Text

Exam Rept 50-352/85-41 on 851111.Exam Results:One Senior Reactor Candidate Failed Written Portion & One Failed Simulator Exam.All Instructor Certification Candidates Passed All Portions of Exam
	ML20141M995
Person / Time
Site:	Limerick
Issue date:	02/14/1986
From:	Keller R, Kister H, Lange D; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20141M992	List: IR 05000352/1985041; ML20141M991;
References
	50-352-85-41, NUDOCS 8603030320
	Download: ML20141M995 (42)
	v • d • e

--

.

U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT

..

EXAMINATION REPORT NO.:

50-352/85-41 (0L)

'

FACILITY DOCKET NO.:

50-352 FACILITY LICENSE NO.:

NPF-39 LICENSEE:

Philadelphia Electric Company 2301 Market Street i

Philadelphia, Pennsylvania 19101 FACILITY:

Lfmerick Generating Station EXAMINATION DATES:

November 11, 1985 CHIEF EXAMINER:

,

2//y/f4

,

David Lansej Reactor EQineer (Examiner)

Date

'

REVIEWED BY:

2.//af/f4 R. M. Keller, Chief, Projects Section 1C

~ Date 2-b APPROVED BY:

_

o H / B. Ki s't'e Chief, Projects Branch No. 1

/ Dite'

SUMMARY: Sixteen written, Sixteen oral and Sixteen simulator examinations were administered to Thirteen Senior Reactor Operator Candidates and three Instruc-tor Certification candidates. One Senior Reactor Candidate failed the written portion of the examination and one Senior Reactor Operator Candidate failed the Simulator Examination.

All Instructor Certification Candidates passed all

'

portions of the examinatio is.

Although the majority of the candidates were generally familiar with in plant components, an overall weakness was noted by all examiners in the experience and confidence level of the SRO candidates ability to direct core alterations and refuel activities. This seems to be a result of a lack of experience com-bined with an inadequately structured on-the-job training program. This con-cern was brought to the attention of the operations manager and a recommenda-tion was made to the training department to incorporate this training into the requalification session immediately preceding the first refuel outage.

B603030320 860220 PDR ADOCK 05000352 V

PDR L

.

REPORT DETAILS TYPE OF EXAMS:

Initial EXAM RESULTS:

R0

SRO I

Inst. Cert l l

Pass / Fail l Pass / Fail i

Pass / Fail I

I i

I I

I

1 I

I Written Exam i

N/A i

12/1

3/0 l

I I

1 1 Oral Exam i

N/A l

13/0

3/0 l

l

1 1.

I I

I Simulator Exam l

N/A i

12/1 l

3/0 l

l l

1 I

I I

I I Overall I

N/A i

11/2

3/0 l

l l

l

1 1.

Chief Examiner at Site:

D. Lange, USNRC 2.

Other Examiners:

F. Crescenzo, USNRC A. Howe, USNRC B. Hajek, Consultant 3.

Summary of generic strengths or deficiencies noted:

Candidates exhibited overall strength locating in plant equipment and using local operating procedures.

,

Weaknesses noted were in the use of Technical Specifications, and overall knowledge of the refuel floor equipment and procedures.

4.

Personnel Present at Exit Interview:

NRC Personnel D. Lange, Examiner F. Crescenzo, Examiner G. Kelly, Senior Resident Inspector

-

- -

-

- -

-

- -

-

- -

.

Facility Personnel E. Firth, Training Coordinator, PECO R. Helt, Senior Instructor, Nuclear Training Section, PECO R. Rhode, Manager, Limerick Training Services, GPC 5.

Summary of NRC Conaents Made at Exit Interview:

The Chief Examiner emphasized the generic strengths and weaknesses noted from the oral exams. Also noted during the simulator portion of exams was a lack of communication between team members.

Simulator performance and training staff assistance was good with the exception of examinations given on December 2,1985 when numerous simu-lator problems were encountered.

6.

Change Made to Written Exam During Examination Review:

See Attachments Attachments:

1.

Written Examination and Answer Key (SRO)

2.

Facility Comments on Written Examinations Made after Exam Review 3.

NRC Resolution of Facility Comments on Written Examinations

,

i

- - - - -

TT/16 l1/71CI?Y l

\\

'

.

)

U.S. NUCLEAR REGULATORY C0pmISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION

'

Facility:

Limerick Generatini st= tion Reactor Type:

Boiling Water Reactor Date Administered:

novemoer u,19o3 Examiner: Brian K. Hajek Candidate:

MAstr /t INSTRUCTIONS TO CANDIDATE:

Use separate paper for the answers. Write answers on one side only.

Staple i

question sheet on top of the answer sheets.

Points for each question are indi-

'

cated in parentheses efter the question.

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

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

}

% of Category

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

25 5.

Theory of Nuclear

,

Power Plant Operation,

.

Fluids, and Thermo-

]

dynamics

'

25

~ 6.

Plant Systems Design, control, and Instrumentation

!

l

25 7.

Procedures - Normal, Abnormal, Emergency, and Radiological Control

25 8.

Administrative Pro-cedures, Conditions, and Limitations 100

.'

Totals

~

Final Grade

.

,

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

Candidate's Signature

.

-

.

,,

-c

-- -

. - - - -, -.

,,m

- - -. -, - - -

y

-

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

-

- - -.

-ep

. - -

,~.v-.

,,

.

,

November, 1985 Limerick Generating Station

,

5.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND u

THERMODYNAMICS (25)

5.1 The following two phrases describe two types of boiling that may occur in the reactor core:

'

(1) Steam bubbles form on the fuel element cladding and break away, moving into the bulk of the fluid, leaving only a small portion of the surface covered at any one time.

(2) Steam completely blankets the cladding ourface, separating it from the bulk of the fluid.

a.

What kind of boiling is described by each of

_

these statements?

(1.0)

b.

Which of these two two types of boiling provides for better heat transfer from the cladding to the fluid?

Give at least two reasons for this better heat transfer.

(1.5)

m 5.2 The source range monitors may be used during accident conditions to provide a backup indication

-)

,

of reactor water level.

If water level would fall to less than one-half core height, what indications would you expect on the source range monitors to indicate this level, and why would you have this indication?

(1.5)

5.3 Following a LOCA and initiation of ECCS, it is possible for the core to experience counter current flow.

Briefly describe (1) what occurs during counter current flow, and (2) what adverse conditions may result.

(2.0)

5.4 Corrosion products in the form of " crud" may become suspended in the coolant and cause operating problems.

a.

List two undesirable characteristics of crud.

(1.0)

b.

What are three examples by which a crud burst might occur?

(2.0)

...

CATEGORY CONTINUED ON NEXT PAGE SRO Examination Page 1 of 14

._

_

.

_ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _

_

.

-

.

November, 1985 Limerick Generating Station

.

5.5 Every jet pump has an Instrument line tapped into the upper end of the dif fuser, and f our (5,10,16, y

and 20) have instrument taps at the bottom end of their diffusers.

These are the calibrated jet pumps.

If one of the upper pressure tap lines should become plugged, resulting in a lower pressure being transmitted from this tap, how will the indicated flow from this jet pump -111-change?

Explain why.

(1.5)

5.6 The reactor has just scrammed from an extended full power operation shortly before m'id cycle.

Because of a Tech Spec LCO, it has been necessary to-be in cold shutdown as soon as possible after the scram.

This was accomplished about ten hours after the scram.

The react r engineer has estimated the reudn. h e wnd.-

^

shut [own -== 7 ine about two percent.

How will

'

-

the shutdown change over the 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months , what will cause the. change (s), and w a fconcerns relative to criticality will need to be considered?

(2.5)

5.7 Emergency injection systems should not be shut off under most conditions when they are required to maintain the core covered.

However, T-117, LQ, Level / Power Control, provides for terminating and

.

preventing all injection into the -RPV, thus permitting level to decrease t>rtcw#the top of the core.

Under what conditions would-this action be desired, and why would it perform the intended function?

(2.0)

5.8 a.

Define the critical power ratio and explain why it must be kept greater than 1.0.

(1.5)

b.

For a constant reactor and bundle power, how and why will the critical power ratio change as (1)

Inlet subcooling increases, and (0.75)

(2)

Core flow increases.

(0.75)

5.9 Explain why reactor pressure instrumentation cannot be used to determine core temperatures following vessel depressurization after a LOCA.

(1.0)

.

CATEGORY CONTINUED ON NEXT PAGE

,

a i

SRO Examination Page 2 of 14

-

_ _

.

__

-

.

. - -.

.

