ML20206N344

Exam Rept 50-361/OL-86-02 for Units 2 & 3 on 860617-26.Exam Results:All Candidates Passed Written Exam & Three Individuals Failed Operating Exam

ML20206N344
Person / Time
Site:	San Onofre
Issue date:	08/06/1986
From:	Elin J, Johnston G, Obrien J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20206N342	List: ML20206N338 ML20206N344
References
50-361-OL-86-02, 50-361-OL-86-2, NUDOCS 8608260236
Download: ML20206N344 (91)

Text

v. ~ .

i-1 L

t s

Report No. -

50-361/0L-86-02 Docket No. 50-361-L'icense No.

LLicensee: . Southern California Edison Company--

P.O. Box 800

.2244 Walnut Grove Avenue Rosemead, California 91770

-Facility'Name: San Onofre Nuclear' Generating Station, t Units 2 and 3 P

Examination Conducted: June 17 - 26, 1986

/ /

Examiners:

G . V.

f Kffh% I hnstof, opera r Licensing Examiner Dagejifigned P. O'Brien, Operator Lic sing Examiner A s/94 Dfte/ Signed Appro*/ed by: . N --

J. Elin, Chief, Operations Section. Dqte{igned Summary: ,

- c >

During this requalification-cycle, twenty percent of-the operating' staff-(21 licensed operators) were examined.1 This included ten Senior Reactor

--_ Operators, and eleven Reactor Operators. -: The NRC prepared a complete written examination and administered'it to the twenty-one participants.

The examiners also administered. operating, examinations to all of the participants. The operating examinationsiindibded simulatbr and oral l walkthrough portions.' All'of 'the participants. passed' the written portion

.of the examination. Three individuals faile'dlthe ope ~ rating examination. - - -

InaccordancewithNUREG-1021,ES-601'RiqualificatkonProgramEvaluations',

the requalification program at' San Onofre Nuclear: Generating Station, '

Units'2 and 3 is evaluated as satisfactory.

e 4

a 8608260236 860812 PDR ADOCK 05000361 V PDR

~

3 4-

- er <x--, --~-wr-m- -ww- g w eeam - -----w.ew s, p,- ne-- -.,,-.m--,,ev,,-wr w,mm.-- ,,--m- e m.n evi. g m -...v--->,,.-ema--ve-re-.----.---r- --,-r.,-,n.-meo

' u ,,, i .s

r:

• l l-DETAILS

1. Persons Examined Afgroup o'f twenty-one operators, ten holding Senior Reactor Operator licenses, and eleven holding Reactor Operator licenses representingfjust over 20% of the licensed staff were examined.

2. Persons Contacted Southern California:

J..Wambold-

• R.- Mette

'*L. Simmons NRC:

L G. Johnston, Operator Licensing Examiner

• J. O'Brien, Operator Licensing Examiner

• Denotes those present at exit-on June 26, 1986.

3. Program Evaluation

' Required for Satisfactory The requalification program was evaluated upon the~ criteria of Examiner Standard ES-601 of NUREG-1021. The requirement for a satisfact5ry program is more than 80% of the evaluated operators

. must pass all operating examinations, all sections of the written

-examination,'and-the written examination overall administered by

the.NRC.

Performance

-The NRC administered examinations to eleven Reactor. Operators and ten Senior Reactor Operators.' One Reactor Operator and two Senior

~

' Reactor Operators failed the operating portion of the examinations, ~

i all of the individuals passed the written examinations. The results for all the operators was a pass rate of 85.7%.

Evaluation Satisfactory.

F I

A 4

r -

t 4 y ,-t, - -c r,. . ___..,-,.,-.7, ._,_,-e_ . , , , , . _ , , . , - + - -, , ,,cr--y .,_. _ ,_,,,_,...e, ,_.v_r_.e_ --,s - ,, .,.w.., , . , . . , .,w_,,,

- 2 5

4. Examination Results Eval 2ation During the conduct of the simulator portion of the operating examination the examinert observed that a significant number of examinees had trouble making reactivity calculations for changes in reactor power. This prrincipally involved use ofxthe inverse '

boron worth curves for determining borations and dilutions of boron. The examiners in follow-up questioning determined that this problem was related more to a lack of familiarity with the curves than a lack of understanding of the process of calculating-a dilution or boration. This area deserves attention'in the ~

requalification training program in the opinion of the examiners.

One individual (a SRO), during the walkthrough portion of.the operat'ing examination did not demonstrate a knowledge of'the location of the remote shutdown panel. While this does not indicate that the operators in general did-not know where the shutdown panel was, it is, in the opinion of the examiners, an egregious example of a lack of familiarity with the physical plant. This in particular deserves close attention in the requalification training program'. This also raises a question as to the effectiveness of ' simulator only' training for requalification training, and also the effectiveness of the training for the Emergency Response Plan that the operators have received.

An evaluation of the-written examinations does not reveal any particular areas of concern because the overall performance was very good. The grades were, in the large part, in the mid 80 percentile to the mid 90 percentile range.

5. Exit Meeting At the conclusion of the site visits, the examiners met with the facility

. representatives denoted in Section 2 of this report to discuss the results of the evaluation to that point.

b d

i s

i i

f

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

Resolution of Facility Comments San Onofre Units 2/3 Requalification 1.0 Facility Comment: Question 2.02 "As Steam Generator pressure drops then Unit load will decrease because our Governor Control systen doesn't control on load, so as pressure drops then the Operator will.have to raise valves to increase load to compensate for decrease pressure."

Resolution:

The examiner will give consideration to actual system _ operation and will adjust the key accordingly.

2.0 Facility Comment: Question 2.03a "The answer is correct, but your answer should also include the H2 monitor and is readily available for the Operators to read in the Control Room."

Resolution:

The examiner agrees and will change the key.

3.0 Facility Comment: Question 2.06b "Need to add to list the Shutdown Cooling Heat Exchangers if CCW is inoperable due to the fact that SDC HX's are used for Containment Spray for pressure reduction of Containment."

Resolution:

Agreed, the examiner will add to key.

4.0 Facility Comment: Question 2.07c

" SONGS 2/3 doesn't have an anti-ejection gripper. Our holding mode applies holding current to the holding gripper to maintain CEA height."

Resolution:

The examiner reviewed the FSAR and determined that the analysis for a rod ejection accident is accounted for on the basis that the internals are designed to sustain the potential for a rod ejection and maint.Jn a coolable geometry. The answer in the key will be changed to "notulug".

si 1 1: , <

J

c. ,

1; l

i Fheility Comment: Question 3.01'

~5.0 .

"SBCS does not control Atmospheric Dump valves as stated in question.

Since procedurely we keep AMI setpoint at 15% and CEDMCS are not used in Auto Sequential some answers for part a. may.be that no signal-will be generated unless assumptions are made."

Resolution:

~

The examiner feels the question is still valid although the lead in ,

paragraph is not correct. The two follow on questions are still correct, in and of themselves.

6.0 Facility Comment: Question 3.09 "As stated in the last part of the answer key the level error signal takes over and restores.the level to the setpoint. So you could expect to have answers saying that S/G 1evel will remain the sam 3 not accounting for the initial change."

Resolution:

If the~ response by the candidate does not consider the short term effects the examiner will consider partial credit in the light of the fact the question did not specifically ask for both short and long term-effects.

7.0 Facility Comment: : Question 3.10

" Answer.to part b. could include that stem lines not designed to hold

-water and S/G reliefs not designed to pass water."

Resolution:

The examiner will accept these as other appropriate answers.

<8.0 Facility Comment: Question 4.05b

, " Requiring the operator to recall a specific level value out of a normal operating procedure exceeds the expected. job performance requirements."

l

' Resolution:

The examiner would expect the candidate to be able to explain the precautions and limitations of a normal procedure. The examiner views the question as worded is sufficient for that purpose but does recognize that specific values are hard to recall. Therefore, the

- examiner will change the key to provide leeway for the value expected.

t l ,

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

. 3 9.0 Facility Comment: Question 4.05c "S023-5-1.3 applies from Mode 5 e.o Mode 3. In the Precautions, heatup rate limits of 8 deg. F per hotr (unit 3 only), 25 deg. F per hour and 50 deg. F per hour are established for different temperature conditions. Also Technical Specifications establish limite of 30 deg. F per hour and 60 deg. per hour for differing temperature conditions.

Requiring the operator to recall specific values for a given condition seems quite demanding."

Reso lution:

Same as 8.0 above.

10.0 Facility Comment: Question 5.12 "The answer key gives 'a.' as the correct answer. I don't believe S/G pressure would increase. Therefore 'b.' would be correct. The reference material does not address this question."

~

Resolution:

The examiner agrees with the comment and will change the key.

11.0 Facility Comment: Question 6.09a

" Low S/G flow trip and Reactor Coolant Flow Low trip are two different trips. The nomenclature in the question is the problem (i.e. both terms are used). RCP Shaft shear is the reason for one and pump seizure is the reason for the other. May have to accept both as the reason."

Resolution:

The examiner will accept either as it would not be clear from reading question which response the examiner.is seeking.

