Text

Exam Rept 50-298/OL-88-01 Administered During Wk of 880215. Exam results:11 of 13 Candidates Administered Test Passed All Portions & Will Be Issued Appropriate License
	ML20148M636
Person / Time
Site:	Cooper
Issue date:	03/28/1988
From:	Graves D, Pellet J; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	;
Shared Package
ML20148M612	List: ML20148M636;
References
	50-298-OL-88-01, 50-298-OL-88-1, NUDOCS 8804060042
	Download: ML20148M636 (99)
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APPENDIX U.S. NUCLEAR REGULATORY COMMISSION REGION IV Operator Licensing Exam Report:

50-298/0L-88-01 Operating License:

DPR-46 Docket No.

50-298 Licensee:

Nebraska Public Power District P.O. Box 499 Columbus, NE 68601 Facility Name: CooperNuclearStation(CNS)

Examination at: Cooper Nuclear Station Chief Examiner:

7Me/$/

D. N. Graves, Examiner, Date' Operator Licensing Section, Division of Reactor Safety c

= _b M W Approved by:

]

J.'L.

Pellet, Chief, Da';e Operator Licensing Section, Division of Reactor Safety Summary NRC Administered Examinations Conducted During the Week of February 15, 1988 (Report 50-298/0L-88-01)

)

Results:

NRC administered examinations to 13 candidatos. Eleven (11) candidates passed all portions of the examination and will be issued the appropriate license.

8804060042 880330

{DR ADOCK 05000g 8

. DETAILS 1.

Persons Examined SR0 M

Total License Examinations:

Pass -

6 5

11 Fail -

1 1

2 2.

Examiners D. Graves, Chief Examiner M. Spencer M. Bishop 3.

Examination Report Performance results for individual examinees are not included in this report as it will be placed in the NRC Public Document Room and these results are not subject to public disclosure.

a.

Examination Review Coment/ Resolution In general, editorial comments or changes made during the examination, or subsequent grading reviews are not addressed by this resolution section. This section reflects resolution of substantive comments made by CNS.

The only comments addressed in this section are those which u re not accepted for incorporation into the examination and/or answer key.

Those comments accepted are incorporated into the master examination key which is included in this report.

Comments may be paraphrased for bre"*;y.

The full text of the comments is attached.

(2.0lb)

"Airborne problem" should be considered an acceptable answer.

The purpose of the condenser is not to prevent high temperatures in the area.

Response

Reject.

High airborne activity does not affect the operation of the RCIC system.

(2.02)

Also accept "perform the appropriate surveillance" since the required answer would be in the surveillance.

Response

Reject.

The candidate must demonstrate sufficient knowledge such that the examiner is convinced the candidate knows flow path and system operation.

. (5.06b)

Also accept Position 10 as acceptable -if accompanied with explanation regarding rod shadowing reducing rod worth at Position 40.

Response

Rej ect. Sufficient information was given in the question for the candidate to determine the core was bottom-peaked and Position 40 was in a region of higher flux than Position 10.

b.

Site Visit Summar (1) At the end of the written examination administration, the facility licensee was provided a copy of the examination and answer key for the purpose of commenting on the examination content validity.

It was explained to the facility licensee that regional policy was to have examination results finalized within 30 days. Thus, a timely response was desired to attain this goal.

(2) At the conclusion of the site visit, the Chief Examiner met with facility representatives to discuss the visit.

The following personnel were present:

NRC Facility D. Graves R. Brungardt R. Bennett J. Dutton l

D. Kuser

[

M. Parrish i

G. Reece J. Surette Mr. Graves opened the meeting by thanking those present for the cooperation received during the site visit and informing those present that current guidelines do not allow the disclosure of preliminary operating examination results. Other items discussed were as follows:

1.

There is no administrative mechanism available to inform the supervisors on shift to know what restrictions apply to the licenses on his shift.

For example, an operator may have a "No Solo" restriction on his license. The supervisor may not be aware of this and allow the operator to be alone in the control rcom under certain plant conditions.

2.

While in the Radwaste Control Room, the examiner observed a i

rigid cardboard tube placed between a valve switch ard a pump switch to hold the pump switch in the START position.

l

s 4-When the operator was asked about this, he stated that the pump trips off if the switch is not held in the START

position, c.

Generic Comments No areas of knowledge were identified as being generically weak, d.

Master Examination and Answer Key Master copies'of the CNS license examinations and answer keys are attached.

The facility licensee comments which have been accepted are incorporated into the answer key.

e.

Facility Examination Review Comments The facility licensee comments regarding tiie written examination are attached. Those comments which were not acceptable for incorporation into the examination answer key have been addressed in the resolution section of this report.

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NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY:

_QQQBgB__________________

REACTOR TYPE:

_BWB-QE4_________________

DATE ADMINISTERED: _QBlQ241h________________

EXAMINER:

_QBeVESt_Qt______________

CANDIDATE:

IUSIBUCIl0US_IQ_CeUQlDeIE1 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 p ass ing 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

__VeLUE_ _IQIok

___SQQBE___

_VeLUE__ ______________CeIEGQBI_____________

_25100__ _25tQQ

________ l.

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 25ADQ__ _25 ADD 2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS

_2510Q__ _251Q0 3.

INSTRUMENTS AND CONTROLS l

_25tQQ__ _25200 4.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL I

1QQ100__

Totals Final Grade All work done on this examination is my own.

I have neither given nor received aid.

Candidate's Signature

I1__E81091ELES_9E_U99LEeB_EQWEB_ELeUI_QEEBoIIQN PAGE 2

t IUEBUQQ186dICSt_BEeI_IBeOSEEB_eUQ_ELUIQ_ELQW l

QUESTION 1.01 (1.00)

Select the correct word'in parenthesis for each of the following statements concerning subcritical multiplication.

(1.0) c.

The closer the reactor gets to critical, the (SHORTER / LONGER) the wait must be to allow the suberitical neutron density to reach equilibrium given equal positive reactivity insertions, b.

Suberitical count rate is (PROPORTIONAL, INVERSELY PROPORTIONAL) to k-effect've.

ANSWER 1.01 (1.00) a.

Longer (0.5) b.

Proportional (0,5)

REFERENCE CNS Reactor Theory Chapter 3 292008K103 4.1/4.0 292008K103

..,(KA's)

QUESTION 1.02 (1.00)

Which of the following best describes the behavior of equilibrium Xenon CONCENTRATION over core life?

(1.0) a.

It decreases because of the increased fuel burn-up b.

It decreases because of the decre'ae in fission yield of Xenon c.

It increases because of the decrease in the delayed neutron fraction d.

It increases because of the increase in thermal flux ANSWER 1.02 (1.00) b (1.0)

REFERENCE CNS Reactor Theory, pg 6-13 292006K114 3.1/3.2 292006K114

...(KA's)

(*****

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

l

It__E81NGIELES_DE_UUCLE88_EQWEB_ELeUI_QEEBoI1QN PAGE 3

i IBEBdQDIUed10St_BEeI_IBeUSEEB_eUQ_ELU10_ELQW QUESTION 1.03 (1.00)

With the reactor critical at 75 on IRM rango 4,

.od withdrawal is used to increase and stabilize power at 75 on IRM range 6.

RCS temperature is 190 deg F.

Select the statement that correctly describes the position of rods, and reason, after the power is stabilized on range 6.

(1.0) a.

The rods will be further withdrawn on range 6 than on range 4 because more fuel must be exposed to the available neutrons to maintain the higher power level.

b.

The rods will be further withdrawn on range 6 to overcome the power defect.

c.

Tha rod position will be the same.

The outward rod motion needed to achieve a given period equals the inward motion needed to return the period to infinity, d.

The rods will be less withdrawn on range 6 due to the increased delayed neutron population associated with the higher power level.

ANSWER 1.03 (1.00) c (1.0)

REFERENCE CNS Reactor Theory, pg 4-36 292008K108 4.1/4.1 292008K108

...(KA'S) 7 QUESTION 1.04 (1.50)

HOW will the Shutdown Margin (Reactivity Margin) just prior to a refueling j

outage compare with the Shutdown Margin following the refueling?

WHY?

Two (2) reasons required.

(1.5) i

I1_ _ E 81 N 91 E L E S _9 E _ U U C L E o B.. E Q W E B _ E L o NI _ Q E i B 6II Q U t PAGE 4

IUEBdQDXUedICS&_HE6I_IB6USEEB_680_.LUID_EL9W ANSWER 1.04 (1.50)

SOM prior to the outage will be larger (0.5) due to f iss ion produc+

poisoning (0.5) and fuel depletion (0.5).

REFERENCE 140TM-TH-4.11-0, Shutdown Margin, pg 5 CNS Reactor Theory, pg 1-35 and 36 K/A 299602 K1.14 2.6/2.9 292002k.14

...(KA'S)

QUESTION 1.05 (2.00)

Why does core thermal power decrease at a much slower rate than indicated neutron power (2 reasons required) follewing a scram from high power operation?

(2.0)

ANSWER 1.05 (2.00)

Thermal power drops at a slower rate due to decay heat (1.0) and the time delay in getting the previously genere'ed heat out of the fuel pellet into the coolant (1.0).

Also accept residual or sensible heat of the core t.omponents as one answer.