- -

-~_.

.

.=

.

..

. r,,

November, 1985 Limerick Generating Station-

'

l

-}

5.10 Exp3 ain' how control rod worth changes with a change

-

in the local neutron flux.

(1.0)

,

a 5.11 Explain why the reactiv'ity.added due to the moderator temperature coefficient causes a major

'

change in total core reactivity for a power change i

from 0.1 percent to 2. percent during a cold

startup, but virtually no change in total core t

reactivity for a power change from 50 percent to

'

90 percent.

(2.0)

.

5.12 Three minutes following a reactor scram, the reactor power indication is 75 on Range 4 of the IRMs.

ANu$sE

]

a.

What will the power /be one minute later?

Show

all your calculations.

(1.5)

'

j b.

Explain the reason for the rate of power decrease you used in Part a.

(1.5)

END OF CATEGORY

"

!

-

i j

I

,

i

i

!

I i

.

?

i i

SRO Examination Page 3 of 14

.

!

,

,-n--,

--a

,,--,--,,,,-wr-

,.o,,v,-,-m,----

r

--rg,---

,----r-

.--a-e--.-.n-,,

e

,-,. -,-

g-e,

+,,-nn-n-,.,

e

,mga e,-r--

-,,

g

--m,--

,e

.

,

November, 1985 Limerick Generating Station

.

6.

PLANT SYSTEMS: DESIGN, CONTROL, AND INSTRUMENTATION (25)

6.1 The Rod Worth Minimizer detects insert and withdraw errors and may apply Insert and Withdraw Blocks, a.

What two actions will cause an Insert Error?

(1.0)

b.

For the cases of.one or more insert errors, what rod blocks will be applied by the RWM?

Describe each case, i.e., one, two, and more than two insert' errors.

(2.0)

6.2 The Standby Liquid Control System is used to inject sodium pentaborate into the reactor system in the event of a failure of the CRD System.

l a.

Where is the standby liquid injected into the vessel?

(0.5)

,

i

.

~

l b.

When SLC is initiated, another system is automatically isolated.

What is this system, and why is it required to be isolated?

(1.0)

w c.

Two heaters are provided to heat the solution in the storage tank.

(1) Why must the solution I

be heated, and (2) what is the specific purpose for each of the two heaters.

(1.5)

6.3 Briefly describe the operation of the system used to monitor for Core Spray Sparger Break Detection.

Be sure to include where a leak can be detected by

,.

this system, and whether the alarm occurs at high low differential pressure (Setpoint not or required.).

(2.0)

6.4 AC Power Bus 1AY160 provides an uninterruptible power source to several plant systems.

What are the three possible sources of power to a.

this bus, which source is preferred, and which are alternates?

(1.5)

b.

If an overvoltage breaker protecting this bus should inadvertently open, name four systems that will lose power from this bus.

(1.0)

CATEGORY CONTINUED ON NEXT PAGE i

.

SRO Examination

~

Page 4 of 14

_

._.

. _ _ _

_.-

.

\\

.

,

t November, 1985 Limerick Generating Station

.

6.5 The recirculation pump speed control system contains two speed limiters in the Manual / Automatic

s Transfer Stations for each recirculation pump.

For each of these limiters, list the conditions that cause the limitar to be in effect, and the purpose of having the limiter enforcing for each condition. (3.0)

6.6 Air pressure is normally applied to the Scram Inlet and Exhaust Valves to hold them closed to prevent a reactor scram from occurring.'

If a trip signal is received in either RPS, System A or B,

" signals" are sent to the Scram Pilot Valves and to the Backup Scram Valves to cause air to be removed from the Scram Inl et Valves.

a.

Are the Scram Pilot Valves and Backup Scram Valves normally energjzed or normally deenergized?

(1.0)

.

A d r,:,:. lop What -RPS " signals'g _cause each Scram Pilot valve b.

and each Backup Scram Valve to change state?(1.0)

,

c.

If one of tthe Backup Scram Valves should fail

'

to operate on receipt of the appropriate.

signal, what assures that the other valve can

)

perform the necessary function?

(0.5).

-

' t.

i 6.7 The Rod Block Monitor receives inputs f rom LPRMs, APRMs, and the RMCS.

For selection of a centrally located control rod, andifor RBM Channel A, discuss

,

which signals are used, and how the input signals are selected for passage to the trip units to assure conservatism.

Be sure to discuss a.

The RBM Averaging and Gain Change Unit, and (1.5)

b.

The Recirc Flow Trip Reference Unit.

(0.5)

6.8 Describe the automatic actions that occur for a Group IB isolation, and the conditions that will cau.se those actions.

(3.0)

CATEGORY CONTINUED ON NEXT PAGE i-

,

y.

T

.)

.

,

.

~

SRO Examination

.

- ' l Page 5 of 14 35'

i s

,

.....

\\.,

-,

, _..., _

,

.

-

.

,

November, 1985 Limerick Generating Station

)

6.9 The SRMs will give rod blocks during a reactor startup.

a.

What four conditions will give rod blocks?

(1.0)

b.

These rod blocks will be bypassed if the

" associated" IRMs meet certain range conditions.

What is meant by the term

" associated", and what are the range conditions that will result in bypassing all the rod blocks listed in your answer to Part a.

(1.0)

6.10 The Automatic Depressurization System will initiate on receipt of five it.put signals.

Two of these are level signals.

,-

a.

If the water level is fluctuating, and falls ('/3 ;

below the low level initiation point, then f. ',

rises to above this_ point, falls again, and h'

g rises again, will ADS initiate?

If it does initiate, will the valves close if level rises again?

Explain your answer.

(1.0)

%

b.

What is the purpose of the low level confirmatory set point?

(1.0)

END OF CATEGORY

~

SRO Examination Page 6 of 14

.

-

.

...

l

.

'

'-

.

i November, 1985-Limerick Generating Station j

,

Y

7.

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

,;

,

7.1 Using the copies of SP/L-1 and SP/T-1 from T-102,

,

Containment Control included as Figure 7.1, a.

Determine the minimum safe Suppression Pool water level if the RPV is at 500 psig and the Suppression Pool temperature is 150 degrees.F.

i Show all work.

(2.0)

'

!

b.

If the minimum safe level cannot be maintained,

i instructions are to initiate an emergency blowdown.

What is'the basis (concern) for-this requirement?

(1.0)

7.2 According to ON-100, Failure of a Jet Pump, Give three sy'aptoms that you could observe in a.

the main control room for a failed jet pump.

(1.5)

b.

If, from the above mentioned symptoms, you suspected a jet pump failure, what additional

"

,

symptom could you check for, and where would you check?

(1.0)

7.3 GP-3, Normal Plant Shutdown, provides instructions for shutdown to three different plant conditions.

For a shutdown to Hot Standby, a.

How and at what pressure is pressure to be maintained before the reactor vessel is isolated?

(1.0)

i b.

Before closing the MSIVs, what two systems

'

should be used to provide all 'the required makeup to maintain a stable water level?

(1.0)

.

c.

After the vessel is isolated, what systems.are i

to be used to maintain the Hot Standby

!

(1) power, (2) pressure, and (3) level conditions?

(1.0)

7.4 According to OT-102, Reactor High Pressure, one of the immediate~ actions on receipt of the High Pressure alarm is to run back recire flow.

What malfunction will this assist in combating, and why

might it work?

(2.5)

)

CATEGORY CONTINUED ON NEXT PAGE

.g

~

SRO Examination Pagel7 of 14

,

_

-,

.--

-

__ _

, _ _.. _

. -

.....

.

...

-.

.-.

.

N;vsnbar, 1985 Li:";3 rick G'in2roting Stction

)

CURVE SP/L-1

,

HEAT CAPACITY LEVEL LIMIT CURVE ~

-

, 30

[ 28

26 d 24

" '4 2

\\

SAF1

3 d 18 N %

o N

I w

a.16

, s

'

UNSAFE N

w i

-

's

" 10

,

20 25 30 35 40 45 50 s

5 to 15

'

AT HC ( F)

3c.

%

)

's CURVE SP/T-1 HEAT CAPACITY TEMPERATURE LIMIT CURVE o

E (

, k.

210

\\,

UNSAFE

.

200 N

-

g 190

%

w 180-m

%

170 m

%

o 160 gg.

E

'150

140 a

$130

" 120

'

,

i i

200 400 600 800 1000 REACTOR PRESSURE (PSIS)

~

-

.

Figure 7.1 Suppression Pool Heat Capacity Curves

}

CATEGORY CONTINUED ON NEXT PAGE

J SRO Examination Page 8 of 14

.