12.0 Facility Comment: Question 7.01a "The possible causes list does not accurately reflect the conditions necessary to lose the S/G inventory._ 'Also, by best estimate analysis, the Steam Generator levels should last approximately 20 minutes minimum following a trip on low S/G level. Since the trip was on on loss of flow, it is somewhat implasible for S/G levels to be 00S low after 15 minutes. I am afraid that due to the somewhat inadequate choices, some examinees may pick number 8 assuming MSIS does not occur."

Resolution:

The examiner acknowledges that number 8 may be appropriate under some circumstances. Also, the examiner only talked with two individuals about this question during the examination, retrospectively, the examinees should have been informed of .the problem.

- . a ,s

• w,

, Q 1,

~

.. . -\ ,

~

e 13.0. Facility _ Comment: Question 7.02b

" Pressure should remain'the same for PZRISat.~ conditions. S023-5-1.3,

'Section 6.4.4 PZR pressure remains constant ordincrease. It increasing

with letdown valve full open, stop; charging and/or turn off heatcrs.

Bubble formation doesn't caase the possible pressure increase it is inventory control.";

Resolution:

The examiner will accept constant or increases.

14.0 Facility Comment: Question 7.05 '

"I looked in the reference procedures, and could find nothing that gave a stated reason for stating p001 before p004. Spray availability would.

not seem to directly affect the length of time necessary for RCP Starts."

Resolution:

The examiner was seeking to see if the examinees were aware that other things affect starting of the pumps other than starting duty. The

--examiner does . recognize that the question does not directly imply that starting _ duty does not have much to do with the choice of pumps.

Therefore, he will give partial credit if.the candidate indicates the starting' duty criteria rather than a discussion of the establishment of~ pressurizer spray flow paths.

15.0 Fncility Comment: Question 7.06 "OtIter sources: Salt Water discharge from CCW Heat Exchangers, S/G Blowdown bypass (fish fry)."

Resolution:

The key for this question does establish a "other" category.

16.0-Facility Comment Question 7.07a

" Tripping the Reactor and tripping the RCP's are not "immediate actions"."

Resolution:

~

The examiner feels'that this does constitute "immediate action",

however, the reviewer clearly felt that these actions were subsequent actions. If the examinees respond differently'the examiner will consider on a case by case.

1

r - -. - -

)

~ ,

s' , 5 17.0 Facility Comment: Question 7.08c "Techical Specification 3.9.12 addresses FHB accident cleanup units rather than exhaust dampers. Losing a damper-in'the exhaust system does not compromise the P. A. cleanup units. Therefore, the 'why' statement in the answer key is inappropriate."

Resolution:

~

The examiner will accept a response that addresses that fact that.the fuel is being moved in the containment, and not in the fuel handling building. Therefore, no suspension of fuel movement in the containment.

18.0 Facility Comment: Question 7.10 "The question asks for three system, yet the answer key refers.co actions. Another action which helps depressurize is to cool down via SBCS or Atm. dumps."

Resolution:

The examiner agrees with the camment and will consider other ' actions'.

19.0 Facility Comment: Question 8.03b "The question asks 'What action is required.' The answer key states specific point values for details of the action statement. If this response is the desired response, the question should state ' State the LCO action statement in detail with specific' actions and times'."

-Resolution:

The examiner is not going-to change the key, but does recognize that the question-does'not clearly ask for the information in-the key., However, in the context of the section of'the examination that the question is in, the examinee should deduce that the response in the key is the one sought.

t P l'

J

k U.S. Nuclear Regulatory Commission Reactor Operator License Examination *

' Facility: SAN'ONOFRE 2/3 REQUALIFICATION Reactor Type: CE-SYSTEM  ;

i Date Administered: JUNE 17, 1986

.,,s

'\ ( O l Examiner: c. JOHNSTON

~~

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

% of Category X of Candidate's Category value Total Score Value Cateoory 1s.0 1. Principles of Nuclear Power Plant Operation, Thermodynamics, g Heat Transfer and Fluid Flow 15.0 e 2. Plant Design Including Safety and Emergency Systems -

M MY 3. Instruments and Controls 25.*r 15.0 M 4. Procedures - Normal, Abnormal, Emergency, and Radiological Control TOTALS Final Grade  %

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

Candtdate's Signature k

e %

3 9

o EQUATION SHEET f = ma y = s/t C w = mg a = v,t + at 2 Cycle efficiency =

E

}

E = mC a = (v - y )/t 0 -At 2 A = AN KE = my vg A = A,e -

= v, + at PE = mgh m = e/t A = in 2/tg = 0.693/tg W = vaP AE = 931Am t b(eff) = (t;0)(tu)

(t +t) b "

Q=$ CAT y.ye 4X l Q=U T y . y UX Pwr = Wg In I=I o 10 *

~

P=P 10 S (t) TVL = 1.3/p P=P o

et /T HVL = 0.693/u SUR = 26.06/T T = 1.44 DT SCR = S/(1 - K,gg) p SUR = 26

/A8ff ) CR, = S/(1 - K,gg )

-g T = (t*/p ) + [(f 'p)/Aeff) 0 f T = 1*/ (p - Q M = 1/(1 - K,gg) = CR g/CR0

"( ~ 0}! eff 0 M = (1 - K,gg)0 ( ~

eff}1 p = (K,gg-1)/K,gg = AK,gg/Keff

, SDM = (1 - K,ff)/K,ff p= [t*/TKygg ] + [H/(1 + A,ggT )] ,

t* = 1 x 10 -5 seconds

~

P = I4V/(3 x 10 0) A,g = 0.1 seconds I = No Idg1"Id22 WATER PARAMETERS Id g =Id2 1 gal. = 8.345 lbm R/hr = (0.5 CE)/d (meters)

I gal. = 3.78 liters R/hr = 6 CE/d (feet)

I ft3 = 7.48 gal. MISCELLANEOUS CONVERSIONS .

3 10 Density = 62.4 lbm/ft 1 Curie = 3.7 x 10 dps Density = 1 gm/cm 1 kg = 2.21 lbm 3

i Heat of va;orization = 970 Etu/lbm I hp = 2.54 x 10 BTU /hr 6

Heat of fusica = 144 Btu /lbm 1 Mw = 3.41 x 10 Btu /hr 1 Atm = 14.7 psi = 29.9 in, l'g. 1 Btu = 778 ft-lbf 1 ft. H.,0 = 0.4333 lbf/in 1 inch = 2.54 cm l F = 9/5 C + 32

• C = 5/9 ( F - 32)

i SECTION 1 Principles of' Nuclear Power Plant Operation Thermodynamics, Heat Transfer, and Fluid Flow QUESTION 1.01 Which one (1) of the statements below is not correct for a nuclear reactor of the San Onofre Unit 2/3 type (1.0)

a. The product of the macroscopic cross section (epsilon) and the neutron flux (phi) gives the neutron reaction rate (interactions per cm(3) per sec).

b. If the thtrmal neutron flux is doubled in one hour, the thermal power produced in the nuclear reactor is doubled.

c. The nee.troa microscopic cross section (sigma) for a certain element varied with neutron energy and is dependent on the isotope of the element.

d. The thermal-rieutron microscopic fission cross section for U-238 ss larger than that for U-235.

• ANSWER 1.01 The answer is (d) . (1.0)

REFERENCE:

San Onofre Rx Theory Notes Ch. 4

h QUESTION 1.02 List four-(4) of the sir (6) factors which must be considered when calculating the shutdown margin.(2.0)

(assume the Tave greater than 200 degrees). ._

• ANSWER 1.02 . (any 4 of the below, 0.5 each, 2.0 max.)

• Baron concentration.

• CEA position.

• RCS temperature

• Fuel burnup.

• Xe concentration.

• Sm concentration.

• REFERENCE

1. San Onofre 2/3 : Safety Technical specifications, pp.

3/4.1-1 to 3/4.1-2.

?

t 4

1 1

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

, - , . ,,--,,,v.- - . , - - - , , , , . . . - . , - -

• OUESTION 1.03 Consider two(2) reactor startup operations that are identical (all rods in, same boron concentration, same RSC temp,, etc) except that the strength of the neutron source for Startup # 1 as seen on the s.tartup power-level meter i s 25 cps, and for Startup #2 is 100 cps.

a. Does this change affect the critical rod position? (0.5)

Explain your answer. (1.0)

b. Does this change affect the power-level at which the reactor would reach criticality? (0.5)

Explain your answer. (1.0)

• ANSWER 1.03

a. No, (0.5) changing the strength of the neutron source does not affect crit. cal rod position. Changing the source strength does not change the reactivity status of the core. The source multiplication effect remains un-affected. (1.0)

b. Yes, the power level at which criticality is obtained is

, different. If the source strength is greater and if the startup rates are the same (and not instaneous), then there will be more neutrons when criticality is reached.

• REFERENCE (1.0)

San Onofre Rx Theory Notes Ch. 9 1

s

l

• OUESTION 1.04 Assume that with a normal flowrate of the primary coolant, enough head wil-1 be generated to reach the DNB point, when the local heat flux reaches 27 kW/ft.

