Text

Exam Rept 50-206/OL-86-02 on 861118-20.Exam Results:All Participants Passed All Portions of Written Exam
	ML20207F417
Person / Time
Site:	San Onofre
Issue date:	12/08/1986
From:	Elin J, Johnston G, Meadows T, Obrien J; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:	;
Shared Package
ML20207F403	List: ML20207F398; ML20207F417;
References
	50-206-OL-86-02, 50-206-OL-86-2, NUDOCS 8701060013
	Download: ML20207F417 (64)
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Report' No. 50-206/0L-86-02 Docket No. 50-206 Licensee:' Southern California Edison Company J P. O. Box 800

, ' 2244 Walnut Grove Avenue Rosemead, California 91770 Facility Name: -

San Onofre Nuclear Generating Station, Unit 1 Examination Conducted: November 18-20, 1986 Examiners: '

_. /2 ~b G. N./Johnsto ,~0perator Licensing Examiner Date Signed h'

/ 2 - 8 4 (p

'P. O'Brien, Operator Licensing Examiner Date Signed O feadows JJ bcN S perator Licensin), Examiner D- E-2C Date Signed Approved: , T) h i J. CF EI1n, Thief, Operstions Section Date,Sig;ned Summary: /

During this requalification cycle, twenty percent of the operating staff.

(ten licensed operators) were examined. This included eight Senior Reactor

. Operators, and two Reactor Operators. The NRC prepared a complete written examination and administered it to all participants. An operating examination was also administered to all ten of the participants. All of the participants passed all portions of the examinatien.

In accordance with NUREG-1021, ES-601 'Requalification Program Evaluations',

the requalification program at San Onofre Nuclear Generating Station Unit I is evaluated as satisfactory.

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-DETAILS 1

1. ' Persons' Examined A group of ten operators, eight holding Senior. Reactor Operator 1 licenses, and two. hold Reactor Operator licenses representing approximately 20 percent of the licensed staff were examined.

2. -Persons Contacted-Southern California Edison:

M. Short

M. Kirby

'*R. Mette NRC:

G. Johnston, Operator Licensing Examiner

J. .

0'Brien, Operator Licensing Examiner T. Meadows, Operator Licensing Examiner M. Royack, Operator Licensing Examiner

Denotes those present at exit on November 20, 1986.

- 3. Program Evaluation Required for Satisfactory:

The requalification' program was evaluated upon the criteria of Examiner i Standard ES-601 of NUREG-1021. The requirement for a satisfactory

program is more than 80% of the evaluated operators must pass all

~ operating examinations, all sections of the written examination, and the L written examination overall administered by the NRC.

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l Performance:-

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The NRC administered examinations to 2 Reactor Operators and.8 Senior l

Reactor Operators. All. ten participants passed all portions of the examination. The result is a pass rate for all the participants of-100%.

[ Evaluation:

Satisfactory.

4. Exit Meeting l The examiners met with the' licensee representatives denoted in Paragraph'2 at the conclusion of the site visit to discuss the results

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of the evaluation to that point.

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Resolution of Facility Comments Facility Comment: 1.04

" Question 1.04: Question and answer OK could.be clearer if question referred to '50% power, equilibrium Xenon....'"

Resolution:

The examiner agrees.

Facility Comment: 1.06

" Question should read "..six factor formula.." instead of "..six function formula...". Answer OK."

Resolution:

The examiner pointed out the error to the candidates during the written examination.

Facility Comment: 1.10

" Reference should be SONGS Unit 1 Reactor Core Physics Book Cycle 9.

Part a: SONGS values different from answer key, see Cycle 9 book, pages 67 and 68.

Part b: SONGS peak occurs 6-7 hours following trip. See Cycle 9 book, pages 67 and 68.

Part c: OK" Resolution:

The examiner will change the key.

1 Facility Comment: 2.01a "High point value (13.3% of section). Add: Feedwater isolation MOV's (MOV's 20, 21 and 22) shut same reference. Same as question 6.01a of SRO exam"..

Resolution:

Point value does not exceed the 20 percent limit imposed by the Examiner j standards. The examiner will cha'ge key but not distribution of points for the indicated answers.

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l l Facility Comment: 2.03b-3.

" SIS Valves, HV851, 852,-853, and 854, are operable on loss of air since valves fail in safety related direction and air is needed to position valve only in non-safety related direction. Reference drawing is in error as it shows air is needed to position valve in safety related direction. Correct drawing is attached. Same as Question 6.05b-3 of SRO exam. NOTE: Determination of operability of components is not a RO function at SONGS but is done,by the SRO on shift."

Resolution:

The examiner stated to the RO candidates that the term used was not OPERABLE / INOPERABLE as in the Tech. Specs. but only as to whether the valves actually do or do not operate. On this basis the examiner will change the question before uploading it to the exam bank but sees no reason to change the key.

Facility Comment: 2.05

" Question 2.05a-2: Answer should say 'through seal water return header' to VCT.

Question 2.05a-3: Answer should read 'approximately 2 GPH' not 2 GPM.

Question 2.05b-1: Answer should be ' Maximum flow rate of approximately 100 GPM.'"

Resolution:

The examiner will change the key.

Facility Comment: 2.06b

" Question is not clear if Administrative Maximum load or Design Maximum load is requested. Procedure S01-10-1, which is referenced, saysimax.

load is 4725 KW with no time limit. Thisuis Admin. limit. Answer key-provides Design limit. Operationally we are limited to Admin. limit by procedure."

Resolution:

The examiner will change key to Admin. limit and time limit to indefinite.

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- Facility Comment: 3.02b'

'"Long term responses may also be included in the answer,Ieg..

PORV may open:after some time with no sprays and max.* heaters, etc."

/ Resolution:

The examiner believes that long term responses will not be likely as the question' refers only to what controls and their response.

' Facility. Comment: 3.03'

" Question is both open ended (Number of control signals is not asked) and Double Jeopardy in that answer to Part b is dependent to answer to Part a. . Answer is correct but credit need be given if only two or three inputs are provided and points should not be deducted twice."

Resolution:

The examiner does not understand the concerns here. The question would not have been considered by the reviewers to have been double jeopardy if the Part a had requested both the inputs and control of director.

The examiner considers this a two part question. Secondly, the examiner considers the control of direction and input to be one answer,-despite

'the format of the question.

Facility Comment: 3.06

" Variable Low Pressure Trip has a nominal value of 1840f with an

. instrumentation setpoint of 1872# to insure that trip occurs before I -design nominal value of 1840f. Copy of' lesson plan covering this is attached. SCE accepts either answer in our exams Pzr HI Level has been changed to 50% recently. Copy of Modification is attached. SIS / LOP

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I should be changed to sequencer since any SIS,-SIS / LOP,. LOP, or' manual Sequencer initiation will trip reactor. See Referenced System Desc.',

page 6."

. Resolution:

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i Will change key.

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Facility Comment: 3.07b

" Low discharge pressure trip setpoint has been changed to 500 psi instead of 675 psi. Copy of FCN is attached. Also, note: this trip is only active in auto."

Resolution:

Agreed, will change key.

Facility Comment: 3.08 "The only. Engine trip on emergency start is overspeed. Differential is a generator trip. Also, question 2.06c asked all trips active on an emergency start and is worth 1.0 points. This question asks engine trips and is worth 0.5 points. This would lead candidate to supply cne answer to this question."

Resolution:

The examiner agrees that the Differential trip is inappropriate, and will drop the answer. The point value of.the answer of overspeed will be increased to 0.5 points to coincide with the 2.06c answer.

Facility Comment: 4.02 Answers and reference (S0123-VIII-4.0 page 6) are for minors. But question does not ask for minors. Answers should be: a) 1.25 Rems /Qtr and b) 7.50 Rems /Qtr."

-Resolution:

Will change key.

Facility Comment: 4.03

" Reference should more correctly state Technical Specifications, Section 4.0.2, page 4-1."

Resolution:

Agreed, will change key.

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Facility Comment: 4.04 "Not a performance based question. SCE does not use this' term in its operation of the facility."

Resolution:

The examiner points out that the NRC holds the licensed operators of a facility responsible for knowing the regulations as they pertain to his duties. This is in spite of the terminology of the facility, the question does ask for the definition of the term in the regulation.

Facility Comment: 4.05b

" Procedure does not specify that immediate action is necessary on.off scale High or increasing rapidly. Also, action in fold out page.

Action listed in answer is correct.

Resolution:

The examiner points out that this procedure on the foldout sheet is an always applicable requirement during the course of the Steam Generator Tube Rupture procedure. That being the case it is, arguably, an immediote instruction.

Facility Comment: 4.07a "Same reference also says that if pump has been running for greater than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , immediate restart may be performed, i.e. no time delay."

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Resolution:

Agreed, will change key.

Facility Comment: 5.02a, 5.02b 9

"5.02a: Question'is almost the same as RO question 1.01a except SRO does not ask candidate to indicate new operating point, yet answer key is just la copy of R0 answer key. Change to answer key for SRO is needed.

