 or equivalent.

Ref:

10 CFR 20 1

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

. 4.2 (2.0)

Concerning Reactor Plant Instrumentation Operation:

(a) At what power levels should the control rod system (2.0) be in manual?

(b) In the event an Intermediate range channel f ails in (1.0) such a manner as to turn ON its companion source range channel, What shall the operator do to the affected SRM while in the Intermediate range?

(a) The control rod system control should be in MANUAL control at power levels less than 15% of full power. (C ZE mw t E)

(b) The affected SRM channel's high voltage power supply shall be turned off.

Ref: SONGS 1 S01-4-34, Reactor Plant Instrumentation Ops.

4.3 (2.0)

The plant is operating near 100% power and a thermal calibration is performed. The indicated power after adjusting the power range instrumentation exceeds 100%:

jdd When this occurs, what action should operator j)N()

take,in addition to anouther calibration?

(b) If a ion provides verifie=+!

that the reactor ':r- grTUTously operating above 1 must be informed witnin i t. - --

ty?

,P()The operator should immediately reduce power level to an indicated 100% and make suitable log entry .

(If the required power reduction exceeds 3%,

report the problemtotheShiftSuperintendent)

(b)If a _ d thermal calibration i

. operation above 103%. r. t ~ er to the UNIT SUPERINTEN .

Ref:

SONG 1 S01-4-34 Reactor Plant Instrumentation Operation g vo s - / z . / - Z-2

4.4 (3.5)

CONCERNING PLANT STARTUP per S01-3-1,Pl an t S/U f rom Col d S/D to Hot Standby:

(a) What is the max imum all owabl e Pressurizer heatup (0.5) rate?

(b) What is the maximum temperature difference al l owed (0.5) between the Pressurizer liquid and Reactor Coolant?

(c) The RCS (except the PZR) temperature and pressure (0.5) shall be limited in accordance with given pressure / temperature curves. What is the maximum RCS heatup rate in any ONE(1) hour period?

(d) Why do operations require that the Reactor Cool ant (2.0)

Pumps be operated con t i nuou sl y (with momentary stops) during the solid water phase of heatup?

l o

(a) 95 F per hour o

1 (b) 200 F ~ 'W

.' O l (c) 60 F in any one hour period (d) To prevent temperature differences that could result in over-pressurization upon RCP restart.

Ref:

SONG 1 S01-3-1, Plant Startup from Col d Shutdown to Hot 4

Standby l

i l

l l 3 l

l r-__- -

i 4.5 (3.0)

Concerning the starting duty requirements for the RCP's per S01-4-3, RCP Operations (a) How many pumps can be started at any one time? (0.5)

(b) What are the limits on restart for each RCP ? (2.5)

(Address, both IDLE & Running Pump limits, number of restarts within a particular time period,and time requirements for idle periods)

(a) Only one RCP is to be started at any one time. (0.5)

(b) If a RCP has failed to achieve full speed after an (2.5) attempted start, restart should not be attempted until the motor has been al l owe d to cool by standing idle for a period of not less than 30 minutes. After a RCP trip, an immediate restart is permitted if the motor has been running under normal load for at least four hours.

For each RCP, the number of starts will be Ilmited to a maximum of 3 within a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months period.A minimum idle period of 30 minutes for cooling will be observed between each restart. When 3 starts or attempted starts have been made within a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months period, the fourth attempted start must not be made until the meter has been allowed to cool by standing idle for at least 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Refs SONG 1 SO1-4-3, RCP OPERATION 4

4.6 (2.5)

EXPLAIN the reason, when preparing for Main Turbine roll, per S01-3-2 Plant S/U from Hot Standby to Minimum Load, reactor power is increased sufficiently to gain 50-75% percent opening on the Steam Dump valves to the condenser.

To ensure that the reactor system (NSSS) can satisfy the turbine load demand without exceeding established safe operating bands (ie:

PZR pressure / level, variable Tave range, Reactor pressure).

Ref:

SONG 1 S01-3-2. Pl an t S/U f rom Hot S/B to Mi n Load, WESTINGHOUSE Reactor Core Control for Large PWR's 5

, e".

4.7 (4.0)

Per S01-1.0-40, Steam Generator Tube Rupture:

(a) After identification of the ruptured S/G and (2.0) verification of proper plant lineup, outline the basic method of placing the plant in a cooled down and depressurized condition.

1 (b) What caution is c!ven for depressurizing the RCS by (1.0) this method?

I (c) What caution is given for starting Reactor Coolant (1.0)

Pumps in the intact loops after depressurization?

(a) (0.25 each)

(1) Establish level in non-ruptured S/G's via aux feed.

(2) Open PORV's to allow SI to establish PZR level at 70%

(3) Start RCP (4) Align two charging pumps to inject via SI col d leg inJoction lines (5) Dump steam to the condenser for a max 100oF/ hour C/D rate (6) Depressurize the RCS via spray or PORVs(RCS 200 psi >

Steam line pressure) (If ruptured S/G overfilling then depressurize to saturation)

(7) Berate RCS to 5% S/D concentration.

(8) When main steam pressure < 350 psig place RHR in service and cooldown to 140 degrees F at less than 50 degrees F per hour.

(b) May result in steam voiding in the upper head and rapid fillin0 of the pressurizer.

(c) Pressurizer level may drop due to the collapsing bubble in the upper vessel head.

Ref: SONG 1 S01-1.0-40, Steam Generator Tube Rupture l

l

! l 6

l

8 4.8 (1.5)

S01-2.1-4, LNPLMNED BORATION, spec i f i es f our(4) symtoms for this abnormal operation. List three of these.

Any 3 of the following (0.5 each):

(a) If rod control is in automatic, rods will travel out with associated " Rod Withdraw Bank #2 High Alarm".

(b) If rod control is in manual, Tave will decrease.

(c) Boric acid flow may be indicated on blend system recorder, FR-1102.

(d) " Boric Acid Tank Lo Level" (65%) al arm may be indicated.