' 1..

a. If the actual local heat flux is 9 kW/ft, calculate the DNBR.(0.5).

b. If the nuclear-reactor power level is increased from ,

that in part "a.", will the DNDR increase, remain the same, or decrease? (0.5)

• ANSWER 1.04

a. DNBR = maximum critical heat flux / actual heat flux

= 27/9

=3 (0.5)

b. The DNBR will decrease because the maximum local heat flux will increase. (0.5)

• REFERENCE

1. San Onofre HTFF Study Guide Part B Ch.4.

2. " Nuclear Energy Training," Module 4, Plant Performance, pp. 8.2-1 through B.2-4, NUS Training Corporation.

\

i h

i I

• OUESTION 1.05 Answer the following TRUE or False :

a. The production rate of indirect Xenon fromilodine is FASTER than the decay rate of Xenon to Cesium?

(% h i

b. Slowing the rate of a power decrease, lowers the ,

height of the resultant Xenon peak? C.% h'#

c. The resultant Xenon peak from a Reactor Trip from 50% power is larger than 'a trip f rom 100 % power?

w=p> (o.d,

d. During an increase in power from equilibrium Xenon conditions, Xenon concentration initially decreases? -

i .- w r o.s

--7

• ANSWER:

1.05

a. True (&;1*5) b.o'

b. True ( O.-3Ert o ,3

c. Fa1se ( 0-:*5) g,

d. True (&r25r) oe

REFERENCE:

San Onof re R:: Theory Notes Ch.10.D a:J ule . Ab-~ 1 ""^

• QUESTION 1.06 Of the following operations, which ONE will have a negative effect on the available Net Positive Suction Headi.(NPSH) for an ideal centrifugal pump: .

a. Throttling open the pump's suction valve.

b. Throttling open the pump's discharge valve.

c. Decreasing the pump's speed.

d. Decreasing the temperature of the fluid being pumped. (1.0)

• ANSWER 1.06

b. is the answer (1.0)

REFERENCE:

San Onofre HTFF notesSection III Part B

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

• OUESTION 1.07 A tank contains water to a level 40 ft. above the bottom of the tank. A Nitrogen cover gas i s at 100 psia. The tank and its contents are at:

70 degrees F, and the density of the water ist j 62.4 lbin/ cubic f t. The pressure.at the bottom of the tank is:

a. 117 psia.

b. 132 psia. .

c. 208 psia.

d. 308 psia. (1.0)

• ANSWER

, e 1.07

a. 117-psia. (1.0) l 4

4 4

i i

t i

5 I

I i

y ++--yw - - -

y w-e.np - wwg- wts, wo ve+- --+-pw-g w iwymy- r-g y --&-c-r- -$7m-N*et---r -m3Wy**

• OUESTION 1.09 If the temperature of the tank and its contents in Question 1.07 was increased, and no water or cover gas was allowed to enter or leave the tank, the

~

pressure of the tank would

a. increase, because the gas volume has decreased and the temperature of the gas had increased?

b. increase, because the water will expand and the pressure will increase due higher water level (water column effect), and also due to the increase in the gas temperature?

c. decrease, because the water density has decreased?

d. decrease, because the cover gas density has decreased? (1.0)

• ANSWER:

1.08

a. (1.0)

REFERENCE:

San Onofre HTFF NotesSection III Part A I

i t

i I

I

i 4

• QUESTION

, 1.09 a. If the pressurizer pressure equals 2250 psia, what is pressurizer Temperature? (0.5)

b. If reactor power is 100%, what is the degree of subcooling in the RCS hot leg. (0.5)

-

• ANSWER 1.09- a. Refer Steam Tabl es - 653 degrees F (+/-5) (0.5)

b. refer to RCS system descriptions degrees subcooled= Tsat-Thot=653-611=

= 42 (+/-5) degrees (0.5)

REFERENCE:

San Onofre HTFF notes, RCS System Description, Steam tables f

i

)

i f

I

, - - - . - , - . -- .n. . _ , , , - . - . - , . , , . , . . , , , , -, - . _ , - _ -,. .n,.--c,..__,, . , _ , , , . . , . . -,.,,,-.c,n , - - . . . , . ,,

• OUESTION 1.10 a. Explain.how the energy of nuclear fission is transfered tn the reactor coolant. (1.5)

b. Explain why it is necessary to continue coo},ing the core following a shutdown from power operations. (0.5)

• ANSWER 1.10 a. Most of the fission energy (93%) is in the form fission product kinetic energy. The fission particler. are slowed down in tlie fuel matrix by inelastic collisions, and this heats up the fuel matrix (1.0). The fuel heats up the clad by conduction and/or convection, which in turn heats up the cooiant. (0.5) K.E. of neutrons, other gamma 2 and beta energies - insignificant.

b. The major contributor is the heat generated by fission product decay in the fuel matrix - decay heat - is sufficient to damage the core if cooling is not maintained. (0.5)

REFERENCE:

San Onofre Rx Theory notes p.60-69

" " RCS System Description l

SECTION 2 Plant Design Including Sa-fety and Emergency Systems

• QUESTION 2.01 If proper Feed System chemisty is not maintained, it is poussible that copper will plate out on the surface of the convergent tiozzle section of the f eed flow venturi,

a. What will be the effect on indicated feed flow with respect to actual flow if this venturi fouling occurs? Explain . (1.0)

" gp? ain the ef f ect this has on Plant Perf ormance?

(1.0)

• ANSWER 2.01 a. Venturi fouling will cause a slight decrease in diameter of the nozzle and increase head loss.

(increase in f riction f actor) . Both of these effects will cause an increase in the measured D/P. Therefore, the indicated flowrate will be greater than the actual flow rate. (1.0)

b. The erroneously high feed tate will cause the calorimetric to be high. Since the plant is running at less than its capibilities, actual plant performance is decreased. (1.0)

• ref er e.,c e: San Onofre Sys. Description - 21 and S.O.Enam Bank 4

i e

• DUESTION 2.02 While the power' plant is operating at a constant 100%

power ~ level, reactor operator causes an inadvertent boration which causes Tc to drop 10 degrees FL Explain how the following parameters will. change

• Steam Generator temperature, (0,5)

• Steam Generator pressure, (0.5)

• Steam Generator Level (0.5)

• Unit Electrical Output (0.5)

Assume that no operator action is taken and that all systems function normally.

• ANSWER 2.02 .The steam temperature and pressure will decrease (0.5 each, 1.0 max). ( Due to the drop in the Tave,)

As SG pressure drops, steam-generator water level will initially increase (Swell in rizer section), then be returned to setpoint by FWRLS. (0.5)

(As SG pressure drops , steam flow will increase, Governor valves open, as unit load is maintained by turbine control system) unit electrical load will remain constant, and if Control Valve Opening Limit (CVOL) is met, Unit Electrical Load will then drop off. (0.5)

• REFERENCE SD-SO23-1RO Main Turbine System Gystem Description 42, RCS Regulating System System Description 21, Feedwater

. , - - , .-. - - , - - - - - , , -..-- - , ----m , - - ,,

QUESTION 2.03 After a large LOCA:

a. Give two (2) methods of determing hydrogen buildup in the containment.  : _. (1.0)

6. State the purpose of the heat-tracing on PASS Containment gas sample lines. (1.0)

• ANSWER 2.03 a. 1. PASS continuous diluted purge sample.

2. Diluted manual sample from PASS containment sample lines.

3. Undiluted grab sample from purge lines.

(South shield wall)

(any 2, 0.5 ea.)

b. To limit plateout of Radioisotopes, which could result from condensation of c,ontainment vapor.

This could result in misleading sample results and future radiation hazzard for personnel. (1.0)

• REFERENCE San Onofre SD-39 Post Acc. Sampling Sys. ,

i h

• 0UESTION 2.04 During100% equilibrium power conditions, CO2 is inadvertently valve into the main generator instead of M2. ._.

a. How and why would this event affect the generator rotor temperature? (1.0)

b. Using the turbine controls, How can the heat 1 cad of the generator be changed without changing the electrical output power? (1.0)

• ANSWER 2.04 a. Rotor temperature would increase because CO2 is not as good a heat-transfer gas as H 2. (1.0)

b. By using Volts Adjust, and reducing the MVAR loading. (1.0)

• REFERENCE

1. SD-SO23-100 Main Gen. and 22 KV System 1

4

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

• OUESTION 2.05 Assume that the controlling pressure transmitter for the Pressurizer Pressure Control System has failed HIGH, while in service. .

a. What two alarms would you expect to receive in the control room? (1.0)

b. What will the Spray Valve (s) and Heaters do? (1.0)

c. If the operator does not take action, how will the plant be affected? (0.5)

• ANSWER

a.

• PZR Pressure HI/LO (50A14) (0.5)

• PZR Preasure Deviation (50A04) (0.5)

b. Spray valves open and all heaters will be turned off. (1.0)

c. Plant Pressure will decrease until SIAS and reactor trip occurs (0.5)

REFERENCE:

SD - 40 PZR and PZR control Sys.

