ML20137M866

Exam Rept 50-206/OL-85-02 on 850709-10.Exam Results:Two Reactor Operator Candidates Passed Written & Operating Exams.One Candidate Failed Written Exam & Received Waiver of Operating Exam

ML20137M866
Person / Time
Site:	San Onofre
Issue date:	08/23/1985
From:	Johnston G, Miller L, Pate R, Shiraki C NRC
To:
Shared Package
ML20137M855	List: ML20137M851 ML20137M866
References
50-206-OL-85-02, 50-206-OL-85-2, 50-20601-85-2, NUDOCS 8509130402
Download: ML20137M866 (61)

Text

_

Examination Report No. 50-206/0L-85-02 Facility Name:

San Onofre Nuclear Generating Station, Unit 1 Docket No.

50-206 Examinations Administered at: San Onofre NGS, San Clemente, California, from July 9-10, 1985 Chlef Examiner:

Iset

& 2 3-d[

W..Johnston, Operator License Examiner Date_ Signed Examiners:

M 7-L} "b

.~ W.

iraki, Op ator License Examiner Date Signed 0

m v m -ss

[4. F. Miller, Chief, Reactor Projects Date Signed Section Approved by:

k dd~

. J. Pate, Chief, Operations Section Date Signed Summary:

Examinations on July 9-10, 1985

~

~ Written examinations were administered to three Reactor 0perator candidates.

Two operating examinations were administered to two of those candidates, all of whom were retakes, one receiving a. waiver of the operating examination.

Two candidates passed the written and operating examinations. One candidate failed the written examinativ rM had received a waiver of the operating examination.

8509130402 B W ADOCK O5000 06 PDR PDR G

A

7 DETAILS 1.

Persons Examined R0 Candidates:

1 Three candidates were examined. Two taking written and operating examinations, the other taking a written examination only with a waiver of the operating examination.

2.

Examiners

• C, W. Johnston, RV C. W. Shiraki, RV L. F. Miller, RV

• Lead Examiner 3.

Persons Attending the Exit Meeting Southern California Edison:

J. R. Tate, Assistant Manager, Operations J. J. Wambold, Training Manager R. J. Mette Supervisor of Operations Training M. J. Kirby, Nuclear Training Administrator NRC G. W. Johnston C. W. Shiraki 4.

Written Examination and Facility Review Written exams were administered as follows:

Three Reactor Operator - July 9, 1985 At the conclusion of the exam the facility staff reviewed the exam and answer key. The facility staff comments are addressed in the enclosed attachment (1). These comments were discussed with the facility staff and appropriate' revisions to the master examination key were made by the lead examiner prior to grading the candidates' responses.

5.

Operating Examinations Facility walkthrough oral examinations were conducted July 10, 1985. No particular generic weaknesses were identified by the examiners during the course of the examinations. The sample size (three), and the fact that the candidates are retakes are factors in this evaluation.

6.

Exit Meeting On July 11, 1985, the NRC representatives met with licensee personnel.

Those individual candidates whoiclearly passed the oral examinations were identified. The NRC representatives discussed the overall performance of the candidates.

]

RESOLUTION OF FACILITY COMMENTS REACTOR OPERATORS EXAMINATION 1.0 Facility Comment: Question 1.03 (a) The reviewer stated that the control group insertion limits were to maintain the reactor suberitical following a reactor trip.

Resolution:

(a) Comment accepted.

2.0 Facility Comment: Question 1.04 (c) The reviewer stated that axial offset would become more positive rather than negative.

Resolution:

(c) Comment accepted.

3.0 Facility Comment: Question 1.06 (c) The reviewer stated that the condition' would he more accurately described as saturated rather than as subcooled.

Resolution:

(c) Comment Accepted.

4.0 Facility Comment: Question 1.08 (b) The reviewer stated that the present Technical Specification limits on cooldown rate for the. pressurizer is 200 deg. F per hour.

Resolution:

(b) The author used the copy in place'in Region V.

Apparently this copy has not been updated. Comment will be accepted.

5.0 Facility Comment: Question 2.06 (b) The reviewer indicated that the reference used did not conform to the mode of operation that the switch position is used for. This is used primarily for heatup after condenser vacuum is established.

(c) Again, the reviewer indicated that this is the position normally used to allow heatup of the plant prior to establishing vacuum.

(d) This position is not normally used for normal heatup as was indicated by the reference, it would only be employed in unusual circumstances.

4 7

g Resolution:

(b) Agreed. The key will be changed to reflect ehe procedures.

s (c) Agreed. T1.e key will be change.d to reflect the procedhres.

(d) Forthiscasetheexaminerwillacce[tthe'explanationthat this position is not normally used.

"~

6.0 Facility Comment: Question 2.09 s

(b) " Discusses reset and block, which is not the question.

See Sol - 1.0 - 12 steps 1 and 2 for example of reset and terminate. Method of termination may vary depending on event."

Resolution:

(b) It appears to the examiner that the term " block" in the key has lead to confusion. The word will be changed to " reset" in the key. This should eliminate the confusion and codform to the procedure Sol - 1.0 - 12.

7.0 Facility Comment: Question 3.05

~s (c) The reviewer pointed out that the power range instruments utilize the intermediate range instrument's compensated ionisation chambers for some channels.

It was also pointed out that the portion of tha key for this question covered the toisic of compensation that does not relate to the operation of the chamber.

Resolution:

(c) The examiner will accept koth comments, the key will be adjusted

~'

for the point value.

8.0 Facility Comment: Question 3.06 (b) The reviewer commented that the term " low differential preenure" is not commons 11y used. The term typically used at the facilf.ty is "high reverse delta P".

Resolution:

(b) Comment accepted, will add "or high reverse delta P" to the key.

9.0 Facility Comment: Question 3.08

'(a) "Should be ' Shutdown Rods Not Withdrawn'."

Res'olution:

(a) The examiner recognizes that the. term ' Shutdown bank deviation alarm' is not in the plant vernacular. Since this is so, the examiner will accept as an alarm function either case.

4

' g s, 7

5

3 10.0 Facility Comment: Question 3.09 (a) For part 5 the reviewer felt that the' feedback from Bistable TC-413B was appropriate to the question.

