Text

Exam Rept 50-352/86-24 OL on 861020-23.Exam Results:Three Reactor Operators,Four Senior Reactor Operators & One Instructor Certification Candidate Passed
	ML20207Q043
Person / Time
Site:	Limerick
Issue date:	01/12/1987
From:	Collins S, Howe A, Keller R; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20207Q039	List: IR 05000352/1986024; ML20207Q038;
References
	50-352-86-24-OL, NUDOCS 8701210192
	Download: ML20207Q043 (97)
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U. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO.'

86-24 (0L)

FACILITY DOCKET NO. 50-352 FACILITY L.ICENSE N0. NPF-39 LICENSEE: Philadelphia Electric Co.

2301 Market Street Philadelphia, Pennsylvania 19101 FACILITY:

Limerick Nuclear Generating Station EXAMINATION DATES: October 20-23, 1985 CHIEF EXAMINER:

I&t.o6, h M

//EY A. Howe, Reactor' En er Examiner date REVIEWED BY:

/b777 R. Keller, Chief, Projects Section 1C date APPROVED BY:

Mff)lfFyn if/

@7 S Aollins, 'Chputy Director,

' date Division of Reactor Projects SUMMARY: Operator License examinations were conducted at the Limerick Genera-ting Station during the week of October 20, 1986.

Examinations were adminis-tered to three (3) reactor operator candidates, four (4) senior reactor opera-l tor candidates (one (1) senior reactor operator candidate waived the written examination) and one (1) instructor certification candidate. All candidates passed the examinations, l

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8701210192 870114'

PDR ADOOK 05000352

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REPORT DETAILS TYPE OF EXAMS:

Replacement X

EXAM RESULTS:

R0 l

SR0 l

Inst. Cert. l l

Pass / Fail l

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I l Written Exam I 3/0

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1/0 l

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l 10ral Exam I

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1/0 l

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l l Simulator Examl 3/0

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1/0 l

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CHIEF EXAMINER AT SITE:

A. Howe OTHER EXAMINERS:

B. Keller

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B. Turner

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B. Raymond C. Marschall T. Lumb D. Jarrell (PNL)

B. Hajek (Consultant)

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1.

Summary of generic strengths or deficiencies noted on oral exams:

Weaknesses:

Some candidates had difficulty locating in plant indications and control room indications which were not simulated on the simulator.

Strengths:

Most candidates were familiar with the content of facility procedures and the procedures available to them.

SR0 candidates exhibited strengths in the use of TRIP's, E, and SE proce-dures.

2.

Summary of generic strengths or deficiencies noted from grading of written exams:

SRO Weaknesses

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Ability to predict ADS response when ADS blowdown is in progress and the logic reset pushbutton is depres-sed.

Knowledge that the Tech Spec suppression pool temperature limit considers ECCS NPSH in its bases.

Strengths

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Knowledge and understanding of HPCI, EHC, and SLC systems. Ability to predict level detector responses under various conditions.

Knowledge of precautions in GP-2 and GP-3.

Knowledge of Administrative proce-dures. Ability to use Technical Specifications.

ILO Weaknesses

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Knowledge of relationship of flow resistance to fluid velocity. Ability to use IRM response to predict reactor water level under accident conditions. Under-standing of how voids can contribute to high rod notch worth during a startup. Understanding of power response during a loss of feedwater transient. Abil-ity to predict RSCS rod blocks given a set of condi-tions. Knowledge that HP-103 required to extend daily dose limit.

Strengths

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Use of steam tables. Level detector response to malfunctions. Procedure for scram reset. Symptoms

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of ON-114, Loss of stator water cooling.

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3.

During the simulator scenarios, the examiners evaluated the candidates'

ability to satisfactorily implement the Emergency Operating Procedure (EOPs). During emergency evolutions they were familiar with their indivi-dual and team responsibilities, they were able to execute the E0Ps with the minimum shift staff identified in the facility Technical Specifica-tions, the candidates did not physically interfere with each other nor did they duplicate efforts (unless required), and they were able to transition from one E0P to another and to enter and exit as required while assuring all necessary precautions and steps were completed.

4.

Personnel Present at Exit Interview:

NRC Personnel A. Howe Chief Examiner S. Kucharski Resident Inspector NRC Contractor Personnel D. Jarrell (PNL)

Facility Personnel J. Doering Superintendent - Operations E. Firth Training Manager L. Hopkins Operations Engineer R. Helt Senior Instructor C. Enpriss Administrative Engineer 5.

Summary of NRC Comments made at exit interview:

The facility staff was thanked for its cooperation during the examination.

The written exam review was productive and some lesson plan changes were agreed to by the facility. The facility was informed where to send the written examination comments.

The oral / simulator examination generic strengths and weaknesses described in Section 1 were discussed.

The minor simulator problems which occurred during the examination such as the inability to retrieve a P-1 edit and Limericks's planned simulator upgrade were discussed.

l The annunciator response card system was discussed since an annunciator

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response procedure at a remote plant panel was found incomplete.

l The facility was notified that results of the examinations would be given in about 30 days.