5.02: This is a completely different question form RO question 1.01b but answer key is for RO question! New answer key to answer SRO question is needed."

, Resolution:

J Agreed, will change key.

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" Answer key incorrect. Answer should be Pressure 2 not Pressure'I. DNB has not occurred at Pressure 2 at this heat flux."

Resolution:

Agreed. will change key.

Facility Comment: 5.06a "See comments for RO question 1.10."

Resolution:

Same changes as RO 1.10.

Facility Comment: 6.01 "See comments on RO question 2.01." j Resolution:

Same changes as RO 2.01.

-Facility Comment: 6.03 The reviewers noted that the Cryogenic unit has been retired in place.

Resolution:

The examiner will leave the question in the exam but delete it before uploading to the exam bank.

Facility comment: 6.04 "See comments on RO question 2.03."

Resolution:

Same as question 2.03.

Facility Comment: 6.06 "See comments on R0 question 3.03."

resolution:

Same as question 3.03 t

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' Facility Comment:. 6.07 "See comments on RO Question 3.02."

. Resolution:

Same as question 3.02.

Facility Comment: 7.01

" Parameters checked for Priority 4 Heat Sink are Auxiliary Feedwater and Main Feedwater. Which reflect Primary and Secondary heat removal capability."

Resolution:

Agreed will change key.

. Facility comment: 7.03

" Point value too high! 20% of the:section for one question. Reference has'been changed. Procedure S01-14-17 has been superseded and replaced u

-byfS0123-0-23.1, copy attached. Answers are still OK."

Resolution:

The. examiner again notes that the-standards allow a 20%. limit for a question.

Facility Comment: 7.04

" Max. Delta T between Pzr. and RC has changed from 200 deg. F to 190 deg. F. Also, Max. Heatup rate of Pzr. has changed from 95 deg. F/Hr t'o.

90 deg. F/Hr. Copy of revised procedure page is attached."

, Resolution: -

Agreed, will change key.

Facility Comment: 7.05a

" Rod position for criticality is usually only related to control bank 2.

By Tech. specs, and admin. controls as well as this startup procedure all other groups are full withdrawn. This is usually assumed and generally not given as part of the answer for this question."

Resolution:

The examiner will eliminate the reference to the other banks. >

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Facility Comment: 7.06

" Reference is wrong. Question asks per procedure S01-3-1 yet reference is to S01-3-2. Procedure S01-3-1 says observe bubble formation when RCS temp. 250 deg. F and PZR 440 deg. F. Procedure does not discuss answer that is in answer key although it is correct.

Resolution:

The key will stand.

Facility Comment: 8.01 This question is false if it is assumed that the rod is immovable "due to binding or friction" per Tech. specs. If immovable for other reason but untrippable then question could be answered true."

Resolution:

The key will stand.

Facility Comment: 8.02

" Question does not ask per Tech. Specs. or administrative 1y.

If Admin. is assumed then if, in Mode 3 and trip breakers are racked in, then 3 reactor coolant pumps required. Reference S01-3-4, Precaution 4.1.6."

Resolution:

It is true the question does not specify Tech. Specs. limitations.

However, the examiner feels no change is required of the key.

Facility Comment: 8.03a, 8.03b

" Question may be answered generically ie. 1.25X surveillance interval, and not just 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months . Typo in answer key b.1. "last done + 15 Hours =

November 29 at 2000 not 2900."

Resolution:

Agreed will change key.

Facility Comment: 8.05

" Reference has been superseded by procedure S0123-0-14. Copy of Attachment 4 to the procedure is attached."

Resolution:

Will change reference.

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,. 4-Facility Comment: 8.06 -

"For answer to'be correct only one of four fuel transfer pumps can be operable. If one per DG is operable then startup may proceed '

with no -

restrictions, ie. Tech. specs. require one per DG.""s r Resolution:

During the course of the examination the examiner informed the. candidates that both fuel pumps to one DG were inoperable. /

Facility Comment: 8.08

" Question is unclear, if by deviations a TCN is met then answer b and d are correct. However, SCE policy is you never deviate (meaning carry out action outside the procedure without preparing a~TCN) from a procedure except in an emergency then a and d are correct. Reference should be 6.8.3 not 6.8.1 and this talks about temporary changes not deviations."

Resolution:

With two possible answers it appears that there is sufficient information for the candidates.

Facility Comment: 8.10b

" Primary responsibility for the EC is the Unit Superintendent. It is true the first recalled EC relieves the SS."

Resolution:

The examiner fails to see any concern.

- . = --

U.S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR REQUALIFICATION EXAMINATION Facility: SONGS 1 Reactor Type: WESTINGHOUSE - 3 LOOP PWR Date Administered: NOVEMBER 18. 1986 Examiner: G. W. JOHNSTON Candidate: 4',. y

' 1 INSTRUCTIONS TO CANDIDATE Read the attached instruction page carefully. This examination replaces the current cycle facility administered requalification examination. Retraining requirements for failure of this examination are the same as for failure of a requalification examination prepared and administered by your training staff.

Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up four (4) hours after the examination starts.

% of Category I of Candidate's Category Value Total Score Value Category 15.0 25.0 1. Principles of Nuclear Power Plant Operation, Thermodynamics, Heat Transfer and Fluid Flow l 15.0 25.0 2. Plant Design Including Safety and Emergency l Systems t

l l 15.0 25.0 3. Instruments and Controls i

15.0 25.0 4. Procedures - Normal, l- Abnormal, Emergency, and !

Radiological Control 60.0 TOTALS l Final Grade All work done on this examination is my own I have neither given nor received aid.

Candidate's Signature -

I e

4 EQUATION SHEET f = me

v = s/t W " 28 s= 2 Cycle efficiency = Net Work (out) 2 o t+ hac Energy (in)

E = aC a = (vg - vg )/t 2

IE = mv V f=v o+ at A = AN A=Aeo ~" '

PE = agh a = 6/t -

A = In 2/cg = 0.693/tg AE = 931Am %(* I) " (t, )(tu) i , (t +t) b Q = p,C,aT y y ,-rx k=UAAT

g ,7 ,-yx Pwr = W '

g a 7,y 10'*

P = P, 10 EI") TVL = 1.3/g -

P = P,e"II HVL = 0.693/p !

SUR = 26.06/T i T = 1.44 DT SCR = S/(1 - K,g,)

2 SUR = 26 IA*ff'I CR g,p x = S/(1 - K,gg )

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T = '(t*/o ) + [(f 'p)/1,ggo ] 1( eff}1.= CR2 (1 - K,gg)2 T = 1*/ (o - T) M = 1/(1 - K,gg) = CR g/CR0

~" eff' 8 " IEeff M = (1 - K,gg)0/II ~ Kaff)1

~I)I eff " #eff/Keff SDM = (1 - K,gg)/K,g, p= ~

[t*/TK, igg.] + [E/(1 + A,ggT )] 1* = 1 x 10 seconds P = I(V/(3 x 10 0) A,gg = 0.1 seconds-1 i

E = Na i

Idly =1d22 WATER PARAMETERS Id =Id2 g

1 gal. = 8.345 lba R/hr = (0.5 CE)/d (meters)

I gal. = 3.78 liters 3 R/hr = 6 CE/d (feet) i 1 fc = 7.48 gal.

MISCELLANEOUS CONVERSIONS l Density = 62.4 lbm/fc 3 1 Curie = 3.7 x 1010 dps Density = 1 gn/ca, I kg = 2.21 lba Heat of varorizations = 970 Etu/lbm I hp = 2.54 x 103 BTU /hr Heat of fusica = 144 Beu/lba 1 Mw = 3.41 x 10 Btu /hr 1 Atm = 14.7 psi = 29.9 in. I's. 1 Btu = 778 ft-lbf 1 ft. H 2O = 0.4333 lbf/in 1 inch = 2.54 cm F = 9/5'c + 32

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'C = 5/9 ('T - 32) -

SECTION i Prinicples of Nuclear Power Plant Operation, Thermodynamics, Heat Transfer and Fluid Flow eQUESTION 1.01 Refer to figure 5.1, a sketch of a typical Auxiliary Feedwater System utilizing two certrifugal pumps of similiar

, characteristics and capacities. The plot of volume Flow Rate versus Pressure shows the system with the "A" Auxiliary Feed Pump in operation as an initial condition.

a) Show the changes from the initial conditions to the curve (s) as the PORV or atmospheric dump opens and reduces Steam Generator pressure by 50 percent. Indicate the new system operating point. (1.0) b) Show the changes from the initial conditions to the curves as the second pump "B" is started. (1.0)

NOTE INDICATE DIRECTION AND MAGNITUDE OF CHANGES.

ANSWER Figure. (2.0)

REFERENCE General Physcis, Volume III, Chapter 2.

QUESTION 1.02 There are two ways to describe flux distribution variations that can occur'in the reactor core. These are typified as flux tilts.

a) What is Axial flux tilt? (0.5) b) What in Racial flux tilt? (0.5)

ANSWER a) Variation of flux along the vertical axis. (0.5) i b) Uneven flux distribution across a horizontal plane view of the core. (0.5)

REFERENCE Westinghouse Trairing Netes.