(b) Part I was also felt by the reviewer to indicate greater specificity than was appropriate as far as the Bistable TC-413B.

In part 2 the reviewer felt the portion referring to the speed signal being fed through an auto / manual group selector switch to be more specific than required by the question.

Resolution:

For both (a) and (b) the comments are accepted.

11.0 Facility Comment: Question 3.10 (b) "The channel has no present control functions."

(c) The reviewer here again was concerned about the level of specificity of the' question concerning the energization of a solenoid valve.

Resolution:

(b) The examiner will accept the answer of no control function.

(c) Agreed, will change key.

12.0 Facility Comment: Question 3.11 (a) The reviewers concetu here was that the candidate may list each permissive and its purpose.

(b) For part 3 the reviewer stated that in this case the portion of the question pertaining to a decrease in power of 5% or greater being detected in any power range channel was not germain to the question.

Resolution:

(a) The examiner would concede that if a candidate answered in that fashion he would be answering the question appropriately, therefore will accept such a response.

(b) Agreed, will strike that portion.

13.0 Facility Comment: Question 4.01 (b) "For a pressure controller Shift Superintendent cannot approve exception."

Resolution:

(b) Agreed, will change key.

4 14.0 Facility Comment: Question 4.03 (a) The reviewer noted a typographical error.

Resolution:

(a) That will be corrected.

15.0 Facility Comment: Question 4.04 For all of the parts the reviewers felt the answers the candidates might give could be other events that would correspond to the symptoms listed.

Resolution:

The examiner agrees and will consider other events as answers so long as they correspond to the symptoms.

16.0 Facility Comment: Question 4.05 (b) The reviewer noted that there are other systems that are required by the T. S.

Resolution:

(b) If the candidate indicates that something in the T. S. would require the shutdown that will be sufficient as an answer.

a k

1

'}'s:-

i I

U.S. Nuclear Regulatory Commission Reactor Operator License Examination Facility:

SAN ONOFRE UNIT I Reactor Type:

vo<tinchoo.5 Date Administered:

Julv 4 10R5 Examiner:

c. v anknornn 4

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 least 70% in each category and a final grade of at least 80%.

Examination papers i

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

~~

% of Category - % of Candidate's Category Value Total Score Value Category i

25.0 25.0 1.

Principles of Nuclear Power Plant Operation, Thermodynamics, Heat Transfer and Fluid Flow

?s.n

?;.0 2.

Plant Design Including Safety and i

Emergency Systems

?S.n 2s.0 3.

Instruments and Controls 25.0

?5.n 4.

Procedures - Normal, Abnormal, Emergency, and Radiological Control 100.0 TOTALS l

Final Grade j

All work done on this examination is my osn.

I have neither given nor received aid.

Candidate's Signature t

i i

EQUATIO:t SHEET f = ma v = s/t Cycle efficiency = (Network out)/(Energy in) 2 w = mg s = y,t + 1/2 at 2

E = mc KE = 1/2 av a = (V7 - y,)/t A = AN A=Ae

~

~

g PE = agh V7 = y, + a t w = e/t i = an2/t1/2 = 0.693/t1/2 1/2'II*EIt1D)It)3 t

h

((t1/2)

• I*0)3 AE = 931 am I = I,e Q = aCpat Q = UA4 T I = I e~"*

g I = I,10-x/TVL l

Pwr = W an y

~

TVL = t.3/u P = P,10 "#I*)

8 HVL = -0.693/9 P = P e /T t

o SUR = 26.06/T SCR = S/(1 - K,77)'

l CR, = S/(1 - K,77x)

SUR = 26s/t* + (8 - p)T CR)(1 - Kd f1) = CR I1 ~ keff2) 2 T=(t*/p)+[(s-p)[1p]

M'= 1/(1 - Kdf)' = CR /CR, g

T = 1/(p - s)

M = (1 - Kd fo)/II ~ Edf1)

T = (a - p)/(19)

SOM = (1 - Kdf)/Edf p = (Kgg-l)/Ke f f = AKeff eff 1* = 10 seconds

/K T = 0.1 seconds-I p = [(t*/(T Kg f)] + [s,77 (1 + AT)]

/

Idgg=1d P = (r+V)/(3'x 1010) 2,2 2 Id gd jj 22 2

I = oN R/hr = (0.5 CE)/d (meters)

Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lbm.

1 curie = 3.7 x 1010dps 1 ga;. = 3.78 liters I kg = 2.21 lbm 1 f t- = 7.48 gal.

1 hp = 2.54 x 103 Btu /hr Density = 62.4 lbm/ft3 1 m = 3.41 x 106 8tu/hr Density = 1 gm/cm3 lin = 2.54 cm Heat of vaporization = 970 8tu/lom

• F = 9/5'C + 32 -

Heat of fusion = 144 8tu/lbm

• C = S/9 ( *F-32) 1 Atm = 14.7 Psi = 29.9 in. Hg.

(

SECTION ONE REACTOR OPERATOR EXAMINATION 1.01 (2.5)

A refueling is being conducted following removal of msny fuel assemblies to locate a f oreign obj ect in the bottom of the vessel.

As the fuel assemblies are replaced, the neutron count rate increases as shown in Fig.

1.

(a)

Define "M"

in Figur e 1.

(0.53 (b)

What is the significance of the point wher = 1/M =

(0.5)

O?

(c) ~Give one reasonable explanation for the concave (0.5) downward shape of the plot.

(d)

The count rate after loading one fuel assembly is (0.5) observed to increase from 100 cps to 125 cps.

If the final keff was

.99, what was the keff prior to loading the asserabl/? (Show your work)

(e)

At what count rate can the reactor go critical, in (0.5) theory?

(a)

Multiplication ratio (fractional neutron increase per generation)

(b)

Estimated fuel loading for criticality (c)

A larger percentage of neutrons are being detected as more fuel is loaded nearer the detector, or the detector was located close to an installed source assembly (either acceptable, may be others).

(d) keff prior to load = 0.9075 using Cl(1-keff1)

=

C2(1-keff2)

(e)

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1.02 (3.0)

Refer to the attached Figure 2 (a)

The net amount of Xe-135 in the core is dependent on its rates of formation and removal.

State the

(.g,,}

four most important processes which produce these rates for Xe-135.