Ref: SONG 1 S01-2.1-4, UNPLANNED BORATION 7

1 4.9 (3.0)

In accordance with S01-14-23," Assignment and Approval of Operations Overtime" :

(a) What is the longest period of consecutive hours (1.0) that a Licenced Reactor Operator may be scheduled to work?

(b) What is the maximum number of hours that a Licensec (1.0)

Reactor Operator may work in a 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period?

(c) What is the minimum time between shifts for a (1.0)

Licensed Reactor Operator working 12-hour shifts?

16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months (excluding shif t turnover time) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (excluding shift turnover time) 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (including shift turnover time)

Ref: SONG 1 S01-14-23 8

4.10 (0.5)

Per S01-14-17, " Valve Operation":

When checking a manual valve's position, To what position must the valve be checked? Who can authorize an exception to this requirement?

When checking a manual valves position, it must always be operated in the closed direction ,

except as authorized by the SRO Operations Supervisor .

Ref: SONG 1 S01/14/17 END OF SECTION 4.0 END OF EXAM 9

NEY s

. l' /4ArrfA kn .

U.S. NUCLEAR REGULATORY COMMISSION -

SENIOR REACTOR OPERATOR LICENSE EXAMINATION

. Facility: SONGS UNIT 1 ;

Reactor Type: PWR WESTINGHOUSE 3 LOOP l Date Administered: JUNE 3, 1986 i Examiner: THOMAS R. MEADOWS, RV Candidate:

INSTRUCTIONS TO CANDIDATE:

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

question sheet on top of the answer sheets. Points for each question are indi-cated in parentheses after the question. The passing grade requires at least '

70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.

% of \

Category % of Candidate's Category '

Value Total __ Score Value Category 25 25 5. Theory of Nuclear '

Power Plant Operation, Fluids, and Thermo-

. dynamics 25 25 6. Plant Systems Design, Control, and Instrumentation 25 25 7. Procedures - Normal,

~-~~--

Abnormal. Emergency, and Radiological Control 25 25 8. Administrative Pro-cedures, Conditions, and Limitations 100 Totals .

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

Candidate's Signature s.

EQUATION SHEET f = ma v - s/t 2 '

w = ag s = v,t + hat Cycle efficiency =

E E = mC a = (vg - y )/t .

A = AN -E KE = mv v g =v +a A = A,e PE = agh w = 0/t 1 = in 2/tg = 6.693/tg W = vaP (t, )(th )

~~

AE = 931am h (tg+t) he$ CAT p ,

I . y ,-IX k=UAAT g . y ,-Wx '

Pwr = W g a 11 3o /Tvt -x P=P 10 8M(t) TVL = 1.3/u t

P=P o e /T HVL = 0.693/u

~

SUR = 26.06/T T = 1.44 DT sCR = S/(1 - K,gg)

SUR=26f*ff*\ g,p CR x = S/(1 - K,gg )

~

T = '(1*/p ) + [(f l p)/x g,p] 1( eff}1 = CR2 (I ~ Eeff)2 y . g*/ (p _ p M = 1/(1 - K,gg) = CR g/CR0

"( ~ 8)! Aaff' M = (1 - Keff)0 /(1 - Keff)1

" ( aff" )! eff * #eff/Keff SDM = (1 - Kaff)/Keff p= [1*/TK,'gg .] + [E/(1 + A,ggT )] ,

1* = 1 x 10 seconds

-I P = I(V/(3 x 1010) gaff = 0.1 seconds E = No

. Idly =Id22 WATER PARAMETERS Id g =Id 22 1 gal. - 8.345 lba R/hr = (0.5 CE)/d (meters) 1 gal. = 3.78 liters R/hr = 6 CE/d (feet)

- 1 fc3 = 7.48 gal. MISCELI.ANEOUS CONVERSIONS .

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

Density = 1 gm/cm 1 kg = 2.21 lba Heat of vaiorization = 970 Etu/lbm 1 hp = 2.54 x 10 3BTU /hr Heat of fusica = 144 Btu /lbm 1 Hw - 3.41 x 106 Beu/hr 1 Atm = 14.7 psi = 29.9 in. I'g. 1 Btu = 778 ft-lbf 1 ft. H 2O = 0.4333 lbf/in 1 inch = 2.54 cm F = 9/5 C + 32

C = 5/9 ( F - 32) em _ .. _ - _ - . _ _ - . . _ _ _ . - . . . .-. . - _ , . . _ , _ _ _ - - , - . < , , , , _ _ . , -_

SECTION 5

.. Theory of Nuclear Power Plant Operations, Fluids and Thermodynamics 5.1 (1.0)

SELECT THE BEST ANSWER. As a subcritical reactor nears criticality, the length of time to reach equal i br i um count rate after an insertion of given fixed amount of positive reactivity, (a) decreases primarily because of the increased population of delayed neutrons in the core.

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

(c) decreases because the source neutrons are becoming less inportant in relation to total neutron population.

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

(d)

Ref:

Westinghouse fundamentals of neuclear physics, (Chapters 6 and 7)

I 1

i i

e 5.2 (2.0)

There are seven(7) startup sources (each producing e in an 500 neutrons per generations) installed assembl y is empty reactor vessel. A new fuel Keff of loaded into the core, establishing a 0.75. What is the MAXIMUM equilibrium neutron level (count rate) achievable? I I

I I

CR=S/1-Keff , CR=3500[1/0.253=14000 neutrons / generation Ref:

Westinghouse fundamentals of neuclear physics, (Chapters 7 and 8) 5.3 (1.0)

Is Keff DEPENDENT or INDEPENDENT of initial source range counts?

EXPLAIN YOUR ANSWER.

INDEPENDENT- Keff (the neutron life cycle) only conciders fission neutrons in the self-sustaining reaction.

Ref:

Westinghouse Fundamentals of Nuclear Physics, (Chapters 7 and 8)

I l ,

l 2

I 5.4 (2.0)

An swer the f ol l owi ng TRUE or FALSE:

(a) During equilibrium power conditions,the production (0.5) rate of indirect Xenon from Iodine is faster than the decay rate of Xenon to Cesium.