+ - - - + ~ ,

QUESTION 2.06 a. List four (4) components which could leak potentially radioactive water into the Component Cooling Water '

system (CCW) 7 ( consider both critical & non-

• critical loops) (1.0)

b. What four (4) saf ety systems would be inoperable if the CCW System was inoperable? (1.0)

c. During a normal plant cooldown with RCS temperature

less than 350 degrees F, what is the major heat load of the CCW System? (0.5)

,

• ANSWER 2.06 a. 1. RCP thermal barrier / seal cooler

2. Letdown heat exchanger

3. SDC HX

4. Primary sample coolers

5. waste gas equipment (unit 3 only)

6. boric acid concentrator (unit 2 only)

7. HPSI Shaft Seal cooler (any 4, 0.25 ea.)

b. 1. Containment' air recirculatiun and cooling units

2. Containment spray pump seal coolers - (2)

3. HPSI pump seal coolers - (3)

4. LPSI pump seal coolers - (2)

(any 4, 0.25 ea.)

c. Shutdown-cooling heat exchanger. (0.5)

• REFERENCE

1. San Onofre SD-7 CCW and SD-45, "SI and S/D Cooling System."

I

• OUESTION 2,07 A lift coil on a CEA nas failed such that no electrical current be pass'ed through it.

a. As a result of this failure, will the CEA drop into the core? Explain your answer.  ; (0.5)

b. With this failure present, will the CEA move in response to'a demand signal? Explain your answer. (0.5)

c. When a trip signal has been properly transmitted to a CEA, what prevents the CEA from being ejected? Assume that there exists an upward force on the CEA that is greater than the weight of the CEA. (0.5) 4

• ANSWER 2.07 a. The CEA will not drop due to the holding action of the upper gripper. (0.5)

b. The CEA will not move upward or downward because the lift coil must be energized to raise the upper gripper. (0.5)

c. The anti-ejection gripper. (0.5)

• REFERENCE

1. San Onofre SD-14 CEDM Control SYS.

T

• OUESTION 2.08 How does the des'ign of the Chemical and Volume Control System minimi=e the potential for thermal transients, to the pressurizer, when Au>tiliary Spray is inLuse? (0.5)

• ANSWER 2.08 Aux. Spray is warmed by the Regenerative heat exchanger.

(Normal Aux Spray thru HV-9201, Manual bypasses RHX)

REFERENCE:

San Onofre SD - 4 CVCS t

r-SECTION 3 Instruments and Controls .

• OUESTION 3.01 The Steam Bypans Control System (SBCS), in addition to controlling bypass and atmospheric dump valves, produces control signals (inhi bi ts, permissives or demand signal s) to other equipment. For both of the f ollowing situations indicate; 1. The control action; 2. The system to which the signal is sent; 3. The purpose of the actions

a. Load rejection from 14% power. (0.75)

b. Steam Bypass demand is present. (0.75)

• ANSWER 3.01 a. 1. Prevents CEA's from being automatically withdrawn or inserted CO.253 (AMI) (AWP signal also produced) 2.

CEDMCS [O.253 3. Allows a quick reloading of the turbine / generator if the loss of load is due to a temporary fault.[0.253. (0.75)

b. 1. Prevents CEA's from being autowithdrawn (AWP) EO.25].

2. CEDMCS [0.25] 3. Prevent raising Reactor power since there is e>: cess NSSS energy. CO.25]

• REFERENCE SBCS System Description 42.

f

• DUESTION 3.02 Given reactor power at 100% with Pressurizer Pressure control system (PPCS) 'and Pressurizer Level control system (PLCS) are in automatic, Tc fails low with Reactgr Regulating selected to T average. There are four control act i ons and alarms associated with the PPCS and PLCS, what are they? (2.0)

• ANSWER 3.02 1. With low Tc, setpoint program goes to minimum (*30%)

which produce a high level error alarm. [0.53

2. Ma>:i mum l etdown f l ow. [0,53

3. Energize backup heaters on high level error if less than 2275 psia. [0.53

4. Stops normal running charging pump.[0.53 (2.0)

• REFERENCE Pessurizer and Pressurizer Control System system description 40.

- - - m - ~ , , ,

r

• DUESTION 3.03 Concerning the Core Protection Calculators (CPC's): One of the inputs to the CPC calculators is neutron power. -

What THREE corrections does this power signal undergo?

(1.5)

• ANSWER 3.03 1. Shape annealing (0.5)

2. Rod shadowing (0.5)

3. Temperature shadowing (0.5)

• REFERENCE Reactor Protection System and Core Protection Calculators system description 43.

l i

i i

• I I

l l

. _ , . . . _ . . . _ _ . . . _ _ . . . . . - . - . . _ , , , , , , r_, _r. - . _,._y

I

• OUESTION-3.04 TRUE/ FALSE -

Tne Geiger Mueller (GM) radiation detector;

a. Uses a Boeon Trifluoride (BF3) coating inside the detector with an argon quench gas. (0.5)

b. Has a short period during its operation where no radiation may be counted (dead time). (0.5)

• ANSWER 3.04 a. False (0.5)

6. True (0.5)

• REFERENCE Glasstone and.Sessonske o

l i

l f

5 i

L-

~

1

• OUEST10N .

3.05 During normal 100 p'ercent power operation, explain how ,

the f ailure of the middle channel NI 75 percent low (reading only 1/4 the correct value) will affect;Will affect RPS channel "A". (1.5)

• ANSWER 3.05 A failed middle NI will cause the CPC to estimate a

~

very high peaking factor EO.53 (the CPC takes three (3) readings and' produces 20 nodes). The failed NI will not affect the total power calculator EO.53. The failed NI will cause a DNBR and LPD trip [O.53. (1.5)

• REFERENCE ,

Reactor Protection System and Core Protection Calculators system description 43. pages 90-92.

4

, - - - - - - , , . , , , ,- - v -

,r, , - , , .,n -, - ---n- , --- --- -. e - ,. - - , - -

[

• QUESTION

.3.06 What THREE (3) conditions.will give you a CEA withdrawal prohibit (CWP) with Reactor power > 1 */.? (1.5)

~

• ANSWER 3.06 1. High pressurizer pressure pretrip (0.5)

2. High local power density pretrip (0.5)

3. DNBR pretrip (0.5)

• REFERENCE CEDMCS system desription 14, page 37.

d

+

y

f

/ g#* M

• OUE ON

~.07 After an extended period of operation copper plates out on the surface of the convergent nozzle section of the feedwater flow venturi due to improper feedwater. system chemistry,

a. What will be the effect on indicated feedwater flow with respect to actual flow due to the fouling of the venturi?

(1.0)

b. What effect would this have on plant performance? (1.0)

• ANSWER 3.07 a. The (ouling of the venturi will cause a slight decrease in diameter and a larger increase in the friction factor.

[0.53 Both of these effects will cause an increase in the detected D/P. Therefore the indicated feed flow rate will be greater than the actual feed flow. [0.53 (1.0)

b. The higher than actual feed flow indication results in a calorimetric that is erroneously high. [0.53 This will cause plant performance to decrease as actual power output decreases. [0.53 (1.0)

(Credit for part b. will be given for the inverse if the candidate determines from a. that feed flow rate is less than actual.)

• REFERENCE RO Requalification Exam, 04/18/86, Question 3.10.

r.

P

• CUESTION 3.'08 Operating Instruction 5023"3-1.1 " Reactor Startup" requires that CEA withdrawal be periodically stopped and a wait until what can be determined? (1.0) _

• ANSWER 3.08 For determining Nuclear Instrumentation trends. (1.0)

-* REFERENCE SO23-3-1.1, PAGE 12.

i-Y i

i

F

• OUESTION 3.09 How would the Steam Generator level change (increase, decrease, or remain the same) if the controlling steam flow signal increased with no change in actual steam f. low ?

Explain your answer. (1.5)

• ANSWER 3.09 Increase CO.53. During the rate of change between steam flow and feed flow there i s an error signal generated, this is removed when the rate of change has ceased. [O.53 After this has occured the the level error signal takes over and restores the level to setpoint. [O.53 (1.5)

• REFERENCE System Description 21, "Feedwater Regulating System".

1 e

i i

1

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

F

• QUESTION 3.10 Regarding the Feedwater Regulating System:

a. What automatic actions occur as a result of a:High Level Override (HLO)? (1.0)

b. What is the High Level Overrride designed to accomplish or prevent? (0.5)

• ANSWER 3.10 a. The HLO closes the feedwater regulating valve and bypass E0.53. And it sets main feedwater pump speed setpoint to minimum. [0.53 (1.0)

b. The HLO prevents moisture carryover to the main turbine.

(0.5)

• REFERENCE Feedwater RegulsSing System description.

END of SECTION 3, CONTINUE ON NEXT PAGE: SECTION 4

r i SECTION 4 Procedures - Normal, Abnormal, Emergency and Radiological Control

• QUESTION 4.01 Answer the following questions TRUE or FALSE in reference to the Reactor Startup procedure.

a. Positive reactivity additions by more than one method is acceptable below the point of adding heat. (0.5)

b. The reactor operator can exceed 2.5 DPM startup rate only with the permission of the shift supervisor. (0.5)

c. Criticality must be anticipated any time CEAs are being withdrawn or boron dilution operations are being performed.

(0.5)

d. Greater than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to achieving Reactor criticality the predicted CEA position must be verified to be greater than the transient insertion limits of Technical Specification figure 3.1-2. (0.5)

• ANSWER 4.01 a. False (0.5)

b. False (0.5)

c. True (0.5)

d. False (0.5)

• REFERENCE SO23-3-1.1 page 10.

l l

t I

l

• OUESTION 4.02 a. What are you required to do if you suspect that you have exceeded an administrative exposure limit? (0.5)

b. Who's approval is required for you to exceed hhCH of the following limits?