REFFRENCE CNS Reactor Theory, pg 7-22 292008K125 2.8/2.0 292008X130 3.2/3.5 292008K125 292008K130

...(KA'S)

ATEGORY 01 CONTINUED ON NEXT PAGE *****)

'v

Ii__E8JNQIELL3_QE_UUQLEeB_EQWEB_ELeNI_QEEBoIIQNt PAGE 5

IBEBdQ2INed10St_UEeI_IBeOSEEB_eUQ_ELVIQ_ELQW QUESTION 1.06 (3.00) a.

Does the magnitude of the initial levei of source range counts affect the critical rod position? WHY?

(1.0) b.

The reactor is brought critical at 40 on IRM range 2 with the short-est permissible stable positive period allowed by GOP 2.1.1, "Cold Startup."

Heating power is determined to be 40 on range 8 of IRM's.

- - - SHOW ALL WORK ****

1.

What is the doubling time if the period remains constant?

(1.0) 2.

How long will it take for powor to reach the point of adding heat if the period remains constant?

(1.0)

ANSWER 1.06 (3.00) a.

No (0.50).

The critical control rod position is a function of Keff or reactivity of the reactor and is not a function of the source count rate (0.50).

b.

1.

From GOP 2.1.1, shortest permiscible stable period equals 50 sec.(0,5).

= 34.7 seconds. (0.5)

Thus Doubling time equals 50/1.44 2.

40 range 2 is equal to 0.04 on range 8 50 eeconds 0.04 P(t)

= 40 Period P(0)

=

P(0) e ^(t/ period)

P(t)

=

0.04 e ^(t/50 sec) 40

=

Time = 345.4 seconds or 5 min. 45 sec (1.0)

(NOTE: Grade method if period is different)

REFERENCE LOTM-TH-4.15-1 CNS Reactor Theory, Chapter 3 292003K108 2.7/2.8 20c108K104 3.3/3.4 292003K103 2920C8K104

...(KA'S)

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

12__E81NCIELES_QE_U90 lee 8_E9 WEB _ELaVI_DEE86IIQNt PAGE 6

IUE80001NedICSt_UE61_I8eBSEEB_eUD ELUID_ELQW QUESTION 1.07 (2.00)

The reactor is operating at 60% power when recirculation flow is increased to increase power.

State the effect (INCREASE, DECREASE, REMAIN THE SAME) the power iricrease has on each of the following (steady state to steady state conditions',:

(2.0) c.

Void fraction b.

Ooppler reactivity coefficient c.

Total Doppler reactivity d.

Feedwater enthalpy ANSWER 1.07 (2.00) c.

decrease b.

decrease (less negative) c.

increase (more negative) d.

increase (0.5 each)

REFERENCE CNS Reactor Theory Chapter 4 CNS Heat Transfer and Fluid Flow 5-47 through 5-58 292008K120 3.3/3.4 292004K108 2.2*/2.4*

292008K120 292004K108

...(KA'S)

QUESTION 1.08 (1.50)

For each of the following events, state which ccaTTicient of reactivity would act first to change core reactivity; (1.5) a.

Loss of extraction steam to feedwater heaters b.

Main turbine trips from 28% reactor power and one BPV fails to open c.

Inadvertent HPCI start (100% power)

It__E810CIELEk_9E_UUCLEeB_E9 WEB _ELeUI 0EEBoIIQUt PAGE 7

IHEBdDQ1UedICSi_UEoI_IBoUSEEB.eUQ_ELUIQ_ELQW ANSWER 1.08 (1.50) c.

Moderator b.

Void c.

Void (0.5 each)

REFERENCE CNS Reactor Theory, Chaoter 4 292008K121 2.9/3.0 292008K121

...(KA'S)

QUESTION 1.09 (1.00)

Which one of the below sets of parameters indicates a water system that is subccoled by greater than 30 F?

TEMP. (F)

PRESS. (psia)

c..

540 1000 b.

560 1500 c.

665 2000 d.

640 2400 ANSWER 1.09 (1.00) b (1.0)

REFERENCE Steam Tables 293003K123 2.8*/3.1*

293003K123

...(KA'S) i QUESTION 1.10 (1.00)

EhPLAIN how steam at 900 psig can be used as the motive N oc e for RCIC injection into the r e a r. +. o r vessel at 1000 psig.

(i.e.,

How can 900 psig steam ruise water pressure to 1000 psig?)

(1.0)

Ix__E81UCIELES_DE_UUCLEeB_EQWEB_EbeUI_QEEBAI1QN PAGE 8

t IUEBdQDIUedIGS&_BEeI_IBeOSEEB_eUD_ELUID_ELOW ANSWER 1.10 (1.00)

(As the steam expands through the turbine), the enthalpy given up in condensation / expansion is more tha. is required to be added to the water (to raise pressure from 1S psia to 1000 ps ig. )

(i.e.,

steam delta b=

1197

- 910 = 287 8tu/lbm > water delta h = 98 - 68 = 30 8tu/lbm) (1.01 REFERENCE LOTM-TH-2.S-0 LOTM-TH-2.10-0 293002K104 2.1/2.4 293003K123 2.8/3.1 293003K123 293002K104

...(KA'S)

QUESTION 1.11 (1.00)

Which one of the f o ). low i ng 8EST describes what occurs if a centrifugal pump la STARTED AND OPERATED with its discharge valve shut as compared to with its discharge valve open. (Assume no vecirculation flow.)

(1.0) a.

Higher / longer starting current and lower running current b.

Lower / shorter starting current and lower running current c.

Higher / longer starting current and higher running current d.

Lower / shorter starting current and higher running current ANSWER 1.11 (1.00) b (1.0)

REFERENCE LOTM-TH-1.4-0 CNS Heat Transfer and Fluid Flow, pg 6-109 291004K107 2.8/2.8 291004K107

...(KA S)

QUESTION 1.12 (2.00)

CNS procedure E0P-1, "RPV Control", requires a reductisti in RPV water level in order to reduce reactor power during an ATWS.

What are two (2) reasons why lowering reactor watar level will help reduce reactor power?

(2.0) r

I1__ESIUQ1ELES_DE_U!!GLEeB_EQWEE EL8UI_QEEBoIIQUt PAGE 9

IBEBdQQ1Ned10Si_UEeI_IBouSEEB_eUQ ELU10_ELQW ANSWER 1.12 (2.00)

Increased voiding (1.0)

Concentrating the boron during SLC injection (1,0)

REFERENCE CNS E0P-1 GE E0P Fundamentals 295037 EA2.02 4.1*/4.2*

295037A202

...(KA'S)

QUESTION 1.13 (1.00)

What effect would isolation of extraction steam to a HP heater have on Recirc Pump NPSH at 854 power?

Explain your answer.

(1.0)

ANSWER 1.13 (1.00)

NPSH would increase (0.5).

Because the feedwater temperature would decrease, decreasing the annulus temperature and the temperature at the suction of the pump (3.5).

REFERdNCE LOTM-TH-1.4-0 CNS Heat Transfer and Fluid Flow, pgs 6-73 through 6-77 733006K110 2.7/2.8 293006K110

...(KA'S)

(*****

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

It__EBINGIELES_9E_UUCLE68_E0 WEB _EbeUI_QEEBoIIQUt PAGE 10 IUEBdQQ1UedIGSt_UEoI_IBeUSEEB_800_ELUID_ELQW QUESTION 1.14 (2.50)

Mctch each of the following statements with the appropriate numbered item:

a.

The limiting parameter that assures PCT will not exceed 2200 degrees F during a design basis LOCA.

b.

Total power passing through a unit length of fuel rod.

c.

APLHGR divided by MAPLHGR limit.

(2.5) d.

LHGR (max)/LHGR (LCO) o.

Power requi.ed to produce CTB/ Actual bundle power 1.

MAPRAT 5.

LHGR 2.

FLCPR 6.

MFLPD 3.

APLHGR 7.

CPR 4.

MCPR 8.

GEXL ANSWER 1.14 (2,50) c.

3 b.

5 c.

1 d.

6 e.

7 (0.5 pts each)

REFERENCE l

LOTM-TH-3.8-0 through 3.10-1 CNS Heat Transfer and Fluid Flow, Chapter 9 293009K105 3.3/3.5 293009K106 3.4/3.8 l

293009K111 2.8/3.6 293009K118 3.2/3.7 293009K105 293009K106 293009K118 293009K111

...(KA'S) l i

I

(*****

CATEGORY U1 CONTINUED ON NEXT PAGE *****)

l l

i L-

'Ax__EBIUCIELES_DEUUCLEeB_EDWEB_EL8NI_9EEBoIIOut PAGE 11 IBEBdQQ18edICSt_UE6I_IB86SEEB_oND_ELUID_ELQW QUESTION 1.15 (1.50)

Following an auto initiation of RCIC at a reactor pressure of 800 psig, reactor pressure. decreases to 400 psig.

Indicate how the following paremeters would change (INCREASE, DECREASE, NO CHANGE) due to the decrease in reactor pressure.

Assume the RCIC System is operating as designed.

(1.5) a.

RCIC Flow to the Reactor b.

RCIC Pump Discharge Head c.

RCIC Turbine RPM ANSWER 1.15 (1.50) a.

No Change b '.

Decrease c.

