ML20127D467

Submits Comments Re Reactor Operator Written Exam on 850319.Preparation of Comments Made More Difficult by NRC Policy Requiring Examiner to Collect Exam/Answer Key Immediately After Review.Related Matl Encl

ML20127D467
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	03/28/1985
From:	Spinney R VERMONT YANKEE NUCLEAR POWER CORP.
To:	Kister H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML20127D270	List: IR 05000271/1985012 ML20127D267 ML20127D467
References
VYV-085-178, VYV-85-178, NUDOCS 8506240238
Download: ML20127D467 (42)

Text

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l VERMONT Y ANK EE NUCLEAR POWER CORPOR ATION i

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t o ia m m (F Wil: Nt il: ilCNT ltOAD VEi!NON. VEltMONT n.UM VYV-085-178 March 28,1985 Harry Kister Regional Branch Chief U.S. PC, Region I 631 Park Avenue King of Prussia, PA 19406

Dear Mr. Kister:

On the basis of the 1984 Amendment to Examiner Standard, ES-201, of MJREG-1021, Operator Licensing Examiner Standards, Vermont Yankee herewith submits comments regarding the Reactor Operator Written Examination given MErch 19, 1985 by Brian Hajek.

These comments are meant to reinforce those made during the facility review of the examination and to provide clarifying documenta-tion of problem areas. As a point for further discupsion, the preparation of these comments was made more difficult by the M C s current policy that requires the exarciner to collect all copies of the examination and answer key immediately following the review.

Thank you for your attention to this matter and please do not hesitate to contact us should you have any questions regarding this submittal.

Very truly yours, VERMONT YA* EE NUCLEAR POWER CORP.

AW Randall W. Spinney Training Manager RWS:nkg b

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VERMONT YANKEE NUCLEAR POWER CORPORATION e'

COMMENTS ON REACTO$ CPERATOR LICENSE EXAMINATION Administered March 19, 1985 7

by Brian Majek Question cl'.5. concerning pump " efficiency"

+

After finding the reference for the answer, which is Thermodynamics, Heat Transfer and Fluid Flow (G.E.,1983), we still feel that the answer suggested by the key is an interpretation of a vague statement. The text does not say that throttling is less efficient but that " changing the pump speed is preferable since throttling invariably leads to a waste of power" (p.7-117). Overall system efficiency is reduced as is implied by this statement but it does not address " pump efficiency".

s.

Question 1.6 concerning Rod Worth The question has no specific direction for the candidate to con-sider the Rod's Worth prior to, during or af ter the rod's movemenf.

The problem arises that when the candidates discussed this topic with,

other licensed operators during control room training, they were intro,

duced to the concept of " rod shadowing." This conceot could lead to gdifferent answers based on whether the candidate asstmed the rod was gr pas not still moving. The accompanying excerpt from Gentral Electric s

, Reactor Theory Text, which was reference material formerly u: red at

' Vermont Yankee, is offered to document this conclusion.

Y i

Question 1[9 conceming "Cargonder" 7

O Increased steaq carryunder from the moisture separators will cause an increase in downecaer,'enthalpy, which could result in:

1) Decreased WSi of the Recirculation Punps (due to incretsed,

saturation conditions)

2) Decreased PSH of the Jet Pumps, and

3) Decreased core. inlet subcooling.

,)

Question 2.3 concerning Residual Bus Voltage Transfer e

All documents referenced dceing the training program do discuss the increased insurge of current that results when the Feec Pump is not tripped during a Residual Bus Voltage Transfer.

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o Question 2.4 concerning Relief Valve Bellows Failure The automatic ADS operation of the Relief Valves is not affected by a bellow failure. The ADS logic energizes the same solenoid-operated air valves as do the ORP 9-3 switches. See the enclosed W CWD (751, 752).

Question 3.2 concerning SRM Retract Permissive As the sama answer key states any one of the following four conditions will preclude a rod block," we would like to add:

"That StM channel bypassed by the CRP 9-5.bystick."

The reverse wording of the question makes it difficult to discern the correct answer from the answer key.

Question 3.5 concerning Loss of Instrument Air The definitive wording in the answer key which QJotes the referenced procedure does not allow for latitude in the candidate s answer.

Plant operating experience shows that the 105 (Start-Up) Feed Reg Valve shuts and the mechanical design of the 55% (Normal) Feed Reg Valves internals cause them to drift open on a loss of air (Reference OP 2172 pg 10). The MS1Vs are equipped with acetsnulators which should hold them open when their source of pneumatics is lost (Reference P+1D 191167).

vy Ro, EXAM 3/1e/ss O

QUESTION 1.s REFERENCE V

The effects of throt tling a centrifugal pump to achieve dif ferent flow rates through a system are shonn in Figure 7-55.

If the speed of the pump remains cc stant, the pum p chcrocteristic curve remains the same. Havever, by closing donn on the throttle volve (which is considered part of the system) the system curve rises. This should be expected since the throttling process is completely irreversible and leads to increased turbulence and head loss in the system. Note that the pump must provide a larger driving head in order to deliver this reduced dopocity. A second method used to vary the flow i

cepocity is by changing pump speed. Figure 7-58 also shows J

how the pump head verses capacity curves change with pump speed. Decreasing pump speed to change flow capacity has the advantage that it does not increase the system head losses and hence the system curve remains unchanged. The operating point also falls lower on the system curve so less pump heod must be developed. Of the two ways to control flow, changing the pump speed is preferable since throttling invariably leods to a waste of power.

Both the methods of throttling and speed control to change flow capacity are demonstrated by the BWR recirculation systems.

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, ouE"JION 1.8 MFERENCE

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3.

Tuel temperature - Since fuel temperature affects primarily fast neutrons which are resonantly captured, and control rods are a ther=al neutron abosrber, fuel temperature and control rod worrh are essentially inde-pendent of each other.

4.

Core age - The' effects of core ' age on control red worth are very complex, curve shown in but essentially rod worth follows the shape of the Kdf Figure 12-19. This can be seen from equation 12-16. Flux changes very little over core life and neither does theraal diffusion length, but the physical size of the core goes up considerably as the control rods are Early in life as withdrawn. Thus control rod worth varies with Edf.

fission products build up control rods are withdrawn, increasing core size C'

and decreasing worth.

As the burnable poisons start to deplete control

.s.

and rod worth rods are again inserted into the core to control Edf increases; finally after the poisons have stabilized and burnable poisons are gone,' fuel depletion becomes the predt=4 mnt factor.

Control rods are withdrawn to keep the thermal utilization factor f relatively con-stant. As the rods are withdrawn core size increases and rod worth de-In summary, then, rod worth varies in t.he same manner as the creases.

I cold, clean E value varies with ageo df 5.