1. >1250 mrem / quarter (0.33)

2. >2500 mrem / quarter (0.33)

3. >4500 mrem / year (0.33)

• ANSWER 4.02 a. Immediately report to the radiation protection office. (0.5)

b. 1. Division Manager (0.33)

2. Division Manager (0.33)

3. Station Manager (0.33)

• REFERENCE SO123 VII-4.8

T

• OUESTION 4.03 Give the reason (s) for the'following " limitations and pre-cautions".

a. The reactor shall not be made critical if RCS temperature is less than 552 deg. F. (TWO required) (1.0)

b. RCS pressure shall not exceed a presuure of 375 psia when Shutdown Cooling is in service.(0.5)

c. Maximum containment pressure shall be less than 1.5 psig. (0.5)

• ANSWER 4.03 a. 1. Ensures moderator temperature coefficient is within its analyzed temperature range. 2, The protective instrumentation is within normal operating range. 3. To ensure consistency with the FSAR safety analysis. 4. The reactor vessel is above its minimum RTNDT temperature.

(1.0)

b. Overpressure protection of SDCS components. (0.5)

c. Insures that containment peak pressure does not exceed design pressure of 60 psig during events analyzed in the FSAR. (0.5)

• REFERENCE TS Bases 3/4.1.1.4 and 3/4.6.1.4 SO23-5-1.3, S023-3-1.1 1

1

.r

• DUESTION 4.04 During a Steam Generator tube-rupture event, what is the concern with the use of the steam-driven auxi1iary Feedwater pump? (1.0)  :-

• ANSWER-4.04 The use of steam-driven Auxiliary Feedwater pu.np woul d result an unmonitored CO.53 release of radioactivity.

[0.53. (1.0)

• REFERENCE SO23-12-4 i

e

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

• OUESTION 4.05 Procedure 5023-5-1'.3 " Plant Startup From Cold Shutdown to Hot Standby" has criteria as to when Shutdown Cooling can be secured in Mode % prior to Mode 4 entry. Suppl.y the indicated values for the followings

a. How many RCP's must be operating? (0.5)

b. What level must be present in the Steam Generators?

(0.5)

c. The heatup rate must not be greater than how many degrees Fahrenheit per hour? (0.5)

• ANSWER 4.05 a. 1 pump (0.5)

b. 50% (0.5)

c. 25 degrees F/ hour (0.5)

• REFERENCE SO23-5-1.3 s

r l

• OUESTION 4.06 List four of the six sympt'ms o for which emergency baration is mandatory in accordance with S023-13-11 " Emergency Boration of the Reactor Coolant System." (2.0) :-

• ANSWER 4.06 . Two or more regulating or shutdown CEAs have not dropped into the core following a reactor trip signal. (0.5)

. Unanticipated reactor cooldown has been initiated.

(0.5)

. One or more regulating groups of CEA's is below the PDIL. (0.5)

. Shutdown Margin < 5.15% delta K/K, Tave >200 deg. (0.5)

. Shutdown Margin < 3.0% delta K/K, Tave <200 deg. (0.5)

. Mode 6 Keff > 0.95 or Baron concentration < 1720 ppm (0.5)

• REFERENCE SO23-13-11, page 2, " Emergency Boration of RCS".

• OUESTION 4.07 Define, and, or describe the following Emergency Action Levels. Include reference to plant, and, or system degredation and radiation releases with each response.

a. Alert. (1.0)

b. Site Area Emergency. (1.0)

• ANSWER 4.07 a. Events which involove degradation of the level of safety of the plant. [0.53 Any releases are expected to be limited to small fractions of the EPA PAG's exposure levels. [0.53 (1.0)

b. Events which involve major failures of plant functions needed for protection of the public. [O.53 Any releases are not expected to exceed the EPA PAG's exposure levels, e:: cept near the site boundary. [0.53 (1.0)

• REFERENCE RO Requalification Exam, 04/18/86, Duestion 4.07 Emergency Plan Summary, Section 3, pages 3-4.

e t

. _ _ _ . . , ._ .- m , .y - ,

• DUESTION 4.08 At the moment a bubble is f ormed in the Pressurizer, how should each of the following parameters respond (increase, decrease, remain the same)  :.

a. The letdown flowrate. (0.5)

b. The Prescurizer pressure. (0.5) c.-The Pressurizer temperature. (0.5)

d. The Pressurizer level. (0.5)

• ANSWER 4.08 a. increase (0.5)

b. increase (0.5)

c. stable (0.5)

d. decrease (0.5)

• REFERENCE SO23-5-1.3, page 16.

1 i

l l

A

=

• QUESTION 4.09 Specify the letter designation of the most correct phrase with which to complete the following sentence concerning the Pressurizer Pressure Control System. To increase the heat supplied by the proportional heaters with the manual / automatic station for the proportional heaters and spray in the manual mode, you must... (1.0)

a. ... increase the controller output,

b. ... adjust the pressure setpoint higher.

c. ... decrease the controller output.

d. ... adjust the pressure setpoint lower.

• ANSWER 4.09 c. (1.0)

• REFERENCE SO23-5-1.3, page 16.

End of SECTION 4 END OF EXAMINATION

• END 1

4

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

e-

- j

, l Je U.S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION l

REQUALIFICATION Facility: SAN ONOFRE 2/3 Reactor Type: cv sy m y

/- , , c , -f c . Date Administered: _T11NF 17. 1986

%. Examiner: c. JonssTon

- -- T 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 quettion are indi-cated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.

% of Category  % of Candidate's Category Value Total Score Value Category 1s.0 25.0 5. Theory of Nuclear -

• Power Plant Operation, Fluids, and Thermo-dynamics 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 Pro-cedures, Conditions, and Limitations 60.0 Totals ~

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

Candidate's Signature

~.

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

. , , . _ . _ _ _ , , _ , , _ ....-,.m_., ,,_, -.,_._ __,,,-,_ ,,, ..-

a e'

, EQUATION SHEET f = ma v = s/t C w = mg a=vt+ at 2 Cycle efficiency = t)

E = mC a = (vg - v 9)/t KE = my vg A = AN A=Ae g ~U

= v, + a t PE = agh m = 0/t A = In 2/tg = 0.693/tg W = VAP AE = 931Am t g(eff) = (t1 )(ts) i (tg+t) 3 Q=$ CAT p I=Ie -Z*

Q = U$AT y ye -ux Pwr = Wg In I=I

~

10 *

~ o P=P 10 SUR(t) TVL = 1.3/u P=P o

et/T HVL = 0.693/u SUR = 26.06/T T = 1.44 DT SCR = S/(1 - K,gg)

[X*gg )0 CR SUR = 26 g_ = S/(1 - K,gg )

T = '(1*/p ) + 1(

~

eff}1 " 2(1 ~ K,gg)2

{(f 'p)fx g,p]

T = 1*/ (p - D M = 1/(1 - Keff) = CR /CR0 g

~ D)! eff P M = (1 - K gf)0 (1 - K,gg)g P " ( eff-1)/K,gg = AK,gg/Keff SDM = (1 - Keff)/Keff p= [1*/TKfgg -] + [H/(1 + A,ggT )] ,

t* = 1 x 10~ seconds

-I P = I4V/(3 x 1010) A,gg = 0.1 seconds i I = No Id yy =Id 22 WATER PARAMETERS Idy =Id 2 1 gal. = 8.345 lbm R/hr = (0.5 CE)/d (meters)

I gal. = 3.78 liters R/hr = 6 CE/d (feet)

I ft = 7.48 gal. MISCELLANEOUS CONVERSIONS .

10 f Density = 62.4 lbm/ft 1 Curie = 3.7 x 10 dps l Density = 1 gm/cm 1 kg = 2.21 lbm Heat of varorization = 970 Etu/lbm I hp = 2.54 x 10 3 BTU /hr 6

, Heat of fusica = 144 Btu /lbm 1 Mw = 3.41 x 10 Btu /hr

, 1 Atm = 14.7 psi = 29.9 in. l~g. 1 Btu = 778 ft-lbf 2

l 1 ft. H 2O = 0.4335 lbf/in 1 inch = 2.54 cm l F = 9/5 C + 32 C = 5/9 ( F - 32)

6 SECTION 5 Theory.of Nuclear Power Plant Operation, Fluids, and Thermodynamics

• OUESTION 5.01 State whether the situations listed below will generate the greatest tensile stress ont the INNER or OUTER wall of the reactor vessel,

a. Heatup at a rate of 80 degrees F/ hour (0.5)

b. Cooldown at a rate of 50 degrees F/ hour (0.5)

• ANSWER

a. Outer (0.5)

b. Inner (0.5)

REFERENCE:

San Onofre HTFF Notes Sect III and Nuclear Rx Eng -

Glasstone; 1967.

i

=

• QUESTION S.02 List 4our (4) of the six (6) f actors which tmist be c o n s. ' red.when calculating the shutdown margin.(1.0)

( a ss- the Tave greater.than 200 degrees).