Decrease (0.5 each)

REFERENCE LOTM-21-1, pg 7 CNS RCIC Lesson Plan 217000K502 3.1/3.1 217000K505 2.4*/2.4*

217000K506 2.7*/2.7 217000K506 217000K505 217000K502

...(KA'S)

QUESTION 1.16 (2.00)

What are four parametens (factors) that affect the point at which boiling transition occurs in a fuel bundle?

(2.0) t i

i I

It__EBIUCIELE3_DE_UUCLE68 20thB_E668I_DEEB6II90t PAGE la IBEBdQQ166dIGSt_UE61_IB6USEEB_600_ELUIQ_ELQW ANSWER 1.16 (2.00) inlet subcooling pressure inlet flow axial power shape local peaking

- bundle power (4 required at 0.5 each)

REFERENCE CNS Heat Transfer and Fluid Flow, pg 9-66 293009K122 - 293009K126 293009K125 293009K124 293009K123 293009K122 293009K126

...(KA'S)

(*****

END OF CATEGORY 01 *****)

b 1

it__EL8NI_DESIQU_INCLUDIUQ_S6fEIY_eND_EUEB9EUCY_SYSIEUS PAGE 13 QUESTION 2.01 (3.00)

In the Reactor Core Isolation Cooling (RCIC) system:

e.

What function does the RCIC lube oil cooler water perform after leaving the cooler?

(1.0) b.

How would failure of the barometric condenser affect operation of the RCIC system?

Cone effect required)

(1.0) c.

How is the turbine exhaust line protected from overpressure during systum operation?

TWO required.

(1.0)

ANSWER 2.01 (3.00) a.

It is used as the condensing medium for the barometric condenser (1.0) b.

System isolation (1.0) due to high area temperature c.

Rupture diaphragms (0.5)

Turbine trip on high exhaust pressure (0.5)

REFERENCE RCIC pg 6, 7,

21 217000K404 3.0/3.1 217000K405 3.2/3.5 217000K405 217000K404

...(KA'S)

QUESTION 2.02 (1.00) i How is SLC system flow capability verified without opening an injection path to the reactor vessel?

(1.0) i l

ANSWER 2.02 (1.00) i Starting the pumps locally does not fire the squib valves (0.5) and the flow is to the test tank (0.5).

REFERENCE SLC-14 211000K408 4.2*/4.2*

l 211000K408

...(KA*S)

(*****

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

(

{

1 c

l

it__EleUI_DESIGU_INGLUDIUQ_S6EEIY_eUD_EdEBQEUGY_SYSIEd3 PAGE 14 QUESTION 2.03 (3.00)

Answer the following with regard to the RHR system and its various modes of operation:

e.

Match the following actions, events, or interlocks in Column A with the pressure in Column 8 that initiates or allows that item.

(2.0)

Column A Column B 1.

Shutdown cooling isolates 50 psig 2.

Allows manual operation of the 75 psig LPCI injection valve 100 psig 3.

Automatically opens the LPCI 135 psig inj ect ion valve 277 psig 4.

Input to ADS 400 psig 432 psig 450 psig i

b.

What is the most limiting failure of a single component in the RHR system with regard to core protection?

(1.0) f I

ANSWER 2.03 (3.00) a.

1.

75 psig 2.

450 psig 3.

450 psig 4.

100 psig or 135 psig (accept either)

(0.5 each) b.

LPCI injection valve failed shut (1.0)

REFERENCE RHR-24, 37 205000K402 3.7/3.0 203000K106 3.9/3.9 203000K402 3.3/3.3 203000A21t 3.4/3 6 205000K402 203000K402 203000K106 203000A211

...(KA'S)

QUESTION 2.04 (1.00)

During startup under Cold Conditions the operator adj usts the Control Rod Drive Pressure Control Valve to.naintain a +260 psid between CR0 and reactor pressure.

Explain how this pressure differential is maintained as reactor pressure increases during the ensuing startup.

i e..,-

m-

..m.

--,--.-,--m

2t__EloNI_ DES 10U_10GL901ND_SeEEI1_6UQ_EdEBQEUQ1_212IEU$

PAGE 15 ANSWER 2.04 (1.00)

The FCV opens up as reactor pressure increases maintaining a constant flow, therafore, a constant pressure differential across the PCV (1.0).

REFERENCE CROH-10 201001 K4.08 3.1/3.0 A1.01 3.1/2.9 201001K408 201001A101

...(KA'S)

QUESTION 2.05 (2.00)

State whether the following conditions or signals WILL or WILL NOT cause initiation of the SBGT system:

(2.0)

NOTE:

Do not consider setpoints.

If the indicated parameters will initiate the system, assume the setpoint has been reached.

a.

High radiation (ARM) on the Refuel Floor b.

High radiation in the reactor building ventilation exhaust c.

High particulate activity in the drywell d.

Low flow in the offgas system o.

High dr>well pressure coincident with high radiation in the drywell f.

Low RPV pressure coincident with high drywell temperature 9

High reactor pressure coincident with a low reactor level h.

Main Steam Line high radiation coincident with low reactor pressure ANSWER 2.05 (2.00) will:

b, e,

g will not:

a, c,

d, f,

h (0.25 each) l l

REFERENCE l

SGT-17, 18 261000K401 3.7/3.8 l

}t__EleNI_ DES 10N_INCLUQ1NQ_SoEEIX_6ND_EUEBQENCY_SXSIEd3 PAGE 16 261000K401

...(KA'S)

QUESTION 2.06 (2.00)

Answer TRUE or FALSE for the following:

c.

The CR0 pump will trip on low suction pressure.

(0.5) b.

The standby CR0 pump auto starts when the running pump trips.

(0.5) c.

CROM Accumulators are charged with air from the instrument air system.

(0,5) d.

Speed Control of the CROM is accomplished by the timing of the directional control solenoid valves.

(0.5)

ANSWER 2.06 (2.00) a.

TRUE (0.5) b.

FALSE (0.5) c.

FALSE (0.5) d.

FALSE (0.5)

REFERENCE CROH-8, 15, 16 201001K412 2.9/2.9 201001K501 2.4/2.4 201001K501 101001K412

...(KA'S)

QUESTION 2.07 (2.50)

List the RHR throttle valves that receive a sealed-in OPEN signal on an auto initiation.

Include HOW LONG the signal is sealed in and WHY these valves receive a sealed in signal.

(2.5)

ANSWER 2.C7 (2,50)

RHR HX bypass (0,5), MO-66, sealed open for 2 minutes (0,5)

RHR injection (0.5), MO-27, sealed open for 5 minutes (0,5)

Signal is sealed in to ensure maximum flow to the vessel is provided following initiation (0,5).

li__El6UI_DESIQU_IUCLUDIUQ 18EEII_6UQ_EdEBQEUQ1_SYSIEUS PAGE 17 REFERENCE RHR-9, 11 203000K4.10 3.9/4.1 203000K410

...(KA'S)

QUESTION 2.08 (2.00)

Indicate whether each of the following are TRUE or FALSE:

(2.0) e.

The pneumatic sLaply for the inboard MSIVs is instrument air with a N2 backup source.

b, The MSIVs pneumatic cylinders are capable of closing the MSIV without assistance from the closing springs.

c.

Only one of the two control solenoids must de-energize to cause the MSIVs to close.

d.

The MSIVs closing springs sre capable of closing the MSIV without assistance from the pneumatic cylinder.

ANSWER 2.08 (2.00) s.

false b,

true c.

false d

true i

RFFERENCE MS-7, 8,

3 239001K601 3.1/3.3 239001K602 3.2/3.2 l

239001K602

...(KA'S) l l

QUESTION 2.09 (4.00) l l

a.

List ten (10) of the valve operations that automatically occur on a j

turbine trip signal.

Assume the plant is operating at power when the turbine trip occurs.

(2.0) f l

b.

What are four (4) of the five (5) protective devices that operate independently on the main turbine to prevent damage to the unit if the turbine was not taken out of service immediately?

(2.0) l l

l l

l

2t__EleUI_DESIGU_INGL9QIUQ_SeEEIl_eUQ_EUEBQEUGl_SYSIEUS PAGE 18 ANSWER 2.09 (4.00) c.

1.

turbine main stop valves close 2.

control valves close 3.

reheat stop valves close 4.

intercept valves close 5.

bypass valves open 6.

feed system startup flow control isolation valves open 7.

feed pump discharge valves close 8.

extraction steam non-return valves trip 9.

extraction' steam dump valves open 10.

feed pump low pressure steam supply valves close 11.

main turbine governor valve drain valves open 12.

MSL drain valves shift from AS/RS heaters to the main condenser (10 required at 0.2 each) b.

1.

overspeed trip mechanism 2.

Iow vacuum trip 3.

low bearing pressure trip 4.

thrust bearing trip 5.

solenoid trip (0.5 each)

REFERENCE MN TURB-16, 17 245000A201 3.7/3.9 245000G007 3.5/3.6 2450000007 245000A201

...(KA's)

QUESTION 2.10 (2.50)

What are five (5) sources or flowpaths that may be Laed to restore fuel pool water level, if necessary.

(2.5) a

it__E'LeUI_DESIGU.INGLVQIUQ_SeEEIl_eUQ_EdEBGEUC1_SISIEd3 PAGE 19 i

' ANSWER 2.10 (2.50) 1.

Skimmer surge tank condensate makeup (normal method) 1 2.