(b) Sl owi ng the rate of a power decrease, lowers the (0.5) height of the resultant Xenon peak.

(c) The resultant Xenon peak due to a reactor trip (0.5) from 50% power is larger than one from 100% power.

(d) During an increase in power from equilibrium Xenon (0.5) conditions, Xenon concentration initially decreases.

(a) TRUE (b) TRUE (c) FALSE (d) TRUE Ref:

Westinghouse Reactor Core Control for Large Pressurized Water Reactors (Chapter 4) 3

5.5 (4.0)

A reactor is initially critical at a power level of 1.0 kw.

At time =0, the Keff of the reactor core was changed to 1.002 by rod withdraw. -

(a) Cal cul ate the reactivity insertion in percent (1.0) millirho (pcm).

(b) Calcul ate the resul tant reactor period (T) . (1.0)

(Assume"Feff=0.0072 and;T =0.1) err (c) Calculate the resul tant startup rate (SUR) . (1.0)

(d) At what time will the reactor power equal 100 kw? (1.0) 5 5 (a) (pcm>= Keff-1)(10 )/Keff= (1.002-1.0)(10 )/1.002

=200 pcm (b) T= Beff- f /4,,9for IBeff T= (0.0072-0.002)/(0.002)<0.1) = 26 sec (c) SUR= 26/T = 1.0 dpm (Whatever T used from above. acceptable)

(SUR)(t)

(d) P=Po10 . . _

t 100 kw =(1 kw)10 t= log (100)= 2 minutes Ref:

West inghouse Fundamen tal s of Nucl ear Reac tor Physi cs, (Chapter 7, Neutron Kinetics) 4

-- - - . . = . _ _

e 5.6 (3.0) an instantaneous, negative reactivity Figure 5.1 illustrates insertion into an already critical reactor core at t=0,followed by ,

a removal of this negative reactivity af ter a stable period is reached at t=1, making the core critical once again.. Assuming no source neutrons:

(a) Show the resulting startup rate (SUR) as' a (1.0) function of time for this event.

(b) Show the reactor power level as a function of time (1.0) for this event.

(c) Expl ain the shape of the reactor power response n (0.5)

(1) at a time Immediately after t=0 (2) at a time Immediately PRIOR to t=1 (0.5)

(a) and (b) attached (c.1) " prompt drop" in total neutron population .

(c.2)" negative constant period" due to the decay of delayed neutron precursors stablizing the decrease in neutron population.

Ref:

West inghouse Fundamen tal s of Nucl ear Reac tor Physi cs, (Chapter 7, neutron kinetics) i 5

FIGURE 5.1

/6 .

Y I I k _

e e t'O 7/Mr : l'I i

e t

)N f;=0 1- 1 ,

3 0 .

M e

g e

s t.0 trt t n u 8 T/MC -

KEY FICURE 5.1 Is ,

D 40 l N .

L e W

T/ME : bl ik S lh4'*

'o O /r) o 4* 1.il\;

2 i e g l --

_ - - - - - _ _ _ _ i C.2 f t'T O*)f) )

L t

(0,2for)

Y (c.zS 1')

y' O.uarj 7""

3 .

a D A 0 **

P<Pg-fr.) e (, f S TAaig pz pino W G.zs,r) / ( r = -c)

%Q c** ri- .

s f pasm r t

rw W

(0,2 5'*0 l

i

' t-0 i

Y ,

B 5.7 (2.0)

Regarding a Main Steam line ruptures Why is a rupture of a Main Steam line at End of Life -(EOL) a much ,

more limiting accident than at the Beginning of Life (BOL) ? l t

The Moderator Temperature Coefficient is less negative at BOL than at EOL. Therefore, at EOL the sudden cooling of the a reactor larger I cool an t upon a Main Steam Line Rupture resul ts in addition of positive reactivity (uncontrolled).

Reft Westinghouse Reactor Core Con' trol for Large PWR's, West inghouse Fundamen tal s of Nucl ear Reactor Physi cs ,

SONGS 1 FSAR and Technical Specifications 6

(

5.8 (1.5)

Answer True or Falsen (a) Heat transfer takes always takes place from hl'gh' (0.5) to low temperatures. .

(b) As water flows around a bend in a pipe, the (0.5) velocity of the water is uniform throughout the diameter of the pipe .

(c) Pressure measured in a static fluid (e.g. reference (0.5) leg of a level indicator) at different levels decreases with increasing elevation of measurement.

(a) True l

(b) Fal se (c) True Ref:

Westinghouse Thermal-Hydraulic Principles and Applications to PWR's (Chapters 1-5) '

E

'l 1

l s

l 7

5.9 (2.0)

Of the following operations,which one will have a negative effect on avai l abl e Net Positive Suction Head (NPSH) of a given centrifugal pumps (a) Trottling open the pump's suct ion val ve.

(b) Throttling open the pump's discharge valve.

(c) Decreasing the pump's speed.

(d) Decreasing the temperature of the fluid (water) being pumped.

(b)

Refn Westinghouse Thermal-Hydraulic Principles and Applications to the PWR, (Chapter 10, Fluid Mechanics in Pumps and Pipes) 8

__. . -= _. -

5.10 (4.0)

The plant is operating at 25% power, rod control in manual , turbi ne control in automatic, when loop 1 RCP trips.

Given no reactor trip or operator actions, indicate whether the f ol l owi ng parameters will be HIGHER or LOWER at the end of ;

the transient as compared to their initial val ues .

Briefly explain your answer, indicating your assumptions.

(a) Th in loop 2 (1.0)

(b) Tc in loop 1 (1.0)

(c) #3 RCS loop flow (1.0)

(d) #2 S/G pressure (1.0)

(a) HIGHER [0.53 (higher del ta T f or each operating loop)[0.53 (b) LOWER [0.53 (reverse flow from other col d legs with l ower T-c ol d> [ 0.53 (c) HIGHER [0.53 (lower pressure drop across core)[0.53 (d) LOWER [0.53 (higher heat transfer for each operating loop /ie,same load demand with two S/G's)[0.53 Ref:

Westinghouse Reactor Core Control for Large PWR's, Thermal- Hydraul i c Pr i nc i pl es(In tegrated Knowl edge) .