?

-

.

l l

.

November, 1985 Limerick Generating Station 7.5 The S97.0 procedure series deals with fuel movements and operation of fuel handling equipment.

Several of these have common precautions.

a.

The word " verify" is used in the precautions and in the steps of the procedures.

What is required when this word is used, for example, in the statement " Verify that no load exists on the main hoist before opening the grapple."?

(1.5)

b.

What two conditions would require the declaration of a Site Emergency?

(1.0)

7.6 In the event of a recirculation pump trip, the only immediate acticn specified in OT-112 if a scram does not occur is to fully drive in deep rods as required to prevent a scram, a.

How are deep rods defined?-

'

(0.5)

b.

Explain why it is necessary to drive in the deep rods.

(1.5)

,

c.

Prior to restarting the recirculation pump,

}

additional deep rods must be inserted.

Why is this necessary?

(1.0)

7.7 The Shift Superintendent or Shif t Supervisor is responsible for issuing high radiation area master keys for emergency use.

However, before these keys may be issued, certain prerequisites must be met.

What are these prerequisites (Three considerations)?

Be sure to discuss the requirements for areas in which the dose rate is greater than 100 mr/hr, and areas where dose rates may exceed 1000 mr/hr.

(2.0)

CATEGORY CONTINUED ON NEXT PAGE

-

.

SRO Examination Page 9 of 14

-

.

-

- - - -

-

- -

-

- -

-

- -

- - - -

- - -

- -

-

- -

-

- -

-

- -

-

- -

-

.

November, 1985 Limerick Generating Station i

7.8 According to ON-107, Loss of CRD Regulating Function, a rapid reactor power reduction must be initiated and the reactor must be in Hot Shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months if CRD flow to the charging water header cannot be established.

a.

What thermal limit for the fuel is of concern in this instruction?

(0.5)

b.

What is the Tech Spec limit and the grounds for this limit'that determines the applicability of.

the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months limit?

(1.5)

7.9 Give f our of the six situations f or which an RWP is required.

(2.0)

7.10 On a Loss of Condenser Vacuum while the reactor is operating at 75 percent power, according to OT-116, a.

What is the single required immediate action, and why will this work to alleviate the condition?

(1.5)

i b.

Give two reasons for not using the mechanical vacuum pump to help maintain condenser vacuum. (1,0)

END OF CATEGORY l

'l

..

SRO Examination Page 10 of 14

.

-

,

-

--

-- -

-. -

-. _.

..

.

. - -.

..

.

-

.

,

'

November, 1985 Limerick Generating Station

'~

8.

ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS (25)

NOTE:

SEVERAL QUESTIONS IN THIS SECTION REQUIRE ANALYSIS OF TECHNICAL SPECIFICATION REQUIREMENTS.

FOR THESE QUESTIONS, USE THE ATTACHED EXCERPTS FROM THE LIMERICK GENERATING STATION TECHNICAL SPECIFICATIONS.

ASSUME YOU ARE THE RESPONSIBLE SHIFT SUPERVISION REPRESENTATIVE WHO MUST MAKE THE COMPLIANCE AND ACTION STATEMENT DECISIONS.

IT IS IMPERATIVE THAT YOU REFERENCE ALL APPLICABLE SECTIONS USED AND PROVIDE YOUR

DECISION PATH FOR GRADING!

8.1 The reactor is operating at 90 percent power on a coastdown near the end of life.

A surveillance has just shown that all four Turbine Stop Valve Closure Setpoints have failed their setpoint adjust tests, and cannot be set down to below 8 percent closed.

I & C reports the f ailure is due to a component found in each trip unit, and that replacement will require about 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months .

It will be at least eight hours before the first trip unit is returned to

,

service.

a.

What is the value for MCPR under these conditions?

Assume that tau is 0.7.

List any other necessary assumptions.

(1.0)

b.

Why is it necessary to increase the value for MCPR under these conditions?

(1.0)

8.2 What are the two Limiting Safety System Settings for the APRMs according to the Technical Specifications, and under what operating conditions are each of them enforced?

(2.0)

8.3 The reactor is operating at 100 percent power in early August when the supply f an in one of the control room emergency fresh air supply systems develops an electrical fault in the motor windings and fails.

Maintenance estimates it will take I

four days to return the system to operation.

One day later, the Control Room temperature rises slowly to 90 degrees and stabilizes.

The operating system is unable to bring the temperature down, and Maintenance attributes the cause to be a faulty cooling unit.

They estimate an 18 hour2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months repair period.

What action statements apply, and what g

action must you take relative to the operation of

./

the reactor?

(3.0)

CATEGORY CONTINUED ON NEXT PAGE SRO Examination Page 11 of 14

.. _

.

.-.

,

,.

-.-

.~

.

+

.

l.

'

November, 1985 Limerick Generating Station 8.4 A truck has arrived on site to pick up a load of

~

radioactive waste drums at 8:00 Saturday night.

At 8:30, while loading the-drums into the truck, one

>

i drum is dropped and cracks.

The contamination is.

spread onto the surface of several-other drums and

!

also onto the bed of the truck.

The Health Physics

.

Technician measures the contamination level to be

^

30,000 dpm per 100 square centimeters, and ' calls-

'

you, the Shift Superintendent, at 9:15.

According.

f to A-31, Procedure for Notification of.the NRC,

i a.

What notification requirement (one hour, four j

hour, 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) must be met?

(0,5)

b.

What plant personnel must be notified

,

!

immediately?

(1.0)

l c.

At what time did the clock start and how long i

do you have to complete notification of the

'

NRC?

(0.5)

I

8.5 The Shift Operations procedure, A-7, states in e

I Section 5.4 that "When fuel is in the Unit 1

)

Reactor, the Unit 1 Reactor Operator shall be "At l

)

the Controls" by remaining in... the west side

!

of the control room..."

How can this be

}

complied with or handled during an emergency that l

renders the Control Room uninhabitable?

(0,5)

8.6 A reactor startup is in progress after an extended

shutdown with the first eight control rods

j withdrawn.

A shift change is occurring, and as you

!

review the panels, you notice that one SRM is in l

Bypass, a second SRM reads 2.2 cps, and the

!

remaining two SRMs read 12 cps.

Can the startup

j continue?

Why or why not?

What actions are I

required?

(2.5)

J

8.7 The Steam Jet Air Ejector discharge radiation i

monitor alarms and continuously indicates a reading l

in excess of the level that requires an Alert to be j

declared.

Investigation of_other indications and

=

the monitoring channel indicates this reading,

'

which has now continued for 10 minutes, to be due

p" ipmentmalfunctionf.Isitnecessaryforyou t

.

hf to classify the event and implement the Emergency p

gfd Plan?

Why or why not?

(1.5)

)

. J CATEGORY CONTINUED ON NEXT PAGE

)

.

SRO Examination Page 12 of 14

..

. -. -. -.

-

..

...

-

-.

.-

..- -.

,

-}

-:.

-

'

-

.

j

.

,

,.

.

November, l'985 Limerick Generating Station

'

,

>

.

8.8 During a monthly surveillance of the suppression

chamber - drywell vacuum breakers, you are informed

.

!

that one of the position indicators failed to

i indicate valve movement during the cycling test.

a.

Has the primary containment integrity been l

lost, and will it be lost if maintenance is -

performed on the position indicator?

-(1.5)

-

i b.

Can the plant continue to operate?

If yes, under what conditions?

If no, why not?

(1.5)

8.9 It is 3:00 a.m. and a surveillance is in progress

,

!

that must be completed prior to 5:00 a.m.

to preclude executing an action step as a result of

,

operating in an LCO action statement.

A required piece of test equipment has also f ailed, and a

,

]

Temporary Change must be made to the surveillance procedure in order to complete the test.

You are i

authorized to sign the temporary change as the h

Shift Superintendent, as required by A-3, but no j

other PORC member is present at the plant site.

$

'

Will you be able to have the surveillance continue?

'

i Explain your answer.

(2.0)

)

l 8.10 Section 3.1.3.1 of the Technical Specifications l

provides guidance for LCOs with inoperable control

'

rods.

It permits for repair of an inoperable i

control rod, or for continued plant operation if i

inoperable control rods are fully inserted and i

disarmed, or trippable if not fully inserted.

However, it goes on to state that "with more than

{

eight control rods inoperable, be in at least Hot Shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ."

If an adequate shutdown

,

j margin could still be maintained with more than j

eight inoperable, but not necessarily fully j

inserted control rods, what is the reason for this

i limit?