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Summary of facility comments and commitments made at exit interview:

The facility stated _that the. simulator upgrade included plans to expand the computer, add more malfunctions, a hardware upgrade to resolve differ-ences between the plant and simulator and the addition of actual control room operator aids in the simulator.

The annunciator response card system was still under development with the cards in the control room nearly complete and those at the remote panels in progress. At this time, only a few control room annunciator cards h.d mandatory use requirements.

The INPO accreditation board was scheduled to review the licensed operator training program for approval the week of October 27, 1986.

The next operator license examinations were planned for November or December 1987.

The facility felt that the written examinations were comprehensive-and challenging and that the simulator examinations were fair with the excep-tion of one scenario which was felt to be beyond the design basis of the plant.

Attachments:

1.

Written Examination and Answer Key - R0

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Written Examination and Answer Key - SRO 3.

Facility Comments And NRC Resolutions of Comments on Written Examinations

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U.

S. NUCLEAR REGULATORY COMMISSION E

REACTOR OPERATOR LICENSE EKAMINATION FACILITY:

LIMERICK 1 REACTOR TYPE:

BWR-GE4 DATE ADMINISTERED: 86/10/20

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EXAMINER:

B.

K. HAJEK

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MAsTa CANDIDATE:

, + ;.7 INSTRUCTIONS TO CANDIDATE:

Une separate paper for the answers.. Write answers on.one' side only.

Staple question sheet on top of the answer sheets.

Points for each question are indicated in parentheses after the question.

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

Examination papers will be picked up six (6)

hours after the examination starts.

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% OF CATEGORY

% OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 25.00 25.00 1.

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER - AND -FLUID-FLOW 25.00 25.00 2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 25.00 25.00 3.

INSTRUMENTS AND CONTROLS 25.00 25.00 4.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 100.00 Totals Final Grade All work done on this examination is my own.

I have neither given nor received aid.

Candidate's Signature

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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS

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ring the administration of this examination the following rules apply:

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1.

Cheating on the examination means an automatic denial of your application and could result in more severe penalties.

2.

Restroom trips are to be limited and only one candidate at a time may leave.

You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.

3.

Use black ink or dark pencil oniv to facilitate legible reproductions.

4.

' Print your name in the blank provided on the cover sheet of the examination.

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5.

Fill in the date on the cover sheet of the examination (if necessary).

6.

Use only the paper provided for answers.

7.

Print your name in the upper right-hand corner of the first page of each section of the answer sheet.

8.

Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write oniv on one side of the paper, and write "Last Page" on the last answer sheet.

9.

Number each answer as to category and number, for example, 1.4, 6.3.

Skip-at least three lines between each. answer.

s.

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11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.

.22. Use abbreviations only if they are commonly used in facility literature.

13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.

14. Show all calculations, methods, or assumptions used to obtain an answer to mathematics 1 problems whether indicated in the question or not.

15. Partial credit may be given.

Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

16. If parts of the examination are not clear as to intent, ask questions of the examiner only.

17. You_must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination.

This must be done after the examination has been completed.

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18. Whsn ysu ccmpleta your cxSninstion, you ch=11:

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Assemble your examination as follows:

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(1)

Exam questions on top.

(2)

Exam aids - figures, tables, etc.

(3)

Answer pages including figures which are part of the answer, b.

Turn in your copy of the examination and all pages used to answer the examination questions.

'c.

Turn in all scrap paper and the balance of the paper that you did not use for answering the questions, d.

Leave the examination area, as defined by the examiner.

If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.

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PRINCIPLES OF NUCLEAR POWER PLANT OPERATIOU, PAGE

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THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

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i QUESTION 1.01 (2.50)

The Main Steam lines contain flow restrictors to limit the naximum flow rate in the event of a steam line break.

These flow restrictors are also used to measure i

the steam flow rates in the main steam lines.

a.

Explain how the flow restrictors are used to

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measure the steam flow rate.

(1.0)

b.

Explain why at normal full power operating conditions, the. flow restrictors do not adversely affee the steam flow, while under main steam lin nditions,.the flow rate is restricted to sati fy design objectives.

(2.5)

f QUESTION 1.02 (3.00)

The reactor has just scrammed from a long term full gock-gp L, power operation.

All rods did not fully insert.

Ten p d,._ J ws g g

g hours after the scram, it has been determined that the-reactor-is shutdown ~by two percent'dk/k.

- 4kh p^h4-a.

Explain any changes that may occur to the degree that the reactor is shutdown, and any adverse consequences for the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

(2.0)

b.

Would it be better to maintain the reactor hot during this 20 hour2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months period, or would it be better tn cooldown?

Explain your answer.

(1.0)

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QUESTION 1.03 (1.00)

When placing the Feedwater Heaters in service, you are cautioned that the

"C" string 3rd heater cascading drain

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line is physically higher than the heater itself.

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Since this is the case, what parameter changes (such as pressure, temperature, flow) must occur before the

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cascading drain system will operate properly during normal operations?

Briefly state why.