9

5.11 (1.5) -

o Letdown water (75 gpm) at 538 F enters the regenerative heat exchanger and exits at 290 degrees F. Assuming normal, steady-state power operation (Total charging flow of 87 gpm at 115 degrees F(32gpm f or RCPseal s),WHAT is the temperature of the charging water entering the RCS from the regenerative heat exchanger? Show all work and state all assumptions.

Q =0 [0.53 letdown chargina ~

m AT= E A T C C A T = AT (5/m )

e C n= 75 gpm E = 55 gpm (charging less 32 gpm for RCP seals) [0.53 C o si T = 248 F (75/55) c o

= 338 F Outlet temperature is therefore:

115+338=453cF [0.53 Ref:

Westinghouse Thermal-Hydraul i c Principles and Appl ications to PWR's,(chapters 3-6) s 10

~

5.12 (1.0)

What is the purpose or bases for the Maximum Safety System Settings for PRESSURIZER HIGH LEVEL and HIGH PRESSURE (setpoints are not required)? .

. In the event of loss of load, the temperature and ptessure of the RCS would increase (reduction of heat removal from S/G's).

These settings are established to maintain the DNBR above 1.30 AND to prevent the loss of the cushioning effect of the PZR steam volume.

Ref Westinghouse Thermal-Hydraul i c Pr i nc i pl es and Appl i cat i ons to the PWR,(Integrated Knowledge), Technical Specifications END OF SECTION 5 s

11

SECTION 6

~

PLANT SYSTEM DESIGN. CONTROL AND INSTRUMENTATION as 6.1 (3.0)

State the valve and pump lineups necessary for Main Feedwater Pump minimum suction requirements for each of the condi tipns bel ow: 4 (1.5) ;

(a) One Main Feedwater Pump in operation. E Two Main Feedwater Pumps in operation. (1.5)

(b)

E (a) Feedwater Pump Suction Val v e open and two Condensate Pumps running for the operation of one Main Feedwater Pump.

(b) Feedwater Pumps Suction Valves open and three Condensate Pumps running for operation of two Main ,

Feedwater Pumps.

Ref: SONG 1 Operating Instruction SO1-7-2

+.

s 12

6.2 (2.5)

The RCS SUBC00 LING MARGIN MONITOR channels are required to have two Pressurizer Pressure inputs and one of three(3) possible l c ambi nat i ons of loop temperature inputs (Thot RTD's,and/or Teore exit Thermal Couples) to be considered operable (Total instumentation).

What is one possible cambination of loop temperature inputs needed to consider the Subcooling Margin Monitor operable?

Any one of the f ol l owi ng f or full credit:

(1) 2 RTD's per HOT leg (2) 1 RTD per HOT leg and 4 Core exit Thermal Couples (3) 8 Core Exit Thermal Couples Ref: SONG 1 S01-4-34, Reactor Plant Instrumentation Operation

\

13

6.3 (1.5)

The Power Range nuclear instrumentation (comparator) compares the four Power Range channels and provides three protect _ive features in order to protect the fuel in the event of a dropped rod. List these three protective features. ,

(a) lkuclear Dropped Rod Stop (0.5)

(0.5) 944 f Auto Rod Wi thdraw Prohibi tm p 5m-cr")

p ir i

+: -

(c) Turbine Runback 3 to 70%( absvw 7 0 %'- (0.5) p en'a n a n d-4 h e drfwat switcn not in dw ? .iP

!Turbin: Ru r.b;c k tc 70% acceptab1: fer 0.5pt:3- -

Ref:

SONGS 1 System Description SD-S01-380 JFv / - z , J - /

14

~

6.4 (3.0)

Consider the LOAD SEQUENCING SYSTEM upon a combined Safety Injection Signal / Loss of Power (SIS / LOP) event at time =0 .

Three of the below listed sequencer responses DO NOT occur.

IDENTIFY the three responses which DO NOT occur. [ 1.0 pts each3 (1) Trip Reactor Trip breakers (2) Closes the PZR PORV's (Interlocked closed)

(3) Deenergize reactor trip breaker UV's (4) Energize L/0 relays for switchgear 1,2,and 3 (5) Energize L/0 relays for MCC's 1,1C,2,2A,and 3 (6) Trip 4160 VAC Bus 1C and 2C Tie Breakers (7) Trip D/G 1 and 2 output bkers (8) Isolates the Steam Generator steam dumps (bypasses interlocked closed)

(9) Block D/G 1 and 2 excitation S/D ekts (10) Reset D/G field (11) Trip lighting transformer (12) Lockout Motor Heater Panels (13) Start D/G's 1&2 (2 ckts per D/G)

(14) Close Feedwater Bypass / control valves (15) Open SI header Iso. valves (16) Close HP & LP Feedwater Hdr. Iso. valves (17) Trip Reactor Coolant Pumps ( immediately )

(18) First out annunciators, Auto alert systems +TSC (19) Trip Heater Drain Pumps (20) Trip condensate pumps

~

(21) Initiate event recorder (22) Trip Feedwater Pumps (23) Initiate Containment Isolation System (24) Open SI loop Iso val.ves (25) Close letdown orifice Iso-valves (26) Close 480 VAC bus 2&3 Tie bkrs (27) Si gnal s Feedwater con trol system to start Feedwater pump after 11 see time-delay.

(28) Trip Turbine Plant Cooling Water pumps (2)

(8)

(17)

Ref:

SONGS 1 Syste- : . scr i p t i on SD-SO1-590 15

6.5 (2.5)

(a) What are two sources of actuating power for opening (0.5) the PORV's (CV-545& CV-546)?

l (b) With the reactor coolant system solid at 400 psig (0.5) and the OMS armed, what is the action of the OMS.if a 200 psig pressure increase occurs?