• ~

• ANSWER 5.02 . (any 4 of the below, 0.25 each, 1.0 max.)

• Baron concentration.

• CEA position.

• RCS temperature

• Fuel burnup.

• Xe concentration.

• Sm concentration.

.

• REFERENCE

1. San Onofre 2/3 : Safety Technical Specifications, pp.

3/4.1-1 to 3/4.1-2.

i

• QUESTION 5.03 Consider two(2) reactor startup operations that are identical (all r'od s in, same boron concentration, same RCS temp., etc) except that the strength of the neutron source for Startup # 1 as seen on the startup power-level meter is 25 cps, and for Startup #2 is 100 cps.

a. Does this change affect the critical rod position? (0.5)

Explain your answer. (0.5)

b. Does this change affect the power-level at which the reactor would reach criticality? (0.5)

Explain your answer. (0.5)

• ANSWER 5.03

a. No, (0.5) changing the strength of the neutron source does not affect critical rod position. Changing the source strength does not change the reactivity status of the core. The source multiplication effect remains un-affected. (0.5)

b. Yes, the power level at which criticality is obtained i s different. If the source strength i s greater and if the startup rates are the same (and not instaneous), then there will be more neutrons when criticality is reached.

(0.5)

• REFERENCE San Onofre Rx Theory Notes Ch. 9 d

., . _ _ - - - _- . . . . ,,,.,-,3-. e-.- ,_.,_m ,,.,-.4 _--- - , _ ..n. .r _ .

_e. ,

J

• DUESTION 5.04 Assume that with a normal flowrate of the primary coolant, enough head wil1 be generated to reach the DNB point, when the local heat flux reaches 27 kW/ft.

a. If the actual local heat flux is 9 kW/ft, calculate the DNBR.(0.5).

b. If the nuclear-reactor power level is increased from that in part "a.", will the DNBR increase, remain the same, or decrease? (0.5)

• ANSWER 5.04

a. DNBR = maximum critical heat flux / actual heat flux

= 27/9

=3 (0.5)

b. The DNBR will decrease because the maximum local heat flux will increase. (0.5)

• REFERENCE

1. San Onofre HTFF Study Guide Part B Ch.4.

2. " Nuclear Energy Training," Module 4, Plant Performance, pp. B.2-1 through 8.2-4, NUS Training Corporation.

l

• QUESTION 5.05 Answer the following TRUE or False :

a .' The production rate of indirect Xenon from Iodine is FASTER than the decay rate of Xenon to Cesi_um? (0.25)

b. Slowing the rate of a power decrease, lowers the height of the resultant Xenon peak? (0.25)

c. The resultant Xenon peak from a Reactor Trip from 50% power is larger than a trip from 100 % power?

(0.25)

d. During an increase i n power from equilibrium Xenon conditions, Xenon concentration initially decreases?

(0.25) t

• ANSWER:

5.05

a. True (0.25)

b. True (0.25)

c. False (0.25)

d. True (0.25)

REFERENCE:

San Onofre Rx Theory Notes Ch.10.D I

e

, . - , - y  % . < , - - ,+y,- - ,, , - - - , v. ---,-,---,-<.-~-e s-- , , e

1

• QUESTION 5.06 A variable speed centrifugal + pump is operating at 1800 rpm with a capacity of 400 gpm at a discharge head of 20 psi, and requires a power of 40 kW. If pump speed is ingreased to 2000 rpm, which of the following best describes the new pump parameters?

a. 444 gpm, 22 psi, 55 kW

b. 444 gpm, 25 psi, 49 kW

c. 444 gpm, 25 psi, 55 kW

d. 444 gpm, 22 psi, 49 kW

• ANSWER 5.06 c. (1.0) 2000/1800 = 1.11 power = (1.11) cubed x 40 = 54.76 Head = (1.11) squared x 20 = 24.64

REFERENCE:

San Onofre HTFF notesSection III Part B l

l 1

1 1

t

• QUESTION 5.07 A tank contains water to a level 40 ft. above the bottom of the tank. A Nitrogen cover gas i s at 100 psia. The tank and i ts contents are at:

70 degrees F, and the density of the water is:

62.4 lbm/ cubic ft. The pressure.at the bottom of.the tank i s:

1

a. 117 psia.

b. 132 psia.

c. 208 psia.

d. 303 psia. (1.0)

• ANSWER 5.07

a. 117 psia. (1.0)

REFERENCE:

San Onofre HTFF notes Sect.III Part A.

?

I l

I i

i l

l 4

4

- , --+-.y, , , - - . - - . . ,--,7-- , - - - ,---c- ,

~

l 4

. l s

• DUESTION 5.08 If the temperature of the tank and its contents in Question 5.07 was increased, and no water or cover gas was allowed to enter or leave the tank, the pressure of the tank would: _

a. increase, because the gas volume has decreased and the temperature of the gas had increased?

b. increase, because the water will e>:pand and the pressure will increase due higher water level (water column effect), and also due to the increase in the gas temperature?

c. decrease, because the water density has decreased?

d. decrease, because the cover gas density has decreased? (1.0)

• ANSWER:

5.08

a. (1.0)

REFERENCE:

San Onofre HTFF NotesSection III Part A

Q:

.- *OUESTION l 5.09 a. If the pressurizer pressure equals 2250 psia, what i s pressurizer Temperature? (0.5)

J

b. If reactor-power is 100%, what is the degrge of subcooling in the RCS hot leg. (0.5)

• ANSWER 5.09 a. Refer Steam Tables - 653 degrees F (+/-5) (0.5)

l b. refer to RCS system descriptions degrees subcooled= Tsat-Thot=653-611=

= 42 (+/-5) degrees (0.5)

REFERENCE:

San Onofre HTFF notes, RCS System Description, Steam tables I

5 4

1 I

i i

i 1

! b 4

L

--, , ---,r- - - . . - - , ,.,,-,,,w,,-,,,,-_,,,.+gn.,,,n_,,.e, ,, ,,,ay,,----g.-y,-..--g- - , - . , , .. . , .

• QUESTION 5.10 a. Explain how the energy of nuclear fission is transfered to the reactor coolant. (1.5)

b. Explain why it is necessary to continue cooling

-the core following a shutdown from power operations. (0.5)

• ANSWER 5.10 a. Most of the fission energy (93%) is in the form fission product kinetic energy. The fission particles are slowed down in the fuel matrix by inelastic collisions, and this heats up the fuel matrix (1.0). The fuel heats up the clad by conduction and/or convection, which in turn heats up the coolant. (0,5) K.E. of neutrons, other gamma and beta energies - insignificant.

b. The major contributor is the heat generated by fission product decay in the fuel matrix - decay heat - is sufficient to damage the core if cooling is not maintained. (0.5)

REFERENCE:

San Onofre Rx Theory notes p.60-69 RCS System Description

9

• QUESTION 5.11 Which of the following is a true statement, if the Safety Channels of NI's were adjusted based on a calculated calorimetric with the given errors, (1.0)

a. If feedwater temperature used.in the calchimetric calculation was lower than actual feedwater temperature, actual power will be higher than indicated power.

b. If reactor coolant pump heat input used in the calorimetric calculation was neglected, actual power will be greater than indicated power.

c. If the steam flow used in the calorimetric calculation was lower than actual steam flow, actual power will be less than indicated power.

d. If one of the Steam Generators were omitted from the blowdown heat rate calculation, the actual reactor power will be greater than indicated power.

d

• ANSWER 5.11 The answer is d. (1.0)

(D= mCp (Ts - Tfw) - (RCP heat) + (B/D heat rate)

REFERENCE:

NUS Trn'g notes Module 4, 5023-3-3.2, S023-3-3.38 SO HTFF notes Sections II & III.

o

• QUESTION 5.12 The reactor is operating- at 50% power with rod control system in manual when a single group A rod drops into the core. Assuming no reactor trip or operataq_ actions occur, choose the answer that best. describes the final steady state conditions.

a. Final power = initial power, Tave less than initial Tave, S/G pressure greater than initial S/G Pressure.

b. Final power = initial power, Tave less than initial Tave, S/G pressure less than initial S/G pressure.

c. Final power less than initial power, final Tave greater than initial Tave, S/G pressure greater than initial S/G pressure.

d. Final power = initial power, final Tave greater than initial Tave, S/G pressure = initial S/G pressure. (1.0)

• ANSWER 5.12 a. (1.0)

REFERENCE:

San Onofre Rx Theory notes Ch. 9

e

• OUESTION 5.13 The largest contribution of hydrogen release to the containment during an accident involving in' adequate core cooling and reactor vessel void formation is from:

a. a zirconium - steam reaction,

b. an aluminum - steam reaction.

c. the release of disolved hydrogen in the coolant and from the hydrogen overpressure on the VCT.

d. radiolysis of the coolant. (1.0)

.