(1.0)

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TEST CROSS REFERENCE PAGE

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.OUESTION VALUE REFERENCE

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nUESTION VALUE REFERENCE

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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE

THE3!10 DYNAMICS, HEAT TRANSFER AND FLUID FLOW e

QUESTION 1.04 (2.00)

'The operator is directed to insert the movable nuclear instrumentation following a reactor scraa.

If a loss of coolant accident had occurred with a loss ~of normal water level indications, how would the IRMs respond to

assist in the determination of reactor water level?

Describe the detector response well below the surface, near the surface, and above the surface of the water.

(2.0)

QUESTION 1.05 (1.50)

For the fluid properties given below, state whether-you would use the Steam Tables or a Mollier Diagram to determine the requested parameter.

(Steam Tables are provided.)

a.

Given a pressure of 800 psia, and an enthalpy of 1150 Btu /lbm, find the percent moisture.

(0.5)

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Given a temperature of 336 degrees F, find the exact saturation pressure.

(0.5)

c.

Given temperature and specific enthalpy, find the specific volume.

(0.5)

I QUESTION 1.06 (2.50)

Corrosion products may become suspended in the reactor coolant system and cause operating problems, a.

List two undesirable characteristics of these corrosion products.

(1.0)

b.

List three different physical ways these accumulated corrosion products can become suspended in the reactor coolant.

(1.5)

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

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1.

THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

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QUESTION 1.07 (2.50)

General Plant Procedure GP-2, Appendix I, REACTOR START-UP AND HEAT-UP, cautions that " extremely short periods have been experienced during reactor startups at other facilities due to high' notch worths."

a-Explain under WHAT two conditions. withdrawal of-

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any control rod, and especially edge rods, may result in high notch worths.(other than first rod in a group), and (1.0)

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b.

Explain WHY these conditions result in high notch I

worths.

(1.5)

QUESTION 1.08 (1.00)

Hydrogen generation under accident conditions can result 9~,4c'

Whdd4 in a significant hazard at Limerick Generating Station, j(t e,,

What two conditions should be avoided to limit hydrogen (1.0);Wdbs*kN"#

I C*' F" generation from a chemical reaction with the fuel v0' 446 element cladding?

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QUESTION 1.09 (2.00)

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l Following an instantaneous loss of all feedwater, state HOW each of the following parameters will change prior i

to a reactor scram (Increase, Decrease, or Remain the l

Same), and WHAT would have directly caused the change.

l a.

Reactor Power (1.0)

b.

Turbino Control Valve Position (1.0)

QUESTION 1.10 (3.00)

l The reactor is operating at 50 percent power.

If HPCI I

initiates, what will be the INITIAL effects on power and

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pressure.

Explain why.

(3.0)

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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE

THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

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QUESTION 1.11 (2.50)

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

a.

What will the power indication be one minute later?

Show all your calculations.

(1.5)

b.

Explain why the power drops at a constant exponential rate starting shortly after the scram.'(1.0')

QUESTION 1.12 (1.50)

The Doppler effect introduces negative reactivity into the core as the fuel temperature increases.

State whether each of the following statements is either TRUE or FALSE.

If a statement is false, restate it correctly, a.

As resonance peaks in U-235 broaden, the Doppler effect adds additional negative reactivity to the core.

(0.5)

b.

Due to the self shielding effect, as the fuel temperature increases, neutrons with energies slightly off resonance will have a greater probability of being absorbed.

(0.5)

c.

The Doppler effect always increases as fuel temperature increases, but the Doppler coefficient becomes smaller as the fuel temperature increases. (0.5)

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(***** END OF CATEGORY 01 *****)

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PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE

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QUESTION 2.01 (2.00)

What are the eight possible steam loads on the Main Steam System.

(2.0)

QUESTION 2.02 (2.00)

Limerick uses a Control Cell Core (CCC) to simplify core operation, improve reliability,,and to obtain better capacity factors, a.

In a CCC, where are (1) the control cell and (2) the non-control cell rods normally positioned (fully inserted, partially inserted, fully withdrawn)?

(1.0)

b.

What is the nominal (relative) enrichment of the fuel in the control cell bundles when compared to

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other bundles, and what advantage does this

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have on changes in core parameters when a control rod is moved?

(1.0)

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QUESTION 2.03 (2.50)

i Two vacuum breakers are located in each SRV discharge line.

a.

What could happen if both vacuum breakers fail to

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j open after an SRV lifts, and what two adverse tp effects could this failure have if the FRV lifted l

again?

(2.0)

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b.

What is the setpoint at which the vacuum breakers

open?

(0.5)

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'2-PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE-

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i QUESTION 2.04 (2.50)

The HPCI system is designed to provide a source of water in the case of a small line break.

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List the'three conditions that will initiate HPCI. (1.5)

b.

Upon a HPCI initiation, what controls the turbine acceleration rate?

( 0. 5 ) -

c.

~At rated flow, what controls the turbine speed?

(0.5)

QUESTION 2.05 (3.00)

Normal operation of an Emergencv Diesel Generator requires proper operation of svveral auxiliary syntems.