(c) What are the three control room alarms directly (1.5) associated with the event described above?

(a) (0.25 each) (0.5)

Instrument air header Nitrogen Bottles

,a /A & -

Cr', v 's }

(b) At approximately 500 psig CV-545 and CV-546.4 (0.5)

(PORV's) open to the PZR relief tank),( relieving pressure (c) The f oll owi ng annunc i ators al arm i n the control room:

OMS HIGH PRESSURE (480 psig) (0.5)

PRESSURE TRANSIENT in PROGRESS (500 psig) (0.5)

PORV Open(Valve Ilmit switch) (0.5) l Ref:

SONG 1 Operating Instruction S01-2.1-11, Reactor Coolant System Instrumentation Study Guide, Technical Specifications s

16

6.6 (2.0)

Concerning the emergency Diesel Generating System, answer the f oll owing TRUE or FALSE:

(a) The D/G speed droop is set to protect the diesel (0.5) from over loading upon bus syncronization.

(b) The D/G Regulator Mode Selector switch is located (0.5) on the D/G Iocal panel.

(c) D/G excitation can only be controlled from the (0.5)

Standby Power Generation Panel during emergency operations.

(d) The MANUAL Vol tage Regulator is normally used when (0.5) parallel ing wi th anuuther generator.

(a) TRUE (b) TRUE (c) FALSE (d) FALSE Ref:

SONG 1 System Description SD-SO1-600, section IX, S01 Diesel Generator and Fuel Oil operating instructions.

s I

17

6.7 (0.5)

What is the only Diesel Generator engine trip enabled when the D/G is auto started by the sequencer?

Engine Overspeed Ref SONG 1 SYSTEM DESCRIPTION SD-S01-600, section X e

18

6.8 (1.0)

FILL IN THE BLNNK:

(a) When a D/G is to be paralleled to the 4kV bus, (0.5) real load (kW) is adjusted by operation of the o -

(b) When a D/G is to be paralleled to the 4kV bus, (0.5) .

the reactive load (KVAR) is controlled by operation of the automatic .

(a) Governor Speed Control Switch (b) Voltage Adjuster (regulator)

Ref:

SONG 1 System Description SD-SO1-600, section X l

1 l

l 19

6.9 (2.0)

What is the purpose or bases, per Technical Specif.ications,for the Auxi1iary Fee & ater System (The emergency safety aux system)?

The operability of the auxiliary feedwater system ensures that the reactor coolant system can be cooled down to less than 350 degrees F from normal operating conditions in the event of a total loss of offsite power .

Ref:

SONG 1 Technical Specifications, 3.4.3 20

l 6.10 (2.0)

(a) What is the minimum water volume, specified by (0.5) )

SONGS 1 Technical Specifications, for the auxiliary. )

feedwater system?

(b) What is the purpose or basis, per Technical (1.5) ,

Specifications, for the minimum water volume limit (LCO) of the Aux Feedwater Storage Tank?

(a) ~150,000 gall ons (minimum) l (b) The operability of the Auxiliary feedwater storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY condi t i ons (i ncl udi ng cool down) for 32 hours with steam discharge to atmosphere concurrent with total loss of offsite power .

Ref:

SONG 1 Technical Specifications, 3.4.4 k

21

6.11 (3.5)

(a) What is the principle of operation of the .

(2.0)

Source Range Nuclear Instrumentation detector?

(b) How is Gamma flux compensated for in the source (1.0) range instrument? .

(c) What inputs to the Reactor Protection and Control (0.5)

System are provided by the Source Range Nuclear Instrument Channels?

(a) The SRM detector is a BF3 gas filled chamber.

Thermal neutrons are absorbed by boron-10 atoms in the gas . These atoms subsequently disintegrate into Li7, emitting a high energy alpha p ar t i r.a1 This Alpha p ar t i cal causes ionization within the gas that is collected by the enamber cathode / anode cl.rcuitry via applied voltage .

The ch ambe r voltage is in the PROPORTIONAL range,therefore,the output event pulse AMPLITUDE is proportional to the magnitude of the ionizing event. The number of pulses is proportional to the number of neutrons reacting with the BF3 gas.

(b) Since the amplitude of the pulses from the detector are proportional to the magnitude of the ionizing event, the pulses from neutron interactions (alpha ionizations) are larger than pulses from gamma ionizations. A pulse height discriminator (gate circuit) in the log level amplifier rejects pulses below a set range (approximately 3.5 Mev) in magnitude, eliminating the gamma pulses from the remaining circuitry.

(c) Rod stop signal at 2 DPM .

Ref:

SONG 1 System Description SD-SO1-380, Technical Specifications, Reactor Plant Instumentation S01-4-34 22 l 1

6.12 (1.5)

During refeuling operations:

(a) What systems are required to be continuously (1.0) l operable to monitor for inadvertant criticality 7, (0.5) l (b) What interlock on the fuel lifting hoist prevents lifting more than one fuel assembly at a time?

1 l

(a) Core subcritical neutron flux shall be continuously monitored during the entire refueling period by not less than two neutron monitors, each with continuous visual indication ar - one with continuous audible indication. .

LBS A y #/

8 (b) Lift ng hoist circui t opens (at >1500 pounds load and stops out motion.

Ref:

Technical Specifications 3.8, SONG Operating Instruction S01-3-7, Foul Handling and Refueling Operations Study Guide END SECTION 6.0 s

23

SECTION 7 Procedures - Normal, Abnormal, Emergency, and Radiological Control 7.1 (3.0) .

In accordance with 10 CFR 20 " Standards for Protection against Radiation":

(a) What are the Radiation Dose Standards for (1.5) i ndi v i dual s in restricted areas per Calender Quarter?

(b) What are the three requirements that must be met if (1.5) the Whole Body limi ts for a Calendar Quarter are to be exceeded?

(a) (0.5 each) 1.25 Rem - Whol e Body; head and trunk; active blood forming organs; lens of eyes; gonads.