Control rod density - The more control rods -inserted into the reactor the lower the worth of each individual rod, because each rod will be absorbing a lower percentage of the avsf1shle neutrons.

This effect

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i is called rod-shadowing.

Consider the example shown in Figure 12-18.

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aussTloN t.6 M:FERENCE.

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Assume initially that all control rods are fully inserted.

Each control rod then can be seen as absorbing (controlling) the thernal neutrons within the four fuel bundles adjacent to the control red. New assu=e that the four exterior control rods are withdrawn.

It can be seen that the single remainins control rod is effectively controlling the neutrons in 16 bundles instead of four.

This implies that the worth of that single inserted rod is much more than when the other four were present.

The reason.for such high worth can also be explained as follows:

• ten centrol rods 3, D, F, and G vere withdrawn, and E was still fully

=

inserted, the " effective" core consisted of not one, but four sna11 four bundle reactors which were essentially unconnected because neutrons

\\.i from one couldn't pass through the inserted control rod blade to cause fission in the others.

Normally, when a control rod is withdrawn, it adds the four fuel bundles which surround it to the effective size of the core.'~ But in this caserby withdrawing the central control red,16 bundles are immediately coupled.

In other words, the withdrawal of the control rod transformed four 4 bundle reactors into one 16 bundle reactor.

'Dius the size of the control rod is effectively much larger than its actual physical size would suggest.

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. CIUESTI ON 3.5 REFERENCE

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y 2.

If the control rod pattern is at or near the 100% load

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li.e, it will be necessary to insert rods to prevent reu bleck and possibly scram.

Insert rods as necessary in reverse order of the rod withdrawal sequence.

3.

If a reactor scram occurs, refer to OP 3100, Reactor Scram.

4 If power is subsequently increased after the transient, observe appropriate PCIOMR limits.

e B.

Feedwater Control Valve Lockup l.

Feedwater control valves (FDh'-FCV-12A or 12B) will lock in the "as is" position on loss of air or loss of signal from the control system. The 10% control valve will shut on loss of air.

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.....A L' hen the feedwater control valves lock,uph' f;..ef:g,

NOTE:

"as is", they do have a Tendancy to d9fft open F' i

2.

Attempt tc reset the lockup by pressing the Reset '.*alve A or Valve 3 pushbuttons en CRP 9-5.

3.

If the valve (s) fails to reset, attempt to avert a reac-tor water level scram (+132") or feed pump trip (+177"):

g,[

Observe air pressure low alarms and, if present, a.

dispatch an operator to correct the situation.

b.

If level is trending downward, atte=pt to control level by opening 10% Control Valve Station FCV-13 and/or reducing reactor power.

If level is trending upward, consider increasing c.

power level or securing one feed pu=p.

4 If water level cannot be controlled by performing 3a, b, or c, scra= the reactor and refer to 0.P. 3100.

C.

Re=oving a High Pressure Heater String from Service

+

NOTE:

Refer to the Turbine Manual, GEK 5585, Vol. 1, Tab. 2, and 0.P. 2160,.for turbine operating considerations when bypassing heaters.

CAUTION: Notify Reactor Engineering if possible prior to bypassing I

heaters. Removing feedwater heaters from service reduces l

core inlet enthalpy resulting in an increase in thermal power level and a shift in power shape. If fuel is operating at or near the pre-conditioned envelope in -

l any region of the core, these power changes can result in PCIOMR violations and consequently :uy cause fuel failures due to pellet clad interaction.

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,Q U.S. NUCLEAR REGULATORY L0letISSION y

REACTOR OPERATOR LICENSE EXANINATION Facility:

Vermont Yankee Reactor Type:

BWR Date Administered: March

. 1985 Examiner:

Br1an K. Ha 1ek l

M STED B dy Candidate:(Print)

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 705 in each category and a final grade of at least 805.

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

E of Category 5 of Candida'te's Category Value Total Scor%

Value Category 25 25 1.

Principles of Nuclear Power ng Plant Operation Thermo-

~

@namics, Heat Transfer and Fluid Flow 25 25

~

2.

Plant Design. Including Safety and Emergency

{

Systems 25 25 3.

Instruments and Controls 25' 25 4.

Procedures - Normal Abnormal, Emergency,, and 4

Radiological Control

{

100 TOTALS Final Grade i

All work done on this examination is my own.

I have neither given nor received aid.

Candidate's 51gnature t (

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March, 19;5 Vermont Y&nkoo D

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.

1 THERMODYNAMICS.

i HEAT TRANSFER, AND FLUID FLOW (25)

I 1.1 0.P.

0100, Reactor Startup to Criticality, warns of the possibility of unexpected high notch worths during withdrawal sequences.

Briefly explain under what conditions these high notch worths can be expected, and in what part of the core you would.

expect tglgliggly higher notch worths to be l

located?

(3.0) 1 1.2 Vermont Yankee procedures state that in order to commence a reactor startup, there must be a minimum count rate on the SRMs.

Where do the neutrons come j

from to produce this count rate?

(2.0) t 1.3 Core orificing is used to assure uniform flow through all fuel elements, somewhat independent of the power variations in individual fuel bundles.

t a.

Explain how the resistance to flow changes as the power increases in a fuel bundle.

(1.0) b.

How does core orificing assure relatively uniform flow independent of plant operating i 45 conditions and location in the core?

(2.0) 1.4 Indicated reactor water level at 100 percent power differs from the actual water level directly above the core (that which is present in the steam separators or within the dryer skirt).

i i -

a.

Which level (actual or indicated) is higher, and by how many inches?

(1.0) l b.

Explain why the above dif f erence occurs.

(1.5) i c.

With the reactor operating at 100 percent l

power, if one recire pump trips, will the water t

level above the core initially increase or i

decrease?

(0.5) l Category Continued on Next Page i

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.RO Examination Page 1 of 12 1

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M rch, 1985 Vermont Ycnkee 1.5 a.

For the centrifugal pump characteristic curve shown in Figure 1.1, show how the pump operating curve will change, and explain why the change occurs, if a valve in the system is throttled one-half closed, such as might be done in the Feedwater Control System.

Be sure to label all points and lines.

(1.5) 4*

b.

The RCIC System uses changes in pump speed to I

control flow rate.

Using the same figure.

y&

explain which flow control system is more Y#

efficient.

(1.5) plO h

Head GPM J

Figure 1.1.

Pump Characteristic Curve 1.6 Consider a single control rod positioned near the radial center of the core.

During a reactor startup near the end of life, this control rod will be the first control rod to be. withdrawn.

After the reactor has been made critical and 100 percent power has been achieved, if this control rod is fully inserted, will its reactivity worth be greater than, the same as, or less than it was at the time of reactor startup?

Fully explain your answer.

(3.0) 1.7 Describe the initial reactivity response (positive or negative) and the resultant effect on core power (increase, decrease, no change) for each of the following changes in plant parameters.