(b)

The reactor trips after steady state operation at BOL from 75% power at 1200.

When will the core be (g,g}

least reactive,- assuming temperature and control rods remain unchanged after the trip?

,c)

How much reactivity will be added (or subtracted)

(

due to Xe-135 at this time, compared to the time

,g)

' of the trip?

(a) til production from fission (0.5),

(2) decay from I-135

(

and Te-135)

(0.5),

13) decay to Cs-135 (0.5),

and (4) neutron absorption (or burnout)

(0.51 1700 (c.5)

(b)

(c)

About

.28%

reactivitty is subtrac at 1700 trip.(0.5(ed relative to the time of the

/

Ref:

Academic program for fluclear Power Personnel, Vol II, Chap.

41 Figure A-13,

" Curve Book-w-

/

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~

Frav G 2

I Reactiety Insertion Due to Xenon at BOL vs Time Following Plant Trip After Steady State Operation at Various Power Levels o

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1.03 (2.5)

(a)

State three reasons for control group insertion (1.5) limits.

(b)

State one reason why flux redistribution Sollowing (0.5) a reactor trip from full power to hot zero power conditions may cause reactivity to increase.

(c)

Explein why the Technical Specifications limit the (0.5) average burnup of tiie core to 21,000 MWD /MTU.

fa)

(1)

To maintain an acceptable power distribution during normal operation (0,5),

(2) Limit effect of a

rod ej ection accident (0.5),

and (3)

Haintain reactor subcritical_durin9 design accident (steam break)

(0.5).

(b)

At zero power, the underator in the top of the core is relatively cooler than at full power (0.25), so with negative NTC (0.25), reactivity is higher.

(c)

Reactivity addition accident analyses assume a

maximum negative value of MTC.

This value may be exceeded (0.25) if the burnup specification is exceeded because MTC beccmen more negative with increasing burnup (0.25). 01 Ref: Tech. Spec.

3.9, 5.5.2,

" Curve Look" pp. 44-45, 66-67 f

e L

6

1.04 (2.75)

(a) Do the Technical Specifications allow operation at (1.0) full power with control bank 2 eighty per cent withdrawn for a month in the middle of the core cycle? Explain your answer.

(b) How will axial offset change immediately after this (1.0) bank is rapidly borated out while maintaining power constant?

Explain your choice.

(c) Four hours after the rod motion, what changes, if (0.75)

any, do you expect to have occurred to a x i.a !

offset?

Explain your answer.

(al No.

(0.5) Technical Specifications (3.5.2) require that the average position of CB 2 be at least 907.

withdrawn (0.25) after 20% of the core cycle (0.25)

(b) Axial offset should Lecome less negative or,

perhaps, s!!9htly positive (0.34),

because the power distribution will shift from a peak in the lower two thirds of the core to a more symmetrical distribution (0.00) and because axial offset is (PTCP - PBOTTCM)/(PTOP + PDOTTOM). (0.33)

(c) Axial offset should become more negative (0.25) because xenon will build up in the reduced flux in the lower two thirds of the core (.25), and burnout in the increased flux in the upper one third of the core (.25) during this time period.

Ref: " Curve Book", TS 2.0.2 e--

O 33a -

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S 0;-

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1 1.05 12.5)

(a)

The plant is operating at equitabrium steady state (1.0) zero power conditions, all rods out, 2934 MWD /MTU when an alert operator notices that TAVE has increased 5 F in the last twenty four hours.

Calculate, using Table 1, what change in boron concentration could have caused this increase.

Show your worit.

(b)

Considering each case independently, does (1.5) differential boron worth increase or decrease as (1)

Baron concentration increases, (2)

C on t r.o !

rods are inserted, and (0) Moderator temperature increases.

tal Change in Doron Conc.

=

NTC X

Change in Temperature X

Boron Worth (0.5),

so Foron concentration must have decreased by 7.0 X 5

X O. * *. = 52,b ppm (0.5)

(b) til Differential boron worth decreases (less negative) as baron concentration increases (0.53 (2)

Differential boron worth decreases as control rods are inserte1 10.5)

(3)

Differential baron worth decreases as moderator temperature increases (0.5)

Ref Academic Prog, for Nucl. Plant Personnel, p.

4-119, West.

Nuclear Trn9 Ctr.

Reactor Theory Notes, Rev. !!!,

p.

65 p

9

~

o

l 1

TA8LE L

END-POINT BORON CONCENTRATIONS, BORON WORTH, AND MODERATOR TEMPERATURE COEFFICIENTS

~

(No Redistribution Effects)

IIIP 2934 Pl.;D/NIU Moderator Tempera!ure Rod Ccndition g fppm)

Boron Worth foam /1000 ccm)

_ Coefficient (pce/'F) l G

APO 1355 150.35 7.0

-]-

Co., trol Group 2 In 991 147.43 12.8 h?

Control Groups 2+1 In 747 144.13 17.2 f5,-

i =

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1.06 (1.5)

(a)

The reactor is shutdown, and pressurizer pressure (0,5) is 1425 psig. At what temperature will the reactor coolant system be 35 F subcooled, according to the Steam Tables?

(b)

A primary to containment atmosphere steam leak is (0.5) occurring from the pressurizer with the pressurizer at 1600 psig.

If containment pressure is 14.7

psia, what kind of steam (saturated, subcooled.

or superheated) should be expected as soon as the steam from the break has depressurized to containment pressure.

(c)

For a

leak identical to that in (b)

above, if (0.5) containment pressure is 50 psia, what kind of steam (saturated, subcooled, or superheated) should be expected as soon as the steam from the break has depressurized to containment pressure?

(a) 555 F (b)

At 14.7 psia, the steam will be superheated (0,5),

(c)

At 50 psia, it will be subcuoted (0.5).

Ref: Steam Tables, fiollier Diagram e

S u

~

1.07 (2.5)

(a)

Define not positive suction head (NPSH).

(0.5)

(b)

Explain

why, for many centrifugal pumps, it is (0.5) recommended that NPSH be some specified amount greater than zero.

(c)

Tha reactor is shutdown with only one reactor (1.5) coulant pump operating.

Then, a second reactor coolant pump is started.

Describe what happens to loop flow in every loop and to flow through the

core, once flow stabilizes.

Calculations are not required.