Explain what would happen (Diesel would start, fail to start, diesel would stop,' diesel would slow down, diesel

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would continue to operate with no change in output,

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etc.) for each of the f(11owing component failures.

Also explain WHY the action you describe would occur.

a.

The electric governor fails during normal

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operation.

(0.75) q4g4 b

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The engine driven fuel oil pump fails to deliver fuel to the diesel due to a gear drive failure during normal operation under full load.

(0.75)

c.

One of the two air start solenoids fails to operate when an emergency start signal is received.

(0.75)

d.

A short has occurred in the control circuitry for the AC powered pre-lube pump which supplies oil to the bearings.

An emergency start signal is received.

(0.75)

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

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PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE

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QUESTION 2.06 (2.50)

The CRD hydraulic system supplies about 3 gpm of purge water to the reactor recirc pump seals, a.

What is the purpose of the seal purge ~ water?

(0.5)

b.

If 0.8 gpm flows by the #1 seal and out the staging line to the DWEDT, where does the rest of the water flow?

(0.5)

c.

You suspect that the #1 seal has failed.

What two indications would you use to confirm your suspicion?

(1.5)

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QUESTION 2.07 (3.00)

When a scram signal occurs at power, describe IN DETAIL how the Control Rod Drive and its associated Hydraulic Control Unit function to insert the control rod.

As a MINIMUM, your answer should include whicn components open, close, energizo, deenergize, what supplies the motive force for the ENTIRE rod travel, and major flowpaths.

LIMIT YOUR ANSWER TO THE CRDM AND THE HCU.

(3.0)

QUESTION 2.08 (2.00)

The Water Fire Protection System contains two water supply pumps, a.

What is the motive force for each pump, and what signals will cause each pump to auto start?

(1.5)

b.

What system do these pumps get their suction from? (0.5)

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CATEGORY O2 CONTINUED ON NEXT PAGE *****)

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PLANT' DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE

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QUESTION 2.09 (2.00)

The DC Distribution system provides power to vital loads.

Normally, power is supplied from the battery chargers, a.

When would power be supplied from the batteries?

Give two occurances.

(1.0)

ld"$., p, b.

What is the minimum time the batteries are designed to provide power to the vital loads assuming they M.g had been fully charged?

(0.5)

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When placing a battery charger.in service, why is it necessary to close the output breaker before closing the input breaker?

(0.5)

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QUESTION 2.10 (2.00)

Consider the Core Spray System as being in a normal lineup during full power reactor operation.

a.

Can you manually open the inboard isolation valve (F005)?

Explain your answer.

(1.0)

b.

If you were performing a full flow test with full

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flow established, and an instrument technician l

inadvertently caused a false Core Spray initiation

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signal, what valve repositioning would occur?

What valve repositioning would not occur?

Explain your

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(1.0)

answer.

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QUESTION 2.11 (1.50)

The Standby Liquid Control System provides an emergency source of negative reactivity for the core.

It is interlocked with or supported by several auxiliary systems.

a.

What is the purpose of the interlock with the Reactor Water Cleanup System?

(0.5)

b.

What two SLC functions will be lost if the Instrument Air supply to SLC is lost.

(1.0)

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END OF CATEGORY 02 *****)

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'3-INSTRUMENTS AND CONTROLS PAGE

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-3 QUESTION 3.01 (1.00)

During a reactor startup, when H. P.

Turbine shell

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warming is being performed at the same time that reactor pressure is being increased, caution must be exercised toLassure that the first stage turbine pressure does not i

exceed 100 psig.

If this pressure is exceeded, how and why could this adversely affect the reactor startup?

(1.0)

QUESTION 3.02 (2.50)

With the reactor operating at 100 percent under steady state conditions in three-element level control, an instrument technician mistakenly isolates and equalizes the pressure across one of the Main Steam Line flow transmitters.

Describe the response of the.Feedwater Control System until steady state conditions are again established.

(2.5)

i QUESTION 3.03 (2.50)

Concerning the recirculation pump RPT breakers, a.

What power is interrupted when the breakers trip, and how does this reduce the transient from an EOC

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turbine trip?

(1.0)

b.

List three automatic trips of these breakers.

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Include the appropriate setpoints.

(1.5)

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

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'3 INSTRUMENTS AND CONTROLS PAGE

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A reactor startup is in progress and the RSCS is enforcing.

Groups 1 and 2 have been pulled to notch 48.

Group 3 rods are being withdrawn.

a.

If all Group 3 rods have been pulled to notch 4, and one Group 3 rod has been pulled to notch 6, how far can the next rod be withdrawn?

Explain your answer.

(1.0)

b.

If all Group 3 rods have been pulled to notch 4,

and one Group 3 rod has been pulled to notch 6, if the reactor power is increasing. faster than A

desired, which rod (s) can be inserted?

(0.5)

c.

If all Group 3 rods have been pulled to notch 20, and one Group 3 rod has been pulled to notch 28, how far can the next rod be pulled?

Can the rod at notch 28 be pulled out any further?' Explain your answer.