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QUESTION 1.03 The flowrate through the Steam Generators on the primary side is twelve (12) times the flowrate through the secondary side.

Explain what parameters cause this apparent difference. (2.0)

ANSWER ,

The difference in the enthalpy on the secondary side is much higher. (1.0) This is so because the sensible heat given up by the primary in the steam generator is taken up to a great degree by the latent heat of vaporization during the change of phase.

(1.0)

REFERENCE

QUESTION 1.04 The Xenon peak that occurs after a reactor trip from 100 percent equilibrium xenon condition is greater than the peak for a trip from 50 percent power due ton (1.0) a) The fission yield for Xenon is higher at 100 percent power.

b) There is more Iodine in the core at the time of a trip from 100 percent power. '

c) There are more thermal neutrons in the core at 100 percent power.

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

ANSWER .

b) (1.0)

REFERENCE VIII-06 ' Principles of Power Control'.

!

QUESTION 1.05 If reactor power increases from 1000 cps to 5000 cps in 30 seconds, what is the Startup Rate (SUR)? (2.0)

ANSWER P = Po (10E SUR(t))

5000 =

000 (10E SUR(.5)) C1.03 SUR e (LOG 5)/(0.5) = (1.699)/(.5) = 1.4 DPM [1.03

REFERENCE VIII-05 ' Reactor Kinetics *.

eQUESTION 1.OS Which of the following six 3'a-+4aa-achov-formula terms increases on a power escalation to allow reactor power to match turbine power?

(1.0) a) The fast fission factor.

b) The thermal utilization factor.

c) The reproduction factor.

d) The doppler effect.

ANSWER b) (1.0) '

REFERENCE

VIII-03 ' Neutron Cycle'.

eQUESTION 1.07 The most serious problem with reaching the critical heat flux (CHF) in a power reactor is caused by: (1.0) a) the poor thermal conductivity of steam.

b) the blockage of flow through the core when steam bubble formation becomes significant.

c) the displacement of boron from the core as steam bubble formation becomes significant.

d) the high pressure surges in the reactor coolant systems l caused by steam bubble formation.

ANSWER

, a) (notes talk to steam blanketing, clad burnout). (1.0)

REFERENCE General Physics HTFF Notes, Chapter 4, Part "B",

Pages 220-230.

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QUESTION 1.08' j

i The Fuel Temperature Coefficient increases (becomes more '

negative from BOL primarily due to (1.0) a) the reduction of fuel to clad gap distance.

b) the r' eduction ir? the moderators boron concentration.

c) the increase in Pu-240 in the core.

d) the increase in thermal neutron flux.

ANSWER c) (1.0)

REFERENCE '

VIII-06 ' Principles of Power Control'.

DUESTIDN 1.09 The reactivity worth of a control rod increases: (1.0) a) as Tave increases from 150 degrees fahrenheit to 500 degrees fahrenheit.

b) as reactor power is reduced from 100 percent to 50 percent.

c) as a result of fission product buildup.

d) when the soluble baron concentration increases.

ANSWER a) (1.0)

REFERENCE VIII-06

Principles of Power Control'.

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eQUESTION 1.10,

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Figure 5.5 is a? sketch of Reacior Power versus Time in hours. At t=0 hours reactor,startup from Xenon free conditions to 100 percent power occurs.At't"50 hours5.787037e-4 days 0.0139 hours 8.267196e-5 weeks 1.9025e-5 months a reactor trip occurs followed by a reactor startup to 100 percent power at t=65 hours.

a) Sketch the Xenon reacti41ty response in the core from this power transient. (Indicate approximate magnitude and duration of each transient.) (1.0) i, b) Indicate the time in hours at which the maximum negative reactivity will be inserted by Xenon. (0.5) 5 c) Indicate th'e' tune in hou s'kr.pproximately) that the maximum rate cf rod insertsen will have to occur in order to overcome gthe slop'e of the Xenon transient (Assume constant Tave and no boration or dilution) . (0.5)

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ANSWER a), b), and c) attached

REFERENCE i ,

General Physics Volume I I ', Chapter 4, Section "D".

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QUESTION 1.11 .

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Listed below a-e :two f actors that can af f ect Departure from Nucleate Boiling Rntio-(DNBR). What will the effect on DNBR (increase or decre'ame) be, if the factors listed below are increased? j1.0) a) Reactor Power.V (O. $)

b) Reactor Co61an, Syste flow. (0.5) .

ANSWER a) DNBR decreases. (0.5) b) DNBR increases. (0.5)

REFERENCE GP HTLFF no*:es.

- - - - 4*********************************************

End of Section i Go on to Section 2

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l SECTION 2 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS

QUESTION 2.01 Concerning the Feed System and Safety Injection Systems

a. What pump and valve changes will automatically occur in the Feed and Condensate system when, at full unit load, a Safety Injection signal is received 7 (2.0)

b. What specific valve interlock prevents the injection of Feed and Condensate water into the RCS 7 (0.5)

c. If during the SIS sequence the SI pump suction isolation valve (HV-853 A&B) f ails to open, will the feedwater pump run ? Explain ! (0.5)

ANSWER

a. - Main feed pump trips and then restarts (0.5)

- Condensate pumps trip (0.25)

- Heater Drain pumps trip (0.25)

- MFP suction valves (HV-854 A&B) shut (0.25)

- MFP discharge valves (HV-852 A&b) shut (0.25)

- Feedflow control valves (FCV-456/7/8) shut (0.25)

- Feedflow Aux control valves (CV-142/3/4) Shut (0.25)

- (SIS Suction Header Isolation Valves (HV-851 ALB) will open.) (N/A)

b. MFP Suction valves (HV-854 A&B) must be shut before the SIS header isolation valves (HV-851 A&B) will open. (0.5)

c. Yes, the FW pump will run for 30 sec. then if HV-853 does not open, the pump will trip. (0.5) reference: SD-SO1-580, pp.16, SO1 exam bank.

SM Wa <W lw 4 [n g(s o 2 0, E I,

QUESTION 2.02' Concerning the CVCS systems

a. Under what three conditions would the Excess Letdown System be Used 7 (1.5)

b. Describe the three Excess Letdown Alignments. (1.5)

ANSWER a.1. inoperability of normal letdown (0.5)

2. used during PZR Steam Bubble formation (0,5)

3. To increase the cooldown rate during the initial cooldown phase of RHR operation. (0.5) b.1. Normal - thru seal water return piping 'into the VCT. (0.5)

2. Thru a section of normal letdown piping and a residual heat exchanger tu the VCT. (0.5)

3. Thru and then directly to the RCS drain tank. (0.5)

REFERENCE:

SD-SO1-310 pp.5, fig. I-1.

DUESTION 2.03

a. List the cources of air that can be used to supply the Instrument Air Header? (1.5)

b. How are the following valves affected by the loss of Instrument air (operable / inoperable) 7 If operable, state why? If inoperable, state status: (failed shut, failed as is, or failed open). (1.5) l 1. Main feedwater flow control valves l 2. Emergency Auxilary feedwater regulators l 3. Safety injection valves HV851,852,853,854 i

j

ANSWER l

a.1. Three Service Air Compressors (0.5)

2. Auxilary Air Compressor (0.5)

3. Diesel Air Compressor (0.5) l l b.1. Inoperable (0.25) - Failed open (0.25)

2. Operable (0.25) - Nitrogen Backup (0.25) .

3. Inoperable (0.25) - alarmed in the contro1 rook, Operable if operator selects Nitrogen backup (0.25)

{

References SD-SO1- 420 & 580 3 SO1-2.4-2 IA Sys Malfunction l

l

QUESTION 2 . O'4 TRUE or FALSE 7 In order to move the carriage in the fuel transfer system, both upenders must be fully down. TRUE or FALSE 7 (1.0)

~

ANSWER TRUE (1.0)

~

Reference SD-SO1-350

QUESTION 2.05

/

During normal operation, about 7 gallons per minute (GPM) of seal water is supplied to each reactor coolant pump.

a. For the following, give the normal flow rate past each, and tell where the majority of that flow goes.

1. Thermal barrier (0.5)

2. #1 seal (0.5)

3. #2 seal (0.5)

4. #3 seal (0.5)

b. in the event of a# 1 seal failure:

1. What is the expected flowrate through the seal? (0.5)

7. . What limits the flowrate to this value 7 (0.5)

ANSWE.R q vw #"

a.1. approx. 5 GPM (0.25), into the RCS (0.25)

2. approx. 2 GPM (0.25), to VCT (0.25)

3. approx. 2 (0.25), to Vapor Seal Head Tank /

GP RCS Drain Tank (0.25)

4. approx.100 cc/Hr.(0.25), to Atmosphere (0.25) tit,$k. Al& VO & QffX.

b.1. 8 00 GPM (0.5)

2. Bushings in the Floating Ring Seals.

Reference SD-SO1-3OO pp.8-11

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

eQUESTION 2. O'6 With regard to the diesel generator systems

a. How long can a diesel generator operate at full power without replenishing the Day Tank ? (0.5)

b. what is the maximum load a diesel generator can supply (0.5),

and how long can it supply this load? (0.5)

, c. List the trips that remain in effect, when the diesel has started automatically due to safeguards actuation. (1.0)

ANSWER 5m tes to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . p ie he

a. a e (0.5)

b. wbbOO kW (0.5) fr 2 ".:ur; ir : 24 '. p;ried (0.5)

c. 1. Generator differential. (0.5)

2. Overspeed. (0.5)

Reference SD-SO1-600 pp.2,3,93 SO1-10-1 END OF SECTICN 2 Go on to Section 3

)

l l

1

SECTION 3 INSTRUMENTS AND CONTROLS

QUESTION 3.01 Which of the following is true concerning the Source Range Channel hi'gh voltage cutoff? (1.0)

a. During a reactor startup either IR channel increasing above the P-6 setpoint will turn off the high voltage.