• ANSWER 5.13 a. (1.0)

REFERENCE:

1. San Onofre Rx Theory notes.

2. San Onofre FSAR

3. CE Degrated Core Perspectives Notes.

4. TM1, Report to Comm. Vol II, Pt.2,p.I-5.27 END of SECTION 5, CONTINUE ON NEXT PAGE: SECTION 6

e SECTION 6 Plant Systems Design, Control, and Instrumentation

• QUESTION 6.01 The Safety Injection System and Shutdown. Cooling System are for providing cooling water to the reactor core in the event of an accident. The basis for the design was to meet the acceptance criteria as outlined in 10CFR 50.46. State i three of the acceptance criteria. (1.5)

• ANSWER 6.01 1. PCT less then 2200 degrees F 2. Maximum cladding oxidation 3. Maximum hydrogen generation 4. Maintaining coolable geometry 5. Provide long term cooling (0.5 each)

• REFERENCE SD 46' SI/SDC i

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

a

• DUESTION 6.02 During an emergency start (SIAS or LOVS) of the diesel generator, what trips remain functional? (1.5)

• ANSWER 6.02 1. Engine overspeed (0.5) 2. Low lube oil pressure (0.5)

3. Generator differential (0.5)

• REFERENCE SD 16, page 29.

a

• QUESTION 6.03 During normal Mode 1 Chemical and Volume Control System lineup, initially, if the charging line loop isolation valves 92O3 and 9202 are shut without securing letdown, the subsequent effects in the Chemical u and Volume Control System may cause overpressurization of the shell side of the Regenerative Heat Exchanger.

a) With no operator action, what mitigates these effects?

(1.0) b) What operator action could be taken to mitigate the effects? (0.5)

• ANSWER 6.03 a) Spring loaded thermal relief to loop 1A (XCV-9216) (1.0) b) Isolate letdown (0.5)

• REFERENCE SD 4, page 27.

• OUESTION 6.04 With respect to the EFAS, which steam generator (s) will receive emergency feed (1,-2, both, or neither) given the f ollowing plant indications? Explain in each case.

a) SG 15% level, 650 psig SG 20% level, 750 psig (1.0) b) SG 15% level, 650 psig SG 20% level, 610 psig (1.0)

• ANSWER 6.04 a) #2 only (0.5) With steam generator level below 23%, feed is initiated unless pressure is below 729 psi. (0.5) b) Neither (0.5) Since both steam generators are less then 729 psi and do NOT satisfy a 50 psid condition to initiate feed to the steam generator with the higher pressure.(0.5)

• REFERENCE SD 2, page 11.

E

. . . , - _ . - - , - , _ . _ _ , _ _- ________.,.,_..._,_,-._m..,m_....m.,-, , . _ _ . , - - _ - - . , , , - - , , , _ . , _ . . . . _ - . , , , - . . _ , - . - ,__.m,- , .-.

7

.s.

• QUESTION l 6.05 a) Describe the condition for the main generator  !

anti-motoring trip to occur. (0.5)

I b) What, specifically, does the Generator Differential trip protect the Main Generator from? (0.5) t

• ANSWER 6.05 a) Trip occurs when a flow of power goes into the generator (equivalent to 3.3 MWe at a PF = 1.) (0.5) ,

b) Phase to phase faults. (0.5)

• REFERENCE SD 32, pages 20, 22.

- _ _ - = _. __ _

k

• OUESTION 6.06 List the process radiation monitors that aree utilized to indicate a primary-to secondary leak.(1.5)

• ANSWER 6.06 1. Main Steam line 2. Condenser Air Ejector 3. M team Generator Blowdown processing system(0.5 ea)

• REFERENCE SDs 41 and 50.

1 l

1 i

i b i

4 i

I J

'I

d

• DUESTION 6.07 The pressuriz.er pressure controller provides singular control of heaters'and spray.

a. How is the output of the pressurizer pressure, master ~~

controller determined when in

1. AUTOMATIC (0.5) 2. MANUAL (0.5)

b. As the controller output increases from 0% to 66.7%,

what takes place to control pressure? (Controller in AUTD)

(0.5)

c. As the controller output increases from 66.7% to 100%,

- what takes place to control pressure? (Controller in AUTO)

(0.5)

• ANSWER 6.07 a.1. Difference between the setpoint (0.25) and the measured press (0.25) of the pressurizer.

a.2. Difference between the setpoint (0.25) and the value set by the operator (0.25).

b. The power supplied to the proportional heaters will go from full power (0.25) to zero power (0.25).

c. (The output of the sprcy valve controller will increase from 0% to 100%) The spray valve goes from full closed (0.25) to full open (0.25).

• REFERENCE

, SD 40, page 13.

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

O E

• DUESTION ,

6.08 Four Safety Injection Tanks (SIT) rapidly flood the core following inadvertent depressurization of the Reactor Coolant System (RCS).

a. How are the SIT 's physically i sol ated f rom thd~RCS?

(0.5) ,

b. How is it ensured that the SIT's will not become '

isolated from the RCS during at power operatiens?(0.5)

c. What two automatic control features are provided to prevent the SIT's fru. being isolated from the RCS dAring power and even during plant cooldown? (1.0)

• ANSWER 6.08 a. A pair of series check valves (held shut by RCS pressure) (0.25) and a motor operated isolation valve.

(0.25)

b. Administrative controls: Motor operated isolation valve is locked open, power is removed and tagged out.- (0.5)

c. 1. An automatice open signal is generated for the motor operated i, solation valve if RCS pressure is greater than 500 psig (i ncreasi ng ) . (0.5)

2. An interlock for the motor operated isolation valve, prevents the valve from being closed if RCS pressure is greater than 376 psig (decreasing). (0.5)

• REFEREi .CE :

Sd 46.

• OUESTION 6.09 In reference to the Low Steam Generator Flow Trips

a. What does the Low Reactor Coolant Flos trip protect the plant from? (1.0)

b. What condition i nitiates the trap? (1.0)

• ANSWER 6.09 a. Reactor coolant pump (RCP) shaft shearing (1.0)

b. When del ta P across primary side of either S/G goes below a variable setpoint. (1.0)

• REFERENCE SD 43.

END of SECTION 6, CONTINUE ON NEXT PAGE: SECTION 7

,-e, - - , ~,mw,wwmwww- -,w wy we-w, y , ~w ww-w w w- w -+,q,"y,rowww'w ,--weve-m -m,-y-9-,ewwe---sm-s--re x, w w~~e'-mw*--r*w'V==~7 -='--"-**=M*'W---W'"**~r-~ *" "M*w' * * - ~ '

SECTION 7 Procedures - Normal, Abnormal, Emergency"and Radiological Control

• OUESTION 7.01 Assume that the scenario described below has occurred for both condtions a) and b). Designate the most probable cause by choosir.g from the list of possible causes. Consi der each of the four conditions separately.

SCENARIO: The power plant has been operating at 100 percent of full power of two (2) months. All four

, (4) RCPs trip, and after 15 minutes your operators inform you that retural circulation flow has NOT developed.

a) The wide-range levels for both Steam Generaters are indicating less than O percent and Tcold is increasing above the Tsat for the Steam Generators. (1.0) b) The Pressuriser level is low, and there is a mismatch oetween the loop RTDs and the core enit thermocouples. (1.0)

I Con e s e voids developing in RCS low path.

2. Failure of both Main Feedwater Pumps.

3. The Auxiliary Feedwater Pumps are operating at maximum flowrate.

4. Inadequate secondsry steam flowrate.

5. Pressurize- relie4 valves stuck closed.

6. Inadequate RCS inventory.

7. More than one CEA stuck in the out position.

8. Steam-line rupture on Steam Generator "A".

• ANSWER 7.01 a) 2. (1.0) b) 6. (1.0) e

• REFERENCE (S):

1. S023-12-1, page 15.

F

• QUESTION 4 7.02 At the moe.ent a bubble is. formed in the Pressurizer, how should each of the following parameters respond?

1._

a) The letdown flowrate (0.5) b) The pressurizer pressure (0.5) c) The Pressurizer temperature (0.5) d) The Pressurizer level (0.5)

• ANSWER 7.02 a) increase (0.5) b) increase (0.5) l c) stable (0.5) d) decrease (0.5) 4

• REFERENCE (S):

j 1.- SO23-5-1.3, page 16.

i 2

i i

n

(

4 l

}

l

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

• OUESTION 7.03 Concerning the Pressurizer Pressure Control System, choose the most correct statemen,t:

To increase the heat supplied by the proportional heaters with the manual / automatic station for th'e Pressurizer heaters and spray in the Manual Mode, you must ... (1.0) a) increase the controller output.

b) adjust the pressure setpoint higher.

c) decrease the controller output.

d) adjust the pressure setpoint lower.

• ANSWER c) (1.0)

• REFERENCE (S):

1. 8023-5-1.3, page 16.

• OUESTION 7.04 In accordance with 5023-12-9 " Reactivity-Priority 1", what ,

are the four react'ivity control success paths?-(2.0)

• ANSWER *~

7.04 1. CEA1 trip. (0.5) p 2. Borate via CVCS. (0.5)

' 3. ' Borate via ECCS. -( 0. 5)

4. CEA insertion. (0.5) ,

• REFERENCE (S):

1. SO23-12-9, pages 2,2,9,13.

1 4

lh 4

1 i

}

i r

• DUESTION 7.05 In preparation for reactor plant heat up, RCP POO2 has been started. If both RCP POO4.and RCP POO1 had tripped on their initial start, what should be the pump starting sequence to establish three running pumps in the shortest time if all subsequent RCF' starts are successful? (RCP POO3 will not be available until RCS temperature is greater than 500 Fahrenheit.)