Include which reactivity coefficient is responsible for the change.

a.

Closure of one MSIV at full power.

(1.5) b.

Loss at feedwater heating.

(1.5)

Category Continued on Next Page O

RO Examination Page 2 of 12

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Vcrcont Ycnkoo

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Mrech, 1985 O

1.8 The reactor is critical at "50" on Range 2 of the IRMs.

A control rod is withdrawn three notches, resulting in a power increase with a stable reactor period equal to the maximum period permitted in DP-0100, Reactor Startup to Criticality.

Heating power is estimated to be at "30" on Range 7.

Show all your assumptions and work for the following calculations.

a.

What is the doubling time?

(1.5) b.

How long will it take to reach heating power?

(1.5) 1.9 What are two problems associated with steam carry-under?

(2.0)

End of Category I

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RO Examination Page 3 of 12

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.Mrrch, 1985 Vercont Ycnkco 2.

PLANT DESIGN, INCLUDING SAFETY AND EMERGENCY SYSTEMS (25) 2.1 While you are perf orming a f ull flow test of the A Core Spray loop, a major leak develops in one loop of the Recirculation System, the reactor pressure quickly decreases to about 500 psig, and then very slowly continues to decrease.

Core Spray receives an immediate auto-initiation signal.

a.

Explain what will happen to Test. Bypass Valve (CS-26A), (2F the Minimum Flow Valve, and (3) the Discharge Valves (CS-11A and 12A).

(1.5) t b.

Will Core Spray inject?

If not, why?

If so, under what conditions?

(1.5) i 2.2 The Advanced Of f Gas System is designed to reduce the SJAE radioactive gaseous release rates to the atmosphere to as low as practical.

a.

Give two reasons for the use of preheaters in this system to raise the discharge temperature from the SJAEs to 300 F.

(1.0) b.

Temperature is monitored at the inlets and outlets of the recombiners as well as at three 43 locations within the beds.

Why is there such a great concern about this temperature profile?

(1.0) 2.3 Normal station service loads and the ESS buses are supplied from 4 KV Buses 1 and 2.

a.

During normal plant operation at full power, from where are these buses supplied?

(0.5) b.

If a loss of the normal supply power should occur, what would cause a block of the normal fast transfer to the alternate supply?

(0.5) c.

If fast transfer is blocked, (1) what kind of transfer will occur, (2) under what conditions will this transfer be permitted, (3) what loads will be shed prior to the transfer occurring, and (4) why are these loads shed?

(2.0)

Category Continued on Next Page I

f~%

RO Examination Page 4 of 12

Mrech, 1985

(])

()

Vermont Ycnkoo 2.4 Four relief valves are used by the Automatic Blowdown System.

a.

How is bellows. leakage detected and what control room indications will you have that leakage has occurred?

(2.0) b.

If bellows leakage occurs, how will it affect the automatic and/or manual operation of the relief valve?

( 1. 0 )~

2.5 There are three different pressures maintained by the Pressure Control Subsystem within the Control h

p Rod Drive Hydraulic System.

ggh "M F

a.

Give approximate values for each of these pressures.

(1.5) b.

What is the primary use of the water at each of the three operating pressures?

(1.5) 2.6 The Diesel Generators receive control power from the 125 VDC system.

a.

How will loss of the DC-1 bus affect the operability of the two station diesel

?y generators?

(1.0) b.

If DC-1 cannot be re-energized, what alternate sources of control power are available to be manually lined up?

(1.0) c.

If a loss of DC-1 occurred, what would be your first indication of whether or not DC-3 was lined up with DC-17 (0.5) 1 2.7 Starting more than one RHR Service Water pump simultaneously may result in a severe pressure transient.

Explain the cause of this potential pressure transient, and list the two precautions that should be followed to preclude the pressure transient.

(3.0)

Category Continued on Next Pagu O

RO Examination Page 5 of 12

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-March, 1995 vermont venu-2.8 The Primary Containment Isolation System is provided with Inadvertent Opening Protective Logic.

a.

What is the purpose of the Inadvertent Opening Protective Logic?

(0.5)

I b.

What actions are required before an isolation I

can be reset? (Any one isolation - specific valves are not required in the answer.)

(1.0) c.

Prior to and immediately following the resetting of a containment isolation signal, what are you cautioned to observe and why?

(1.0) 2.9 For each of the following statements regarding the High Pressure Coolant Injection System (HPCI),

2 (1) indicate whether the statement is TRUE or FALSE, and (2) give an explanation to prove your answer.

I a.

If a high steam tunnel temperature is sensed during HPCI operation, the system will immediately shutdown and isolate.

(1.0) b.

If the HPCI turbine trips due to high turbn ne exhaust pressure, it cannot restart until the

, O pressure f alls below the setpoint and the trip t

signal has been reset.

(1.0) c.

If the HPCI turbine trips due to an overspeed condition, it will restart phen.the speed coasts down to between 3000 and 4000 RPM.

(1.0) 2 End of Category 1

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RO Examination Page 6 of 12 l

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Vermont Ycnkoo March, 1985 i

3.

INSTRUMENTS AND CONTR (25)

/

3.1 List th.e_two.tr ntrol functions and the setpoints associated with the reactor f eedpump suction venturies.

(1.0) t 3.2 The Nuclear Instrumentation is designed to provide for channel overlap during startup and shutdown operations.

Various interlocks are provided that i

will lead to either rod blocks, reactor scrams, or i

that preclude operation of detector drives.

a.

For the SRMs, two interlocks are part of the system.

For the Retract Permissive Interlock, (1) what three signals are monitored, and (2) what four conditions must be violated simultaneously for a rod block to occur?

(3) Will the interlock stop detector movement? (1.6) b.

When the Mode switch is transferred to RUN,

)

what IRM ggC3g functions are bypassed?

(0.4) c.

What four trips will occur from the APRM system if the transition to RUN is not made 4

within the necessary power range?

(1.0) i MV 3.3 Consider the situation in which the reactor is operating with both recirc pump M/A stations in manual because the Master Controller has f ailed full scale high.

Reactor level is within the normal hand, and may be considered to stay there for purposes of answering this question.

Assume an operator inadvertently transfers the "A" Recirc Pump M/A~ Transfer Station to AUTO.

Explain how the Recire Pump speed will change (increase, decrease, or remain the'same), what the l

final stable speed will be, and the reason for this speed change if a.

The total feed flow is < 20 percent, and (1.0) b.

The total feed flow is > 20 percent.

(1.0) 3.4 List-five automatic actions that occur as a result of High Radiation sensed in the Main Steam Lines.

(2.5) r Category Continued on Next Page l

l LO

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RO Examination Page 7 of 12 I

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

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March, 1985 Vermont Ycnkee

3. 5' Instrument Air is required to position or maintain the position of a large number of valves in the plant.