(a)

NPSH is the difference between the auction pressure and the saturation pressure of the fluid being pumped (Mathematical explanation also OK)

(b)

To prevent cavitation.

(c)

Loop flow in the originall/ operating loop drops slightly (0.25),

loop flow in the newly operating loop rises to the same valuo (0.05),

loop flow in the idle loop is zero or slightly reversed (0.5),

and flow through the reactor increases to sit 9htly less than twice its original value (0.5).

Reft General Physics HTFF Fundamentals Gect III, Pt.

B, Chap 1 1

P'

3 1.08 (3.0)

Refer to Figure 4 (from Tech. Spec. 3.1.3)

(a)

Explain why, in theory, the curves in the figure (1.5) should be expected to shift to the right for the second 6.0 effective full power years, the third 6.0 effective years, etc.

(b)

What are the pressurizer heatup and cooldown rate (0.5) limits set by the Technical Specifications?

(c)

The reactor coolant system is at 1600 psig and 300 (1.0)

F due to a rapid cooldown transient.

In what direction should pressure and temperature be changed to minimize the chance of overstressing the reactor vessel? Explain.

(a)

The shift in the curves corresponds to a change in RTNDT (a measure of the fracture toughness of the

vessel, or of its degree of embrittlement) during that period.

(0.75)

RTNDT is expected to increase with reacter vessel fluence, according to the Tech.

Specs.,

which Hill sttadily increase during these intervals, since fluence is neutron flux times time.

So the curves shift to the right. (0.75)

(b) 100 F/

hr heatup (0.25) and 200 F/

hr cooldown (0.25)

(c)

Pressure should be decreased while maintaining t eraper a ture constant (0.5).

This will minirai:e thermal stress, and therefore total stress on the reactor vessel (0.5).