(1.5)

!

l

,

8.11 The Shift Operations procedure, A-7, states that five "of ficial logs shall be maintained" and temporarily filed in the Control Room.

,

i j

a.

What are these five logs?

(2.0)

l b.

If a change to an entry in one of these five i

logs is required, how must that change be

!

made?

(1.0)

>

CATEGORY' CONTINUED ON NEXT PAGE

%

!

SRO Examination Page.13 of 14

)

.

. _.

. --.

..~

.

,

-

.

,

j

.

November, 1985 Limerick Generating Station s

}

8.12 During RCIC testing, the suppression pool temperature was inadvertently permitted to increase above 110 degrees F.

The reactor was scrammed in

accordance with Tech Specs, and is now in Hot Shutdown.

Suppression Pool Cooling is on and the suppression pool water temperature has been reduced

,

to 98 degrees F.

Can you perform a reactor restart?

Be sure to include all the considerations required for your decision.

(2.0)

END OF EXAMINATION

_

.

J

)

i a

'I

.

i i

u

s.

i

.

SRO Examination Page 14 of 14

-

.

_-

._,.-.

. _ -

, -, -

--.

- - -. _ - _ _

.-

-,

- _..

--.

,.

._

_. _

_ _

_

._.. _ _.

_

, _

_

__

..

.

Er,g 3," 33 gy

.

..

,

f = ma v =.5/t-

-Cycle efficiency = (Network

out)/(Energy in)'

.

w = mg s = V,t + 1/2 at

,

i

.E = mc f

KE = 1/2 mv a = (Vf -.V,)/t A = AN A=Ae"

-

g PE = mgh

,

.

Vf=V + at

- w = e/t 1.= in2/t1/2 = 0.693/t1/2

-

1/2'II*b(*bII*b))

t

-

i

[(t1/2) * I*o)3

,

aE = 931 am i

1=Ieg Q = mCpat i

)

Q = UA c T I * I e~"*

'

{

n I = 1, 10-x/TVL Pwr = w an f

,

.

P = P 10 ""III

_

TVL = 1.3/u

-

'

g HVL = -0.693/u P = P e /T t

o

.

]

SUR = 26.06/T SCR = S/(1 - X,ff)

~

CR = S/(1 - K,7fx)

x

SUR = 26o/ t* + (a - o)T CR (1 - K,ff j) = CR (1 - keff2)

j

T=(t*/o)+[(8-p)bo]

.

M = 1/(1 - K,77) = CR /CR j

g T = t/(o - 8)

M = (1 - Keffo)/I1 - Keffl)

j T = (a - o)/(lo)

SOM = (1 - K,ff)/K,ff

' " IKeff-II/Eeff * #eff/K t* = 10 seconds

'

eff T = 0.) seconds-I

,_

o = [(t*/(T Keff)] + [I,f f (1 + AT))

/

'

Idi=Id

.

j P = '(IoV)/(3.x '1010 ),.'

I d".2,,2 2 g d ' - '

j

'

',

,

I = oN-R/hr = (0.5 CI.)/d (=eters)

'

}

Water Parameters Miscellaneous Conversions

1 gal. = 8.345.lbm.

I curie = 3.7 x 1010eps

,

i Igai. = 3.78 liters I kg = 2.21 lbm.

j 1 fte = 7.48 gal.

I hp = 2.54 x 10 Btu /nr:

,

!

Oensity = 62.4 lbm/ft3 1 ms = 3.41 x 10 Btu /hr

<

!,,

Density = 1 gm/cm3 lin = 2.54 cm

-

!

Heat of vaportration = 970 Btu /lem

F = 9/5*C + 32 i

Heat of fusion = 144 Btu /lbm

C = S/9-(*F-32)

1 Atm = 14.7 psi = 29.9 in. Hg.

-

.

J

..

.

m

.

. -

.

-

.

,

N;v;cber, 1985 Litarick COncrating Statien

.

)

5.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS (25) - ANSWERS 5.1 a.

(1) Nucleate boiling (2) Film boiling b.

Nucleate boiling.

(0,5)

(1)

The thermal conductivity of steam is very low.

Therefore, the steam layer increases the resistance to heat flow if film boiling is occurring.

(2)

In nucleate boiling, the physical movement of the bubbles into the fluid

,

assists in the removal of heat.

(3)

Nucleate

'

boiling also increases the fluid flow rate, j

further increasing the heat transfer rate.

!

Credit at 0.5 each for any two of the three items or other appropriate items.

REFERENCE:

LOT-1340, ppg. 2 - 7.

5.2 (1) The count rate would be lower in the voided

"

part of the core than below the water level.

(0.5)

(2) The neutron detectors are designed to detect thermal neutrons.

In the volded sections of the core, there will be less or no thermalization, and the detectors will show a lower flux.

(1.0)

,

REFERENCE:

LOT-1700, ppg. 6 - 7.

i

].k SRO Examination Answers Page 1A of 16A

,

..

-

- - -

-

. - -

-

-. -

-

- - - - -

-

.

-

.

November, 1985 Limerick Generating Station

.

5.3 (1) During low or no forced convection flow -

g

.j conditions, there is no flow from below the core.

(0.25)

Flow from above the core must replace the steam loss of inventory.

(0.25)

(2) If the steam loss rate is excessive, it may preclude or' reduce the flow from the upper plenum.

(0.25)

As the steam boils of f, the fuel will heat up, and as the steam flow decreases due to the lack

'

of water being present, water will flow back into the channel.

(0.25)

,

This will result in rapidly changing fuel temperatures (oscillations), and excessive thermal stresses.

(1.0)

!

REFERENCE:

LOT-1690, pg.

7, and T-LOT-1690-7.

5.4 a.

(1)

Can be transported throughout the coolant i

system.

-

.*c (2)

Fouls heat transfer surfaces.

(3)

Collects inside the RPV, becomes activated, and becomes another source of radiation.

(4)

Once activated,' can collect in low flow areas if transported outside the RPV,

.

Credit for any two.

b.

Chemical shock

}

(A) % {vt CtJ h h 1 8 d h d

Thermal shock

,

w

{

g g y,_,,

Mechanical shock f

i REFERENCE:

LOT-1010, ppg. 10 -12.

i

5.5 The indicated flow will increase.

(0.5)

^

The differential pressure between the two pressure taps is used to determine the flow rate.

The upper tap is the low pressure tap, and the lower tap is the high pressure tap.

If the upper tap indicates

,

l ow, the dP will increase, and the indicated flow rate will increase.

(1.0)

'

g REFERENCE: LOT-0010, ppg. 31 - 32, and

J LOT-1280, ppg. 2 - 6.

SRO Examination Answers Page 2A of 16A

.

- -

.

November, 1985 Limerick Generating Station

.

5.6 (1) SDM will decrease continuously (0.5)

(

(2) due to the decay of Xenon (0.5)

(3) since the Xenon peak has already occurred.

(0.5)

(4) Recriticality will be a major concern since peak Xenon is about (4.7 percent) following a scram from full power operation.

(0.75)

(5) Other measures will need to be taken to assure shutdown.

(0.25)

REFERENCE:

LOT-1510, ppg.

6, 7, & 12.

r 5.7 If an ATWSlhad occurred and the reactor could not be shut down by other means.

(1.0)

M1f b M2 6,Mg 3,w g dgfe[.unm

-av f

,, 9,y w

Uncovering of the core woul'd pr,evenI*7 0 If the9 d thermalization of neutrons and thus provide another g*J

' means of add ing negative reactivity to the core.

(1.0)

()/ ~[o/ -

5 Lwn% vid.a Q)hvoud %n t& mYs'es

-

REFERENCE:

'6-117, Power control path, and LOT-17 pg.

6, Item 33.

.

-

7.pdv er. & h 4pw 'n.W, q

ritical power)1.0 ko $

5.8 a.

CPR is the ratio o to some

}

operating power.

(0.5)

It must be kept greater than one because at a CPR of 1.0, transition boiling will occur

.

leading to rapid temperature fluctuations of Ir0 2.W w, *,j the cladding surface temperature which will

?,

cause stresses leading to cladding

perforations.

(1.0)

b.

(1)

CP increases --y C. T' F.

'% cAAA<,u.

(0.25)

because a greater enthalpy rise is required to bring the coolant to saturation conditions.

(0.5)

(2)

CP increases -% C. Pvt. *mtALA444 (0.25)

because more power input is required to raise

.

the coolant enthalpy to saturation conditions. (0.5)

REFERENCE:

LOT-1360, pg.

6, and LOT-1370, pg.