~

b. If one IR channel fails low while at power, both Source Range high volatage will be reenergized.

c. If a IR channel increases above 2 X 10 E(-9), it will turn off its respective source range's high voltage,

d. During a reactor shutdown either IR channel decreasing below the P-6 setpoint will turn on the high voltage.

ANSWER

c. (1.0)

REFERENCE SD-SO1-380 Excore NI's

QUESTION 3.02 Pressurizer Pressure Transmitter PT-430 fails low during normal at power operations.

a. Name one alarm that you expect to annunciate in the control room as a direct result of the PT f ailure ?

(0.5)

b. What pressurizer systems or controls will be affected and how will they respond? (1.0)

ANSWER !

a. Safety Injection Train A Channel 1 Alarm or VLPT Channel 1 Alarm (0.5)

b. 1. Spray Valves will close if they are open (0.5)

2. All heaters in Automatic will turn on (0.5)

Refercrce Re g?3 Exam Bank

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

eQUESTION 3.03

a. What 'are the control signal inputs to the Automatic Rod Control System? (1.0)

b. For each of the above, State if the input controls rod speed and/or rod direction. (1.5)

~

ANSWER a.1. T avg (0.25) b.1. Direction and speed (0.5) ,

2. T ref (0.25) 2. Direction and speed (0.5)

. 3. Nuclear Flux (0.25) 3. Speed (0.25)

4. P-P ref (0.25) 4. Speed (0.25)

REFERENCE SD-SO1-400, and Requal Exam Material -

eQUESTION 3.04 What are the two (2) Turbine "Runbacks", and to what value will each run back the Turbine? (2.0)

ANSWER

1. Dropped Rod (0.5) -

runback to 70 % (O 5)

2. Low Frequency (0.5) -

runback to 85 % (0.5) a Reference SD-SO1-2OO l

i

QUESTION 3.05 Concerning the Sub-Cooling Monitoring Systems

a. What signals are inputs to each channel of the Sub-Cooling Monitoring System 7 (1.0)

b. How is the margin to saturation computed 7 (1.0)

c. At what margin does it alarm? (0.5)

ANSWER

a. Temperature 4 Core Exit Therr.ocouples (0.25) 3 Hot Leg RTD's (0.25)

Pressures 2 Pressurizer Pressure Xmitters(ChA) (0.25) 1 for Ch. B -

(0.25)

b. The highest of the 7 Temperature signals is compared to the saturation temperature calculated from the lowest pressure signal input for that Ch.

This difference is the output or margin to Saturation. (1.0)

c. 40degreesF(+/-2deg) (0.5)

Reference SD-SO1-390 pp.- 20.

D

_- - . _ .- =

QUESTION 3.06 List 5 of the 11 Reactor Trips, include in your answers name of the trip, setpoint(s), logic, and initiating devices. (3.0)

ANSWER -

Trip. Setot. coincidence Device

/loaic

1. -Hi SUR 5 DPM /,g, lA[j;re 1/2 'I . R . NIS.

2. Va. low Press. 1872 min.)si 8 2/3 PZR Press.

- 3. Turbine Trip 45 PSIG 2/3 A.S. Oil

4. *Two Loop RCS 85%or BKR 2/3 loop Flow /BKR low Flow position ,

contacts

5.

Single Loop RCS 85%or BKR 1/3 Loop Flow /

Low Flow position contacts

6. PZR Hi Level 70 % 2/3 level channels

7. PZR Hi Press. 2200 psig 2/3 press. channels

8. Overpower 25%,85%,109% 2/4 P.R. NIS

9. SF/FF Mismatch SF>FF by 25% 2/3 Channels

10. O f 9 ft-S P7.5 <..- e -w(5 $/4*D,W,545 / #d 1/2 Sequencers

11. Manual 1/2 Pushbuttons

May be listed as one trip depending on assumptions concerning P-8 Status

( any 5 lines 0.6/line or O.15/ item)

REFERENCE SD-SO1-570 and Requal exam bank 1166 e,

)

f 1

6

, , . . . - - - . ~ , . - .-.,.-----w--,,,--,----v, c.-ww.-y,-.,,.,c , , - , - ~ . - ~ . . - - - . - , , - - - - - , , , , , ,,..-----.---...-.,% e , ,7 -.- - - . - -,.w.y

QUESTION 3. 0'7 Concerning the Auxilary Feedwater System (AFW):

a. What will Auto Start the AFW pumps?

( include in your answer coincidence logic and setpoints ). (1.0)

b. What will trip the AFW pumps 7 (1.0)

ANSWER

/

a. Low S/G Le0el 5*/. N . R . ( 0. 5 )

2/3 S/G's per train (0.5)

(ch.A = Motor driven ch.B = Turbine driven)

b. Low Suction Pressure 0.6 psig for 10 med. on

auto start - both pumps ao

Ch.A - Low Disch Press. - ~ psi (0.25)

Ch.B - Overspeed (0.25)

Reference SD-SO1-620 and requal exam bank

QUESTION 3.08 For the Diesel Generators

a. List the Engi.ne trips that are in service on an Emergency start. (0.5)

ANSWER

a.

engine overspeed ;C. 25J(& d)

=

9:nrrat:r dif'-- ntiel '^ 25) .

Reference Requal Exam 1175 End of Section 3 Go on to Section 4 S.

Section 4 Procedures - Normal, Abnormal, Emergency, and Radiological Control

QUESTIDN 4.01 Regarding SO1-1.4-1,

Response to Imminent Pressurized Thermal Shock Condition *:

(J.O) a) After termination of an overcooling transient that results in a Safety Injection actuation why is it important to maintain RCS temperature and pressure stable?

ANSWER a) To avoid potential pressure increase that could result in potential overstress of the pressure vessel. (1,0)

REFERENCE SO1-1.4-1, page 19 eQUESTIDN 4.02 What are the Regulatory Limits for exposure for t'he following areas of the body in Rems per Quarter?

a) The hands and forearms. (0.5) b) Skin of the whole body. (0.5)

ANSWER %.1[ ,

a) ##*?" Rems /Qtr. (0.5) b) fh7,50 Rems /Qtr. (0.5) i The above limits are all that is required for full credit.

REFERENCE SO123-VII-4.0 page 6 l

i

eQUESTION 4.00' Regarding SO1-2.1-9 " Loss of Residual Heat Removal System".During refueling operations with the Reactor Coolant System drained down the operating train of RHR experiences conditions that indicate a loss of flow in the system.

a) What are two symptoms (not Alarms) that could provide that indication? (1 0)

ANSWER a) Any two ($.$) :

1. Increasing RHR temperature (any temp. indicator).

2. RHR pump fluctuating current.

3. Decreasing RHR flow or Lo flow.

4. Lack of temperature difference across RHR heat exchangers. '

REFERENCE SO1-2.1-9, page 2.

QUESTION 4.09 Why must the Volume Control Tank pressure be maintained between 14 to 20 psig when the Reactor Coolant Pumps are in operation?

(1.0) *

ANSWER To ensure the lower side of No. 2 seals are maintained wet. (1.0)

REFERENCE SO1-4-3, PAGE 2.

End of Section 4 End of Examination

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QUESTION 4.06 Regarding S01-4-11, " Letdown Domineralizer Operations":

a) The procedure includes a precaution about placing an unsaturated mixed-bed demineralizer in service. What concern can arise if an unsaturated mixed-bed domineralizer -

is inadvertantly placed in service? (A.0)

ANSWER a) Unintentional dilution of boron (Speaks to 80 ppm dilution being anticipated.) CL. 0 )

REFERENCE SO1-4-11, page 2.