18.75 Rem - Hands and forearms; feet and ankles 7.5 Rem - Skin of whole body.

- (b) (0.5 each) 3.0 Rem per calender quarter-maximum 5*(N-18) total accumulated dose to the whole body where N is the individuals age in years at his last birthday.

Form NRC-4 or equivalent.

Ref:

10 CFR 20 s

24 i l

7.2 (1.5)

Concerning Reactor Plant Instrumentation Operation:

(a) At what power levels should the control rod system (0.5) be in manual?

(b) In the event an Intermediate range channel f ails .in (1.0) such a manner as to turn ON its companion source range channel, What shall the operator do to the.

affected SRM while in the Intermediate range?

(a) The control rod system centrol should be in MANUAL control at power levels less than 15% of full power.

(b) The affected SRM channel's high voltage power supply shall be turned off.

Ref: SONGS 1 S01-4-34, Reactor Plant Instrumentation Ops.

7.3 (1.5)

4. / The plant is operating near 100% power and a thermal calibration is performed. The indicated power after adjusting the power range instrumentation exceeds 100%:

,^ gfhjf (a) When this occurs, what action should operator take,in addition to anouther calibration?

(1.0)

(b) If a second calibration provides verification (0.5) that the reactor had been previously operating above 103%, Who must be informed within the Facility?

(a>The operator should immediately reduce power level to an indicated 100% and make sui table log entry .

If the required power reduction exceeds 3%,

report the problem to the Shift Superintendent.

(b)1f a second thermal calibration verifies operation i above 103%, report this matter to the UNIT SUPERINTENDENT.

Ref: ez,/

SONG 1 S01-$>4.4, Reactor Plant Instrumentation l '

! Operation 25 i

4 7.4 (2.5)

CONCERNING PLAPR STARTUP per S01-3-1,P1 ant S/U from Cold S/D to Hot Standby:

(a) What is the maximum allowable Pressurizer heatup (0.5) rate? -

(b) What is the maximum temperature difference al l owed (0.5) between the Pressurizer liquid and Reactor Coolant?

(c) The RCS (except the PZR) temperature and pressure (0.5) shal l be Ilmited in accordance with given pressure / temperature curves. What is the maximum RCS heatup rate i n any GME(1) hour period?

(d) Why do operations require that the Reactor Coolant (1.0)

Pumps be operated continuously (with momentary stops) during the solid water phase of heatup?

o (a) 95 F per hour (b) 200 F J_. ._=r o

(c) 60 F in any one hour period (d) To prevent temperature differences that could result in over-pressurization upon RCP restart.

Ref:

S01-3-1, Plant Startup from Col d Shutdown to Hot SCNG1 Standby 26

e

? .

h

'r,

. t'3

'Ei 7.5 (3.0)

'.' . } Concerning the starting duty requirements for the RCP's per S01-4-3, RCP Operation:

(a) How many pumps can be started at any one time? (0.5)

.4 (b) What are the limits on restart for each RCP ? (2.5)

,,' (Address, both IDLE & Running Pump limits, number of restarts within a particular time period,and

, l g time requirements for idle periods) 9,

, ,.7 4

.'. (a) Only one RCP is to be started at any one time. (0.5)

R.

(b) If a RCP has failed to achieve full speed after an (2.5)

I attempted start, restart should not be attempted

t. .I; until the motor has been al l owe d to cool by in standing idle for a period of not less than 30

'S[,' minutes. After a RCP trip, an immediate s . . l. .

restart is permitted if the motor has been running

.ph under normal load for at least four hours.

E,1

.tB , For each RCP, the number of starts will be limited to a maximum of 3 within a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months period.A

'$.9

(,r : minimum idle period of 30 minutes for cooling will

'Fa be observed between each restart. When 3 starts or attempted starts have been made within a

. l. 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months period, the fourth attempted start must not

'! be made until the moter has been allowed to cool by

i. ', !' standing idle for at least 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

,4 d.'

'g

. .; Ref SONG 1 SO1-4-3, RCP OPERATION

. ,; .1 2

b',

y ,

i vf g ., .

k.,.,. 27 f-4 4

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

7.6 (1.5)

Explain why the operator in preparing for Main Turbine roll, per S01-3-2 ( Plant S/U from Hot Standby to Minimum Load), increases reactor power suf f iciently to gain 50-75% percent opening on the Steam Dump valves to the condenser.

To ensure that the reactor system (NSSS) can satisfy the turbine load-demand without exceeding established safe operating bands (ie; PZR pressure / level, variable Tave range, Reactor pressure).

Refs SONG 1 SO1-3-2, Pl an t S/U f rom Hot S/B to Mi n Load, WESTINGHOUSE Reactor Core Control for Large PWR's I

l 28

7.7 (3.0)

(a) What are the maximum Reactor Coolant Activity (1.0)

Limits?

(b) What is the purpose or bases for the maximum (1.0)

Reactivity Coolant Activity Limits? -

(c) Why is the reactor required to be subcritical with (1.0)

Tave< 535 degrees F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months if the Dose Equivalent I-131 exceeds 40 u Ci/gm ?

(a) (0.5 each) 1

< or = 1.0 u Ci/gm dose equivalent I-131

< or = 100/ E u Ci/gm 5 = the avg of the beta and gamma energies per disintergration( in Mev) for isotopes, other than Iodine and Tritium with t1/2 > 15 min.

(b) Limits result in a two hour dose rate at the site boundry less than 10 CFR 100 limits following a steam generator tube rupture with a primary to secondary leak rate of 1 gpm.

(c) With Tave less than 535 degrees F the release of activity should a S/G tube rupture is prevented since the saturation pressure of the RCS is below the lift pressure of the atmospheric steam relief valves.

Ref Technical Specifications s

29

7.8 (4.0)

Per S01-1.0-40, Steam Generator Tube Ruptures (a) After identification of the ruptured S/G and. (2.0) verification of proper plant lineup, outline the basic method of placing the plant in a cooled down and depressurized condition.