Connecting hoses to the service box condensate and demin water

]

connections 3.

Fire hoses 4.

Croestie the RHR system with the fuel pool cooling system to take a i

suction on the CST.

S.

Crosstie service water to the RHR system which is then crosstied to the f uel pool cooling syster.t.

(5 at 0.5 each) t REFERENCE FPC-18 233000K406 2.9/3.2 233000K406

...(KA'S) i 1

QUESTION 2.11 (2.00)

Match the following pla-t areas (a - g) with the type (s) of fire protection system (1 - 4) that is(are) available in that area:

(2.0)

NOTE:

MORE THAN ONE TYPE OF SYSTEM MAY APPLY TO EACH AREA a.

Reactor feed pump room 1.

Fire Water System i

b.

Service water pump room 2.

Carbon Dioxide System c.

Diesel generator day tank rooms 3.

Halon 1301 System d.

Fire pump house - diesel fire pump room t.

No Automatic System Available l

e.

Control building, cable spreading room (918')

f.

Turbine generator bearings 1,

2, 3

3 Main control room l

i.

i I

1 __EI.6NI_DE2196_INGLUD180_S6EEII_680_EdE80EUDX_SXSIEd3 PAGE 20 i

ANSWER 2.11 (2.00) a.

1 b.

3 J

c.

2 d.

1 e.

1, 2

t.

2 9

4 (c - g at 0.25 each)

REFERENCE FP system description 286000G004 3.8/3.9 286000G004

...(KA'S)

\\

i i

i f

(*****

END OF CATEGORY 02 *****)

'bA..[USIB9dEUl$.6U0.GQUIBQLS PAGE 21 1

QUESTION 3.01 (3.00) e.

STATE whether the solenoids associated with the following valves are NORMALLY Energized or Deenergized.

NO SCRAM SIGNAL EXISTS.

(1.0) 1.

Back-up Scram Valves 2.

Scram Discharge Volume Vent and Orain Valves b.

Repositioning the Mode Switch from STARTUP/ HOT STANOBY to RUN causas certain reactor scram functions to be bypassed and others to be effective.

LIST the three (3) scram functions (or setpoints) which are bypassed AND the three (3) scram functions (or setpoints) which become effective when toe Mode Switch is taken to RUN.

(2.0) t Ai4 SWE R 3.01 (3.00) f a.

1.

Deenergized (0,5) 2.

Energized (0.5) b.

Activated in RUN:

- MSIV Closure ( *. $ 3. )

4 1

- Companion IRM/APRM (0.33)

- APRM flow biased scram (0.33)

Bypassed in RUN:

IRM Inop (0.33)

- IRM Upscale (0.33)

APRM 15% HIGH Flux (0.33) l REFERENCE CRDH-17, 19 RPS-14, 16-18 212000K108 3.0/3.1 212000K412 3.9/4.1 21?000A216 4.0/4.1 212000K412 212000K108 212000A216

...(KA'S) j a

i i

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'it__I'USIBubEUIS_eUD_CQUIBQLS PAGE 22 s

QUESTION 3.02 (1.50)

Indicate at what reactor water level each of the following actions is directly initiated.

If more than one level applies, indicate all of the cpplicable levels.

(1.5) a.

Direct reactor scram b.

Standby Gas Treatment System starts c.

RCIC starts d.

HPCI isolation o.

Recirculation pumps trip f.

Main Steam Line isolation (MSIVs)

ANSWER 3.02 (1.50) a.

+12.5" b.

+12.5" c.

-37" d.

+56.5" e.

-37" f.

-145.5" (a - f 0.25 each)

REFERENCE SGT-4, NBI Table 1 216000K101 3.9/4.1 216000K102 3.8/4.0 216000K103 3.4/3.6 216000K114 3.8/4.1*

216000K123 3.3/3.4 216000K103 216000K102 216000K101 216000K123 216000K114

...(KA'S) i

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F

~, ~ -

3t__lU31BudENIS 6UQ_QQUIBQL3 PAGE 23 QUESTION 3.03 (1.50)

Your shift is performing a reactor startup.

Criticality is achieved with a 120 second period at a moderator temperature of 180 degrees F.

Due to a personnel error during maintenance on the LPCS initiation circuitry, the LPCS system starts and injects to the vessel.

Assume NO OPERATOR ACTION.

a.

If a reactor scram were received during this event, what reactor protection system function would initiate it?

Assume the scram signal was not directly caused by the personnel error and all instrumentation is functioning properly.

(1.0) b.

How would the system respond if the operator had j ust closed the suppression pool suction valve prior to the initiation signal and the CST suction valve was already shut?

Address suction path only.

(0.5)

ANSWER 3.03 (1.50) a.

IRM Flux Hi-Hi (1.0) b.

ro suction path would be open (0.5).

CST suction is manual valve and the suppression pool suction valve does not automatically open.

REFERENCE CS-4 209001K406 2.6/2.9 209001G015 3.8/4.2*

20900lG015 20900lK406

...(KA'S)

QUESTION 3.04 (1.50)

A reactor startup is in progress.

The "A"

SRM is bypassed so the Instrument Technicians can troubleshoot the power supply.

The tech mistakenly takes the "B"

SRM OPERATE switch to STAN0BY and starts troubleshooting its power cupply.

a.

WHAT specific plant / system TRIP did this cause?

(0.5) b.

HOW did this trip spec if ically affect the plant startup?

(0.5) l c.

On WHAT IRM range would the above trip have been automatically b y p a s s e. d ?

(0.5)

$1 _l'USIBudENIS_6NQ_G9UIBQLS PAGE 24 ANSWER 3.04 (1.50) c.

SRM Inop. Trip.(also accept rod block) (0.5) b.

Inop trip on SRM's causes a Rod Block (0,5) c.

Range 8 or above ( > Range 7 acceptable) (0.5)

(If rod block is given in a.,

accept discussion of S/U delay L'til rod block cleared in b.)

s REFERENCE SRM-21, 26 215004K103 3.0/3.0 215000K401 3.7/3.7 215000K406 3.2/3.2 215000K406 215004K103 215000K401

...(KA'S)

OVESTION 3.05 (2.50)

LIST five (5) automatic reactor scram functions that are NEVER bypassed.

(NOT individual channels)

(2.5)

ANSWER 3.05 (2.50) 1.

high drywell pressure 2.

high reactor vessel pressure 2.

Iow reactor water level l

4.

high main steam line radiation 5.

APRM Inop.

Accept neutron monitoring system (individual inputs may be bypassed, but there is always some type of NMS scram)

(0,5) each REFERENCE RPS-14, 15, 16, 17 212000K412 3.9/4.1 212000K412

...(KA*S) i 1

l f

f

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CATEGORY 03 CONTINUED ON NEXT PAGE *****)

[

i

hi__I'USIBudEUIS_800.QQUIBQLS PAGE 25 i

QUESTION 3.06 (2.00)

Given the following data for APRM channel C:

LPRM level:

A B

C D

LPRMs assigned:

5 4

4 4

LPRMs bypassed:

1 3

1 0

h n.

If APRM Channel Selector Switch on the local instrument is placed to the COUNT position, what would be the expected meter reading?

Describe HOW you arr ived at your answer.

(1.0) b.

Based on the above information, is APRM C operable?

Answer YES or NO and EXPLAIN WHY.

(1.0)

ANSWER 3.06 (2.00) j o.

(12 operable channels)(5% per operable channel) = 60%

OR 4.8V (60/125)(10v)

=

Either percentage or voltage is acceptable (1.0) b.

Inoperable (0.5) due to < 2 LPRMs per level (0.5)

REFERENCE Technical Specifications, pg 31 i

APRM-17, Figure 3 215005K104 3.6/3.6 215005A208 3.2/3.4 215005G011 3.4/4.1 215005K104 215005G011 215005A208

...(KA*S)

QUESTION 3.07 (1.00)

How far above the top of the active fuel (TAF) is RPV level INSTRUMENT i

l ZERO (with the exception of the wide range yarways)?

(1.0) l

ht__1NSIBudEUIS eUD.GQUIBQLS PAGE 26 ANSWER 3.07 (1.00) 164 inches (accept 152 to 176 inches) (1.0)

REFERENCE NBI Figure 10 216000K122 3.6/3.8 216000K122

...(KA'S)

QUESTION 3.08 (1.50)

Assume both recirculation pumps are running at 80% speed.

State how the recirculation pumps' speed is affected by each of the conditions below.

Consider each case separately.

(1.5) a.

The operator closes the 8 recirculation pump discharge valve to the mid position.

b.

Two reactor feed pumps are operating and a feedwater problem causes RPV level to temporarily decrease to 20".

c.

One of the two operating reactor feed pumps trips and the reactor scrams on low level.

ANSWER 3.08 (1.50) a.

(The 8 MG set trips and) the B recirc pump coasts to a stop (0.5) b.

No effect (0.5).

c.

Both pumps runback to 45% speed (0.5).

REFERENCE Recirculation System, Figure 14 202001A211 3.7/3.9 202001A212 3.6/3.8 202001A223 3.2/3.2 202001A211 202001A223 202001A212

...(KA'S)

QUESTION 3.09 (2.00)

List the FOUR MSIV isolation signals that are never bypassed during normal plant ope"ation.

3etpoints NOT requited.