For a loss of instrument air header pressure, how will each of the following be affected?

(2.0) a.

MSIVs b.

Feedwater Flow Control Valves I

c.

Safety Relief Valves Turbine Extraction Steam Valves (P e Q M M d.

e.

Turbine Extraction Steam Dump Valves 3.6 The Rod Block Monitor System prevents the power in fuel bundles surrounding a moving control rod from approaching MCPR limits.

a.

Explain the function of the gain change circuit.

(1.0) b.

Give two reasons for requiring the gain change circuit to act as it does.

(1.0) c.

List three ways the RBM may be bypassed.

(1.0)

JmA M Uike Mb 3.7 With the reactor operating at 50 percent p wer, explain how a malfunction in the Pressure Control Unit that cause's the dE tg igil in ibn ILgwet!

ggnitign on the Control witch will affect turbine and plant operation.

(2.5) 3.0 While the reactor is operating at 100 percent power in three element level control (See Figure 3.1.),

one of the steam flow inputs is lost.

Discuss the effect this will have on the reactor level.

Be sure to include in your discussion how the various inputs are derived, how error signals are developed and used, and what the final approximate indicated and actual parameter values (reactor power, steam flow, FW flow, and level) will be.

(3.0) l 3.9 List the four automatic protective actions that occur as reactor pressure increases above normal.

Include the setpoints for each action.

(2.0)

Category Continued on Next Page O

l RO Examination Page 8 of 12

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Vermont Ycnkee March, 1925 3.10 The Backup Scram Valves and Alternate Rod Insertion (ARI) Valves have been installed to provide redundant assurance that the reactor will scram when the Reactor Protection System detects unsafe operating conditions.

a.

State how many Backup Scr i Valves and ARI Valves there are, and briefly explain how these valves function to assure scram action.

Be sure to include an explanation of the purpose of the bypass check valves.

(1.5) b.

What three actions will initiate ARI?

(0.5) 3.11 The RBCCW System provides coolant to reactor auxiliary equipment under all station operating and accident conditions.

i a.

How is the temperature of the water in the RBCCW system regulated?

Name two changes in RBCCW-System operating conditions that would require temperature regulation changes to be made.

(1.0) b.

Under what conditions will the non-running RBCCW pump start if the running pump trips due

~

to an electrical fault?

(1.0)

End of Category 1

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RO Examination Page 9 of 12 1

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Vermont Ycnkee March, 1925 l

4.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL CONTROL (25) 4.1 During plant heatup after reaching critical, according to 0.P.

0101, i

a.

How and when is warmup of HPCI and RCIC gepleteopar g/,M (1.0) b.

Satisfactory MPR operation is to be demonstrated after pressure has reached about 900 peig.

However, steam flow cannot be read accurately at these rates.

How is it estimated?

(0.5) c.

After transferring to RUN, the IRM detectors are to be withdrawn.

How can you be sure that proper withdrawal is occurring?

(0.5) d.

A hold point at a reactor coolant temperature of 212 degrees is required until a certain plant parameter falls below a specified value.

What is this parameter, and how is it monitored?

(0.5) e.

The stack release rate must be kept at 50 percent of the Tech Spec limit as long as the mechanical vacuum pump is operating.

What is the reason for this restriction?

(0.5) 4.2 According to 0.P. 3105, Relief Valve Stuck Open Emergency Procedure, a.

What are four getemgtet changes you would be able to check to verify that a relief valve was open?

Do not list alarms or lights that initiate or change state.

(1.0) b.

What immediate actions are you required to take according to this procedure?

Be sure to-I include any time consi,derations.

Do not include any actions that may be required after a reactor scram may occur.

(2.0) r i

l Continued on Next Page P

RO Examination Page 10 of 12

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Vermont Ycnkoo M rch, 1905 O.

4.3 In accordance with OP-3100 Reactor Scram Emergency Procedure, there are five areas of immediate concern to Control Room personnel that the immediate actions address.

What are these five areas of concern (Ngt the immediate actions. ) ?

(2.5) 4 4.4 The Precautions section of OP-2110, Reactor Recirc System, recommends that operation of the recirc p

(F pumps be minimized at suction pressures below 300 I

a.

When are the recire pumps normally shut down?

(0.5)

I e

l M8 b.

'bmen the pumps are secured, the procedure calls, for the discharge valves 'o be shut, and for the discharge bypass valv=s to be opened.

Why i

are each of these actions called for?

(1.5) c.

If the recire pump is to be isolated, its Seal t

Purge System must be removed from service.

Why is this required?

(0.5) 4.5 If a less of coolant accident should occur, two j

emergency procedures are provided for your use.

These are OP-3116, Loss of Reactor Coolant, and 1

OP-3124, Loss of Reactor Coolant Outside Primary Containment.

j l

a.

What are the key events for entry into l

OP-31247 (1.0) i b.

What immediate actions are required if the high pressure emergency coolant injection systems are unable to maintain reactor water level, and

{

why are these actions necessary?

(2.0) l 1

4.6 It is approximately 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after a reactor l

shutdown.

The reactor coolant temperature has been j

controlled using the Shutdown Cooling System for j

the last three hours.

According to OP-2124, l

Residual Heat Removal System, i

a.

List the six symptoms and automatic actions that may occur on a Loss of Shutdown Cooling.

(2.0) b.

Why must'you be especially careful to not

~

misposition motor operated valves while j

determining the cause f or the loss of the Shutdown Cooling System?

(1.0)

Continued on Next Page P

RO Examination Page 11 of 12

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Vermont Ycnkoo March, 1985 4.7 It is a hot summer day about three in the afternoon when the Startup Transformer is lost.

According to OP-3119, Loss of Startup Transformers, a.

What plant operating getaggtet will change as a result of this loss?

(1.0) 6.

What are the immediate operator actions?

(1.5) 4.8 According to OP-3121, Fuel Element Failure Emergency Procedure, automatic actions occur to isolate various systems as a function of radiation levels that may exist for extended time periods.

a.

List the isolations that occur, any time delays that may be associated with each of them, and the corresponding instrument setpoints.

(2.0) b.

If a reactor scram does not occur, what immediate action is required to avert a scram? (1.0) 4.9 The Abnormal Operations section of OP-2112, Reactor Water Cleanup System, cautions in several places that if the system isolates, measures must be taken i

to preclude entry of resins 1nto the Reactor Vessel.

4O>

a.

Why is it possible for resins to enter the normal flow path to the vessel?

(1.0) b.

What should the valve lineup be when a pump is restarted, and for how long should this flow path be maintained?