(1.0)

l QUESTION 3.05 (2.50)

"

Three runbacks are associated with the EHC Logic System.

'

For the Load Reject Runback, state a.

What will initiate it, (1.0)

b.

Its effect on turbine operation, and (1.0)

c.

Whether a reactor scram will occur.

(0,5)

QUESTION 3.06 (2.00)

The Rod Block Monitor receives inputs from LPRMs, APRMs,

'

and the RMCS.

For selection of a centrally located control rod, and for 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, (1.2)

b.

The Recirc Flow Trip Reference Unit, and.

(0.5)

c.

When an LPRM signal is automatically bypassed?

(0.3)

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

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- -. -. - -... -

. - -.

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'

'3 INSTRUMENTS AND CONTROLS PAGE

,- QUESTION 3.07 (2.50)

For each of the following operating conditions, faults, failures, or malfunctions affecting the reactor level measurement or transmitter components, state whether the INDICATED level will increase, decrease, or remain the same.

a.

A leak occurs in the reference leg.

(0.5)

b.

A break occurs in the variable leg.

(0.5)

c.

An equalizing valve opens.

(0.5)

d.

A hole develops in the diaphragm.

(0.5)

e.

Drywell temperatures are elevated during a LOCA.

(0.5)

QUESTION 3.08 (2.50)

The Containment Instrument Gas System provides a source of clean, dry nitrogen to pneumatic devices inside the Primary Containment, including the ADS valves.

<

a.

What controls the loading of the compressors for

!

this system?

Include in your answer what pressures are sensed, and how the lead / lag arrangement of the compressors is setup.

(1.5)

b.

If-the compressors fail to load properly, and header pressure drops, what automatic and/or manual actions assure sufficient pressure to the ADS valves?

(1.0)

QUESTION 3.09 (1.50)

If the radiation level in the Reactor Enclosure Exhaust rose to 4 mr/hr, what three automatic actions would occur to limit radioactive release to the environment?

(1.5)

l l

QUESTION 3.10 (3.00)

List the rod withdrawal blocks and setpoints for the IRM system.

State the conditions for when the rod blocks are bypassed.

(3.0)

!

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

l I

- _...., _ _ -.

_.__m

..,o

_,,.__._ _.,.._

. _.

.

'3.

INSTRUMENTS AND CONTROLS PAGE

,

o-QUESTION 3.11 (1.50)

The RECW Emergency Operation Shutoff Valve (HV-102)

isolates the RWCU heat exchangers and the RWCU pump seal coolers.

If.this valve was inadvertently closed, how

-

would the RWCU system automatically respond (three actions required)?

Assume no operator action.

(1.5)

.

QUESTION 3.12 (1.00)

During a reactor startup, you withdraw a control rod and notice a short period.

Before the timer function times out, how can you immediately insert the rod?

(Excluding scram.)

(1.0)

)

!

-

l l

i (***** END OF CATEGORY 03 *****)

. -.

.

..

. -

.

-

.

-.

..,. -. - -, - -

-, -

-.__,. - - -.

-. _ _. -

..

C

'4.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

,

RADIOLOGICAL CONTROL

,-

QUESTION 4.01 (3.00)

According to E-10/20, Loss of Offsite Power, what initial actions are to be taken or verified to assure proper coolant flow to plant systems.

Assume that the diesel generators have started AND THAT SAFEGUARD POWER IS AVAILTtBLE, and that the Instrument. Air Compressors

-

have appropriately started.

List only required system lineups and startups, such as verify flow to.

or

..,

start pump in.

(3.0)

..

QUESTION 4.02 (2.00)

What four items should you check to assure that a procedure is a valid procedure before using it to operate a plant system?

(2.0)

w QUESTION 4.03 (1.50)

A~ fire ~has erupted in the main Control Room, and it has become uninhabitable.

According to SE-1, Remote Shutdown, what are the Immediate Operator Actions that must be taken prior to leaving the Control Room?

(1.6)

QUESTION. 4.04 (3.00)

What are the LIMERICK STATION whole body Radiation Exposure Guides for a licensed Reactor Operator with a completed NRC Form 4 on file?

Include quarterly, daily, and annual guidelines.

Also include any special requirements for exceeding these guidelines.

(3.0)

QUESTION 4.05 (2.00)

What are the entry conditions for T-102, Containment

.

Control?

(2.0)

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

-.

_

-

_ - _. -. _

_ _ _ _ _.. _. _ _ _ _ _ _ _

..... -

__

. _, -. - _

".4.

PROCEDURES - NORMAL, ABNORMAr., EMERGENCY AND PAGE

RADIOLOGICAL CONTROL

.,--

hUESTION 4.06 (3.00)

According to ON-103, Control of Sustained Combustion in the Offgas System, a.

What are the symptoms that require entry into this procedure?

(2.0)

b.

Why does the procedure recommend a power reduction?

(1.0)

y QUESTION 4.07 (1.00)

One of the Immediate Operator Action steps in OT-100, Reactor Low Level, is to run back recirc flow.

Why will this action help to correct the low level condition?