5.

5.9 Reactor pressure instrumentation gives results indicative of reactor temperatures for a saturated condition.

Following a LOCA with vessel depressurization, saturated conditions do not exist.

REFERENCE:

LOT-1692, pg. 14.

SRO Examination Answers Page 3A of 16A

.

m

.

..

.

..

.

..

.

._.

-.

_

--

.

,

-

.

,

November, 1C85 Limerick Generating Station

.

5.10 The worth of a particular rod is a function of the

)

'

square of the local flux (divided by the square of

.

the average flux.]

y,pd A d

-

.

REFERENCE:

LOT-1490, pg.

4.

5.11 The moderator temperature coefficient adds negative

reactivity as a function of changes in moderator j

temperature.

(1.0)

'

,

i As a heatup progresses, the moderator temperature increases, thus causing the addition of negative reactivity.

(0.5)

l

?

As power increases from 50 to 90 percent, the l

i i

reactor continues to operate at saturated j

conditions with ef fectively no change in moderator temperature.

Thus no reactivity change occurs.

(0.5)

<

l REFERENCE:

LOT-1450 W

H ina f. ~ M = ewe m 5.12 a.

P=P e

o

'

m ~,. m ' o.oo,e % p m 60/(-80)

!

= 75 *

is s < > 's. m M i.o y 1 @ o. o n v 't.

,

e

,

l

= 75 * O.472 35.42 on Range 4

=

b.

For a large negative reactivity insertion,. the neutron decay rate is a function of the longest lived delayed neutron precurser group, which is about 55 sec.

( 0.75 )

The short lived precursers decay in about two minutes, so it is appropriate to use only this

,

single value for this problem.

( 0.7 5 )

REFERENCE:

LOT-1430, pg. 16 and T # 11.

.

i

!

END OF CATEGORY

,

!

$

i

.

SRO Examination Answers Page 4A of 16A

,

.

-

- -

-

- - - - -

-

- -

_

..

_

-. _ _. _ _. _ _. - _ _ _ - _ _... _

_ _ _ _ _

_ _

_... - _.. _ - _ _

.-. _ -

_

-

.

-

.

November, 1985 Limerick-Generating Station

,

i

'

s, 6.

PLANT SYSTEMS: DESIGN, CONTROL, AND INSTRUMENTATION (25)

- ANSWERS 6.1 m.

(1)

In-sequence rod inserted one notch past -

the alternate insert limit.

(2)

Rod in lower group inserted (one even]

.

i notch past h alternate withdraw

'd '.

limit.

,, [

s, '

s

,

-

g' i. -

c-

,..,

( Alternate limit is always one even notch

.

h;*(/Al.f further inserted past a withdraw or insert limit.}

-

b.. cone insert error - Block applied to rod causing de d** *v the error, o f p,

c gp

,

ALTwo insert errors - Block applied to the two 4.iJ}f

,y rods causing the errors, p, sr.

More than two insert errors All rods will have,an insert error except ' o,7

,

those with withdraw errors.

.

All rods, except those with insert errors, 3,c

,

will have withdraw blocks.

REFERENCE:

LOT-0090, ppg. 11 -12.

6.2 a.

Via the

"B" Core Spray header.

'

b.

RWCU (0.5)

To prevent removal of injected boron.

(0.5)

c.

Heaters are necessary to prevent precipitation.(0.5)

The A heater maintains normal tank temperatures (between 75 and 85].

(0.5)

Both heaterp are %used when, adding chemicals.

(0,5)

(#,2?< b 4 h.d<g REFERENCE:

LOT-310, ppg.

4, 5,

11.

6.3 This system consists of a dP cell connected inside

,

the Drywell between the two loops.

(0.5)

It will detect a break or leak between the vessel wall and the shroud penetration (per. LP) or I

actually anywhere back to the inboard isolation valves - including connecting lines from NPCI and SLC (per Tr-# 1).

(Credit for either answer.)

(1.0)

The alarm is on high dP (4.4 psid].

(0.5)

Ih)

}

REFERENCE:

LOT-0350, ppg. 12 - 13 y

.

SRO Examination Answers Page SA of 16A

'

,

L

-

- -

-

- - - - - - -

- - - - - - - - - - - - - - -

- - - - - - - - - - - - - - - -. - - - - -

.

November, 1985 Limerick Generating Station

.

)

6.4 a.

Preferred Source:

1DA 250 volt DC

) ~N '/

W.

First Alternate:

114 D-C-F (or TSC UPS)

e,x,uo& W *'

Second Alternate:

114 A-G-F g,m a 6

d' g

-

(Alternates are 440/480 AC)

b.

RWCU and Recirc Controls Process Radiation Monitoring Area / Process Monitoring

.

Inboard Valve relay panel Area / Process Rad Recorders Turbine EHC RPS Trip System A C

t for any four at 0.25 each.

REFERENCE:

LOT-0300, pg.

3, T# 2.

,r.~

j 6.5 (1) 28 percent limiter c. ') ' ',s Total feed flow (,20 % to assure NPSH (0.5)

enti. vt.

-

!

Vessel level (12.5 " to anticipate RPS scram and aid in power reduction (0.5)

Recirc pump discharge valve [T M It.

ully open to protect pump from overheating.

(0.5)

(2) 75 percent limiter (runback)

Lev e l (, 2 7.5" e_r i ividual feed purp flows eM 4 20 %

- to help restore level by reducing reactor power and thus steam flow (0.75)

d r%

To ta l f eed f l ow 3P 9 0 % _ov condensate pumps -"-h;g A*4 ee(8pt to a

prevent feed pump trip on low NPSH.

(0.75)

REFERENCE:

LOT-0040, ppg.

6-7, T-2.

%.y,3,, W V)U h VJ. % 1 VD Id "%M AiWN T d (fl0 N*f 3Y hw eM

s

.

J t

SRO Examination Answers Page 6A of 16A

.

-

- -

- -. - -.

- - - - - - - - -

- - - -

-

.

... -

.

- -

- - -

-

- -

-

- - - - -

-

- -

- -.

- - - -

-.-..

.-

'

,

.

,

.

November, 1965 Limerick Generating Station

.

l T

6.6 a.

Scram pilot valves are normally energized.

Backup scram valves are normally deenergized.

'

b.

Scram Pilot Valves - Each trip system controls one solenoid - one changes state on loss of signal f rom RPS A, and the

'

other on loss of signal from RPS B.

i Backup Scram Valves - Must receive signals from both RPS A and B to energize.

c.

Check valves ensure that a single failure of one valve will not prevent scram function.

REFERENCE:

LOT-0300, ppg.

2, 9,

10.

6.7 a.

The RBM receives inputs from the four LPRM

'

strings (Levels A and C) around the selected control rod.

(0.3)

.

It also receives a core average input from APRM

,

'

C [ Alternate E].

(0.3)

'

If the LPRM average is less than the APRM.

~

signal, the cAPRM signal is W o.d Med +o e. Q +(thftW a nA(.0.45)

L P N y

If the LPRM average is equal to or greater t)Yan

the APRM signal, the LPRM average is passed.

(0.45)

b.

RBM A receives inputs from Flow Units A and C.

(0.2)

'

,

It passes only the signal of lowest v..lue.

(O.3)

'

CTr;n (d c.4 9 e it n, fop 8]

'

REFERENCE:

LOT-0280, ppg. 4 - 5.

6.8 On receipt of a reactor low water level (-38"] or MSL high rad [3xNFPB], the following valves close: (0,5)

1.

Main Steam Drain Sample Inboard and outboard (0.5)

i

,

2.

Recirc Sample Inboard and Outboard (0.5)

On receipt of only the MSL High Rad, the following (0,5)

actions occur:

,

1.

Mechanical Vacuum Pump trips (0.5)

2.

Mechanical Vacuum Suction Line Valve closes.

(0,5)

,

REFERENCE:

LOT-0180, pg. 6.

,

_

.

SRO Examination Answers Page 7A of 16A

.

. -

.

-

.

-

.

November, 1985 Limerick Generating Station

,

t 6.9 a.

Upscale gg

> 10 ep, INOP Dowr. scale er

< 3 cps i

Not fully Tnserted and < 100 cps b.

For SRMs A and C, the associated IRMs are A, C, E,

& G.

(or for B & D: B, D, F, & H.)

(0,5)

They must be on Range 8 or above.

(0.5)

,

i REFERENCE:

LOT-0240, ppg.

6-7, T-8.

. - l)

d5 6.10 a.

ADS will not initiate until the low level p, c. v t initiation point has been held for at least 106 0 d 5'

sec.