Ref: TS 3.1.0 e

4 D*

t-2400 ff I

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2200 2000 MAXIMUM ALLOWABLE HEATUP RATE-60 *F/Fft 31800

~~~~

1600 MT '

.6, e

E 1400 W

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(

d

/

800

$EATUP RATES

• F/ HR 600 0M 20-400 60-CRITICALITY LIMIT--=

200

~~

O 50 10 0,

ISO 200 250 300 350 400 INotCATED TEMPERATURE (*F)

I

, SCE REACTOR COOLANT SYSTEM HEATUP FIGURE UMITATICNS AFPUCABLE FOR FIRST 6.0 Enc.CTIVE

=

FULL POAER YEARS.

TEF60R = 10

• F,

P ERROR = 60 PSIG change No. 1 J

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camCn.rrr uurt - --

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eO im.

ieO 200 zw 300 3w a

INDICATED TEMPERATURE (T)

FIGURE 3.1.3b SCE REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPL} CABLE FOR FIRST 6.0 EreciME l

FULL POWER ' TEARS.

TERROR = lO

• F.,

P ERROR = 60 PSIG ego;,)g4 c

i

1.09 (2.25)

(a)

State three conditions which indicate that natural (1.75) circulation of the reactor coolant sy s tera is occurring (assume no forced circulation).

(b)

If a DNB ratio of 1.00 exists in~ the core, what is (0.5) the probable effect, if any?

(a)

(1) RCS DELTA T < or = to Full Load DELTA T (0.5)

(2) RCS or CET temperatures constant or decreasing (0.5).

(3)

Steam generator pressures constant or decreasing at a ' rate equivalent to the rate of decrease of RCS temperatures (0.5) while

- maintaining steam generator level constant with continuous auxiliary feedwater (0.25).

(b)

Failure of some fuel cladding (0.5)

Ref:

TS 2.1,,

Westinghouse Mitigating Core Damage Training

Manual, p.

51 D

4 r.

1 1.10 (2.5)

The reactor is operating at 85 % reactor power when Control Bank 2

is shimmed out 10 steps.

Sketch (on the attached Figure 5) reactor

power, fuel temperature, T HOT, T COLD, and T STEAM versus time until new final values are reached.

Numerical estimates are not

required, but carefully show the relative relationships of peaks and trends of these parameters.

Assume no reactor trip occurs.

END OF SECTION ONE See Attached Figure (0.5 pts. per variable)

Ref:

Academic Program for Nuclear power Plant Personnel, Volume

'III, Nuclear Power Plant Technology, Chapter 3, Section C I

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SECTION 2 REACTOR OPERATORS EXAMINATION Plant Design - Including Safety and Emergency Systems 2.01 (2.0)

For the Reactor Coolant Pump Seal Water Circuit:

(a)

What is the primary function of the Reactor (1.0)

Coolant Pump Seal Water Circuit in conjunction with the pump seals it provides?

Ob )

In the event of a failure (loss of seal water (1.0) flow) how will seal water be supplied provided the loss of flow is not froma ruptured line?

2.01. Answers (a)

To effectively prevent the leakage of reactor (1.0) coolant along the Reactor Coolant Pump shaft.

0b )

The Test pump in the CVCS is sized to supply (1.0) adequate flow to all three RCP's.

Reference:

Study Guide No.

4, pagen 13, 14

17. and 24.

e e

1

'2.02 (2.0)

The pressurizer' has two code safeties and two power operated relief valves.

(a)

During operation while at power it is determined (140) that there is leakage coming from a relief valve on the pressuri:er..How can an operator determine which specific valve is. leaking among the four valves?

(b)

If the leakage was determined to be past a power (1.0) operated relief valve, and exceeded the Technical Specification limits, would the plant be required to be shutdown? Explain.

2.02 Answers' (a)

Each safety valve has its own temperature element, (0.5) an indication of higher than ambient temperature

.would allow the operator to determine the valve.

The case for the power operated reliefs is (0.5) different, they have a common temperature element.

However they have individual isolation

valves, which can be closed to determine which valve is leaking.

(b)

No (0.5),

The power relief valves have block valves and can be isolated so that operation can continue (0.5).

Reference:

Requalification Exam No. 0031, page 7.

e I

i t

i 2

2.03 (2.0)

Regarding the Pressuri:er:

(a)

Give three reasons why a 1 gpm flow rate is (1.5) maintained around the spray valves into the Pressurizer.

(b)

The Pressurizer level program has a ramp function (0.5) from 25% at no load to 37% at full load.

Why is pressurizer level required to vary as a function of power?

2.03 Answer:

(a) 1)

To reduce thermal stresses and thermal shock (0.5) when the spray valves open.

2) To help maintain uniform water chemistry (boron (0.5) concentration).

3)

And to maintain the desired temperature in the (0.5) surge line.

(b)

To allow for surges into and out of the (0,5)

Pressurizer during load transients.

Reference:

Study Guide No. 16.

N

2.04 (3.0)

For the Diesel Generators:

(a)

List six trips for the Diesel Generator engineg (3.0) that are in service for a manual start.

2.04 Answers (a)

Any 6 (0.5) each.

1) Engine Overspeed.

2) Generator Differential.

3) High Crankcase Pressure.

4) Low-l ow Engine Oil Pressure.

5) Low-low Turbo Oil Pressure.

6) High-high Jacket Water Temperature.

7) High Vibration Engine or Turbo.

8) High-high Lube Oil Temperature.

9) High Main Bearing Temperature.

Reference:

Study Guide No. 98, page 9.

2.05 (3.5)

For the Reactor Coolant Pump No. 1 Seals:

i (a)

An operator is about to start a Reactor Coolant (1.0)

Pump, system pressure is 375 psig. An annunciator window is illuminated "No.

1 SEAL LO FLOW".

ceg the pump be started? Egglain.

(b)

What 5 conditions must be met to start a Reactor (2.5)

Coolant Pump?

2.05 Answers (a)

Yes (0.5).

At pressures below 1000 psig the No. I seal flow may be too low to clear the alarm (a

flow of at least 0.3 opm is considered adequate for the condition)

(0.5).

(b)

1) Oil lift pump permissive light on.

(0.5) 2).No.1 seal delta-P greater than 275 psig.

(0.5)

3) No.1 seal leakoff greater than 0.25 gpm.

(0.5)

4) Thermal barrier delta-P about 20 inches.

(0.5)

5) Reactor Coolant System Pressure greater than or (0.5) equal to 350 psig.

References Study Guide No.

3, page 12.

~

4 l

5 O-

-3 o

s 2.06 (4.0)

On the "J" console there is a 5 position switch associated with the Steam Dump System.

For the following swi tch posi tions under what conditions is each position used?

And how does the system

_ operate in each position?

(a)

Position 2

" Automatic" (1.0)

(b)

Position 3 "Prescure Control

- Atmosphere and (1.0)

Condenser" (c)

Position 4

" Pressure Control - Atmosphere Only" (1.0)

(d)

Positicn 5

" Pressure Control - Condenser Only" (1.0) 2.06 Answer:

(a)

Normal operating position when above 20%

power (0.5).

System will operate when a load reduction of.10% to 40% occurs and Tave i s greater than Tref by 5 degrees F (0.5)

(b)

The switch is placed here when Automatic control is not availiable (0.5).

S tri c tl y by pressdre control only (from PC-418A pressure controller.)

(0.5).

(c)

This is used when the condenser is not avalliable (0.5). Pressure control is as in (b) above (0.5).

(d)

Used for normal startup and