(1.0)

QUESTION 4.08 (2.00)

According to GP-11, Reactor Protection System - Scram Reset, the first step is to place the DISCH VOL HIGH

WATER LVL BYP Keylock Switch to the " Bypass" position.

However, this action may NOT always be appropriate as the first step.

Under what conditions should other action be taken first, why, and what is this other action?

(2.0)

l QUESTION 4.09 (1.50)

What symptoms require entry into ON-114, Loss of Stator Water Cooling Runback?

(1.5)

QUESTION 4.10 (2.00)

The two Immediate Action Steps (Other than if a scram l

occurs.) in OT-101, High Drywell Pressure, are to L

maximize drywell cooling, and to terminate drywell inerting.

Describe specifically how you would accomplish and assure these two tasks.

(2.0)

l l

!

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

I

,

+.

--

-,

,.

, -.

.-.

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

.j.

.

'4.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

'

RADIOLOGICAL CONTROL

. - -

i QUESTION 4.11 (1.50)

According to A-7, Shift Operations, Section 5.5, Controlling Access to the Control Room and Control Areas, a.

When is a verbal request required for crossing-the

" red" line in the Control Room?

-(0.5)

b.

Who is authorized to order the immediate departure or removal from the control room of all people not required for safe plant operation when a plant emergency exists?

(1.0)

QUESTION 4.12 (2.50)

The first indication of a problem in the Scram Discharge Volume (SDV) is likely to be the "SDV NOT DRAINED" alarm.

This requires entry into OT-105, SDV High Level.

a.

What immediate actions are required?

(1.0)

b.

What automatic actions should be verified?

Include the setpoints.

(1.5)

(***** END OF CATEGORY 04 *****)

(************* END OF EXAMINATION ***************)

.

-

--... -.

--.

- ---...

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-

..

1 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE

THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

_'

) ANSWERS -- LIMERICK 1-86/10/20-B.

K.

HAJEK ANSWER 1.01 (2.50)

giqWik Ey a.

Pressure taps are located at the point of minimum

/

s s

bT %.+ Il - b y+*g*-

pipe diameter and at an upstream location of

)

Y/

greater diameter or convergence [0.5].

The differential pressure between these two points

,,4

g g,, pg y, is related to flow rate [0.;5].

., AS A4 di

(gg

b.

As the flow rateaincreases through the venturi, the vg4 4 y 4 g ut resistence to flow increases.

This resistence to

- - f'

"

flow is proportional to the square of the flow

'

rate [0.5],

Thus at normal full power flow, the resistence to flow is not too great [0.5] [about 9 psid], while under accident conditions, the resistence to flow is high, limiting the max flow to 200 percent of the design flow rate [0.5].

REFERENCE LOT-1280, ppg 4 - 5.

Objective 1.

LOT-0120, ppg.

L, 12.

Objective 2.d.

ANSWER 1.02 (3.00)

a.

Since the time at which the shutdown reactivity was measured was near peak xenon [0.5], the reactivity of the core will increase over the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months as the xenon decays. [0.5]

Since peak xenon

.

,

reactivity is greater than two percent, the reactor will restart without further operator action. [1.0]

I b.

Since the moderator temperature coefficient is negative (0.5], it would be better to maintain the reactor hot during this period. [0.5]

REFERENCE LOT-1450, objectives 1,

3.

LOT-1510, objectives 5, 6.

ANSWER 1.03 (1.00)

l The heater pressure must increase (0.25] so it is sufficient to j

overcome the drain line head pressure. [0.75]

Sern m *k5 Y O c & VMvv m a>4 & dre O Y A'

.

-

.

-

.

-

.

...

_

'~

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE

THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

) ANSWERS -

LIMERICK 1-86/10/20-B. K. HAJEK REFERENCE SO2.1.A, Precaution 7.2 LOT-1270, Objectives 2, 3,

7.

.

j-ANSWER 1.04

.(2.00)

.

Expect normal detector response below the surface of the water (0.5].

As level approaches the IRM,. boiling around the IRM will result in an increase in the. thermal rudTg5 WM s

diffusion length', neutrons will travel farther, and the g'

,4,'c gg,.

detector output will increase. [0.75] [This detector

output may be confused with a core recriticality event.]

Q44 4aohdL-As the level continues to drop below the IRM, decreased M;ofrud,g thermalization will result in a rapid decrease in IRM u

output (0.75]

-

belta5(A A 3oo

)

REFERENCE d*4 @

LOT-1692, ppg. 16 -17.

Objective de.

GE Mit7gd b % 6nt. b c Ed /7Q E-3 - T-Y ANSWER.

1.05 (1.50)

a.

Mollier Diagram Steam Tables cq, Moth Kp (owe a bon)

b.

c.

Steam Tables

,

REFERENCE

'

LOT-1150, Objective 1.

LOT-1160, Objective 1.

.

f i

)

. _ _

~ _.,

.

_ _._

..

.

.m

,

..

.

...

..

.

...

-

.

1.

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE

THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

~

) ANSWERS -- LIMERICK 1-86/10/20-B. K. HAJEK ANSWER 1.06 (2.50)

a.