After this, if it does initiate, the initiation signal will seal in.

The low level

]

initiation point is not itself a seal in j

signal.)

b.

The low level confirmatory signal prevents inadvertant initiation following a single level instrument failure.

REFERENCE:

LOT-0330, ppg. 14 - 15, T-6.

,

END OF CATEGORY

!

A LJ I Nole : If [g bl,p D/$/h JOI h ; b b

M3

)

m ;), le L a $ o.o m A S @ i W, 7

,

-4

!

,

i F

1..

i

4 l

'

SRO Examination Answers Page SA of 16A

-

.

a

.

November, 1985 Limerick Generating Station

.

!

7.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL CONTROL (25) - ANSWERS 7.1 a.

From SP/T-1, the Heat Capacity Temperature Limit for a reactor pressure of 500 psig is 180 F.

This gives (180 - 150) a delta T Heat capacity of 30 degrees F.

Then from SP/L-1, the minimum safe level is about 13,5 f eet.

b.

If the pool level is too low, there may be insufficient volume to condense the steam.

REFERENCE:

T-102 figures, LOT-1560, pg. 18.

7.2 a.

An unexplained decrease in reactor power.

An unexplained decrease in core flow indication.

w.

An unexplained increase in recirc drive flow to the loop containing the defective jet pump.

)

b.

An unexplained decrease in indicated dP on the jet pump sharing the same riser with the defective jet pump.

This would be found [on Panel C619) in the aux equipment room.

REFERENCE:

ON-100 and LOT-1550, pg.

3.

7.3 a.

At 920 psig with the bypass valves.

b.

RCIC and CRD c.

(1)

Control rods (2)

RCIC Y ft. SR.4 -

(3)

RCIcem.

REFERENCE:

GP-3, ppg. 17 - 18.

t 4}c,GP.36 m m't%4-noYcaet w e. /1 A58**-

Ake und n 'h ym vek -u %Adrd,JDenth > u>ry's l

)

'

SRO Examination Answers Page 9A of 16A

-

.

November, 1985 Limerick Generating Station

.

T N(1) The high pressure condition might be caused by

'

7.4 ijs an EHC malfunction.

(0.5)

p

%\\

.

f/

'r (2) If the EHC is attempting to maintain an

'

6 (,@'j abnormally high pressure at the turbine inlet, (0.5)

.

,.)

g?'p P /

reducing power and thus total steam flow will (0,5)

result in a smaller pressure drop in the steam (0.5)

geg

.

t y

lines.

This will result in a lower reactor pressure being required to maintain the turbine fp inlet pressure.

(0.5)

,

'

REFERENCE:

OT-102, Bases.

7.5 a.

(1)

The platform operator

'/fg p (AL and

-

(2)

the Fuel Handling Director 3 --

<

,

(3)

are required to visually examine the relevant component.

-

l l

b.

(1)

Observation of major damage.to spent fuel, i

or (2)

Water loss below fuel level in the spent e

fuel pool.

)

REFERENCE: S97.OB, C,

and others, Precautions, 7.6 a.

Deep rods are defined as rods positioned t

I between 08 and 26. o rt, c o - ff n 19 @ W C ' * *'

/p.-t41 tiata, p t.'m ugr 4 h eroe A o 'f 7],(,.

e'

-

e

,

b.

(1)

The margin between actual-reactor power and the APRM setpoint is reduced as recirc flow is reduced with a fixed rod pattern.

The rods are inserted to increase this margin, s..)

-(., %g ra gcre roh 4 M <a ku pp l

A 10 percent flux spike can be expected f@ rom g c.

the first opening jog of the recirc pump discharge valve, and additional margin to the scram limit is required.

REFERENCE:

OT-112, Bases, ppg. 1 - 3.

piC !V eje4W - r-t rs No Vbt f F'

>

W Ag te,&~ L\\m *,.. lw s %4 n A rt"- WO >

ap a%swit'..

i

)

!

l

SRO Examination Answers Page 10A of 16A

,

-.

-

.

November, 1985 Linerick Generating Station T'

7.7 (1) An RMP_must_hage_been approved, f or the area in which the work is to Ee perforned.

(0.5)

(2)

If the dose rate is > J_00 mr/hr following must be pr'ovidh, one of the

~

g wA radiation instrument that continuously n'

indicates the dose rate in the area.

0 b

(v

.

y#[y ga (,

9 An alarming dosimeter if the dose rates have

,y

,

been established and the personnel have been

,y('S *

'

informed of the survey data.

[A,V $j.

gO cA qualified HP Tech who controls access and

.

/s

y c performs periodic surveillance.

[i

,

..f

/

4[d,d

-

F (3) If the dose rates are in excess of 2000 mr/hr whole body at 18 inches, the area shall be roped off, posted, and a flashing red light

'

used or an individual provided to prevent unauthorized access.

(0.5)

REFERENCE:

HP-109, ppg. 1 - 2.

'

7.8 a.

MCPR

,

b.

Inop accumulators (0.5)

or excessive scram times (0.5)

are cause for declaring rods inop.

(0.5)

(Could go on to discuss separation by at least two control cello - not required.]

REFERENCE:

ON-107, Bases, and T.5W. pg.

3-1.

7.9 (1) Whole body rad levels are 30 mr/hr or greater.

r; (2) Average removable contamination >about 10,000 p

dpm/200 sq cm beta-gamma or 100 dpm/ 100 sq cm q

alpha.

'6 (3) Area is posted no Airborne Rad Area.

(4) Any measurable neutron exposure exists.

,

I (5) Any area for which radiological conditions are

')

unknown.

~v (6) Area is designated by llP as "RWP required for entry" (*)) B reA<Ata) f.onbmin Mcd gsle.

4,

)

REFERENCE:

IIP-310, pg.

2.

(O 4 tiding, befg n y;ndh j tedA'* ;a* H d dUfl

'

l% Ly w wjy J.<,n. 4 AA w.'s ^$ m ;,

(10) Aa r o.4,O d o r,.w 64 D%5/O Q *- a ~ +6 8-1 '"3 d'

SRO Examinatio'Jn Answers P, age 11A of 16A

%

.

-

.

-

--

- -

-

-.

- - - _ _ _ _ - _ - _. _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ - _ - _ _ - _ _ - _ - _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _

.

'

.

November, 1985 Limerick Generating Station

-

I 7.10 a.

Reduce reactor power.

(0.5)

.

'

C

'

This will reduce the number of non-condensibles introduced to the condenser, thus slowing or

'

terminating the vacuum loss, and giving the operator time to take other action.

b.

(1)

Could result in an excessive north vent stack release rate since the hold-up pipe will be bypassed, i

(2)

When the reactor is above five percent, could have a detonable mixture of hydrogen and

!

oxygen.

REFERENCE:

OT-116, Bases.

l END OF CATEGORY

,

i

>

!

l i

i.

!

,

.

I i

i j

SMO Examination Answers P, age 12A of 16A

.

,

November, 1985 Limerick Generating Station

-

8.

ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS (25)

.

- ANSWERS C

8.1 a.

Since the reactor is in coastdown, and it is probably the intent to maximize the power, 3 3.4.t assume the flow is at 100 percent.

This will s') C..t o c result in a K f actor of 1.0.

Then from T.S.

.; y 3,1,3 4 Figure 3.2.2-1, MCPR is about 1.30.

A wl. F f. Caml.{ li e y as X, t i.o yT c m eet *,#. 3 f, b.

With the EOC-RPT lnoperable, the limit is a function of the scram time, which will be the dominant factor in turning power.

(If the EOC-l rpt were operable, the recirc pump trip would l

initially turn power.)

l

'

REFERENCE:

TS ppg. 2-8 - 2-11 and 3-46 - 3-51.

LOT-1390, pg.

5.

ld u k U ')

8.2 (1) When the mode switch is in RUN, 0.66 W + 51 %

1. A rtM Jed (CnM d.c /f. W M t 56/, u k S4 k yt4,.if u L4 ui.4.

~i W *f'"'

(2) When the mode switch is in STARTUP or HOT 9, L,f STANDBY, 15 percent of rated power.

REFERENCE:

LOT-1390, pg.

2.

t Tu b S g t.t Td.h 2.2.1 - 1

)

8.3 (1) With only one operating system, 3.7.2.a.

applies and the reactor can be operated for up l

to seven days with the system inop.

(0.5)

l

'

(2) However, with the second system 0.0.S.,

this action statement no longer applies and the case is not specifically covered.