(1.5) l I

End of Examination l

4 d

RO Examination Page 12 of 12

Fi JI Rev. O Feedwater Control System Simplified Diagram 8/B4 Rx. Level Rx. Level Feed Flow Feed Flow Steam Flow Steam Flow Steam Flow Steam Flow A

B A

B A

B C

D p_q Level A

B '

Select L.,_ j g

Switch Total Total W

Feed Steam Flow Flow l

i Turbine i

l f ~~ ]3

~

Steam Mode Flow Select 1

Flow Sicnal Switch (

j Comparator (Mismatch)

Three Flow Element Asp Mismatch i

Master Level E-Controller l

I Man / Auto Man / Auto Man / Auto Steam Leak Transfer Transfer Loading Detection Station Station Station System Figure N.1 dl

f a ma

. - s/t C7 Ele efficiency = (Network out)/(Energy in) 2 w = mg s = V t + 1/2 at o

2 E = mc KE = 1/2 my a = (Vf - V,)/t A = AN A=Aeg PE = mgh Vf = V, + a t w = e/t 1 = in2/t1/2 = 0.693/t1/2 t

eff = [(tin)(t)3 1/2 h

[(t 1/2I * (*bIl AE = 931 am I=Ien Q = mCpat Q=UA4T I'= I e'"*

n I = 1,10-*l Pwr = W an f

TVL = 1.3/u P = P,10 ""I*)

HVL = -0.693/u 5

P = P e /T t

o SUR = 26.06/T SCR = S/(l'- K,ff)

CR, = S/(1 - K,ff,)

S SUR = 26o/t* + (a - o)T CR (1 - K,ffj) = CR II - keff2) j 2

T=(t*/o)+[(8-9)/Io].

M = 1/(1 - K,ff) = CR;/CR,

T = 1/(a - 8)

M = (1 - K,77,)/(1 - K,ff;)

T = (8 - a)/(lo)

SOM = (1 - K,77)/Keff a = (X,ff-1)/K,ff = d,ff/K,ff t* = 10-seconds T = 0.1 seconds-I o = [(t*/(T K,ff)] + [r ff (1 + AT))

/

I d) = 1 d j

P = "(I+V)/(3.M '1010 ),..

I d '.$.=2 2' I d " ' -

y 22 2

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

Water Parameters Miscellaneous Convers' ions I gal. = 8.345 lbm.

I curie = 3.7.x 1010dps 1 ga;. = 3.78 liters I kg = 2.21 lbm.

I fte = 7.48 gal.

I hp = 2.54 x 103 8tu/hr Density = 62.4 lbe/ft3 1 m = 3.41 x.106 Btu /hr O

Heat of vaporization = 970 Stu/lem Density = 1 gm/c..r5 lin = 2.54 cm

• F = 9/5'C + 32 Heat of fusion = 144 8tu/lbm

• C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. Hg.

e

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March, 1985 Vermont Yankee

'(~

M hsTE e An)S C E 1 l 1.

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER, AND FLUID FLOW (25)

ANSWERS F

1.1 (1) Relatively higher notch worths would be found in core regions that had relatively low neutron i

flux levels during the previous reactor operation (or a trip after a high power run).

This would be in the peripheral areas of the core, and particularly near the top of the core where boiling had been occuring.

(Other reasonableeqtanationsalsoacceptable.)

(1.5)

(2). Xenon production is a f unction of local power or flux level.

After shutdown, it will peak highest in these previously high flux regions.

This will tend to d.epress the flux in these same regions during the subsequent startup, and cause the flux to be higher in previously low flux regions to maintain an equivalent power

' level.

Since control rod worth is a function of the square of the flux, these regions will-have relatively higher worth n'tches.

(1.5) o

REFERENCE:

0.P.

0100, pg. 3 Reactor Theory, Chapters 31 and 33.

D 1.2 (1) Spontaneous fission of uranium - 235 h 238 (2) Photoneutrons f rom the gamma,n reaction with deuterium in the water (3) Neutrons f rom the alpha,n reaction between the decay alphas from the transuranics and the 0-18 l

in the oxide fuel CO-18(alpha,n)No-213 t

REFERENCE:

Hot license Program Quiz, 7/30/84 1

I 4

)

s RO Examination Answers Page 1A of 17A l

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o March, 1985 Vermont Yankee 7 "',

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x i

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1.3 a.

As power increases, the amount of boiling in a 1

flow channel, increases, and causes an increase in the resistance to flow - that is, core dP increases.

b.

(1)

Two regioes are used for core orificing.

The central region consists of all but the outermost ring of perimeter fuel with the smaller or fewer orifices in the outer fuel positions.

(2)

The orifices installed in the fuel supports add a relatively large dP to the total dP across the core.

The dP added by boiling is then insignificant compared to the dP of the orifices, and so the dP is nearly the same (as is flow) to all fuel cells under any condition of power.

REFERENCE:

System Lesson Plan, Reactor Vessel and Internals, ppg. 13 - 14.

1.4 a.

(1)

Indicated level is higher.

(0.5)

(2)

Approximately 7 - 10 inches.

(0.5) b.

The indicated water level'is sensed outside the f

dryer skirt.

(0.75)

Steam flow at 100 percent power (0.25) causes a backpressure on the dryers.

(0.50) c.

It will increase.

REFERENCE:

System Lesson Plan. Reactor Vessel i

Process Instrumentation, Note from Pete Alter dated 1/28/84.

Category Continued on Next Page f

p 1

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it RO Examination Answers Page 2A of 17A 1

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M rch, 1985 Vcreant Ycnkee

/; ~

1.5 q

y%t a. new s[t> cwrde.

4L Head

+_

% g y m a y ymking point, t

GPM

. Figure 1.1.

Pump Characteristic Curve

,a.

Throttling the valve causes the friction losses in the system to increase, and the system curve to move to the left.

b.

Changing pump speed is more efficient since a Mk lower flow rate can be achieved without M

increasing the pump head.

Increasing the pump M

head would require more power.

~

M. Da "

jf41 j

REFERENCE:

Hot Liconse Progeam Qui T6f' VY Fluid Flow Student Handout, Fluid Flow, Chapter 4.

1.6 Its reactivity worth will be less.

(0.5)

• @[

Rod worth is proportional to the (local flux divided by the average flux) squared.

With all k.

,e rods inserted, the average flux is low.

Then when Y)((V[

the rod is withdrawn, it causes the local flux t o

be relatively high, resulting in a relatively high r

reactivity worth.

If the same rod is inserted from the fully 0

withdrawn position when all other rods are mostly withdrawn, the flux depression caused by the inserted rod will result in a small value of (local flux / avg flux)^2.

The worth of the. rod in this case is thus much less than in the above case.

REFERENCE:

VY Reactor Theory II Student Handout, ppg. 31 31-5.

Nh f

W&

&$ Imp /

klen1

[

1.

C R W bd dr. A1.

+g,o W I

%. b ^^ ^0 hm I..

i

- 8e4.