1.

Can be transported throughout the coolant system.

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.

1.

Chemical shock such as increasing the oxygen level or decreasing pH 2.

Thermal shock such as heatup or cooldown 3.

Mechanical shock such as a reactor trip or recirc pump start or stop Credit for the shock or the example, as long as have at least one from each of the three groups.

(Note that the only acceptable answers in plant quiz were the three shocks.)

REFERENCE LOT-1010, ppg. 10 - 12.

Objectives 1,

2, 4,

& 5.

Week 11 quiz.

.

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

. - - - - - -, - - - - -. - -

- - - - - - -

- - - - -

.

'

'1 FRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE

THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

~

ANSWERS -- LIMERICK 1-86/10/20-B. K. HAJEK ANSWER 1.07 (2.50)

a.

1.

Peak xenon 2.

No voids in the core b.

Control rod worth is a function of the square of the local flux [0.25].

During normal reactor operation, the highest reactor flux is in the center of the core [0.25].

Following a scraa, the xenon will peak where the flux was highest [0.25].

Then, during a subsequent startup, since..the xenon tends to suppress the flux, the flux will be pushed to the previously low flux regions (0.25].

This effect will also occur relative to the voids, since during normal operations, the voids tend to push the flux to the bottom of the core [0.25].

With no voids present, the flux may be higher than expected at the top of the core [0.25].

Key points to be covered:

1.

Effect of flux on reactivity.

2.

Effect of flux on xenon and xenon on flux.

3.

Effect of voids on flux.

REFERENCE General Plant Procedure GP-2, Appendix I LOT-1510, Objectives 1,

2, 4.

LOT-1460, Objectives 1,

3.

LOT-1490, Objectives 3, 4,

6, 7,

9.

ANSWER 1.08 (1.00)

High temperatures [above 1600 degrees F]

"Ilt 40 cm Availability of steam Q4 g REFERENCE LOT-1693, ppg.

2, 3,

6.

Objectives 1,

2, 4.

l

.

l l

l

-- -

-

--

.

-. -. -.

'

.

1-PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE

THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

.'

ANSWERS -- LIMERICK 1-86/10/20-B. K. HAJEK ANSWER 1.09 (2.00)

a.

Power will decrease [0.5) due to a loss of core inlet subcooling [0.5].

b.

The control valves will close down [0.5) due to the resultant pressure decrease (0.5].

-

REFERENCE LOT-1691, ppg. 1 - 2.

Objectives 1 and 3a.

ANSWER 1.10 (3.00)

qg y)lll pvidt, DT-10'I Check at Facility for answer verification.

j Pressure will decrease [0.5) due to the cold spray FSA A, condensing the steam above the reactor core. [1.0)

hiseund Power will decrease (0.5) due to the increase in voids (Df*L W d"

.

caused by the lower pressure. [1.0)

JWA@

s ($> pau4-t 4 du 4o REFERENCE HPts G+akg ik Nw-LOT-1170, Objectives 1,

3.

LOT-1460, Objectives ~1, 3.

LOT-0340, Objectives 4, 5.

ANSWER 1.11 (2.50)

a.

P = P(orig) e**t/ tau (0,5)

= 75 * e**60/(-80)

(0.5)

= 75 * O.472 = 35.42 on Range 4.

(0.5)

b.

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

(0.5)

The short lived precursers decay in about two minutes, [so it is appropriate to use only this single value for this problem.]

(0.5)

REFERENCE LOT-1430, ppg. 14 - 15.

Objective 4.

,

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2 -

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE

,

..

ANSWERS -- LIMERICK 1-86/10/20-B.

K.

HAJEK ANSWER 1.12 (1.50)

a.

False, in U-238 broaden,

...

b.

True c.

True REFERENCE LOT-1470, pg.

2.

LOT-1480, ppg. 7 - 9.

Objectives 1, 2..

ANSWER 2.01 (2.00)

1.

Main turbine Al'*

S k VIVS 2.

Steam jet air ejectors AlkfVivs 3.

,HPCI turbine

g 4 g. "I#y,9 4.

RCIC turbine 5.

Steam seals 6.

Reacto.- feed pump turbines 7.

Off Gas recombiner 8.

Hotwell spargers REFERENCE 4. 6Objective 4.

LOT-0120, pg.

ANSWER 2.02 (2.00)

a.

Non-control cell rods are normally positioned full

,

out.

(0.5)

Control cell rods are the only rods normally inserted in the core.

(0.5)

!

I b.

The relative enrichment of the fuel in a control cell is low relative to the other fuel bundles in the core.

(0.5)

This has the advantage of decreasing the worth of these control rods, and thus decreasing the effect

.

their movement has on core paramenters.

(0.5)

r REFERENCE LOT-0020, ppg. 30 - 34.

Objective 8.

.

_

-. _. _

_

.

'2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE

-

ANSWERS -- LIMERICK 1-86/10/20-B. K. HAJEK

,_

.

ANSWER 2.03 (2.50)

a.