QUESTIDN 4.07 Regarding Operating Instruction 01-3-4 " Reactor Coolant Normal Operation":

a) What criteria must be met before the restart of any Reactor Coolant Pump after it has tripped? (1.0) b) How many consecutive starts are allowed for a,two (2) hour interval? (1.0) ,

ANSWER -

a) The pump must stand idle for at least 30 minutes. (1.0) b) Three. (1.0)

REFERENCE SO1-4-3, page 10.

j So I

.- . . . . . - ,. _. - . - -._..-_. ,_ ,__---__ ,,_ - . . _ __ _,_ _ ._, .._ - . .._,,.,_-..-_._,,____.,_n.,-,.._,-.- -

eQUESTION 4.03 During the performance of a monthly surveillance test you

-notice that the previous test was conducted on December 1, 1985.Today is January 5, 1986. This is an interval of 36 days. The test was conducted previously on November 2, 1985 and pctober 1, 1985 (Calender on f ollowing page. )

a) Does this situation constitute non-compliance with the Technical Specifications? EXPLAIN. (2.0)

ANSWER a) No (d.@).The Technical Specification allow a surveillance interval to be extended by 25 percent (0.95). With the interval for 3 consecutive surveillances not to exceed 3.25 times the interval (O.d5).

REFERENCE Technical Specifications 4 01 p

y-/ ,

eQUESTION 4.04 Regarding 10 CFR 55 ' Operators Licenses's a) What are ' Controls as defined by 10 CFR 557 (1.0)

ANSWER a) Controls are defined as aparatus and mechanisms the manipulation (0.5) of which affect the reactor or reactor power (0.5).

REFERENCE 10 CFR 55.4 AND 55.10.

eGUESTIDW 4.05 Regarding S01-1.0-40 " Steam Generator Tube Rupture":

a) Under what condition may the RHR System alignment be made when conducting the cooldown phase of the Steam Generator Tube Rupture procedure if the RCS temperature is greater than 350 deg. F.? (1.0) b) What action must be taken immediately if the ruptured Steam Generator level is OFF SCALE HIGH or INCREASING RAPIDLY?

(1.0)

ANSWER ..

a) If the RCPs are running. (1.0) b) Depressurize the RCS to approximately the main steam pressure. (1.0)

REFERENCE S01-1.0-40

T i

I e, .. ---..-

i i U.S. NUCLEAR REGULATORY COMMISSION l SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION

, Facility: SONCS 1

~

Reactor Type: WESTINGHOUSE - 3 LOOP PWR Date Administered: NOVEMBER 18. 1986 Examiner: .T . P_ O'RRTFM Candidate: KEV INSTRUCTIONS TO CANDIDATE Read the attached instruction page carefully. This examination replaces the cuterent cycle facility administered requalification examination. Retraining requirements for failure of this examination are the same as for failure of a requalification ====ination prepared and administered by your training staff.

Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up four (4) hours after the examination starts.

% of Category % of Candidate's Category Value Total Score Value Category 15.0 25.0 . 5. . Theory'of Nuclear Power Plant Operation, Fluids, and Thermodynamica

, 15.0 25.0 6. Plant Systems Design, Control and Instrumentation 15.0 25.0

7. Procedures - Normal, Abnormal, Emergency, and Radiological Control 15.0 25.0 8. Administrative Procedures, Conditions, and Limitations 60.0 TOTALS Final Grade All work done on this examinacion is my own, I have neither given nor received aid. ,

Candidate's Signature e

\

o i

. ie

. EQUATION SHEET f = ma -

y e/e w = mg a = v,t + at 2 Cycle efficiency = Net ,

I rk (out)

E = aC a - (vg - V )/t KE = my vg A = AN

= v, + a A = A,e" E PE = agh a = 8/t A = in 2/tg = 0.693/tg W = v4P AE = 931Am t g(eff) = (c, )(th )

(t + ty Q = k AT ,

I . r .-EX Q = UA T I . r .-UX Pwr = Wg la ~

I=I 10 *!

. P=P 10 SUR(t) yyn . 1,3fy .

t P=P o

e /T HVL = 0.693/u SUR = 26.06/T T = 1.44 DT SCR = S/(1 - K,gg) fA ggo \

SUR = 26 CR

= S/(1 - K,ffx) 6-p x CR 1 (1 - Keff)1 = CR 2 (l - Keff)2 T = '(L*/p ) + [(f 'p)/A,ggp ]

T = 1*/ (p - D M " 1/(1 - Eeff) = CR /CR g 0 "I - 8)! eff 8 M = (1 - K gg)0/ (1 - K,ff)g 8"I eff ~I)I eff " #eff/K,gg

SDM = (1 - Keff)/Keff p= -

[L*/TK;gg.] + [H/(1 + A,ggT )] ,

1* = 1 x 10 seconds P = I4V/(3 x 1010) -1 geff = 0.1 seconds I = No Idlg=Id22 WATER PARAMETERS Id =I022 g

1 gal. = 8.345 lba R/hr = (0.5 CE)/d (meters) 1 gal. = 3.78 liters 3

R/hr = 6 CE/d (feet)

I fc = 7.48 gal.

MISCELLANEOUS CONVERSIONS .

Density = 52.4 lbm/ft 3 1 Curie = 3.7 x 1010dps Density = 1 gm/cm i kg = 2.21 lbm Heat of varorization = 970 Etu/lba 1 hp = 2.54 x 103BTU hr

, Heat of fusica = 144 Beu/lba 0 1 Mw = 3.41 x 10 Beu/hr 1 Atm = 14,7 psi = 29.9 in. I'g. 1 Btu = 778 ft-lbf I ft. H oy = 0.4333 lbf/in 1 inch = 2.54 cm F = 9/5 C + 32

C = Q/8LR*rLa gtL -. .__ . . .

0 Section 5 Theory of Nuclear Power Plant Operation, Fluids, and Thermodynamics

0UESTION 5.01 The reactor is determined to be shutdown by 6 percent delta K/K with indication in the source range of 30 counts per second.

a. What is the Keff when the reactor is shutdown by 6 percent

. delta K/K? (0.5)

b. What would the count rate be if Keff is increased to 0.987 (0.5) -

c. What would the count rate be if Keff is increased to 0.997 (0.5)

ANSWER delta K / K = 0.06 1 - K / K = 0.06 1 = K + 0.06(k) 1 = 1.06(K)

a. KEFF = 1/1.06 = 0.94 (0.5) 1 - K1 / 1 - K2 = CR2 / CR1 and 0.06 / O.02 = CR2 / CR1

b. CR2 = (30) (3) = 90 cps (0.5)

. Conversly CR3 / 30 = 0.06 / O.01

c. CR3 = (30) (6) = 180 cps (0.5)

(Rule of thumb, doubles)

REFERENCE Glastone and Sesonske, " Nuclear Engineering".

p.,n_., . - - - , - - . , -.-__-.._-_-,...--_,,ey ,_ m., g _. ,. --.---- ,- . .

a

QUESTION 5.02 Refer to FIGURE 5.1, a sketch of a typical auxiliary feedwater system utilizing two centrifugal pumps of similar characteristics and captities. The plot of Volume Flow Rate versus Pressur e shows the system with the."A" auxiliary feed pump in operation as the initial condition.

a. Show, on the Figure provided, how the curve (s) will cht.nge as'the FORV opens and reduces Steam Generator pressure by 50 percent. (1.0)

.. b. Show, on the Figure provided, how the curve (s) will change when the discharge valve for the "A" pump is partially shut. (1.0)

ANSWER

a. Attached (1.0) ..

b. Attached (1.0)

~* REFERENCE Section III, Part B, General Physics manual on Heat Transfer and Fluid Flow.

QUESTION 5.03 Refer to the FIGURE 5.2 that follows this page. The figure is of " Heat Flux" versus " Temperature Difference between a Wall and the Bulk Fluid" for an' operating reactor. Note that there are two curves represented for two pressures ( P1 less than P2).

a. What is the principle type of heat transfer that is occuring at pressure P1 and 1.OE4 BTU /Hr-ft between the wall and the bulk fluid? (0.5)

b. What is the principle type of heat transfer that is occuring at Pressure P2 and 1.OE4 BTU /Hr-ft between the wall and the bulk fluid? (0.5)

c. What pressure will yield a lower fuel centerline temperature at 3.OE5 BTU /HR-FT7 (0.5) /

d. Asuuming bulk temperature well below saturation, Will decreasing the pressure affect the bulk fluid temperature at a heat flux of 3.OE5 BTU /Hr-ft? (0.5)

ANSWER .

a. Nucleate boiling. (0.5)

b. Convection (other terms may be used). (0.5)

c. Pressure K. (0.5)

d. No acceptable. The bulk fluid temperature remains constant (in independent of pressure below saturation.) (0.5)

REFERENCE General Physics Nuclear Technology, Section E, Pages 2-144, 2-151- 2-159, and 2-164.

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0 10 10 0 10 10 10 I

l Temperature Difference Between Wall and Bulk Fluid ( F )

FIGURE 5-2.

9 Se I(q. .

o

DUESTION 5.04 Regarding a Main Steam line ruptures

a. Why is a rupture of a Main Steam line at End of Life-(EGL) a much more limiting accident than at the Beginning of Life (EOL)? (1.0)

ANSWER

a. Th6 Moderator Temperature Coefficient i s less negative at EOL (f.03). Than at EOL.This difference in magnitude increases the severity of the addition of positive

.. - reactivity O M. due to the sudden cooling of the reactor coolant from the Main Steam line rupture. (WFTT

REFERENCE General Physics, Volume III, Chapter 3, Section 3.