(b) What caution is given for depressurizing the RCS by (1.0) this method?

caution is given for starting Reactor Coolant (1.0)

(c) What Pumps in the intact loops after depressurization?

i (a) (0.25 each)

(1) Establish level in non-ruptured S/G's via aux feed.

(2) Open PORV's to al l ow SI to establish PZR level at 70%

(3) Start RCP (4) Align two charging pumps to inject via SI col d leg injection lines (5) Dump steam to the condenser for a max 100cF/ hour C/D rate (6) Depressurize the RCS via spray or PORVs(RCS 200 psi >

Steam line pressure) (If ruptured S/G overfilling then depressurize to saturation)

(7) Borate RCS to 5% S/D concentration.

(8) When main steam pressure < 350 psig place RHR in service and c ool down to 140 degrees F at less than 50 degrees F per hour.

(b) May result in steam voiding in the upper head and rapid fl11ing of the pressurizer.

(c) Pressurizer level may drop due to the collapsing bubble in the upper vessel head.

Ref: SONG 1 S01-1.0-40, Steam Generator Tube Rupture s

30

7.9 (1.5)

S01 -2.1 -4, LNPLAPNED BORATI ON , sp e c i f i e s f ou r ( 4 ) symtoms for this abnormal operation. List three of these.

Any 3 of the following (0.5 each): .

rods will travel (a) If rod control is in automatic, #2 High out with associated " Rod Withdraw Bank Alarm".

(b) If rod control is in manual, Tave will decrease.

(c) Boric acid fl ow may be indicated on blend system recorder, FR-1102.

Lo Level" (65%) alarm may be ,

(d) " Boric Acid Tank indicated.

Ref: SONG 1 S01-2.1-4, UNPLANNED BORATION 1

31

7.10 (2.5)

Technical Specifications 3.1.4 Basis identifies two(2) basic kinds of LEAKAGE from the Reactor Coolant:

(a) Name the two possible kinds of RCS leakage.(0.5 e'ach) (1.0)

(b) What is the acceptable value(gpm) for a source 'of (0.25) leakage not identified?

(c) What is the reason that this value set for not (1.0) identified leakage, be allowed to increase when the source is discovered ?

(d) What is this new acceptable increased leakage (0.25) value?

(a) (0.5 each)

(1) To other CLOSED systems (2) Directly to the CONTAINMENT (b) The acceptable value of leakage not identified is set at one gpm.

(c) Once the source of leakage has been identified, it

- can be determined if operation can safely continue.Under these conditions,the al l owabl e leakage rate i s all owed to increase. This is BASED on the contingency of sustained loss of all off-site power and failure of the on-site generation.

With this new acceptable increased leakage value, decay heat removal can safely be accomplished for a period in excess of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Within the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months period, the reactor cool an t system can be DEPRESSURIZED.

(d) 6 gpm Ref: Technical Specifications 3.1.4 s

32

7.11 (1.0)

Technical Specification 3.1.2 requires that both PZR safety valves must be operable when the reactor is critical, but requires that only one PZR safety be operable when the reactor is shutdown and the head is on the reactor vessel. .

What is the reason or basis for the relaxed requicements while shutdown?

1 One PZR safety is sufficient to prevent overpressurization when the reactor is subtritical since it's relieving capacity is greater than the available heat sources:

Decay heat Pump Energy Pressurizer heaters Ref: SONG 1 Technical Specifications 3.1.2 END OF SECTION 7.0 -

l s

33

SECTION 8 Adminestrative Procedures, Conditions, and Limitations 8.1 (2.0)

Concerning reactor core SAFETY LIMITS:

Overheating of the fuel cladding is prevented by , (SELECT TWO OF THE MOST CORRECT STATEMENTS BELOW) (1.0 pt. each)

(a) restricting fuel operation such that the reactor coolant never reaches saturated conditions.

(b) resticting fuel operation to within the nucleate boiling regime.

(c) limiting the actual power and temperature to below the locus of points for approximate RCS pressure of ,, _,-

Figure 8.1(attached) ,La _t.isn sne matery i smii s. waswwdvd. ~

(d) limiting the actual power and temperature to above the locus of points for approximate RCS pressure of Figure 8.1(attached),!= -hich *ha 2f e tr ' : . t l; ;: e- e de d J*~~'

(b)

(C)

W Ref: SONG 1 Technical Specifications 2.1, Westinghouse Thermal-Hydraulic Principles and Applications to the PWR l -

34 ge---- - - - - - . - - - - ---y--w.-----, - - . - . - + - - . - - - - , . - - - - - , , - - - - -y----- - - - - - - - .--- , c ,3-mm--a. . . . , , , . g- --.-,-,y----,--.---.,.,%--- - , .-_,,,.-,__,

Figure 8.1 !

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1._, . . _ _ _ _.

8.2 (3.0)

In accordance with S01-14-23," Assignment and Approval of Operations Overtime" : .

(a) What is the longest period of consecutive hours (1.0) that a Licenced Reactor Operator may be scheduled to work?

(b) What is the maximum number of hours that a Licensed (1.0)

Reactor Operator may work in a 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period?

(c) What is the minimum time between shifts for a (1.0)

Licensed Reactor Operator working 12-hour shifts?

16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months (excluding shift turnover time) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (excluding shift turnover time) 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (including shift turnover time)

Ref: SONG 1 S01-14-23 i

35

8.3 (2.0)

(a) Who is responsible for design, construction, (0.5) operation and maintenance of SONG 1, and all site support functions?

(b) Who is responsible for the Control Room command (0.5) function? -

(c) Define, " Control Room Area". (1.0)

(a) Vice President and Site Manager, Nuclear Generation Site (b) Shift Superintendent (c) " Control Room Area" is defined by the control room and the Shift Superintendent's office.

Ref SONG 1 Technical Specifications 6.1 1

I w

36

~.

8.4 (3.0)

Per the Technical Specifications for Adminestrative Controls, Section 4.2, " Organization":

(a) What is the minimum crew composition required on (-h5P M Fs site in Mode 3 ? .

(b) What is the minimum crew composition required .on 7I site in Mode 5 ?