cooldown, and is placed here for normal operation when in pressure control (0.5).

Pressure control as in (b) above (0.5).

Reference:

Study Guide No. 53, pages 7 and 8.

v s

k

. 2.07 (3.5) r For the Pressurizer Level Control System supply the. associated -

f

-alarm and/or controlLactuations for the following setpoints.

i (a) 70% level.

(0.5) f e

l (b)

+4% of. programme'dLlevel.

(1.0) v l

(c)

-4% of programmed level.

(0.5)

[

i

.. '10%~ level.

(1.5)

(d) i

,~

2.07. Answer:

i (a)l High Pressurizer water level Reactor trip.

(0. 5).

I (b)' High Pressurizer water level alarm.

(0.5) i Heaters off.

(0.5) t

- (c)- Pressurizer low level alarm.

(0.5) f l

4 l

(d)

Pressurizer low-low alarm.

(0.5)

'[

Heaters off.

(0.5)

Isolation of letdown valve.

' (0. 5)

=l

Reference:

Study Guide No. 11, page 16.

q.

w w

w 4"

6 t

~7 -

2.08 (2.0)

For the Containment Spray System and the Containment Spray Actuation System (CSAS):

(a)

What is the sequence (order) of events when an (1.0) actuation signal (CSAS) comes in to start Containment Spray?

(b)

What setpoints and logic must be satisfied to (1.0) initiate an actuation of Containment Spray from the CSAS (automatic onl y) ?

2.09 Answer:

(a)

Refueling Water Pumps start first.

(0.25)

The Spray Header valves open.

(0.25)

The Hydrazine injection pump stars.

(0.25)

The Hydra:ine valves open to admit hydrazine to (0.25) the Containment Spray header.

(b) 2/2 Safety Injection (1/1 sequencer per train)

(1.0) and 2/3 High Sphere pressure - 10psig.

Reference:

Study Guides No.s 16 and 17.

w a

8

2.09 (3.0)

Concerning the Safeguards Load Sequencing System (SLSS):

(a)

The SLSS recieves six signals from various sources (1.5) including one from a test switch.

What are the remaining five signals?

(b)

What actions must an operator take to tecminetg (1.5) and reset an Automatic actuation of Safety Injection?

2,09 Answer:

(a)

1) Pressuriner Pressure (0.3)

2) Containment Pressure (0. 3)

3) Undervoltage signal from both 4160 busses (1C

( 0. 3 )

and 2C).

4) Safety Injection Black (0.3)

5) Standby Diesel Generation (Voltage frequency (0.3) signal and circuit breaker status signal.)

(b)

First he must block the automatic actuation signal by depressing the reset switches on the sequencer surveillance panel (0.5),

and then in d

ot et turning off the feedwater pumps (0.5) and the cafety injection pumps (0.5).

Reference:

Study Guide No. 17 pages 11, 12, and 15.

END OF SECTION 2 e

)

SECTION 3 REACTOR OPERATORS EXAMINATION Instrumentation and Controls 3.01 (2.0)

Regarding the steam generator level control system:

(a)

The steam generator level control system is (1.5) commonaly referred to as a three element control system, what are those three elements?

(b)

If there are three elements used for control in (0.5) this

system, why then is steam pressure (not one of the three ' elements') also an input to the system?

3.01 Answer (a) 1.

Feedwater flow.

(0.5) 2.

Steam flow.

(0.5) 3.

Steam generator level.

(0.5)

(b)

To provide input for density compensation in the (0.5)

Steam Flow Computer.

Reference:

OT-1045 and OT-1064 " Steam Generator Water Level Control", Study Guide No. 62.

v I

1

I 3.02 (1.5)

For the diagram (figure 3.1) on the following page describe:

1 (a)

Which side of the manometer will rise when a fluid (0.5) i is flowing in the direction shown?

i

~

(b)

What. ' relationship (i. e.

equation) is used to (0.5) determine the flow rate in the pipe?

(c).

What other type of instrumentation is normally (0.5) provided in the plant to perform this function (of the manometer)?

3.02 Answer

.(a)

The leg connected to the restriction of the (0.5)

. venturi..

(b)

The Bernoulli equation (Flow rate is proportional (0.5)

'.to the square root of the differential pressure).

-(c)

Typically a Bourdon tube type transmitter

(.5)

(D/P. cell).

Reference:

General. Physics

" Heat Transfer Thermodynamics and Fluid Flow Fundamentals" w

~

3.03 (3.0) l 1

9 i

I I,

Li

SECTION 3 REACTOR OPERATORS EXAMINATION Instrumentation and Controls 3.01 (2.0)

Regarding the steam generator level control system:

(a)

The steam generator level control system is (1.5) commonaly referred to as a three element control system, what are those three elements?

(b)

If there are three elements used for control in (0.5) this

system, why then is steam pressure (not qne of the three ' elements') also an input to the system?

3.01 Answer (a) 1.

Feedwater flow.

(0.5) 2.

Steam flow.

(0.5) 3.

Steam generator level.

(0.5)

(b)

To provide input for density compensation in the (0.5)

Steam Flow Computer.

Reference:

OT-1045 and OT-1064 " Steam Generator Water Level Control", Study Guide No. 62.

r i

3.02 (1.5)

For the diagram-(figure 3.1) on the following page describe:

(a)

Which side of the manometer will rise when a fluid (0.5) is flowing.in the direction shown?

.(b)

What relationship (i. e.

equation) is used to (0.5) determine the flow rate in the pipe?

(c)

What other type of instrumentation is normally (0.5) provided in the plant to perf or m this function (of the manometer)?

3.02 Answer (a)

The leg connected to the restriction of the (0.5) venturi.

(b)- The Bernoulli equation (Flow rate is proportional (0.5) to.the square root of the differential pressure).

(c)

Typically a Bourdon tube type transmitter (0.5)

(D/P cell).

Reference:

General Physics

" Heat Transfer Thermodynamics and Fluid Flow Fundamentals" e

P*

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6

3.03 (3.0)

For the curve on the following page (Figure 3.2) identify the following parts of the curve:

(a)

I (0.5)

(b)

II (0.5)

(c)

III (0.5)

(d)

IV (0.5)

(e)

V (0.5)

(f)

VI (0.5) 3.03 Answer (a)

Recombination (0.5)

(b)

Ionization (0.5)

(c)

Proportional (0.5)

(d)

Limited Proportional (0.5)

(e)

Geiger-Mueller (0,5)

(f)

Continous Discharge (0.5)

Reference:

General Physics Volume II, Chapter 3, Section E e

l l

8 I

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Curve 1: Radiation event of lower specific ionization.

Curve 2: Radiation event of higher specific ionization.

Fig u re 3.2 w

O

3.04 (2.5)

During solid plant operations the following alarms come in:

" OMS Hi Pressure"

" Pressure Transient in Progress" (a)

What condition would actuate these alarms (i. e.

(0.5)

System setpoints)?

(b)

What function does the Overpressure Mitigation (0.5)

System serve?

(c)

After securing charging flow as the first (1.5) immediate action

takea, what valves must an operator check to insure are open to complete the immediate actions in response to these alarms?

3.04 Answers (a)

The " OMS Hi Presvire" alarm comes in at 480 psig.

(0.25)

" Pressure Transient in Progress" at 500 psig.

(0.25)

(b)

To orotectect the RCS from potential (0.5) overpressurination during solid operations.

(c)

Insure LCV-1112, letdown isolation, open.

(0.5)

PCV-1105, letdown control valve open.

(0.5)

And, CV-525 and CV-526.

containment isolation (0.5) valves, are open.

Reference:

Line Diagrams. 501-2.1-11 "Overpressurination Mitigation System Actuation", and Study Guide No. 6 l