QUESTION 5.05 Periodically chemical agents are added to the Reactor Coo'lant System (RCS).For each of the following chemicals added to the RCS, indicate why the chemical 1s added and (what is it designed to control):

a. Hydra:ine (0,5)

6. Hydrogen (0.5)

c. Lithium Hydroxide (0.5)

ANSWER

a. Control of 0::ygen (at low temperature). (0.5)

b. Control of oxygen (at power). (0.5)

c. Control of pH in RCS. (0.5)

REFERENCE System Description " Chemical and Volume Control System" W

6 e

a

-eamm,= ,m-4 , , - -w-g--- - - - - - p w- me y , - ,- , , - ,,_ - ,v,,-~ -

o *

QUESTION 5.06 Figure 5.5 is a sketch of Reactor Power versus Tin e in hours.

At t=0 hours reactor startup from Xenon free conditions to 100 percent power occurs.At t=50 hours a reactor trip occurs followed by a reactor startup to 100 percent power at t=65 hours.

a. Sketch the Xenon reactivity resporse in the cor e f rom this power transient. (Indicato appro:<imata magnitude tpercent delta k/k3 and duration of each part of the transient.) (1.5)

b. At what time in hours will the maximum rate of rod insertion have to occur in order to overcome the Xenon transient. (Assume constant Tavg and no boration or dilution.) (0.5)

c. What are the production and removal mechanisms for Xenon?(1.0)

ANSWER

a. (Attached) (1.5)

b. 65 to 70 hours8.101852e-4 days 0.0194 hours 1.157407e-4 weeks 2.6635e-5 months (0.5)

c. Xenon Production - Iodino decay (.25)

Direct yield from fission (.25)

Decay (.25) .

Burnout (.25) -

REFERENCE

" Westinghouse Training Notes", ' Introduction to'PWR Control' and

Fission Product Poisoning *.

h

e FICURE 5.5 4

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rms a 'Io n. (Q (o.s) x e

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_-....._...._-.m._ , - . _ _ _ . . . _ , . . . . _ _ , . _ , ,,_,____-__,,%,_,.__.____. m. . . - . . . - _ . - _ . - - . _ . ,

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QUESTION 5.07 For the'fs!'iowing'andicate which-is thu,most acc. urate statement: ,

a. Tr t' tNin conde'nser functions,by: (0.5) a 1.Remosleg the latent heat of vapor.'.:ation at a constant terperature to allch the steam to condense. .

7

2) P"oviding a low pressure volume that allows the steam to condense. ,,

c. t. ;

'~ '

i 3. 'Co? ling thu steam to the foint where is at natuaration F.

1

', to npt.r ature .

t

'.r 1

.\

e b. Condensate depression ist (0.5) t

(. Paintained'by adequate backprecsure on th'a low pressure turbine exhaust.

l 2. U3ed to main'tain constant temperature profile on the condons2r tubeshteets.

3. Used to mair,tain adequake Net Positive Suction Head f or

'the co7densato pumps.

o 1

4. Used 'to maintain adequate !!st Positive Suction Head f or '

the main feedwater pumps.

t,

ANSWER

a. 1. (0.5)

o. 3. (0,5) !

\

REFERENCE ,

i 1( '

General Physt:s "He.it Jr.?nsfer aind Fluid Flow"

0L,t'.G T:CN 5. 00 At, Dv41nning of' Life .(BOL) tl=e p4 ant nuclear instrumentation t.oi .pt rate wi ll ina.rease as plant tamperature increases causing a'heactor Coolant Systre 4RCS) coolant density decrease.

a. Assuming there is ne rod movamunt or changes in RCS boron concentration. What two factors contribute to this phenomana? (1.0)

,

ANSWER I

e. 1. tlautrons travoi forther, more leakage. (0.5) 2, , l.owor boron donsi ty, less' baron abuorpiten. (0.3) i

/ **

l # REFERENCE General Physics, Volume II, Chaper 4, Sections D & D s

b u__________.__._.__. 1 ._ _

r i O

0UESTION 5.D9 Over the life of a core, the composition of the fuel changes.

a. How and why does this change in composition overr core lif e affect the reactor period for equal reactivity additions?

(1.01

ANSWER

a. The decrease in Uranium 235 and the increase-in Plutonium r %

239 cause the Beta Bar Effective to decrease. . (0.5) -

This Beta Bar Effective change causes the reactor period to decrease. (0.5)

REFERENCE General Physics, Volume II, Chc.pter 5, Sect. E.:

QUESTION 5.10 -

Regarding the feedwater heaters:

If the level in a feedwater heater is allowed to increase, what will happen to the temperature of the f eedwater exiting ~

the heater, increase, remain the same, or decrease? Why?

(1.0)

ANSWER Decrease (0. The level increase exposes 1ess surface area intheheate$5).

r tubes. There is less latent heat removed from the, steam, and therefore less heat transfered to the feedwater resulting in a lower exit temperature for the feedwater. (0 55)

REFERENCE General Electric " Thermodynamics, Heat Transfer, and Fluid Flow", Chapter 8. !

End of Section 5 CONTINUE NEXT PAGE

- k y

e.

N.

h T

i J

f, .

s

.I

'3

, ( SECTION 6 ff" PLANT SYSTEM DESIGN, CONTROL AND INSTRUMENTATION

OUESTION 6.01 Concerning the Feed System and Safety Injection System:

a. Wh'at pump and valve changes will automatically odcur in the Feed and Condensate system when, at full'uniy load, a Safety Injection signal is received ?

- / (2.0)

6. What specific valve interlock prevents the injection of Feed and Condensate water into the RCS 7 (0.5)

c. If during the SIS sequence the SI pump suction isolation valve (HV-853 A&D) fails to open, will the feedwater pump run ? Explain ! (0.5)

ANSWER .

a. - Main feed pump trips and then restarts (0.5)

- Condensate pumps trip (0.2G)

- Heater Drain pumps trip (0.25)

- MFP suction valves (HV-854 A&B) shut (0.25)

- MFF discharge valves (HV-852 A&b) shut (0.25)

- Feedflow control valves (FCV-456/7/B) shut (0.25)

- Feedflow Aux control valves (CV-142/3/4) Shut (0.25)

- (SIS' Suction Header Isolation Valves (HV-851

--~~ss A&B1'will open.) (N/A) b; MFF(Suction valves (HV-854 A&B) must be shut before the SIS header isolation valves (HV-851 A&B) 4 will open. (0.5)

c. Yes,0the FW pump will run for 30 sec. then if HV-853

. does not open, the pump will trip. (0.5) reference: SD-SO1-580, pp.16, 501 e: am bank.

t > $g / j'$0e Un t/D 1

4 1 .

+,

e

QUEGTION 6.02 Concerning the CVCS system: .

a. Under what three conditions would the Excess Letdown System be Used 7 (1.5) b." Describe the three Excess Letdown Alignments. (1.5)

ANSWER

- - a.1. inoperability of normal letdown (0.5)

2. used during PZR Steam Bubble formation (0.5)

3. To increase the cooldown rate during the initial cooldown phase of RHR operation. (0,5) b.1. Normal - thru seal water return pipi~n'g into the VCT. (0.5)

2. Thru a section of normal letdown piping and a residual heat exchanger to the VCT. (0.5)

3. Thru and then directly to the RCS drain tank. (0.5)

REFERENCE:

SD-SO1-310 pp.5, fig. I-1.

QUESTION 6.03 What is the purpose of the Cryogentic Waste Gas Treatment System? (1.0)

ANSWER To Cryogentically ( low temp. & Charcoal ) remove and store Xe and Kr from the Radioactive waste gas prior to release.

(1.0)

REFERENCE SD-SO1-530

QUESTION 6.04 Which of the following is true concerning the Source Range Channel high voltage cutoff? (1.0)

a. During a reactor startup either IR channel increasing above the P-6 setpoint will turn off the high voltage.

b. If'one IR channel fails low while at power, both Source Range high volatage will be reenergized.

. - c. If a IR channel increases above 2 X 10 E(-9), it will turn off its respective source range's high voltage.

d. During a reactor shutdown either IR channel decreasing below the P-6 setpoint will turn on the high' voltage.,

ANSWER

c. (1.0)

REFERENCE .

SD-SO1-380 Excore NI's

QUESTION 6.05

a. List the sources of air that can be use'd to supply the Instrument Air Header? (1.5)

b. How are the following valves affected by the loss of Instrument air (operable / inoperable) 7 If operable, state why? If inoperable, state status: (f ailed shut, failed as is, or failed open). (1.5)

1. Hain feedwater flow control valves

2. Emergency Auxilary feedwater regulators

3. Safety injection valves HV851,852,853,854

ANSWER a.1. Three Service Air Compressors

/ (0.5)

2. Auxilary Air Compressor (0.5)

3. Diesel Air Compressor (0.5) b.1. Inoperable (0.25) - Failed open (0.25)

2. Operable (0.25) - Nitrogen Backup (0.25)

3. Inoperable (0.25) - alarmed in the controlroom, Operable if operator selects Nitrogen backup (0.25)

Reference:

SD-Sol- 420 & 580 ; SO1-2.4-2 IA Sys Mali' unction

. e ---r- - , ,

w g m s-p -

c-

2dESTION h.06

a. What are the control signal inputs to the Automatic Rod Control System? (1.0)

b. For each of the above, State if the input controls

^ rod speed and/or rod direction. (1.5)

ANSWER a.1. T avg (0.25) b.1. Direction and speed (0.5)

2. T ref (0.25) 2. Direction and speed (0.5)

.. -- 3. Nuclear Flux (0.25) 3. Speed (0.25)

4. P-P ref (0.25) 4. Speed (0.25)

REFERENCE SD-SO1-400 - '

QUESTION 6.07 Pressurizer Pressure Transmitter PT-430 fails low during' normal at power operations.

a. Name one alarm that you expect to annunciate in the control room as a direct result of'the PT failure ?