(2.0)

31__fuSIBudEUIS_600_G98IBQLS PAGE 27 ANSWER 3.09 (2.00) low reactor level

- high steam line space temperature high MSL radiation high MSL flow (0.5 each)

REFERENCE MS-9 223002K404 3.2/3.6 223002K404

...(KA's)

QUESTION 3.10 (3.00)

For each of the conditions listed below, indicate in which direction the GOVERN 0R VALVES and BYPASS VALVES will respond (answer with OPEN, CLOSE, NO CHANGd, or words to that effect).

Assume the reactor is operating at power end the DEHC is in MODE 4.

(3.0) o.

Raising the pressure control signal above the load reference.

b.

Reducing the valve position limiter setting to below the pressure control signal, c.

A loss of the speed loop (speed signals) occurs.

ANSWER 3.10 (3.00)

GOVERNOR VALVES BYPASS VALVES a.

No Change Open More b.

Close More Open c.

No Change No Change (6 at 0.5 each)

REFERENCE DEH-6, 10, 11, 12 241000K106 3.8/3.9 241000K108 3.6/3.7 241000K615 2.3/2.4 241000K106 241000K615 241000K108

...(KA'S)

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31__IUSIBUUEUIS_eUD_CQUIBQL3

_PAGE 28 QUESTION 3.11 (2.00) a.

The NORMAL / ISOLATE switches for a emergency diesel generator are in the ISOLATE position.

How does this affect diesel generator operation locally and from the control room?

(1.0) b.-

With the diesel generator running, state what effect taking the diesel generator GOVERNOR CONTROL switch to RAISE has on the machine:

1.

BEFORE the output breaker is closed.

(0,5) 2.

After the output breaker is closed CEOG paralleled with normal bus supply.)

(G.5)

ANSWER 3.11 (2.00) n.

The EDG will respond to local manual signals only (1.0).

b.

1.

increase generator frequency (or diesel speed) (0.5) 2.

increases generator load (0.5)

REFERENCE 0G-24, 32 264000K401 3.5/3.7 264000K402 4.0/4.2 264000K406 2.6/2.7 264000K402 264000K401

...CKA'S)

P j

QUESTION 3.12 (2.00)

J a.

The plant is being started up with reactor level being controlled by one RFP with manual control on its M/A station and its startup flow valve in automatic on its Master Controller.

The feed flow detector for the operating feed pump fails downscale.

Explain how this failure 4

j affects reactor level control?

(1.0) 1 b.

What two (2) conditions will cause a Reactor Feed Pump lockup?

(1.0) i e

i

,...r-

._...rm._.y....__.._,.r-w

..y-m.

...._._-~-,,..e-4-o, r-.,,,.__,

.----e,w,----,.----,v.

-.. w

,m_..- - - - - - -

l 32__INSIBubEUIS_800_CQUIBQL3 PAGE 29 l

ANSWER 3.12 (2.00) a.

The startup flow control valve uses only single element control (vessel level only) thus the loss of the feed flow signal will have no effect (1.0).

b.

- low current control signal to the RFP speed control unit (0.5) of

< 6 ma.

loss of the selected level input to the level control system (0,5).

REFERENCE FCS-9, 10 259002K406 3.1/3.2 259002K604 3.1/3.1 259002K409 3.1/3.1 259002K604 259002K409 259002K4 M

...(KA*S)

QUESTION 3.13 (1.50)

What parameters /s ignals does the RSCS use to determine the position of a control rod when:

(1.5) o.

Greater than 50% rod density?

b.

Less than 50% rod density up to the auto bypass?

ANSWER 3.13 (1.50) a.

RSCS uses the full-in, full-out signals from the RPIS probes for each individual rod (0,5).

b.

The RSCS uses the rod selected (0.33), the direction of movement requested (0.33), and the settle function of the RMCS timer (0.33).

REFERENCE RSCS-6, 12 201004K405 3.2/3.2 201004K405

...(KA'S)

(*****

END OF CATEGORY 03 *****)

it__fB9CEDUBES_:_UQBdekt.6EUQBueLt_EMEBQEUQ1_6NQ PAGE 30 B6D10L00106L_CQUIBQL i

l QUESTION 4.01 (1.50) a.

When operating limitorque motor operated valves, how long should one wait (minimum) after releasing the switch before operating the valve in the opposite direction?

(0.5) b.

A limitorque motor operated valve is operated using the manually engaged lever and handwheel.

What should be done to the valve prior to returning the system to service?

(1.0)

ANSWER 4.01 (1.50) a.

3 seconds (0.5) b.

Stroke the valve electrically (1.0) to verify proper operation.

REFERENCE S0P 2.2.9, Rev 28, og 6 291001K108 3.4/3.5 291001K109 2.7/2.7 291001K109 291001K108

...(KA'S)

QUESTION 4.02 (2.00)

Following a trip of the main turbine, WHY must the valve positioner of one feed pump turbine be run to the upper limit?

Your answer should include what this step accomplishes and why it is necessary to be performed.

(2.0)

ANSWER 4.02 (2.00)

Low pres ure steam to the RFP turbines will be lost due to the turbine trip (1.Gj, and running the valve positioner to the upper limit allows the use of high pressure steam to the RFP turbine (1.0), ensuring feed flow.

REFERENCE S0P 2.2.28, Rev 43, pg 7 259001K120 3.1/3.2 259001A209 2.6/2.7 259001K120 259001A209

...(KA'S)

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CATEGORY 04 CONTINUE 0 ON NEXT PAGE *****)

1 S1__dBQGEQUBES_:_UQBdeLi_oBUQBdobi_EdEBGEUQ1_oNQ PAGE 31 86010LQQ106L_QQUIBQL

. QUESTION 4.03 (1.50)

I f

HOW and WHY does increased air inleakage into the Offgas System affect the j

indicated Offgas radiation level?

(1.5)

ANSWER 4.03 (1.50)

Indicated radiation level increases (0.5) due to the higher offgas flow reducing the holdup time for activated product decay (1.0).

REFERENCE AP 2.4.7.1, Rev 12, pg 4 271000A215 2.7/2.9 271000A215

...(KA'S)

QUESTION 4.04 (3.50) a.

The control room is filling with a noxious vapor from en undetermined source.

The Shift Supervisor decides to implement the "Toxic Gas in Control Room" procedure.

What 3 IMMEDIATE ACTIONS are required of operating personnel by this procedure?

(1.5) b.

The above actions are determined inadequate and the control room is to be evacuated.

What actions or verifications should be performed, if possible, prior to leaving the control room by the Control Room Operators?

(2.0)

NA__IBQGEQMBES :_UQBd6kt_6BdQBd6Lt_EdgBQEUQ1_6UQ PAGE 32 BeDIDLQQIGoL CQUIBQL l

{

l ANSWER 4.04 (3.50) c.

- Start the Control Room Ventilation Booster Fan (0.5) BF-0-1A.

Stop the Control Room Ventilation Supply Fans (0.5) SF-C-1A & 18.

- Essential control room personnel obtain self contained breathing apparatus and use as necessary (0.5),

Scram the reactor b.

- Leave Mode Switch in RUN Verify all control rods inserted After rods verified in, TRIP:

Main turbine ONE feed pump TWO condensate booster pumps

- TWO condensate pumps

- Ensure that both RFP turbine turning gear control switches are in AUTO

"" is 0.25)

(each item marked with a REFERENCE AP 2.4.8.5, Rev 5, pg i EP 5.2.1, Rev 16, pg 1 295016G010 3.8*/3.6*

295016G010

...(KA'S)

QUESTION 4.05 (1.50)

Who may perform the independent verification of control rod movements performed with the Reactor Mode Switch 1., STARTUP or RUN?

THREE C3) required for full credit.

(1.5)

ANSWER 4.05 (1.50)

- licensed operator Reactor Engineering representative

- qualified STA (0.5 each)

REFERENCE GOP 2.1.10, Rev 15, pg 2 201001G001 3.7/3.7 201001G001

...(KA*S)

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

ht__iBQQEQUBES.:_U9Bd6Lt_eBBQBdeLc_EUEBQENQ1.6UQ PAGE 33 BoDIQLQQIQ6L_QQU1BQL QUESTION 4.06 (2.00)

What THREE (3) Group Isolations are exp;cted to occur anytime an automatic or manual scram occurs from a normal operating water level, and WHAT causes them to occurt Assume power operation prior to the scram.

(2.0)

ANSWER 4.06 (2.00)

Groups 2, 3,

6 (0.5 each) due to level shrink / void collapse as a result o' the scram (0.5).

REFERENCE GOP 2.1.4, Rev 27, pg 6 295006K301 3.8/3.9 223002G015 4.1/4.3*

295006K301 223002G015

...CKA'U)

QUESTION 4.07 (3.00) a.

To avoid operation in the instability region of the power / flow curve, reactor power should not exceed the

___(1)___

rod line when total core flow is ___(2)___.

(1.0) b.

What panel indication will alert the operator that flow instabilities are occurrin0?

(0.5) c.

If instabilities occur, what two methods are available to suppress the instabilities (excluding scram) and which is preferred?

(1.5)

~

s t _ _ E'B QGE Q U B E f _:_UQ Buo L t _ e Q UQ B d 6 L t _ E U E B Q ENGl_68Q PAGE 34 B6Q10LQQlCAL_GQUlBQL f

l i

ANSWER 4.07 (3.00) c.