)

l 4

3.05 (3.0)

For the following nuclear instrumentation detectors discuss the operation of the detector chambet:

(a)

Intermediate Range Instruments.

(1.0)

(b)

Source Range Instruments.

(1.0)

(c)

Power Range Instruments.

(1.0) 3.05 Answer:

(a)

The intermediate range instrument chamber is a

Compensated Ion Chamber (0.5).

This chamber is essentially two chambers, one coated on the inside with a Baron coating (0.25).

the other not.

The chambers are wired together such that the outputs cancel out the current that incident gamma radiation contributes leaving a current output representative of the neutron population (0.25)

(b)

The chamber is filled with Baron Trifluoride gas (0.5).

Thermal neutrons are absorbed by the Boron 10- atoms which then disintegrate into Lithium 7

atoms and emit high energy alpha particles (0.25).

This generates an ionization that is collected by the el ec tr odes producing a pulse output (0.25) proportional to the number of neutrons.

(Discrimination is accomplished in the circuitry in the channel drawer.)

(c)

This channel uses an Uncompensated Ion Chamber (0.5).

Which is coated on the inside with a Doron coating (0.25).

Because of the power range in which the instrumentation operates there is no need to compensate for gamma radiation because it presents a very minor contribution (0.25).

Reference:

Study Guide No. 32: General Physics Volume II, Chapter 3 Section E.

v M-

l I

3.06 (2.0) l For the f ollowing components how would the failure affect their indication and why?

(a)

The failure of a

Delta-P diaphragm on the (1.0) pressuri z er level detection system.

(b)

A rupture of the reference leg on the Steam (1.0)

Generator level detection system.

1 3.06 Answer:

1 1

(a)

The failure of a diaphragm in this case would I

present a

condition where there is no indicated differential pressure (0.5).

This would cause an indication of high l evel (0.5).

(b)

If a reference leg ruptured the indication would be of a low differential pressure (0.5).

This would also indicate a high level (0.5)

Reference:

Requalification Exam 940, page 13, question 10.

w D

9 3.07 (1.5)

For the Main Turbine:

(a)

Besides the Overspeed trip that is provided what (1.5) are three other trips associated with the Turbine.

3.07 Answers (a)

Any three (0.5) each.

Low bearing oil pressure.

Solenoid Trip.

Thrust bearing trip.

Low vacuum trip.

Reference:

Study Guide 51, pages 29 and 30.

+

7

3.08 (3.0)

For the following alarm indications indicate what condition exists for the alarm to be in.

(a)

Shutdown bank deviation alarm.

(1.0)

(b)

Control bank deviation alarm.

(1.0)

(c)

Rod bottom light on for one control rod.

(1.0) 3.08 Answer:

(a)

The alarm will come in whenever a

rod in a

(1.0) shutdown bank deviates belgw a greget ggsitign.

(b)

In this case the alarm would come in because a rod (1.0) has deviated by more than a preset amount from the bank gggitign.

The bank positon comes from the demanded indication circuit from the Digital Detection System.

(c)

This would come in if a

rod dropped below (1.0) approximately 25 steps.

Reference:

Study Guide 13, page 2.

e 8

)

e-3.09 (3.0)

For tne Control Rod Drive Summing Computer:

(a)

The computer receives inputs from 5 sources, name (2.0) 4 of those sources.

(b)

The output from the computer goes to two places,.

(1.0) what are those places and what function does the signal that is provided perform?

3.09 Answer:

(a)

Any 4 (0.5) each:

1) Average Tavg.

2) Tref.

3) Primary pressure.

4) Nuclear flux.

5) Feedback from Eistable TC-413B indicating demanded rod postion.

(b)

1) Bistable 413B'(0.25),

to move rods in or out (0.25).

2) Pod speed control' er (0.25).

Which produces a

l demanded speed to be fed to the rod drive system through an auto / manual group selector switch (0.25).

Reference:

Study Guide 12. page 4.

9

~

t

}

3.10 (1.5)

For the following Operational Radiation Monitoring System channels what type of de ;ec'.or is used, and what control funtions do they provide? (Do not or ovide alarms. )

~ (a)

Channel R-1211 - continuous air sample from either (0.5) the vapor container or the plant stack.

(b)

Channel R-1214 Main Stack Gas Monitor.

(0.5)

(c)

Channel R-1216 Steam Generator Blowdown-Liquid (0.5)

Sample monitor.

3.10 Answer:

(a)

Scintillation dotector.

(0.25)

No direct control function.

(0.25)

(b)

Geiger-Mueller detectors.

(0.25)

High level alarm deenergizes SV-99 which closes to (0.25) stop all gas discharge form gas decay tanks.

(c)

Scintillation detector.

(0.25)

High radiation level alarm energines solenoid (0.25) valve (SV-84),

causing CV-100, 100A, and 100B to close.

This stops discharge to blowdown tank and to circulating water outfall.

(Candidate only has to indicate that discharges to blowdown tank and outfall are isolated.)

Reference:

Study Guide 31; pages 2, 4,

8, 9,

11, and 13.

10

3.11 (2.0)

The reactor Protection System is provided with 8

" Permissive Circuits":

(a)

What are the gutposes of these permissive (0.5) circuits?

(b)

What function do the following permissive circuits provide?

1) Permissive Circuit No.

5 - Shutdown Margin Low (0.5)

Alarm

2) Permissive Circuit No. 4 - Steam Dump Automatic (0.5)

Mode Cutout

3) Permissive Circuit No. 3 - Rod Drop Rod Stop, (0.5) 3.11 Answer:

(a)

To provide increased reliability, and protection.

(0.5) and to provide compatibility between control systems.

(b)

1) This would warn operators of the possibility of (0.5) inadequate shutdown margin as determined by the shutdown margin computers.

2)

This would block automatic operation of the (0.5) steam dumps when Tave-Tref setpoint deviations are reached unless the MWe load decrease is greater than 45 MWe in less than 10 seconds.

3) This would prevent automatic rod withdrawal and (0.5) would initiate a turbine load limit runback to 70%

power when a decrease of 5% or greater in power is detected in any power range N.I.

channel.

Reference:

Study Guide 12, pages 7, and 8.

End of Section 3

{

11

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J l

SECTION FOUR l

REACTOR OPERATOR EXAMINATION 4.01 (2.5)

(a)

Under nonemergency conditions, give three general (1.5) types of activities for which procedures are required to be used.

(An example would be

.a n evolution affecting plant reliability).

~(b)

It is desired to perform minor troubleshooting of (1.0)

- a pressure controller.

One valve will be shut to perform the troubleshooting, and then it will be reopened.

No in-place procedure is required for this activity.

May this activity be performed without equipment control?

Explain.

(a)

Any three of the following (0.5

each, max.

of 1.5):

manipulation of all safety related equipuent, maj or (complex) evolutions, infrequently performed evolutions, manipulations of maj or (nonsafety) equipment, or evolutions l

affecting plant plant safety.

(b)

In general, no, since equipment control is required for most evolutions not controlled by procedures.

(0.5)

However,