RO Examination Answers Page 3A of 17A

,I

n o

o Mrech, 1905 Vercent Yankee

,-~

\\

1.7 a.

Reactivity increases Power increases i

Void coefficient b.

Reactivity increases Power increases Moderator temperature coefficient

REFERENCE:

Hot license program quiz, 8/3/84,

.and VY Reactor Theory II Student Handout, Chapters 26 - 28.

1.8 a.

Max permissible stable period = 30 sec (0.5)

DT = period /1'.445 or 1.443

= 30/1.445 = 20.8 sec (1.0) b.

t/T P = Po e P/Po = 3000/5 = 600 In 600 = t/T t = 30

• In 600 = 192 sec = 3.2 min

REFERENCE:

VY Reactor Theory II Student Handout, Chapter 25 OP-0100 t$ sfm b>OI48 ~

1.9 (1) Steam bubbles in the recirc suction lines aaggqh greatly increase the chance for cavitation g,, 9 J _J

_s

~..

(2) Steam in the downcomer adds heat to the water y g lyh$

before it enters the fuel region, encouraging f, g,g earlier boiling, and contributing to adverse power ration.

REFERENCE:

VY System Description, Reactor

~

Vessel and Internals, pg. 16.

End of Category i

s if" N;

w RO Examination Answers Page 4A of 17A

l o

o March, 1985 Vermont Yankee

-~

2.

PLANT DESIGN. INCLUDING SAFETY AND EMERGENCY SYSTEMS (25)

- ANSWERS i

2.1.The Core Spray system is already operating, in the full flow test mode.

Therefore the test bypass valve is open and the min-flow valve is closed.

a.

(1) The test bypass valve CCS-26A3 will close with no delay.

(0.5)

(2) The min-flow valve will open when low flow Class than 300 gpm] is sensed in the flow element. When C300 gpm3 flow'is sensed by the flow element, the min-flow valve closes.

(0.5)

(3) When reactor pressure decreases to 350 psig, the discharge valves CS-12A'and CB-11A will be signaled OPEN. (Note that C3-11 is already throttled open.)

(0.51

?

b.

No.

(Context of answer will be considered.)

(0.5)

, When reactor pressure decreases to below the CoreSpraySystempressure,injectiontotpe begins.(,O p h d g g M M vessel (1.0) l

REFERENCE:

0.P.

2123, Core Spray, pg. 3 System Lesson Plan - Core Spray pg.

6.

2.2 'a.

(1)

Moisture has a quenching effect on the catalytic reaction and so must be eliminated.

i (2)

The reaction is enhanced r.t temperatures

[above 250 FJ that are elevated.

b.

The temperature profile through the bed is i

4 (1) a good indication of recombiner effectiveness, (2) as well as hydrogen concentration.

REFERENCE:

VY System Description, ADG, ppg.

6, 7, 16.

j I

l

!,e C./

,y

'f

,i, RO Examination Answers J

Page 5A of,17A I N f

March, 1985 Vermont Yankee

.-s 2.3 a.

From the Aux. Transformer Ethrough breakers 12 and 223.

b.

If the synchrocheck relay does not see synch, the fast transfer is blocked.

c.

(1) Residual voltage bus transfer.

(0.4)

(2) When voltage has decayed to about 1000 V (0.4) gh (3) RFP breakers are tripped prior to transfer 4pg being permitted (0.4)

,,y j

(4) To (1) insure that Breaker 13(23) will not F

• F trip subsequently from inrush current, (0.4)

SPM and (2) it speeds voltage decay.

(0.4) b,hb

,W

REFERENCE:

OP-2142, pg.

1.

h#

2.4 a.

Bellows leakage is detected by a pressure switch (0.5)-

that causes a control room alarm [ Blowdown Valve Bellows Leaking]

(0.5) and a yellow light above the control switch to illuminate (0.5) if pressure external to the bellows increases,

@ bout 10psig)

(0.5)

[

b.

Manual operation w not be affected.

(e,5 ). '

T.iSL However, automati eration will be precluded. (0.7) h)$w@rud:b

}

REFERENCE:

OP-2122, ppg. 1 and 4.

VY System Description, ADS, pg.

4.

2.5 a.

(1)

Nomirally at pump discharge pressure, at about 3750

• head (1630 psig)

Ealarm provided at 1410 psi 3 (2) 260 psi above reactor pressure (3) 20 to 40 psi above reactor pressure b.

(1)

Accumulator charge maintenance (2)

Drive water (3)

Cooling water

REFERENCE:

VY System Description, CRDH, ppg. 4-7 ll/~~

,L.

l l

RO Examination Answers Page 6A of 17A

i O

O March, 1985 Vermont Yankee 2.6 a.

Control power will be lost to the Diesel I

o Control [PNL-1B Ckt 123 for DG B.

DG A will

/W/h not be affected Csince it is supplied from Bus P

DC-2AS3, I

y i

b.

Diesel Control PNL-1B may also be supplied from

. p k'

either DC-2 or DC-3 [with concurrence of the M

Maintenance Department and the Shift r

[

Supervisor 3.

p gg

• hV c.

Since DC-3 supplies the Control Room d,j annunciators, the first indication of this 6,

g lineup would be the loss of the annunciators.

I

REFERENCE:

OP-2145, ppg. 415, and 11.

g(/)N S %

2.7 Starting two pumps simultaneously may decrease g'

station service water pump discharge pressure suf ficiently to start the standby pump.

If more than one pump is in standby, the pressure transient could be severe.

(1.5)

• (([o preclude the pressure transient,

-~ (1) pause at least ten seconds between each pump start, and (0.75)

(2) prior to starting an RHR SW pump, insure that only one station SW pump is in STANDBY.

(0.75) k!)5

REFERENCE:

OP-2124, pg.

6, Precaution # 3.

2.8 a.

The purpose is to prevent any isolation valve from automatically opening when the PCIS logic is reset.

b.

(1)

Place the control switches for the valves in the affected group Clisted in Appendix D of the procedure 3 to the CLOSED position.

(0.75)

(2)

Check Cat CRP 9-53 that the Containment Isolation reset Permissive Light for the affected group is lit.

(0.25) c.

Observe the radiation monito'rs Con CRP 9-2, 10, 11 and 50 closely]

(0.5) to ensure that no undesirable venting of radioactive effluents occurs.

(0.5)

REFERENCE:

OP-2115, ppg. 3 and 16.

/*

RO Examination Answers Page 7A of 17A

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March, 1995 Vermont Yankee s

2.9 a.

False (0.25)

There is a 30 min time delay associated with the isolation.

(0.75) b.

False (0.25)

The turbine trip signal does not seal in.

CThis will result in the turbine cycling on and off when the signal clears as it will because the turbine operation is the cause of the condition.3 (0.75) c.