1.

80% (0.5) 2.

< 45% or < 33 M1bm/hr (either at 0.5) b.

Accept LPRM or APRM oscillations or increased noise band on APRM i

recorder (0.5).

c.

- inserting control rods (0.5)

- increasing core flow (0.5) preferred method is to reverse the actions that caused the flux oscillations (0.5).

REFERENCE GOP 2.1.10, Rev 15, pg 4, 10 202001G010 3.5/3.7 l

295001G011 3.9/4.2 295001G011 202301G010

...(KA'S)

QUESTION 4.08 (3.00)

For each of the following conditions (a - d), state which E0PCs), if any, should be entered.

(3.0) a.

Reactor water level 11" Orywell pressure 1.87 psig Orywell temperature 138 deg F b.

Suppression pool temperature 97 deg F Reactor Building Sump A (NW Quad) indicates 40" (HI-HI at 34")

Turbines (main, RFP, HPCI, RCIC) tripped / ceased inj ect ing on high reactor level c.

Reactor pressure 400 psig Reactor water level

-45" Suppression pool level

-1.0" Suppression pool temperature 92 deg F RCIC isolated due to high temperature in steam pipe area d.

Reactor water level 15" Reactor pressure 1000 psig Orywell pressure 1.9 psig

hi__dB00EDUBES_:_UDBdekt_eBUQBdebt_EdEBGEUG1.66Q PAGE 35 GeD10LQQ1 gel _CQUIBQL ANSWER 4.08 (3.00) a.

E0P-1 RPV Control b.

E0P-2 Primary Containment Control E0P-3 Secondary Containment Control c.

E0P-1 RPV Control E0P-3 Secondary Containment Control d.

none (6 at 0.5 each)

REFERENCE CNS E0Ps 295026G011 4.4*/4.6*

295031G011 4.2/4.6*

2950320011 4.1/4.2*

295036G011 3.8*/4.15 295036G011 295032G011 295031G011 2950260011

...(KA'S)

QUESTION 4.09 (2.50)

A loss of all site AC power has occurred. Answer the following questions l

concerning E0P 5.2.5.1, Loss of All AC Power Station Blackout.

a.

What reactor water level indication (s) are available in the control i

room following this event?

(1.0) b.

What reactor water level indication (s) are available outside the control room following this event?

(0.5) j l

c.

What are TWO (2) negative consequences or concerns regarding excessive drywell temperatures during this event?

(1.0)

I J

\\

'it__fB00E09BES_:_UQBU6Lt eBUQBdelt_EdgBQgNQ1_eUQ PAGE 36 88DIQLQQ1CeL_CQUIB0L ANSWER 4.09 (2.50) a.

The 3 GEMAC's and ossociated recorder on panel 9-5 (1.0).

j b.

The Yorwayo may be monitored locally in the Reactor Building (0.5).

c.

- erroneous reactor water level indications failure of electrical components, 1.

e.

wiring, solenoids, etc.

unequal expansion of refueling bellows flange (2 required at 0.5 each)

REFERENCE CNS E0P 5.2.5.1, Loss of All Site AC Power Station Blackout, Rev 4, pg 3 CNS AP 2.4.8.4.2, Rev 11, pg 2 l

295003A202 4.2*/4.3*

295028K102 2.9/3.1 295028K102 295003A202

...(KA*S)

QUESTION 4.10 (2.00)

Indicate whether each of the following statements is TRUE or FALSE:

(2.0) a.

A reactor startup is NOT permissible under natural circulation flow conditions.

b.

A reactor startup is NOT permissible with only one recirculation pump in operation.

c.

If the reactor is operating at power (both recirculation pumps in operation) and one recirculation pump trips, reactor operation may continue for no more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

d.

A reactor recirculation pump mey not be started if the reactor is in natural cieculation flow and reactor power is greater than 1%.

(

i

'at__$80GE09BES_:_UQBdelt_eBUQBU6L&_EUEBGEUGY_eUQ PAGE 37 BeQ1DL901CeL_CQUIBQL 1

ANSWER 4.10 (2.00) c.

True b.

False c.

False d.

True (0.5 ench)

REFERENCE S0P 2.2.62, Rev 27, pg 13, 14 CNC Technical Specifications 3.3.F.

202001G005 3.4/4.2*

2020010011 3.4/4.2*

202001G011 2020010005

...(KA'S)

QUESTION 4.11 (2.50)

A condition arises which requires entry into a high radiation area.

The operator entering the area will receive a whole body dose of 40 mrem.

The following personnel, with their related personal information, are available to do the work:

(2.5)

NOTE:

Each exposure below (qtr, yr, life) includes the exposure above it.

Candidate 1

2 3

4 Sex male male female male Age 27 38 24 20 Today's exposure 50 mrem 10 mrem 10 mrem 20 mrem Wkly / exposure 260 mrem 150 mrem 150 mrem 250 mrem Qtr/ exposure 870 mrem 600 mrem 485 mrem 920 mrem Yr/ exposure 2200 mrem 29G5 mrem 500 mrem 2810 mrem 54730 mrem 5200 mrem 9770 mrem Life exposure Remarks history 4 months i

unavailable pregnant Each candidate is technically competent and physically capable of perform-ing the task.

Emergency limits do not apply and time constraints do not permit obtaining authorization for en exposure limit increase.

Which condidates have acceptable exposure margins to perform the task?

Indicate the reason (s) for rejecting a candidate for the job, if applicable.

l l

3t__eB00EQuBES_:_NDBdebt_6EU9Bdokt_EdEBQEN01.6ND PAGE 38 BoDIQLQQIC8L_CQUIBQL ANSWER 4.11 (2.50)

Candidate #1:

Acceptable (0.5)

Cendidate #2:

Rej ected (0.25) since he will exceed the yearly limit of 3000 MREM (0.5).

Candidate #3:

Rejected (0.25) since she will exceed 500 MREM / GESTATION PERIOD (0,5).

Candidate #4:

Acceptable (0.5)

REFERENCE CNS HPP 9.1.1.3, Rev 23, pg 13, 14 CNS HPP 9.1.2.1, Rev 18, pg 5, 6

294001K103 3.3/3.8 294001K103

...(KA*S)

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NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION 1

FACILITY:

_QQQPgB._________________

REACTOR TYPE:

_BWB-qed _________________

DATE ADMINISTERED:_ Bald 2fik________________

EXAMINER:

_QBe2Elt_Qt______________

CANDIDATE:

lukIBUCIl001_IQ_CeUQ10eIEL Use separate paper for the answers.

Write snswers 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 giade of at least 80%.

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

% OF i

Subject:

Comments - NRC Examination Administered February 16-19, 1988.

Dear Mr. Graves:

Attached please find our comments on the questions and answers associated with the subject examinations given on February 16, 1988. We believe that our concerns and comments should be considered in the grading of the written exams administered and, in some cases, serve as a basis for refining several of the examination questions and answers.

Please contact us if you would like to pursue our comments further or if additional clarification is desired.

Sincerely,

/

$1%

'. G. K nci Manager - Nuclear LGK/sjt/LETS27 Attachment cc:

G. A. Trevers w/ attachment i

G. R. Horn w/ attachment

~ *!'5 s

J. W. Dutton w/ attachment v-K. C. Walden w/ attachment R. Brungardt w/o attacheent G. H. Reece w/o attachment l

K. K. Finn w/o attachment l

TS File l

t i

I I

I l

QUESTION 1.05 (2.00)

Why does core thermal power decrease at a much slower rate than indicated neutron power (2 reasons required) following a scram from high power operation?

(2.0)

ANSWER 1.05 (2.00)

Thermal power drops at a slower rate due to decay heat (1.0) and the time delay in getting the previously generated heat out of the fuel pellet into the coolant.

(1.0)

REF CNS Reactor Theory, Pg. 7-22 292008K125 2.8/2.9 292008K130 3.2/3.5 292008K125 292008K130

...(KA's) 1.05 COMMENT The residual heat of components in the vessel will also reduce the rate of thermal power decrease.

RECOMMENDATION Also accept residual heat of components.

REF l

NPR Student Text Pg. 14

QUESTION 2.01 (3.00)

In the Reactor Core Isolation Cooling (RCIC) system:

a.

What function does the RCIC lube oil cooler water perform after leaving the cooler?

(1.0) b.

How would failure of the barorietric condenser affect operation of the RCIC system?

(One effect required)

(1.0) c.

How is the turbine exhaust line protected from overpressure during system operation? Two required.

(1.0)

ANSWER 2.01 (3.0) a.

It is used as the condensing medium for the barometric condenser.

(1.0) b.

System isolation (1.0) due to high area temperature c.

Rupture diaphragms which exhaust to RCIC room (0.5)

Turbine trip on high exhaust pressure (0.5)

REF RCIC Pg. 6, 7, 21 217000K404 3.0/3.1 217000K405 3.2/3.5 217000K405 217000K404

...(KA'S) 2.01.b COMMENT An airborne problem would occur. The student text states the purpose of the barometric condenser is "to prevent any leakaga of the radioactive steam into the environment." The purpose does not include preventiag high temperatures in the area.

REC 0KMENDATION Accept airborne problem as an answer for part b.