(0,5)

l (3) Since the second system cannot be returned to l

service within 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months , and the temperature is

!

already high, it also cannot meet the

'

(

surveillance requirement of 4.7.2.a.

(0.5)

i (4) With two systems 0.0.S.,

section 3.0.3 applies,(0.3)

and action must be taken within one hour (0.3)

to be in Startup in the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,and (0.3)

,

l Hot Shutdown 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months later, and (0.3)

l Cold shutdown in the next 24 hourn.

(0.3)

l REFERENCE:

Tech Specs and Unses for 3.0.3 l

!

l SRO Examination Answers Page 13A of 16A

-

.

a November, 1985

-

j*

Limerick Generating Station

,3 '" ~ <?.

.s*

One hour - i"; gov *<Y

'

8.4 a.

JD ad pr'#b b.

Immediately attempt to notify the Station

!

Superintendent, Assistant Superintendent, or designated alternate.

c.

The clock started at the time of occurrence of the event, leaving only 15 min to notify NRC.

REFERENCE:

A-31, ppg.

1-2, 7.

.

8.5 (During these conditions, the restriction does not

apply.

Also, during transients, shift supervision j

may alter this requirement.]

!

In that event, the "at the control" location is the emergency shutdown panel for that unit.

(0.5)

REFERENCE:

A-7, Figure 2 footnote."

^

8.6 (1) With the mode switch in Startup, three SRMs are required to be operational with counts > 3 cps

,

per 3.3.7.6.a (0.5)

(2) If the signal to noise ratio can be shown to be

)

>

2, then 0.7 cps is acceptable, per 4.3.7.6.c (0.5)

and the startup may continue.

(0.5)

i (3) This must be confirmud, or one SRM must be i

returned to operability within four hours.

(0.5)

l l

(4) If this cannot be done, the reactor must be in Hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

(0,5)

REFERENCE:

Tech Specs page 3/4 3-88.

.

8.7 No.

(0.5)

l i

Even though the indicated radiation level exceeds

,

l the implementing guidelines, it has been found tra l

!

be a spurious indication, and EP-101 states if the

!

trigge s known to be spurious, do not classify the event.

(1.0)

REFERENCE:

EP-101, ppg.

2, A-13.

SRO Examination Answers Page 144 of 16A

. -

-

- -

- - -

-

-..

-

- -.

- - - -

-

-.

.

..

.

November, 1985 Limerick Cenerating Station

-

s 8.8 a.

PCI has not been lost initially.

(0.5)

.

d

It won't be lost if the repair procedure for f,i e

,,,

maintenance on the failed position indicator I

.

does not not violate PCI, or if no repair is (* J performed.

.

(1.0)

,

s d*#j r)?

i b.

The plant can continue to operate if (0.5)

,I

,

l % ' p ' g '/ 1

!

..r*

1.

the other vacuum breaker in the pair is jr

.

4",A verified to be operable within two hours and at least once/15 days thereafter, or (0.5)

VS (6.V.I.o.t)

2.

the vacuum breaker with the failed indicator is verified to be closed by a pressure test within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and once/15 days thereaf ter. (6,y.l. C t)

(0.56 (Otherwise be in Hot Shutdown within the next 12 (

hours, and Cold Shutdown within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .)

.

l REFERENCE:

Tech Specs Sections 3/4.6.4.1

,)

8.9 Yes.

(0.5)

Approval may be obtained by telephone (0.75)

L with the approvers initials entered by the change l

along with the words " Telephone Approval."

(0.75)

REFERENCE:

A-3, Section 5.3, pg.

3.

8.10 The occurrence of eight inoperable rods could be indicative of a generic problem and the reactor

,

l must be shutdown for investigation and resolution.

i REFERENCE:

Tech Specs Bases, pg. B 3/4 1-2.

8.11 a.

(1)

Reactor Log Unit 1

(2)

Chief Operator Log

}

(3)

Control Supervisor Log l

l (4)

Radwaste Log

!

(5)

STA Log b.

Changes or deletions should be made by a line

<

through the entry so as not to make the original entry unreadable (0.5)

and initialed.

(0,5)

REFERENCEt A-7, ppg. 24 - 25.

,

"

SMO Examination Answers Page 15A of 16A

. -

-

_ -.

_-

.

._-

.

..

,

.

November, 1985 Limerick Generating Station

'

-

8.12 No.

(0.5)

You must have the SP tempavature less than 95 degrees F (0.75)

because Tech Spec 3.0.4 does not permit entry into an operational condition while in an Action Statement.

(0.75)

REFERENCE:

Tech Specs 3.0.$ and 3.6.2.1.

-

END OF EXAMINATION i

i i

l

,

(

!

a r

,

I I

i i

t i

i f

i

\\

!

-

SRO Examination Answers Page 16A of 16A

-

- -

-

- -

-

.

k 1 T f7 C f7 m e 11 fc2.3

'

.

NRC SRO EXAM - 11/11/85 - PECO COMMENTS

.

,

5.3 Question uses buzz-word of " counter-current" flow which might be confusing to the operators.

5.4 Use of buzz-word of " crud burst" might be confusing to the operators.

Additionally, as por attached reference crud burst answer op key should be expanded to include indicated exampics as in LOT-1010, pg. 10 and 11.

5.7 Question improperly worded in that level is never decreased below TAF as per T-ll7.

Several different answers should be accepted:

(1)

Entry conditions for lowering level as per T-ll7.

(2)

If candidate realized that question asks for h[

lowering level below TAF in violation of y>*gf.

procedure, his answer would be that this action E@

would never be desired.

(3)

Lowering level to TAF provides less natural p

circulation producting voiding in the core. (see

.

(

attached TRIP proc. bases)

g

<

b (4)

Better concentration of Boron in solution due to.V 4

P less volume in vessel above TAF.

p*

(Y' Y 5.8 (b) Question desired change in Critical Power Ratio /and the answer addresses Critical Power change only. /

p g..%4 "4 g -

.

'

.

This should allow the students to provide additional answers of CPR increasing for both b(1) and b(2)

.

J.#.

r-d assuming that core bundle power is kept constant since f

CPR is ratio of critical power divided by bundle power %

of interest.

pg. $4 5.11 Answer indicates definition of moderator temperature coefficient needed for 50% credit on question.

This was go gv/,

never asked as part of question.

Grading should be distributed accordingly as per explaining reactivity changes in the two situations mentioned.

~)Cbi&

cJxr

.

y

,

.

O

-- - _.

-

_

-

.,

.

NRC SRO Exna - 11/11/85 - PECo Commgnto pig 2 2 of 5

.

,

5.12 (a) Answer could also be t Rx power as velh as power indication of Range on IRM.

4,gc.

-

h

-

(b) Explanationforthisquestioncould(e!,lsobediscussion

[t of - 80 sec period which occurs after any normal y C#" aq,.

Reactor scram.

This would indicate that short lived

/

precursors are non-existent and should not be

~~

considered in the grading.

,

pgoft.f 6.1 (b) Answers for one insert c'tror and two insert-orrors ares incorrect.

The rod block' does not occur until the

-

third insert error.

Lesson plan will be corrected to vf, indicate as such.

p

,ps For more than two insert errors all rods have an insert M

D block.

All rods except,those with insert errors have a FN d$

withdraw block.

'3M d

Additionally, a computer malfunction could cause an --oF" j

, jj a,

w o,

insert error.

e p

6.2 (c) Part (2) of this que'stion asks for specific purpose fo ek5 hi cach of the two heaters.

The attached reference LP LOT-0310, pg. 5, para. 7.a should bn an alternate answer to this question. t y,g

,

M4 6.3 Candidates could have interpreted this sparger o P to mean x

+/- on A P instrument.

Depending on which line breaks, the instrument will go hi/ low A P.

Therefore alarm could be on high or low AP (-4.4 psid or +4.(ipsid).

gg, S

'

6.4 (a) For full credit answer gould also be:,

Preferred source:

250 VDC i

'

of- -

),

"'

1st Alternate:

TSC UPS

'

2nd Alternate:

440 VAC (see LP-LOT-0330, pg. 3, Support Information)

j (b) Additional acceptable answers are:

\\

d (i)

NSSSSF LG S Lo d And sk - noR L? -

g (ii) APRM's /

,

I

'

<

'

I'

})

,

b r-

,

,,..

-

,

....

,

.

'

NRC SRO Exam - 11/11/85 - DECO Commsnts PIgn. 3 o f 5

.

(Ref:

See LGS Load Analysis)

6.5 Answer for limiters to be in effect are:

g, 'Tklo k tre. 3 (1) 28% Limiter y d -)$ A 4 Lt. A W ^

N

$"

Total feed flow <20% or

.a..,k 4s.