~the Shift Superintendent may.' approve a ninor maintenance item, -not important to safety, without equipment control. (0.5) l Ref* SO!-14

-42, 20!-14-12 pp. 3-10 i

9 4

4 D--

4.02 (3.6)

List the twelve immediate actions for a reactor trip with safety inj ect ion initiation.

Valve numbers and positions, and setpoints are not required.

~(1) Verify reactor trip

.(2) Verify turbine trip (3) Verify electrical busses energized (4) Check ~_ if SI initiated (5) Verify RCP's tripped (6) Verify SI system valve alignment

'(7) Veri +y SI system pumps running (8) Verify Charging pump suction aligned to RWST (9) Verify SI flow (10)~ Verify containment isolation (11) Verify Auto AFW Initiation (12) Chech Containment Pressure (0.3 pts. each)

Ref: SO1-1.0-10 w

p.

.s c

w

~

4 4.03 (3.0)

(a)

List the four highest priority critical safety (2.0)

. functions.

(b)

Explain how red and orange conditions in critical (1.01 safety function trees require different priority secuences.

1 1

'(a)

Maintenance or control of-subtriticality, core

- cool i ng, reactor. coolant inventory, and heat sink I

(0.5 pts. each, correct order not required)

- (b)

Both require departure frcm-any other EDI in

use,

[

but a

red condition requires immediate 4

implementation of the EDI (0.5),

whereas for an p

)

orange condition it is required to complete the i

3 current pass through all of the s'.3tus trees (to l

look for higher priority red conditions)

(0.5).

h Ref: S01-1.0-1 pp. 11-14 I

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4.04 (2.01 According to the Emergency Operating Instructions, what events should the following symptoms correspond to?

(a)

Reactor Trip and a Steam / Feed Nismatch Reactor (0.5)

Trip Alarm ON (b)

Reactor Trip, Safety Inj ec t ion, Electrical Power (0.5)

Available, Emergency Systems Operating as

Required, and Steam Generator

'A' level is rising uncontrolled.

(c)

Six core exit thermocouples indicating 700 F (0.5)

(d)

Same as (b) except steam generator levels are (0.5) under control and containment radiation monitor R1255 is above its alarm setpoint.

(a)

Loss of Secondary Coolant (b)

Steam Generator Tube Rupture (c)

Potential Loss of Core Cooling (d)

Loss of Reactor Coolant Ref: S01-1,0-30, 301-1.0-40 301-1.2-2, S01-1.0-20 e

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I 4.05 (2.0)

Abnormal Operating Instruction S01-2.1-10, Loss of Component Cooling

Water, requires that for a total loss of component cooling water which cannot be promptly

restored, a natural circulation cooldown be performed using the test pump

only, to control pressurizer level.

(a)

Explain why natural circulation is required by the (1.0) procedure.

(b)

Explain why a

cooldown is required by the (0.5) procedure.

(c)

Explain why the use of letdown to control (0.5) pressurizer level is not permitted.

(a)

The

total, unrestored -loss of CCW requires a

reactor trip (0.5) and trip of all RCPs (0.5),

so only natural circulation is available to cool the Core.

(b)

A cooldown is implied by Tech.

Specs.

(0.0 and 3.1.2.E) with no'RCPs operating. '

(c)

There will not be adequate CCW flow to the nonregenerative heat exchanger to cool letdown.

Ref: TS 3.0, 3.1.2.E, 301-2.1-10

+

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)

4.06 (2.5)

(a)

Upon confirmed loss of which vital or utility bus (0.9)

(or buses) is an immediate reactor trip HOT required?

(b)

Which AC Distribution buses are required to be (1.1)

OPERABLE by the Technical Specifications, assuming the plant is in Mode 1, and T40T in any Action statements of the Specifications?

(c)

What equipment is recuired other than fuel tanks (0.5) by the Technical Specifications for the emergency diesel generators.

(a)

Vital buses 3A, 5 and 6 (0.3 each)

(b) 4160 Buses 1C and 2C, 430 Buses 1,

2, and 3, and Vital Buses 1,

2, 2,

CA, 4,

5, and 6.

(0.1 pts each)

(c)

Fuel transfer pumps Ref: 501-2.6-3, TS 0.7

+

D

4.07 (3.0)

The plant is being started up from hot standby to miniuum load.

(a)

What rod positions are procedurally required for (1.0) criticality unless otherwise directed by the Shift Superintendent?

(b)

When, according to-the procedure, should (1.0) criticality be anticipated?

(c)

What is.

the limit on Start Up Rate during the (0.5) startup?

id)

What is the limit on turbine bachpressure prior to (0.5) rolling the turbine?

(a)

Shutdown Banks 1 and 2 and Control Bank 1 at 318 steps. Control Bank 2 &t 100 steps (0.25 pts each).

(b)

At any t irae when control rods are being withdrawn, or when baron dilution is in progress (0.5 pts each)

(c) 1.0 dpm (d) 5.5" Hg Ref:

SO1-3-2 i

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,+__.,.y_,,-m..,,,y_

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4.08 (2.0)

(a)

Define a high radiation area.

(1.0) 4 i

L (b)

What are the SONGS quarterly Administrative Limits (1.0) for radiation exposure?

i

+

i (a)

.A high radiation area is any area which is i

accessible (0.1) in which a maj or portion of a

f l

person's body (0.1) could receiv in excess of 100 meern in one hour (0.8) t j

t (b) 900 mrem whole body (0.34), 3750-mrem skin (0.03),

f and 4700 mrem extremities (0.30) i 4

i Ref: SONGS Radiation Training Handout l

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I 4.09 (2.5) i i'

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With respect to boration and dilution of the reactor coolant I.

system

• I 7

(a)

Why is boric acid inj ect ion pump discharge (0.5) j flow required to be limited to 7.5 9pm?

J' 1

5 (b)

Why is it required to have an RHR or RCP operating (0.5)

[

while borating?

j l

(cJ What is the required setting for the Boric Acid (0.5)

Blend System when not diluting or borating?

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(d)

How is pressurizer boron concentratio'n maintained (1.0)

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within specified tolerances af the reactor coolant

[

f-system boron concentration?

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(a)

To. prevent pump cavitation f

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(b)

To provide adequate mixing of the baron

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l (c)

AUTO MAVEUP 4

L (d)

Placing pressurizer-heaters in MANUAL (0.5), causing automatic spray operation to increase the turnover i

e rate of coolant in the pressurizer (0.51

)

j 4

Ref:

SO1-4-la 1

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4.10 (1.9)

(a)

Which a d-t h e lowest Emergency Action Level which (0.6) is characterized by maj or failures of plant functions needed for protection of the public?

(b)

What are the imined i ate operator actions for a

(1.3) severe earthquake which occurs while operating at power?

END OF. SECTIO!! FOUR f

(a)

Site Area Emergency (b)

(1)

If any equtpuent failure ts.

Indicated, trip the reactor. ( '). 6 5 )

(2)

Trip the Circulating Water Pumps if the reactor is tripped. (0.65)

Ref: SO1-2.5-1, SONGS Emergency Plan m

P e-

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