(0.5)

b. What oressuri er s'ystems or controls will be affected and how will they respond? (1.0)

ANSWER

a. Safety Injection Train A Channel 1 Alarm or VLPT Channel 1 Alarm (0.5)

b. 1. Spray Valves will close if they are open (0.5)

2. All heaters in Automatic will turn on (0.5)

Reference Requal Exam Bank End of Section 6 CONTINUE 14 EXT PAGE **

< - ---r-m e g- -y-.

SECTION 7 PROCEDURES: NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL CONTROL

OUESTION 7.01 SONGS 1-Emergency Procedures utilize a "

Critical Safety. Function -

Status Tree" (CSFST) concept.

a. List the six (6) CSFST's in order of priority and briefly describe what parameters are checked for each.

(3.0)

ANSWER

a. Priority Tyge Parameter 1 Subcriticality Checks power' levels on the PR, IR and SR 2 Core Cooling Checks core exit TC's and Hot Leg RTD's 3 RCS Integrity Checks cooldown rates and P/T limits g{o (#,d Mhe 4 Heat-Sink Checks , primary and secondary heqt removal prameters 5 Containment Checks Containment pressure, sump level,and radiation levels.

6 ,

RCS Inventory Checks Pressurizer level, priority: (0. lea.) Type: (0.3 ea.) parameters: (0.1 ea.)

REFERENCE S01-1,0-1

OUESTION 7.02 TRUE or FALSE:

On your first pass through the CSFST's, you note an " orange" priority 3 condition and a " red" priority 5 condition. In Accordance with S01-1.0-1, you would address the " orange" priority 3, first, because it is of higher priority.

TRL'E c ~ cA'.cr? (1.0)

ANSWER FALSE (1.0)

REFERENCE SO1-1.0-1

QUESTION 7".03 j Per 501-14-17, What are the three of four actions that should be taken when a gagging device is installed or a relief or safety valve 7 (3.0)

ANSWEB ,

1

1. Document the valve gagged. (Safety Related valve or Component Control Form f or Saf ety Related Valves, CO's Log, Procedural Steps, Etc.)

2. Initiate a maintenance order, if the gag as installed due valve failure, and hang a defiency tag on the gagged valve.

3. Evaluate the affected system or component for an operability assessment.

4. Notify engineering for a techinical evaluation to determine if continued operation is acceptable.

( 3 of 4 required, i.O ea.)

REFERENCE SO1-14-17

OUESTION 7.04 (3.0)

Fill in the following blanks in accordance with S01-3-1:

a. Maximun Delta T between the Pressurizer liquid and the Reactor Coolant is 7 '(0.5)

6. Fully withdrawn position of all control rods shall be _

_ ____7 (0.5)

c. RCS Pressure and Temperature should not exceed:

_______psig (0.5) and ___

_____ degrees F (0.5) l when RHR system is in service.

d. Maximum heatup rates are _______ for the RCS, and (0.5)

_____ for the Pressuri:cr. (0.5)

ANSWER go

a. degrees F (0.5) -

b. 318 (0.5)

c. 400 (0.5), 400 (0.5)

d. 60 deg./hr, (0.5) 45 deg./hr (O.D)

QQ

- e.

Reference

QUESTION '7.05 The plant is being started up from Hot Standby to Minimum Load.

a. Wnat rod position are procedurally required for

- criticality unless otherwised directed by the Shift Superintendent ? (0.5)

b. When, according to the procedure, should critically be anticipated ? (0.5)

. - /

c. What is the limit on Startup Rate duringthe startup?

(0.5)

d. What is the limit on turbine backpressure prior to rolling the turbine ? . (0.5)

ANSWER

a. 0;mtda-o Look. 1 oud 2 crd ra-tr:1 CacJ 1 .t 310 SLep=.

Control Bank 2 at 100 steps (0.5)

b. At any time when control rods are being withdrawn, cr when baron dilution is in progress. (0.5)

c. less than 1.0 DPM (0.5)

d. 5.5' Hg (0.5)

REFERENCE SD1-3-2

QUESTION 7.06 In accordance with S01-3-1, Plant Startup from cold S/D to Hot STBY, a steam bubble being formed in the pressurizer is indicated when 7 (1.0)

ANSWER

a. Charging flow is less than letdown flow with RCS pressure constant (1.0)

-

Reference SO1-3-f l

QUESTION '7.07-Answer the following concerning Cold Leg Injection and Recirculation:

a. At.what RWST level is it that the Cold Leg Injection

-and Recirculation Procedure is entered? (C.5)

b. ' How long' after reaching the RWST level is it that the operator has to reset SI 7 (0.5)

,, .. c. At what RWST level must the cold leg recirculation path be completed by, (0.5) and what are the consequences if the alignment is not completed by this level 7 (0.5)

ANSWER -

a. 21 %. (0.5)

b. 30 seconds (0.5)

c. 7% (0.5)

At 7 % level adequated NPSH to the charging pumps cannot be assured (0.5) possible formation of vortices could result and air entrainment of the pumps.)

REFERENCE S01-1.0-23 End of Section 7 CONTINUED ON NEXT PAGE

~

. s a

Section 8 Administrative Procedures, Conditions, and Limitations

OUESTION 8.01 TRUE or FALSE.

Technical Specification 3.5.3 does not require any action if one control rod is immovable, provided the immovable rod is within +

or - 35 steps of'it's group step counter demand position. (1.0)

ANSWER False. (1.0) ,

REFERENCE Technical Specification 3.5.3

OUESTION 8.02 During MODE 3: .

a. How many Reactor Coolant Loops must be in operation? (0.5) i

b. How many Reactor Coolant Loops must,be opera'ble? (0,5)

ANSWER

a. 1 (0.5)

b. 2 (0.5)

REFERENCE Technical, Specification 3.1.2

/

,w + -, -

.-,,y - , --e. + c- g --- rw . w -e---e- -

OUESTION 8.D3 Each RPS Pressurizer Pressure channel is required by Technic-1 Specifications to undergo a channel check on a sh'iftly basis (at least once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ). Some extensions of the basic interval are allowed by the Technical Specifications.

Records.show that this was done on:

November-27 at 0000 November 27 at 1500 November 28 at 0000 November 28 to 1400 November 29 at 0500

a. What is the maximum allowable interval between channel check surveillances? (1.0)

b. When is the next channel check surveillance due? $.0)

ANSWER f.

sv'[f ge.c g, g X i N' I

a. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months + 25 percent = 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months (1.0)

b. 1. Last done + 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months = Nov 29 at 2000

2. Last 3 + 3.25.x 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months =

Nov 28 G 0000 + 39 hours4.513889e-4 days 0.0108 hours 6.448413e-5 weeks 1.48395e-5 months = Nov 29 @ 1500 Therefore IT must be done no later than: Nov 29 0 1500 (1.0) .

REFERENCE Technical Specifications 4.02

QUESTION 8.04 When a fire suppression Spray / Sprinkler System is declared inoperable for a portion that protects an area containing redundant safety-related equipment, the required action is to:

(pick one) (1.0)

a. commence a Unit shutdown within one hour.

b. establish an hourly fire patrol for the affected area.

c. establish a continous fire watch with backup fire suppression equipment in the affected area within one hour. -

d. establish a backup suppression system in one hour or be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ANSWER

c. (1.0)

REFERENCE Technical Specification 3.14

k

- e 4

QUESTION 8.05 Station Order SO123-0-14 Attachment 4 contains certain Operations Reporting Requirements: ;

a. State three conditions of which the. Unit i Superintendent 4

(or his designee) and the on-duty Shift Technical Adviser must be notified by the Shift Supervisor as soon as practi- .

cal. (1.5)

,

ANSWER

1. An unplanned trip or load reduction. ,

2. An unplanned or unexplained major system or component failure.

3. Any significant event which cannot be corrected by oper-l -

ators, or any abnormal operating event, which, in the a judgement of the Shift supervisor,. warrants notification of the Unit 1 superintendent and/or on-duty sh'ift Technical Advisor.

-4 . Any event that would be reported as a Reportable Occurance.

5. Any item listed in Attachment 4.

(Any three (0.5) each.)