(1.0)

• ANSWER 7.05 1. Start RCP POO1 then RCP POO4 (1.0)

• REFERENCE (S):

- 1. So23-3-1.7 page 4.

2. 5023-5-1.3 page 10, step 4.5.5.

4 f

i

• OUESTION 7.06 Identify four sources of liquid effluent, besides the Steam Generator blowdown processing system neutalization sumps (BDP), for which a batch or continuous liquid effluent release permit would normally be generated. (1.O)

• ANSWER 7.06 1. Primary plant makeup storage tanks.

2. Radwaste primary tanks.

3. Radwaste secondary tanks.

4. Miscellaneous waste condensate monitor tanks.

5. Other (Any 4 (0.25) each)

.

• REFERENCE SO123-III-5.11.23.

l t

I l

i

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

4

• QUESTION 7.07 If the plant was operating i n Mode 1 and instrumentation in the Control Room indicated rising Reactor Coolant Pump thrust bearing temperature in all 4 pumps in excess of 225 Fahrenheit:

w

a. What action must be taken immediately? (1.0)

b. What condition would have-caused this event? (0.5)

• ANSWER 7.07 a. Trip reactor and RCP's carry out SO23-12-1 actions.

(1.0)

b. CCW flow lost to bearings. (0.5)

I

• REFERENCE (S):

1. 5023-13-7, page 4.

N 4

4 i

i I O I

-* - , - - , ,,,n-, , - - , - - , - - e -- -----n- ,,--,m---,.,,n _m,.-- - ~ . , ~ , - , - - - - - , - - , . - - ,,.----,w.,, .- g

1

a e

• QUESTION 7.08 State if and.why each of the following events would or would not require' suspension of movement of irradiated fuel inside the containment.

a) A maintenance person opens the i.nner door o(~the

> air lock to exit the containment. The outer door of the air lock is shut. (0.66) b) Securing of the operating shutdown cooling loop which leaves both loops secured but operable.

(0.66) c) The damper of a spent fuel ventilation exhaust fan fails shut due to a loss of instrument air to the damper. (0.66)

• ANSWER 7.08 a) Would not (+0.33) only one door of the air lock is required to be closed (+0.33).

b) Would not (+0.33) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> without SDC is allowed I by Technical Specifications (+0.33) c) Would not (+0.33) one exhaust fan would still be operable and that is all that is required

(+0.33).

• REFERENCE (S):

1. Technical Specification, 3/4.9.12, 3/4.9.8, 3/4.9.4.

i

?

1 1

i

)

l l

f, 4

l

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

• OUESTION 7.09 Assume that.a Steam Generatc.r tube rupture has been verified at a leak rate in excess of 100 gpm. Emergency Instruction 5023-12-4 " Steam Generator Tube Rupture" cautions the operator to reduce the RCS Th to less than 525 degrees Fahrenheit before isolating the affsEted Steam Generator. What is the reason for this caution? (1.0)

• ANSWER 7.09 To minimize the possible lifting of the Steam Generator safety valves. (+1.0)

• REFERENCE (S) :

1. CEN-152, rev. 3, section 6, page 4.

1 -

2. S023-12-4 f

a 1

i e

l

• QUESTION 7.10 Following a, steam generator tube rupture, the operator is instructed to control the.RCS pressure, maintaining it below 1000 psia. What three (3) systems are available to the operator to effect this control of the pressure? (1.5)

• ANSWER 7.10 . Main spray depressurization (0.5)

. Auxiliary spray depressurination (0.5)

. Throttling of the HPSI pumps (0.5)

• REFERENCE (S):'

1. S023-12-4 END of SECTION 7, CONTINUE ON NEXT PAGE: SECTION 8 i

SECTION 8

. Administrative Procedures, Conditions, and Limitations z.,

• OUESTION 8.01. Complete the following table to indicate the minimum shift crew composition in Mode 1. (2.0)

Crew members Number

a. SS ____

b. SRO ____

c. RO ____

d. AO ____

e. STA ____

• ANSWER S.01 a. 1

6. 1 C. 2

d. 2

e. 1

((0.4) each category)

• REFERENCE T.S. TABLE 6.2.1

~e

• QUESTION

.8.02 TRUE OR FALSE Independent verification, by a second qualified operator is not required for...

a. ...those portions of a system within a radiation area which will result in greater than 5 mrem whole body dose. (0.5)

b. ... components that are located,in posted airborne contamination areas. (0.5)

• ANSWER

. 8.02 a. False (0.5)

b. true (0.5)

• REFERENCE 5023-0-36, PAGE 8.

(T e

• QUESTION B.03 The Technical specifications specify a limiting condition for operation (LCO) with respect to the Auxiliary Feedwater system.

a. For what does the OPERABILITY of the Auxiliarp-Feedweter system provide assurance? (1.0)

b. With one Auxiliary Feedwater pump inoperable, What action is required? (1.0)

• ANSWER 8.03 a. The operability of the AFS ensures that the r eactor -

coolant system can be cooled down to less than 330 Fahrenheit from normal operating conditions in the event of a total loss of offsite power. (1.0)

b. With one AFS pump inoperable restore the required AFS pumps to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LO.53, or be in Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Hot Shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. CO.53 (1.0)

• REFERENCE T.S. 3.7.1.2, BASES 3/4.7.i.2.

.o

• QUESTION 8.04 List the letter designations of those statements chosen from the following which are correct. the statements are in response to, " Temporary changes to procedures may be made provided:" (2.0) :_

a. Critical operation of the unit shall not be resumed until authorized by the Commission (NRC). (0.5)

6. The intent of the original procedure is not altered. (0.5)

c. The change is approved by two members of the plant management staff, at least one of whom holds a SRO

- license in the affected unit. (0.5)

d. The change is documented, reviewed and approved by the Station Manager within 7 days of implementation.

(0.5)

• ANSWER O.04 b. and c. --

correct

a. and d. - incorrect (0.5) each (Those not listed will be counted as incoreact.)

• REFERENCE T.S. 6.8.3.

e i

• QUESTION 8.05 Regarding administrative limits for radiation protections

a. What are the quarterly limits for the whole body, extremities, and the skin of the whole body? (1.5)

De*&. .

-eT'To whdt level can the quarterly limit for whole body dose be extended? (0.5)

• ANSWER 8.05 a. whole body - 900 mrem (0.5) extemities - 4700 mrem (0.5) skin - 3750 (0.5)

5. 225_ .... ... / q t r ' J, . S i

• REFERENCE SO123-VII-4-0

o

• QUESTION 8.06 Technical Specification 3/4.1.1.4 state the lowest loop operating temperature for the RCS Tave shall be greater than 520 Fahrenheit when the reactor is critical. Explain what four (4) things this specification ensures. (2.0)

• ANSWER 8.06 3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY l This limitation is required to ensures l 1. The moderator temperature coefficient is within its analyzed range. (0.5)

2. The protective instrumentation is within its normal operating range. (0.5)

3. The pressurizer is capable of being in an OPERABLE

. status with steam bubble. (0.5)

4. And the reactor pressure vessel is above its minimum RT/NDT temperature. (0.5)

• REFERENCE 1.S. BASES 3/4.1.1.4

l

[o -

J

*DUESTION B.07 What are the BASES for restricting the movement of loads, t in excess of 2000 pounds, over the fuel assemblies in the storage pool? (1.0)

• ANSWER 8.07 In the event of a drooped load the activity release will be limited to that contained in a single fuel assembly. (0.5) i And any possible distortion of fuel in the fuel storage racks will not result in a critical array. (0.5)

• REFERENCE T.S. 3.8. 1 i

1 i 3 s i

l l

1 1

I I

I t i

l l.

f 1

f r

E 4

J j

i r

J i

i l

I.

--m- -s < ge p-e _e,m,---.--.em+. c--~,-----gwn,,--,c--w va--n,-,y- -w - m e , rw-v "v r*"

e t

c>

• OUESTION 8.08 SO123-VIII-1 " Recognition arid Classi fication of l Emergencies" requires throp actions to be performed within {

15 minutes of off-normal conditions. What are those three j actions? (1.5)  :-

• ANSWER 8.00 1. Review event category tabs. (0.5)

2. Classify emergency. (0.5)

3. Declare the emergency. (0.5)

• REFERENCE SO123-VIII-1 i

r -

s t

e

• QUESTION 8.09 Technical Specification 3.1.3.6 establishes insertion limits for the regulating.CEAs. What are the BASES for these limits? (1.0)

• ANSWER 8.09 1. The minimum Shutdown Margin is maintained. (0.5)

2. The potential effects of a CEA ejection accident are limited to acceptable levels. (0.5)

• REFERENCE T.S. BASES 3/4 1-5.

A I

o

• OUESTION 8.10 During Mode 6 refueling operations can the SRO in charge of overseeing fuel movements be the Shift Supervisor? Explain.

(1.0)

• ANSWER O.10 No. (0.5) The SRO must be restricted to fuel handling operations only, with no colateral duties. (0.5)

• REFERENCE T.S. 6.2.2.d.

END of SECTION O END OF EXAMINATION l