Rx Vessel level <l2.5" or U

Recirc pump discharge valve not full open (Purposes are correct as stated)

(2) 751 Limiter Rx level <30" and any individual feed flow <20%

or Total feed flow >90% and any condensate pump breaker

.

not closed.

.

(Purposes are correct as stated)

6.6 (b) RPS " signals" was interpreted as RPS Logic by many k

candidates.

, fjl A.d4 u sW-6.7 Correct answer for case where LPRM average is less than APRM signal is that LPRM signal is conditioned and passed.

A'

h, nMy 6.8 An acceptable answer for valves that automatically close could also be " Main Steam samples" and "Recirc Samples"

which implies inboard and outboard valves.

On 3XNFPB isolation actions most candidates stated that mechanical vacuum pump trips.

It is implied that MVP og suction line valve closes because_that is h _happens-whenever MVP trips.

(i.e., [ is an interlock).h

6.10 Candidates may answer this discussing the 420 see timer M t associated with logic for High drywell pressure and LoLoL

1. Y level (See LP-LOT-0330 for details).

7. 2 (a)

Becaus,e question indicates you are looking for a failed g

jet pump in the control room, the words " unexplained"

,

do not necessarily need to be added to the symptoms.

gpolly

[> W ('

hh(e d

%. r e

+

,.. -,.

,- -

,-

---n

l

-

. -.

_....

....

.........

.

NRC SRO Exam - 11/11/85 - PECo Comments Page 4 of 5

'

7.3.c.(2) Because of the present lack of decay heat after a LGS Rx shutdown, S/RV's are used in pressure control after the vessel is isolated.

GP-3 will be changed by the Operations Engineer to reflect this.

Therefore, an alternate correct answer to glL this question would be S/RV's.

,

7.4 Alternate answers for malfunctions would be any positive ox '4 % c reactivity insertion or stop valve, control valve, or MSIV y e,oM closure as per OT-102 Bases.

g 4.u 4. 2 6

~

Grading.for the second part of this question should be revised accordingly if EHC malfunction not assumed in firs H vuIn fl*

part of question.

7.6 (a) Definition of " deep rods" has an alternate answer of ow., b d N

f-W r*M de2IM@

SNE manual attached).

16 or 18 (as per G.E.

A fn, Li4'd toTos/

(c) Candidates stated additional deep rods drive @-[*g

"*^Q-H***

prevent power increase when starting pump and opening.*MM"'

It should not be necessary to state '".

discharge valve.

10% Rx power spike for full credit.

7.7 Because this question mentions " emergency use", then an gg alternate acceptable answer should be to keep the

.

i'

k g I, individuals who enter the high radiation area _from Hf lobo

dose and so much rem Thyroid gg&*re.

g ) exceeding so much rem W.B. dose as per the LGS Emergency Plan pro

-

L ((attached).

q q O/gg.>'> >

,

M 7.9 Attached list of HP-310 (pps. 2 and 3) indicates 10 f

dd

possible answers for this question versus the 6 listed in present answer key.

pg.

ydt.

8.2 There are 4 possible correct answers for the desired LSSS gj gg,.

values.

The attached T.S.

table 2.2.1-1 indicates these

.

-

Any two are acceptable. - a a*ded kg 2.2.,1

)

"i O

answers.

LY ohe ugtu&nn /r dournSA h kc$u A IVM kp3A 8.3 Point values fur grafing indicate specific steps required for full credit of question.

In actuality, candidates would more likely answer this question by referencing CREFAS Tech Spec and then realizing T.S. 3.0.3 was the applicable action.

Grading should be adjusted ba' sed upon candidate's ability to show proper flowpath for his answer. OL B.4 (a) An alternate correct answer for this question is no g

action required because A-31, pg. 7, paragraph 5.1.9

-

y'v i

\\

. (.O

.

'

4)

,

a3p e

.

-.

.

_

_ _ _ _ _ _

.

,

.

..

. - _....... _.. _ _ _

_.

__

.._._

. _ _ _,,.

,

'

.

NRC SRO Exam - 11/11/85 - PECo Comments Page 5 of 5

states delivery of rad. material and not loading of a truck as question states.

(b) Plant personnel to be notified could also be all personnel in the localized area as per EP-303 (attached)

g P

'

(c) Scme confusion exists as to when clock actually starts.

-g The occurrence of the significant event could be

_,[

j/

interpreted to be when the HP reported to SSVN at 9:15

cm2.in Radwaste.

This would mean that SRO would have M f that radiation contamination exists at 30,000 dpm/100 an hour from 0915 to report to NRC.

An additional comment is that pg. 2 of A-31, paragraph 5.0 specifically states "Use of Attachment 2 is recommended for quick reference.

..

This indicates.

"

.

that an SRO should not have to commit this listing of notification to memory (see attached reference).

Grading of this question should be adjusted as such.

Au.4[vfJ gJ wM Mo-fa D 13 p i LA s

,

2yna A J6.LAJk4.;IJw.

"

d Gv h h e A-3I p A s l gas ~ U4lhs~%V& A'1A

" P" P

.

t

.

w

,

-v w,

,,

y

-

Wfl0 f77 C l2 NRC Resolution of Exam Reviaw Comments and Changes

.

Noted to be Necessary during the Gradinc

Question Number Comment / Resolution 5.3 The referenced " buzz word" came straight from the referenced lesson plan.

5.4 The referenced " buzz word" came straight from the referenced lesson plan.

Examples from the referenced lesson plan were included as correct answers during the grading.

However, the various examples were required to be different and not just variations of each other.

5.7 Full credit was given for any two of the following:

1.

Reflection, 2.

Increased voiding, 3.

Increased boron concentration, and 4.

Reduced steam flow to the suppression pool.

Interpretation of the question was not a problem for the candidates.

5.8 This comment was well taken.

The answer sheet was corrected.

The second part of the comment was addressed in the question.

5.11 The answer in the answer key does-not require a definition, but gives half credit for simply stating that void coefficient is negative.

5.12 Comment accepted.

.6.1 This question was graded per the lesson plan, but full credit was not deducted if the lesson plan answer was not given.

One half of the candidates answered exactly as stated in the LP.

Plant comments indicate the LP will be updated.

6.2 This comment does not change the answer given in the key.

It only provides alternate wording.

6.3 Comment accepted.

6.4 Comments accepted.

Only descriptive identifications were required.

6.5 Comment accepted.

The answer given in the key was the negative of the correct answer.

This was noted during the exam review.

..

.

-

-.

-

- - -. -

- -

-

- - - - -

-

..

-

- -

-

- -

.

-

.

6.6 This interpretation was announced during the examination, and the words "from the RPS logic" were

added to the question.

All candidates should have had this interpretation.

~

6.7 Comment accepted and answer sheet corrected.

6.8 Comment accepted.

6.10 Comment accepted and graded according to the candidate's stated assumptions.

7.2 The exact wording of the answer key is never required.

The specific woding in the key is directly from the procedure.

.7.3 RCIC was added as an alternative answer.

Operations is correcting the procedure.

7.4 Comment accepted as long as the answer is internally consistent, and the second part responds to an event that recirc flow reduction will mitigate.

7.6 The alternate answer was accepted for 60 percent credit because the question referred to the procedure requirements.

The 10 percent spike value was not required.

However, it should be noted that inserting the rods will not prevent the spike.

7.7 Either answer was accepted.

The alternate answer provided by the plant was required to grade only two exams.

7.9 Additional answers were added to the key.

8.2 The additional two answers from the Tech Specs were added to the key.

The LP was wrong.

8.3 Comment considered during the grading.

8.4 All candidates answered this question per the referenced procedure.

Thirteen gave the one hc r answer that was required by the answer key.

The other three gave four hours.

Oredit was not deducted due to the plant comments.

Parts b and c were graded according to the procedure.

-.

-

v t e Inspection Report - Limerick - 1985041
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \| 25 \| 26 \|
1985 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 11(0) 98(0) 13(0) 15(0) 17(0) 19(0) 20(0) 21(0) 23(0) 24(0) 25(0) 26(0) 27(0) 28(0) 29(0) 30(0) 31(0) 32(0) 33(0) 34(0) 35(0) 36(0) 37(0) 38(0) 39(0) 40(0) 41(0) 42(0) 43(0) 44(0) 45(0) 46(0) 48(0) 99(0) Security Operator Exams Emergency Planning Other

IR 05000352/1985041

Text

Navigation menu

Search