REFERENCE $6/23~ '

001 0 ! ^O p :- . P!. 2 , 2d 7C J"_ ' . '

QUESTION 8.06 i The plant is in MODE 4, preparing for a routine p'lant startup. As

a result of routine surveillance, it is determined that only one Diesel Generator Fuel Transfer pun.p is operable. The Maintenance Supervisor has told you, that the cause of the failure is known, parts are on site, and that repairs and operability checks will be completed in less than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . Explain why the startup can

or cannot proceed. (See enclosed Technical Specifications.)

(1.5)

~

i

ANSWER

The startup cannot proceed, (0.5) because " entry into an l Operating Mode is not allowed if an action statement must be relied upon to do so. (1.0) l (The plant can remain in MODE 4, but cannot enter MODE 3.)

REFERENCE Technical Specification 3.0.4 and Technical Specification 3.7 i .

i e

l

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

l .

3. 7 AUKILIARY ELECTRICAL SUPPLY

i APPLICABILITY: Applies to the availability of electrical power for the operation of the plant auxiliaries. .

OBJECTIVE: To , define those conditions of electrical power availabil- l ity necessary (1) to provide for safe reactor operation, )

(2) to provide for the continuing availability of -

)

engindered safeguards, and (3) to ensure that the station 72 can be maintained in the shutdown or refueling condition 5/3/83 for extended time periods.

~

SPECIFICATION: I. In Modes 1, 2, 3 and 4 the following specifications 84 shall apply:

11/14/84 A. As a sinf=== the foIlowing shall be OPERABLE:

1. One Southern California Edison Company and one 84 San Diego Cas & Electric Company high voltage 11/14/84 transmission line to the switchyard and two transmission circuits from the switchyard, one immediate and one delayed access, to the

onsite safety-related distribution system. '

This configuration constitutes the two required offsite circuits.

2. Two separate and independent diesel generators each with:

a. A separata day tank containing a diniana of 290 gallons of fuel,

b. A separate fuel storage systes containing 34 a minimum of 37,500 gallons of fuel, and 4/1/77

c. A separate fuel transfer pump.

' 3. AC Dietribution

a. 4160 Volt Bus 1C and 2C,

b. 480 Volt Bus No. 1, Bus No. 2 and Bus No.

3, and

c. Vital Bus 1, 2, 3, 3A, 4, 5 and 6.

j 4. DC Bus No. I and DC Bus No. 2 (including at 1 east one full capacity charger and battery 84 ,

, supply per bus). 11/14/84

5. The two Safety Injection System Load Sequencers.*

The automatic load function may be blocked in Mode 3 at a pressure ,$

1900 psig.

3-80 Revised:

. /A. O>'

11/28/84

B. Action 84 11/14/84

1. With one of the required offsite circuits C inoperable, demonstrate the OPERABILITY of the

, remaining AC sources by performing Periodie 34 Testing Requirements A and B.I.a of Technical 4/1/77

Specification 4 4 within one hour and at Isaat once per eight (8) hours thereaf ter; restore an additional offsite circuit to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in COLD SHUID0iiN l within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . !

2. IfonedieselgeneratorisdeclaredinopeN 84 able, demonstrate the OPERABILITY of the two 11/14/84 l pffsite transmission circuits and the !

/ remaining diesel generator by performing l Periodic Testing Requirements A and B.I.a of 34 Technical Specification 4.4 within one hour 4/1/77 and at least once per eight (8) hours there-after; restore the inoperable diese1' generator to service within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in COLD SHUfDGiN within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ,

3. With one offsite circuit and one diesel gener- 84 ator of the above required AC electrical power 11/14/84 sources inoperable, demonstrate the OPERABILITY of the remaining AC sources by performing Periodic Testing Requirements A and 34 B.I.a of Technical Specification 4 4 within 4/1/77 one hour and at least once per eight (8) hours thereaf ter; restore at least one of the inoperable sources to OPERABLE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months o'r be in COLD SHUID0iiN within the j next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . Have at least two offsite j circuits and two diesel generators OPERABLE .

i within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from the time of initial loss or be in COLD SHUIDoliN within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

4. With two required offsite circuits inoperable, 84 demonstrate the OPERABILITY of two diesel 11/14/84 generators by performing Periodic Testing Requirement B.1.s of Technical Specification 4.4 within one hour and at least once per eight (8) hours thereafter, unless the diesel generators are already operating; restore at 34 l

least one of the inoperable sources to 4/1/77 j OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at '

j least HOT STANDBY within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

! With only one of the required offsite circuits l restored, restore the remaining offsite circuit to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 84 from the time of initial loss or be in COLD 11/14/84 SHUID0liN within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

, ..f r ,

..Y l

3-81 Revised: 11/28/84

sg

5. With two of the above required diesel I

(

generators inoperable, demonstrate the OPERABILITY of two offsite circuits by performing Periodic Testing Requirement A of l

I Technical Specification 4 4 within one hour and at least once per two (2) hours there ~ !

af ter; restore at least one of the inoperable diesel generators to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in COLD SEUIDOWN within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . Restore both diesel generators to 0PERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from time of initial loss or be in COLD SEUIDOWN within the

. next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

,6. With less than the above complement of AC buses OPERABLE, restore the inoperabia bus within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in COLD SEUIDOWN within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . '

7. With one required DC bus inoperable, restore the inoperable bus to OPERABLE status within 2 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months or be in COLD SHUIDOWN within the next 11/14/84 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

8. With a required DC bus battery and both of its chargers inoperable, restore the inoperable battery and one of its chargers to operable C status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in cold shutdown within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

9. With one Safety Injection Load Sequencer inoperable, restore the inoperable sequencer to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in COLD SHUIDOWN within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

II. Additionally, in Modes 1, 2 and 3 the following specifications shall apply:

A. As a minimum, the following shall be OPERABLE:

1. The MOV850C Uninterruptable Power Supply (UPS).

B. Action

1. With the MOV850C UPS inoperable, restore the '

UPS to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUIDOWN within the l

l following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . *-

/ -

3-82

./ d )o.

i Revised: 11/28/8h

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

s *.

0UESTION 8.07 Provide the minimum number of individuals required by Technical Specifications for the following pcsitions to operate the plant at full power (mode 1):

a. - =--

Senior Operating Licenses (SRO)

b. -i Operating Licenses (RO)

c. ------

Non-licensed Persons (NPEO)

d. ------

Shift Technical Advisors (STA)

~

ANSWER

a. 2 (0.5)

b. 2 (0.5)

c. 2 (0.5)

d. 1 (0.5)

REFERENCE Table 6.2.1 and Technical Specifications 6.2.2

~* QUESTION G.08 Deviations from a procedure during normal plant operations:

a. are not allowed.

b. can be made if the original intent of the pr,ocedure i s satisified. .

c. can be made with verbal approval of two licensed operators.

(one of which is licensed as a SRO).

d. can be made only in an emergency when immediate action is needed to protect the public health and safety.

b. (1.0) A

WI O 'T

For normal procedure B would be more correct.

REFERENCE Technical Specification 6.8.1

QUESTION 8.09 TRIP / TRANSIENT REVIEWS per SO123-0-25 are performed by:

a. Any personnel licensed by the NRC.

b.- Shift Supervisor and the Shift Techncial Advisor.

c. Shift Technical Advisor.

d. Shift Techncial Advisor, only if they -hold a valid SRO 1icense. (1.O)

ANSWER

b. (1.0)

REFERENCE SO123-0-25 page 3.

QUESTION 8.10 EPIP SO123-VIII-10 defines the Emergency Coordinator as the person who is designated to tako charge of all emergency control measures, and has ultimate autho;-ity over all onsite activities and personnel.

a. Who initially fills this role in the early stages of a event? (0.5)

b. Who relieves this first person and has the primary responsibility for this rolo? (0.5)

ANSWER

a. the Shift Superintendent (0.5)

b. the first Recall EC (0.5)

REFERENCE SO123-VIII-10

I 1

OUESTION 8.11 Which of the following is NOT required of the A.C. and D.C.

Electrical power sources by Technical Specifications in mode 5 and 67 (1.0)

a. Two circuits-between the offsite transmission network and onsite ESF Electrical System.

b. One operable Diesel generator set.

c. One energized and operable 125-volt D.C. bus aligned to its associated charger.

d. One energized and operable 480-volt AC bus.

ANSWER

a. (1.0)

REFERENCE Technical Specification 3.7

QUESTION 8.12 If control power is last to a pressurier power operated relief valve (PORV) while in Mode 1, to continue to operate:

a. no action is required by Technical Specifications provided another PORV is operable and all pressurizer ccde safoty valves are operable,

b. Technical Specifications require the associated block valve to be verified open and then its power supply to be removed, if the PORV is not made operable within one hour.

c. Technical Specifications require the associated block valve to be shut and maintained in the closed position if the PORV is not made operable within one hour,

d. Technical Specification requires action to be initiated within one hour to place the plant in at least HOT STANDBY within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , if the PORV is not made operable.

ANSWER

c. (1.0)

REFERENCE Technical Specification 3.1.5 END OF SECTION 8 END OF THE EXAM o

+*********BE SURE TO SION THE FRONT COVER SHEET ***********

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