REF Reactor Core Isolation Cooling Studen'c Text Pg. 7.

QUESTION 2.02 (1.00)

How is SLC systen flow capability verified without opening an injection path to the reactor vessel?

(1.0)

ANSWER 2.02 (1.00)

Starting the pumps locally does not fire the squib valves (0.5) and the flow is to the test tank (0.5).

REF SLC-14 211000K408 4.2*/4.2*

211000K408

...(KA's) 2.02 COMMENT The surveillance starts the pumps locally and goes from test tank to test tank.

RECOMMENDATION:

Also accept perform the appropriate surveillance.

REF Procedure 6.3.8.1 1

l l

?

i I

t

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

a QUESTION 2.03 (3.00)

Answer the following with regard to the RHR system and its various modes of operation:

a.

Match the following actions, events, or interlocks in Column A with the pressure in Column B that initiates or allows that item. (2.0)

Column A Column B 1.

Shutdown cooling isolates 50 psig 2.

Allows manual operation of the 75 psig LPCI injection valve 100 psig 2.

Automatically opens the LPCI 135 psig injection valve 277 psig 4.

Input to ADS 400 psig 432 psig 450 psig b.

What is the most limiting failure of a single component in the RHR system with regard to core protection?

(1,0)

ANSWER 2.03 (3.00) a.

1.

75 psig 2.

450 psig 3.

450 psig 4,

100 psig (0.5 each) b.

LPCI injection valve failed shut (1.0)

REF l

RHR-24, 37 205000K402 3.7/3.8 203000K106 3.9/3.9 203000K402 3.3/3.3 203000A211 3.4/3.6 205000K402 203000K402 203000K106 203000A211

...(KA's) 2.03..

COMMENT The setpoint for the input to ADS is required to be set > 100 psig and i

< 165 psig.

Therefore either 100 psig or 135 psig is an acceptable answer.

REC 0KMENDATION l

Accept either 100 psig or 135 psig or both for number 4 in column A.

l l

L

e

-1 QUESTION 2.03 (3.00)

(CONTINUED)

REF Technical Specification Table 3.2.B NPR Student Text Pg. 14 2.03.b COMMENT The injection valve is a normally shut valve therefore if it fails it will be in the shut position.

RECOMMENDATION Do not require "shut" for full credit.

REF Introduction to CSCS Student Text Pg. 11.

QUESTION 2.09 (4.00) a.

List ten (10) of the valve operations that automatically occur on a turbine trip signal.

Assume the plant is operating at power when the turbine trip occurs.

(2.0) b.

What are four (4) of the five (5) protective devices that operate independently on the main turbine to prevent damage to the unit if the turbine was not tcken out of service immediately?

(2.0)

ANSWER 2.09 (4.00) a.

1.

turbine main stc.) valves close 2.

control valves close 3.

reheat stop valves close 4.

intercept valves close 5.

bypass valves open 6.

feed system startup flow control isolation valves open 7.

feed pump discharge valves close 8.

extraction steam non-return valves close 9.

extraction steam dump valves open 10.

feed pump low pressure steam supply valves close 11.

main turbine governor valve drain valves open 12.

MSL drain valves shift from AS/B5 heaters to the main condenser (10 required at 0.2 each) b.

1.

overspeed trip mechanism 2.

low vacuum trip 3.

low bearing pressure trip 4.

thrust bearing trip 5.

solenoid trip (0.5 each)

REF MN TURB-16, 17 245000A201 3.7/3.9 245000G007 3.5/3.6 245000G007 245000A201

...(KA'S) 2.09.a COMMENT The extraction steam non-return valves trip and act as check valves on a turbine trip vice close.

RECOMMENDATION Accept that the extraction steam non-return valves trip instead of extraction steam non-rett in valves close.

REF p.

14.

Extraction Steam and Heater Drains Student Text g

QUESTTON 2.11 (2.00)

Match the following plant areas (a - h) with the type (s) of fire protection system (1 - 4) that is (are) available in that area:

(2.0)

NOTE: MORE THAN ONE TYPE OF SYSTEM MAY APPLY TO EACH AREA a.

Reactor feed pump room 1.

Fire Water System b.

Service water pump room 2.

Carben Dioxide System c.

Diesel generator day tank rooms 3.

Halon 1301 System d.

Fire pump house - diesel fire pump 4.

No Automatic System room Available e.

Control building, cable spreading room (918')

f.

Turbine generator bearings 1, 2, 3 g.

Main control room ANSWER 2.11 (2.00) a.

I b.

3 c.

I d.

1 e.

1, 2 f.

2 p.

4 (a - g at 0.25 each)

REF FP system description 286000G004 3.8/3.9 2860000004

...(KA'S) 2.11 COMMENT The answer to part c should be 2 (Carbon Dioxide) vice 1 (Fire Water System).

RECOMMENDATION Accept 2 instead of I for part c.

REF Fire Protection Student Text Pg. 15.

i l

L l

i QUESTION 3.01 (3.00) a.

STATE whether the solenoids associated with the following valves are NORMALLY Energized or Deenergized. NO SCRAM SIGNAL EXISTS. (1.0) 1.

Back-up Scram Valves 2.

Scram Discharge Volume Vent and Drain Valves b.

Repositioning the Mode Switch from STARTUP/ HOT STANDBY to RUN causes certain reactor scram functions to be bypassed and others to be effective. LIST the three (3) scram functions (or setpoints) which are bypassed AND the three (3) scram functions (or setpoints) which becone effective when the Mode Switch is taken to RUN (2.0)

ANSWER 3.01 (3.00) a.

1.

Deenergized (0.5) 2.

Energized (0,5) b.

Activated in RUN:

- MSIV Closure (0.33)

- Companion IRM/APRM (0.33)

- APRM 118% HIGH Flux (0.33) l Bypassed in RUN:

- IRM Inop (0.33)

- IRM Upscale (0. 33)

- APRM 15% HIGH Flux (0.33)

REF CRDH-17, 19 RPS-14, 16-18 212000K108 3.0/3.1 212000K412 3.9/4.1 212000A216 4.0/4.1 212000A216 212000K108 212000K412

...(KA'S) i 3.01 b COMMENT The APRM High Flux setpoint should be a flow bias formula instead of 118%.

RECOMMENDATION Accept the formula for the APRM flux.

REF l

RPS Student Text Table 2 f

QUESTION 3.08 (1.50)

Assume both recirculation pumps are running at 80% speed. State how the recirculation pumps' speed is affected by each of the conditions below.

Consider each case separately.

(1.5) a.

The operator closes the B recirculation pump discharge valve to the t

mid position, b.

Two reactor feed pumps are operating and a feedwater problem causes RPV level to temporarily decrease to 20".

c.

One of the two operating reactor feed pumps trips and the reactor scrams on low level.

ANSWER 3.08 (1.50) a.

(The B MG set trips and) the B recirc pump coasts to a stop (0.5) b.

No effect (0.5) c.

Both pumps runback to 45% speed (0.5)

REF Recirculation System, Figure 14 202001A211 3.7/3.9 202001A212 3.6/3/8 001A 2b2001A223 202001A212 (KA'S) j 3.08 COMMENT On a scram and the subsequent generator trip, the recirculation pump on 1

the normal transformer will trip.

Also feedvater vill be reduced to below 20% causing the running recirculation pump runback to 20%.

RECOMMENDATION The answer as stated is correct initially for the conditions stated.

i however if the candidate carried through with scenario he vi 1 give more information. We have just recently changed one of our exam bank questions to reflect this also.

REF Procedure 2.1.5 i

l f

,_,m.

_,,--.__.,_,.,_,._,...__.,._.e_..

_,,_-y

,,,, _m,

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

QUESTION 4.09 (2.50)

A loss of all site AC power has occurred.

Answer the following questions concerning E0P 5.2.5.1, Loss of All AC Power Station Blackout, a.

What reactor water level indication (s) are available in the control room following this event?

(1.0) b.

What reactor water level indication (s) are available outside the control room following this event?

(0,5) c.

What are TWO (2) negative consequences or concerns regarding excessive drywell temperatures during this event?

(1.0)

ANSWER 4.09 (2.50) a.

The 3 GEMAC's and associated recorder on panel 9-5 (1.0) b.

The Yarways may be monitored locally in the Reactor Building (0.5) c.

- erroneous reactor water level indications (0.5)

- failure of electrical components, i.e.,

viring, solenoids, etc.

(0,5)

REF i

(NS E0P 5.2.5.1, Loss of All Site AC Power Station Blackout, Rev 4. Pg. 3 295003A202 4.2*/4.3*

2950.93K102 2.9/3.1 295028K102 295003A202

...(KA'S) 4.09 COMMENT As stated in the abnormal "Ventilation System Failure - Loss of Coolers in the Dryvell" a concern with high dryvell temperatures is unequal expansion of the refueling bellows flange.

RECOMMENDATION Also accept unequal expansion of the refueling bellows flange.

REF Abnormal Procedure 2.4.8.4.2 t

I r

{

o QUESTION 5.02 (2.50) a.

HOW will the Shutdown Margin (Reactivity Margin) just prior to a refueling outage compare with the Shutdown Margin following the refueling? WHY? Two (2) reasotis required.

(1.5) b.

At what point in cycle life is compliance with the Shutdown Margin Technical Specification verified?