True (0.25)

The overspeed trip device dumps oil pressure off the stop valve.

When the turbine has coasted down, oil pressure is restored and the stop valve reopens.

(0.75)

REFERENCE:

OP-2120, ppg. 2 - 3.

End of Category C+..i RO Examination Answers Page BA of 17A l

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March, 1985 Vermont Yankee t

3.

INSTRUMENTS AND CONTROLS (25) - ANSWERS 3.1 1.

Trip of RFP after 8 sec on low suction flow of 300,000 lbm/hr f or two seconds. -

CTime delay to allow for pump start.]

2.

Open/close 91gnal to minimum flow valve to recire flow at 490,000 lbm/hr (980 gpm).

REFERENCE:

VY Question Bank 8/24/84 Quiz and Feedwater SD ppg. 4 & 8.

t 3.2 a.

(1)

SRM Log Count Rate (0.2)

SRM Detector Position (0.2)

IRM Range Switch Position (0.2)

(2)

Any one of the f ollowing four conditions will preclude a rod block.

That is, all f our must be violated simultaneously f or the rod block to occur.

qu@%g h

SRM detector full in (0.2)

SRM log count rate above'100 cps (0.2)

V#

A All IRM range swktches on or above Range 3(0.2)

Mode switch in RUN with APRMs on scale (0.2)

(3)

Interlock does not stop detector movement.(0.2) b.

(1)

Hi Hi C120/125 of scale 3 (0.2)

(2)

INOP CH.

V.

Iow. Function Sw., Module (0.2) unplugged 3 c.

(1)

< 5 % in RUN gives Rod Block (2)

> 12 % not in RUN gives Rod Block (3)

> 15 % not in RUN gives scram (4)

< 5 % in RUN with corresponding IRM upscale or inop. will give a scram

REFERENCE:

VY System Description - NIs, ppg. 24-25, 36, 98-99.

FSAR Fig 7.2-8 provided during SRO exam review (for Part c)

'I(

RO Examination Answers Page 9A of 17A

O O

March, 1915 Vermont Yankee 7m

(

3.3 a.

The pump speed will stay at the minimum speed or go to the minimum speed because the low speed limiter is controlling speed as long as total feed flow is less than 20 percent.

b.

The pump speed will go to the high speed limit of the Master Controller Speed Demand Limiter because with feed flow > 20 percent and the pump discharge valve full open, the low speed limiter is bypassed.

REFERENCE:

VY System Description, Fig. 11.

3.4 (1) MSIVs close SteamLineDrainsshuth I TLIS-(2)

(3) Recirc Sample Valves shut (4) Main Condenser vacuum pump trips and isolation valve closes (5) Steam Packing Exhauster tr'ips and isolu zion valve closes

.(6) Reactor scrams Only five required.

0.5 each.

(0.25 for valve, 0.25 for action)

Items #4 and #5 are each considered as one item in total per VY answer key.

REFERENCE:

VY Question Bank, 8/27/04 Cuiz Shut-bd neY r s'haeo MMMM 3.5 a.

MSIVs b.

Feedwater Flow Control Val ves Fail as i s -will dvi eye-c.

Safety Relief Valves Loss of remote manual or ADS actuation d.

Turbine Extraction Steam Valves Shut e.

Turbine Dump Valves Open

REFERENCE:

OP-2190, pg. 8

.J j

RO Examination Answers Page 10A of 17A l

1

i March, 1985 Vermont Yankee S.

i 3.6 a.

The gain change circuit adjusts the gain of the averaging amplifier so that the RBM output will be at least equal to or greater than the referenced APRM.

i b.

(1)

Local power may be significantly lower than the core average because of the rod having a high reactivity wortj[

('2)

Several of the highest reading LPRMs may be bypassed, resulting in the local average reading being lower than it should be.

c.

(1)

< 30 percent power on reference APRM (2)

Edge rod selected (3)

Manual Joy stick

REFERENCE:

VY Question Bank, 10/1/84 Quiz VY System Description, NIs.

ppg. 53, 74 -76, 88 3.7 (1) Whan the MPR f ails to lower, this increases the MPR stroke, calling for the control valves to open, and reducing the MPR pressure setpoint to 19tfRC RCRERWCE.

9(

(2) When the MPR setting falls below the EPR setting, it will take control of the pressure torque bar and increase the pressure signal to the control valve relay.

(3) The Control Valves will open, increasing turbine load, and decreasing steam line pressure.

(4) The decreased pressure will increase core voids, and reactor power will decrease, causing a further pressure decrease.

p p ' hTec.k (5) At 950jpsig in RUN, a Group I isolation will p g occur, closing the MSIVs, and causing a reactor scram.

• M*%

0.5 for each of the above items or appropriate variation of each l

REFERENCE:

VY Question Bank, 8/31/84 Quiz

(

VY System Description, Turbine l

Control System, ppg. 7 - 9.

l

\\-L' 1

RO Examination Answers Page 11A of 17A l

l

- - ~.. _

oo M m

O O

~

Mrrch, 1985 Vermont Yankee

=

-(g 3.0 (1) (a) The steam flow signals from the four main steam lines are summed, and (b) the FW flows from two parallel FW lines are summed.

(c) d#

With loss of one of the four steam signals, a

~

25 percent mismatch will exist.

(0.6)

• )

,e

~

3 (2) (a)

This mismatch will produce an error signal Q

1. r which (b) will reduce feed flow to correct the jr error signal.

(0.4)

I M

(3) (a) Since the actual steam flow has not been U

v2h d changed, (b) the level will decrease, and (c) a

)

second error signal will be generated.

(d) Since the system is level dominant, (e) a 7

signal will be sent to open the FCVs and increase feed flow to maintain level constant. (1.0)

(4) The-final indications will ber (1.0)

Reactor power: 100 %

Steam flow: 100 % actual, 75 % indicated FW flow 100 %

Levels Constant, but lower than prior to the fault CA 100 % loss of steam. signal is worth about a 7 inch level change, so a one-quarter loss will result in about a 2 inch change.3

REFERENCE:

System L'esson Plan - FW Control Plant discussion, 1/22/05 3.9 Automatic Action Setpoint Reactor Scram 1055 psig RPT/ARI 115@ psig p

Relief Valves Open 1000, 1090, 1100 psig Safety Valves Open 1240 psig CTS 3 0.25 per auto action 0.25 per setpoint set

REFERENCE:

VY Question Bank, 10/11/84 Quiz t

l l

RO Examination Answers

.Page 12A of 17A i

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March, 1985 Vermont Yankee b,

k_

3.10 a.

Both pairs of valves function in a sirmilar way.(0.5)

Upon receipt of an ARI or RPS scram signal, the respective valves will gnetglag, causing them (0.25) to block the incoming instrument air and vent the scram valve air header.