(c) Is the Shif t Technical Advisor exempt from overtime 05.c5/~rIZfj guidelines?

pr/~C 4,P G zny in --

,wL Q-2 8 cenced Senior Reactor Operator

[f )

2- Licenced Reactor Operators f_ # Non-licensed Auxiliary operator y_g /rf 1- Shift Technical Advisor 5- Fire Brigade members 1- Health physics technician (when fuel is in the reactor) rrA &,,.

(b) 44T25 eacnI- "" ' "- C 1-Licensed Senior Reactor Operator 1-Licenced Reactor Operator p, 7 f .r r 1- Non-Licensed Reactor Operator 5- Fire Brigade members

~" " #

1- Health physics technician (when fuel is in the reactor)

(c) YES -

/. Z r M' Ref: SONG 1 Technical SpecifIcatiens 6.2 37 l

l

8.5 (3.0)

Per 10 CFR 55, " Operator Licenses" a (a) The" Exemptions from License" provisions of the Code (1.0) of Federal Regulations (10 CFR 55), al l ow what individuals to operate the reactor controls without a license? -

(b) As defined in 10 CFR 55, when is an individual (1.0) deemed to be operating the controls of a nuclear facility?

(c) What are the " controls" defined in 10 CFR 557 (1.0)

(a) An individual may manipulate the controls as part his training to qualify for an operator license under the direction and in the presence of a licensed operator or senior operator.

(b) An individual is deemed to operate the controls of.a nuclear facility if he directly manipulates the controls or directs anouther to manipulate the controls.

(c) " controls" -apparatus and mechanisms, the manipulation of which directly affect the reactivity or power level of the reactor.

Ref 10 CFR 55 s

l 38 l

l

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

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

8.6 (2.0)

Per 10 CFR 20, " Standards for Protection Against Radiation":

(a) What is a RADIATION AREA ? (1.0)

(b) What is a HIGH RADIATION AREA ? (1.0)

(a) Area (accessible to personnel) where a major part of the body could receive:

5 mrem in one hour (0.5) or 100 mrem in 5 days (0.5)

(b) Area (accessible to personnel) where a major part of the body could receive:

100 mrem in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (1.0)

Ref: 10 CFR 20 39

8.7 (3.0)

Temporary changes to procedures as descr i bed by Technical Specifications d.8.1, may be made provided what three(3) criteria are met? (1.0 pts each) -

(a) The intent of the original procedure is not (1.0) altered.

(b) The change is approved by two members of the (1.0) site / station management staff exercising responsibility in the specific area and unit or units addressed by the change, and at least one of whom holds a SRO license on the unit affected.

(c) The change is documented , reviewed and approved by (1.0) responsible management (Vice President and S i,t e Manager, Station Manager,or equivalent per SONG 1 Admin procedures-acceptable) within 14 days of implementation.

Ref: Technical Specifications 6.8 40

8.8 (1.5)

Per Technical Specifications 3.14, Bases, What three systems and/or components make up the " Fire Suppression Systems"?

(0.5 pts each)

(a) The water system (b) Spray and/or sprinklers (c) Fire hose stations i

Ref: Technical Specifications 3.14 O

9 l

l 41 l

l l

8.9 (3.0)

The f oll owi ng questions pertain to Technical Specification 3.8

" Fuel loading and refueling" (1.0 each):

(a) What is the basis for maintaining a minimum water (1.0) elevation of 40' 3" in the refueling pool and in the spent fuel storage pool?

(b) What is the basis for restricting the movement of (1.0) loads in excess of 1500 pounds over the fuel assemblies in the storage pool ?

(c) How long must the reactor be subcr i ti cal before (1.0) movement of irradiated fuel in the reactor vessel can commence?

(a) (1) 23 feet of water is available to remove 99% of (0.5) the Iodine gap activity assumed to be released .i n the event of a dropped and damaged fuel assembl y .

(2) at least 12 feet of water above the top of the (0.5) fuel rods of a withdrawn fuel assembly so as to limit dose rates at the top of the water.

i (b) In the event of a dropped loads (1) the activity rel. eased will be limited to that (0.5) contained in a single fuel assembly.

(2) any possible distortion of fuel in the storage (0.5) racks will not resul t in a critical array.

(c) At least 148 hours0.00171 days 0.0411 hours 2.44709e-4 weeks 5.6314e-5 months (1.0)

Ref SONG 1 Technical Specifications,3.8 s

42

e 8.10 (2.5)

Per S01-14-17, " Valve Operation": -

(a) When checking a manual valve's position, To what (0.5) position must the valve be checked? Who can authorize an exception to this requirement?

(b) What are the minimum requirements for " Independent (1.0)

Verification of a valve position (people involved and items checked)?

(c) The instruction states a precaution guarding against overheating MOV motors:

(1) State the maximum amount of starts allowed per (0.5)

< minute without a cooling period.

(2) What is the recommended cooling period? (0.5)

(a) When checking a manual valves position, it must always be operated in the closed direction (0.25),

except as authorized by the SRO Operations

. Supervisor (0.25).

(b) Both the positioner and verifier i t.depende n t l y

~

determine both the valve number and position.

(c) (1) 5 starts per minute (2) 15 minute cooling period i

Ref SONG 1 S01/14/17 l

END OF SECTION 8.0 END OF EXAM 43

. - _ - --

ML20203L055

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1.4 Comment

1.7 Comment

1.8 Comment

2.6 Comment

2.7 Comment

3.1 Comment

3.3 Commest

3.5 Comment

3.6 Comment

3.9. Comment

4.2 Comment

4.4 Comment

4.5 Comment

4.7 Comment

5.5 Comment

6.3 Comment

7.7 Comment

8.1 Comment

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1.4 Comment

1.7 Comment

1.8 Comment

2.6 Comment

2.7 Comment

3.1 Comment

3.3 Commest

3.5 Comment

3.6 Comment

3.9. Comment

4.2 Comment

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4.7 Comment

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6.3 Comment

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8.1 Comment

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