(0.5) c.

What reactor conditions must be present for the verification to be reasonably accurate?

(0,5)

ANSWER 5.02 (2.50) a.

SDM prior to the outage will be larger (0.5) due to fission product poisoning (0.5) and fuel depletion. (0.5) b.

initial fuel loading or refueling (0.5) c.

cold, xenon free (0,5) 2 REF LOTHM-TH-4.11-0, Shutdown Margin, Pg. 5 CNS Reactor Theory, Pg. 1-35 and 36 CNS Tech Spec 3.3.A Bases K/A 292002 Kl.14 2.6/2.9 292002K114

...(KA'S) i 5.02 b C0KKENT i

Fuel loading occurs at the beginning of the cycle.

RECOMMENDATION Accept BOL

QUESTION 5.03 (3.00) a.

Does the magnitude of the initial level of source range counts affect the critical rod position? WHY?

(1,0) b.

The reactor is brought critical at 40 on IRM range 2 with the shortest permissible stable positive period allowed by C0P 2.1.1, "Cold Startup." Heating power is determined to be 40 on range 8 of IRM's.

- - - SHOW ALL WORK ****

1.

What is the doubling time if the period remains constant?

(1,0) 2.

How long vill it take for power to reach the point of adding heat if the period remains constant?

(1.0)

ANSWER 5.03 (3.00) a.

No (0.5).

The critical control rod position is a function of Keff or reactivity of the reactor and is not a function of the source count rate.

(0.5) b.

1.

From GOP 2.1.1, shortest permissible stable period equals 50 sec.

(0.5)

Thus doubling time equals 50/1.44 - 34.7 seconds (0.5) 2.

40% range 2 is equal to 0.04% on range 8 P(0) = 0.04 P(t) = 40 Period = 50 seconds P(t) = P(0) e"(t/ period) 40

= 0.04 e*(t/50 see)

Tine = 345.4 seconds or 5 min. 45 sec.

(1.0)

(NOTE: Grade cethod if period is different)

REF LOTM-TH-4.15-1 CNS Reactor Theory, Chapter 3 292003K108 2.7/2.8 292008K104 3.3/3.4 292008K!04 292003K108

...(KA'S)

QU"STION 5.03 (3.00)

(CONTINUED) 1.06 a 5.03 a COMMENT This could The question asked for initial level of source range counts.

be interpreted as either the initial counts when the startup begins or source strength.

At the beginning of startup the initial level of source The k vill range counts will be based on source strength and k affectboththeinitialcountsatthebeginningofastartupa$bI the point that the reactor is critical. Thetefore if the candidates discussed the initial level of source range counts they would have answered yes with an appropriate explanation. The students could base their answer on whether the question was asking for source strength or initial counts on the startup.

RECOMMENDATION Also accept yes if the candidate provides an appropriate explanation.

REF Reactor Physics, page 3-7 and 3-8

A 4

QUESTION 5.06 (2.00)

For each of the pairs of conditions listed below, state WHICH condition would have the GREATER differential rod worth and briefly. EXPLAIN WHY.

a.

Reactor moderator temperature of 150 F or 500 F (1.0) b.

For a rod position 10 or position 40 of a core operating at 100%

power (assume BOL)

(1.0)

ANSWER 5.06 (2.00)

At 500 F (0.5) As moderator temperature increases, neutron leakage a.

out of the fuel bundles in increased, thus the control rod is exposed to higher neutron flux and rod worth increases.

(0.5) b.

At 40 (0.5) the core will be bottom peaked and the rod will be traveling through on area of high flux.

(0,5)

REF CNS Reactor Theory, pgs 5-27, 28 292005K110 2.8/3.3 292005K110

...(KA'S) 5.06.b COMMENT Students could say position 10 if they consider the fact that all other rods will probably be deep (00-16) which would reduce the worth of any shallow rods (32-48) due to shadowing.

One section of the Reactor Theory text describes shallow rods as being shaping rods and deep rods as power rods.

If the students interpret the question to ask only about a single rod at BOL when flux is bottom peaked, they could answer with position 40.

RECOMMEKDATION Accept either position 10 or position 40 if they provide an appropriate explanation.

REF Rx Theory Text Rev 1 CH 5 Page 22, 25 l

T l

l

(

l i

l

4 4

QUESTION 7.10 (2.50) a.

What is the purpose of the "Emergency Rod Movement Sheet"?

(0,5) b.

List four (4) examples or conditiona which may require the use of an "Emergency Rod Movement Sheet".

(1.0) c.

If a power reduction from high power is required, what is the i

preferred method?

(0.5) d.

When using control rods to reduce power, which rods should be selected, if possible? Answer with regard to the axial position of the rods selected.

(0.5)

ANSWER 7.10 (2.50) a.

To document non-scheduled, out-of-sequence rod movements (0.5) b.

- alleviate thermal limit problems

- power reduction due to transient

- recover from rod drift

- recover from single rod scram (0.25 each) c.

reduce recirculation flow (0.5) d.

deep rods /more than 2/3 inserted into the core (0.5)

REF NPP 10.13, Rev 15 Pg. 2 201003G001 3.6/3.7 201003G001

...(KA'S) 4 7.10 b COMMENT Procedure 10.13 gives examples of when "power reduction due to transient" applies.

The candidates may have given examples such as recirculation i

pump trip, reduced feedwater heating, or excessive river debris clogging i

intake screens.

i REC 0KMENDATION i

i Also accept any of these individual events as transient answers.

REF Procedure 10.13, page 2 I

i f

I

1 QUESTION 8.01 (3.50) a.

List the four general conditions that would require the issuance of a Special Work Permit (SWP). Values NOT required.

(1.0) b.

What is the NORMAL maximum length of time that a SWP will be in effect?

(0.5)

P c.

What is the maximum extension that can be granted for a SWP7 (0.5) d.

TRUE or FALSE.

Health Physics personnel are exempt from the SWP issuance requirement during the performance of the radiation and c

contamination surveys for SWP evaluation.

(0.5) e.

Once a SWP is initiated and authorized, copies are kept at WHA 1 FOUR (4) locations?

(1.0)

ANSWER 8.01 (3.50) a.

- High area radiation t

- High airborne contamination 4

- High surface contamination

- Industrial hazards (0.25 each) b.

1 month + 7 days (0.5) c.

7 days (0.5) d.

True (0.5) l I

e.

- job site

- SWP board

- HP office 4

- SS office (control room)

(0.25 each)

REF i

HPP 9.1.1.4, Rev 16, Pg. 4, 6, 7 1

294001K103 3.3/3.8 l

294001K103

...(KA'S) 8.01 e COMMENT i

The SWP board is mounted.in the laundry supply area on elevation 903, i

RECO.9tENDATION Accept either SWP board or laundry supply area.

l i

1 i

4

a QUESTION 8.08 (3.50) e a.

List five (5) devices / actions that would be considered temporary modifications per Conduct of Operations Procedure 2.0.7, Plan:

Temporary Modification Control.

(2.5) b.

What FOUR (4) persons, by title, may perform the Safety Evaluation / Technical Review for temporary modifications?

(1.0)

ANSWER 8.08 (3.50) a.

- jumpers

- lifted leada

- fuse removal

- blocked relay

- booted contacts

- installed breaker test blocks / actuator links

- mechanical jumper

- installed / removed blank flanges (5 required at 0.5 each) b.

The System Engineer (0.25) or on duty STA (0.25) in collaboration with the System Engineer and the CRS (0.25) or another SR0 (0.25) when the CRS is not on site.

REF Control of Operations Procedure 2.0.7, Plant Temporary Modifications Control, Rev 4. Pg. 3, 6 294001A103 2.7/3.7 294001A103

...(KA'S) 4 j

8.08 COMMENT Some candidates may provide specific devices that require PTMs instead of listing the categories listed in the procedure.

The PTM procedure also lists a category of "other."

RECOMMENDATION Also accept specific items such as spool pieces as devices that require PTMs.

REF Procedure 2.0.7 page 6

t e

.a QUESTION 8.10 (2.00)

During refueling operations individual responsibilities are assigned in Procedure 10.21, "Special Nuclear Materials (SNM) Control and Accountability."

a.

Who MUST direct the SNM handling operations involving SNM movement BETWEEN Item Control Areas?

(0,5) b.

Who MAY direct the SNM handling operations involving SNH movement WITHIN Item Control Areas?

(0.5) t c.

Who functions as the SNM Executor?

(0.5) d.

Who functions as the SNM Checker?

(0,5)

ANSWER 8.10 (2.00) a.

An individual holding an SRO license (0.5) b.

An individual holding an R0 license (0.5) c.

Control Room Refueling Monitor or Control Room Operator (accept either 0.5) d.

Refueling Floor Supervisor or SRO on the Refuel Floor (accept either 0.5)

REF Nuclear Performance Procedure 10.21, Special Nuclear Materials Control l

and Accountability Instructions, Rev 1 Pg. 3 234000G001 3.4/3.8 234000G001

...(KA'S) r 8.10 COMMENT f

I Procedure 10.21 also states that the SNM checker can also be any individual who did not function as SNM Executor.

RECOMMENDATION e

Also accept any RO or SRO who did not function as the SNM Executor.

REF 1

Procedure 10.21, page 3 l

_ - _ _ _ - _ _,.

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