(0.25)

The bypass check valves function to assure that j

the scram air header will vent properly if any J

one of the valves in either pair should fail.

(0.5) b.

(1)

High reactor pressure - C1150 psig3 (0.125)

(2)

Low reactor water level at -82.5 " with a (0.125) 10 see time delay (0.125)

(3)

Manual (0.125)

REFERENCE:

VY System Description, CRD Hydraulic System, ppg. 16 - 17, Figure of Scram Discharge Volume VY Lesson Plan, RPS, ppg. 8 - 9.

1 3.11 a.

The RBCCW System temperature is regulated by throttling the service water outlet valve or the outlet bypass valve.

(0.5)

Conditions which could require such throttling s

include 43 (1)

A change in service water temperature (2>

A change in system loads (3)

Tube f ouling 4

0.25 each for any two b.

The standby pump will start if it is in auto (ON) and if the the discharge header pressure drops below about 74 psig.

(08)

REFERENCE:

VY System Description, RBCCW, pg. 6 DP-2182, pg. 5 End of Category i

RO Examination Answers Page 13A of 17A

- - - - - -, - ~, -

~'-

o o

March, 1985 Vermont Yankee m

4.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL CONTROL (25) - ANSWERS

[friortothereactorpressurereaching100 4.1 a.

psig,]keylockswitchesConCRP9-39and9-303 i

are positioned to HPCI and RCIC WARMUP, respectively, to open the steam inlet valves CHPCI/RCIC 15 and 163.

These low pressure bypass switches are to be restored to NORMAL prior to the reactor pressure reaching 150 psig.

b.

It is estimated by observing Bypass Valve No. 1 to be positioned lhetween 7 and 15 percent open]

c.

By selecting each recorder to IRM and observing that the flux level is decreasing.

d.

Until coolant oxygen Cis.below 200 ppm 3 (0.25) as determined by a. Chemistry sample.

(0.25) e.

The vacuum pump releases directly to the stack and not through AOG.

REFERENCE:

0.P.

0101,Section I.A.

Normal Operation a.

Steps 5 and'11.

i b.

Step 19.

a.

1 c.

Section I.B.1.d.

d.

Step 6.

e.

Cauti.on in Step 9 and plant discussi.on, 1/23/85 4.2 a.

(1) 12 - 15 % drop in generator output (2) 12 - 15 % drop in steam flow (3) Suppression pool temperature increasing (4) Suppression pool level increasing (5) Feedflow adjusts to maintain level (6) Distinct audible level change in Reactor Building (7) Change.in D/W-Torus dP (goes low)

(8) Other reasonable parameters -

(0.25) for each of any four i

b.

(1) Determine which relief valve and cycle the control switch from AUTO to CPEN to AUTO.-

If successful, no f urther action required.

(0.5)

(2)

If not successful, (a) transfer station loads to S/U i'

Transformer

• (0. 5)

(b) and manually scram the reactor (0.5)

(c) within 2 min of the relief valve lifting.

(0.5)

REFERENCE:

0.P.

3105, ppg. 1 - 2.

l RO Examination Answers Page 14A of 17A l

-.. ~...

~

O O

March, 1985 Vermont Yankee m

- k_

4.3 (1) All rods fully inserted.

(2) Continuity of station power Cincluding tripping the main turbine].

(3) Stabilize reactor water level between 137 - 167 inches.

(4) Establish reactor pressure control.

(5) Limit radioactive gas release to within appropriate TS and state regs.

REFERENCE:

OP-3100, pg.

1.

4.4 a.

When Shutdown Cooling has been established b.

(1)

Close the discharge valve to insure the pump coasts to a complete stop (2)

Open the discharge bypass valves to insure that the pumps cooldown at the same rate as the balance of the primary system.

c.

Remove seal purge to prevent lifting the seal purge FCV station relief valve.

,sj

REFERENCE:

OP-2110,Section I.

B.

4, 5, & 6.

4.5 a.

(1)

Isolation of a valve group of PCIS and i

t (2)

No concurrent high drywell pressure b.

It will be necessary to manually initiate ADS, LPCI, and Core Spray (1.5)

[

because high drywell pressure does not exist.

(0.5)

REFERENCE:

OP-3124, pg.

1.

1 RO Examination Answers Page 15A of 17A

r?. -..

1 o

o I

[

March, 1985 Vermont Yankee l

A k[.

4.6 a.

(1)

RHR valves [17, 18, 32, and 333 indicate shut or de-energized.

(2)

Reactor Pressure Low Permissive annunciator Con CRP 9-33 clears.

-(3)

Drywell Pressure High annunciator alarms.

(4)

Reactor Level Low annunciator alarms.

(5)

RHR and/or RHR SW pumps being used for S/D Cooling stop (annunciator alarms).

(6)

Primary coolant temperature / pressure increase.

b.

Mispositioning valves will cause rapad changes in inventory by allowing water to drain to the torus or radwaste.

REFERENCE:

OP-2124,Section II.B., pg. 25.

4.7 a.

Degradation of condenser vacuum.

b.

(1)

Shift to open cycle [per OP-21803.

(2)-

Reduce power to 20 -25 percent

[per OP-01023.

(3)

Notify REMVEC and appropriate plant management.

REFERENCE:

OP-3119 4.8 a.

(1)

FCV-11 closes (0.3) if timer times out (0.1) a.

in 30 min udth AOG operating (0.1) 6.

in 2 min i. th AOG bypassed (0.1) on high rad C. Ci % bar gamma 3 (0.1 )

(2)

FCV-516A(E) cic Os (0.3) if timer tames out (0.1) a.

in 15 min (0.1) at 0.3 Ci/sec (0.1) b.

in 1 min (0.1) at 1.5 Ci/sec (0.1)

(3)

MSIVs close (0.3) at 3 x normal (0.1) with no time delay (0.1) b.

Reduce reactor power Efirst by reducing recire flow, and then by rod insertion] to maintain the radiation levels below 3 x normal.

i

REFERENCE:

OP-3121, ppg. 1 - 2.

l RO Examination Answers

.Page 16A of 17A

.r O

O 4

3 March, 1985 Vermont Yankee "N,

4.9 a.

With no flow in the system, the elevation of j

the F/Ds is such that a siphon effect can suck the resins backwards out of the F/Ds, down to the bypass line, where resin can enter the vessel on the next pump start.

b.

The pump sh d be started with th bypass line open and th rains to radwaste open.QpThe return to the vessel valve should be only slightly open so that the pump will start.

(1.0)

~

(3)Continuetooperateinthismodeuntil Chemistry verifies Cby millipore analysis] that there are no resins in the system.

(0.5)

REFERENCE:

OP-2112,Section II.A, B.,

and G.,

p?g. 14 -16.

End of Examination s'

RO Examination Answere p.g.174 og g74 l

__ -, _._-