Water could be drawn up into the discharge pipe

[1.0],

and thus cause a higher pressure at the SRV outlet which would result in a lower flow rate

[0.5],

and cause excessive water hammer which would result i{pipedam (0.5 6 A.eJ/

b.

They open when drywell pressure exceeds SRV discharge pipe pressure by 0.5 lbs (psid).

REFERENCE LOT-0330, pg.

6.

Objective 2, 4.

ANSWER 2.04 (2.50)

a.

1.

Low level [-38

"],

OR 2.

High drywell pressure [1.68 psig], OR 3.

Manual.

0.5 points each b.

The ramp generator c.

The flow controller

REFERENCE LOT-0340, ppg. 20 - 21.

Objectives 5, 6,

9.

,

l l

_

_ _ _ _ _ _ _. _ _ _..

_ _. _ _. _ _ _

_ _ _ _ _ _. _ _ _ _ _, _ _. _ _ _

.

,.

.2 -

BLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE

.

-

ANSWERS -- LIMERICK 1-86/10/20-B. K. HAJEK

!

.

ANSWER 2.05 (3.00)

a.

The mechanical governor will take over fev64 and the DG will run at the speed setting of the mechanical governor, b.

The DG will continue to run because the backup DC fuel oil pump will auto start on low pressure [10 psig decreasing).

c.

Diesel would start because only one of the two air start system 4 s required.

d.

Engine would start normally because the AC powered pre-Jube pump doesn't operate during an emergency start anyway.

It is operated only prior to manual starts. (The DG is pre-lubed for three minutes after the manual start switch is taken to start.]'

O.25 for each operation, 0.50 for each reason REFERENCE LOT-0670, ppg.

5-9.

Objectives 3, 4.

d

~ sgh g' ap07 ANSWER 2.06 (2.50)

To prolong seal life by providing relatively cool' M I F Y a.

water to the seal cavities.

L jp^cric!f M

-

auc 1p M A

,

b.

To the reactor through the breakdown bushing.

,

c.

The #2 (or upper) pressure increasing to the same as the #1 pressure (or 1000 psig] [0.75], and a i

j flow switch alarm (at 0.9 gpm] should come in.

[0.75]

'

REFERENCE

.

LOT-0030, ppg. 6 and 28.

Objectives 2, 4,

5.c, 6.

.

_. _ _ _ _. - _

. -..

. - _ _ _ _.__

-

_ _ - _ _ -..

.-- _

.

.2 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 253

__

ANSWERS -- LIMERICK 1-86/10/20-B. K. HAJEK

.

ANSWER 2.07 (3.00)

A scram signal deenergizes the scram pilot valves (0.5),

venting air from the scram inlet and outlet valves, allowing them to open (0.5).

This vents water from the overpiston area of the CRD to the SDV (0.5) and applies HCU accumulator water to the underpiston area of the CRD (0.5).

This dp provides the initial motive force for the rod (0.5).

As accumulator pressure drops below reactor pressure, a ball check valve in the CRD opens to apply reactor pressure to the CRD to complete the scram stroke (0.5).

REFERENCE LOT-0070, pg. 29.

Objectives 3, 8.

ANSWER 2.08 (2.00)

a.

The electric pump [0.5] [is the main pump].

It will start on low pressure in the fire main [at 100 psig]. [0.25]

The diesel pump [0.5] [is the backup pump].

It will start at lower pressure [95 psig) in the

. fire. main. [0.25].

,

b.

The [14 inch] suction connects to the circ water NHA'-'ibN

'

line upstream of the LP condenser.14 is 4@f A WcAg c.colI b

$N mt

$

> f4, REFERENCE

'

LOT-0733, ppg. 10 - 11.

Objectives 3, 4,

5.

l ANSWER 2.09 (2.00)

a.

1.

Loss of AC power to the chargers.

2.

Failure of the chargers.

l

!

b.

Four hours.

c.

If the AC input breaker is closed first, the battery charger will trip on High output Voltage.

REFERENCE LOT-0690, pg. 14, Quiz - Week 15, Q# 13.

Objectives 2, 3,

5.

l

.

-

-n---

n

,,

,

,,

. - - -, _

-,-c-

-,, - - - - - - -

.

.2 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 25b

.

-

ANSWERS -- LIMERICK 1-86/10/20-B. K. HAJEK

,

.

ANSWER 2.10 (2.00)

a.

No. [0.25]

The inboard and outboard isolation valves are interlocked so only one can be open with no injection P-s1 present. [0.75]

b.

The fu* 1 flow test valve will auto close on receipt of the, initiation signal. [0.5)

.

The inbaard isolation will not open because reactor pressuru is not below 455 psig.

g REFERENCE LOT-0350, ppt;. 10 - 11.

Objectives 8, 9.

ANSWER 2.11 (1.50)

a.

To prevent removal of boron injected into the vessel, b.

1.

Air mixing during chemical addition 2.

Tank level indication REFERENCE

-

LOT-0310, ppg.

5, 7,

12.

Objectives 2, 0,

11.

!

l

l

.

-.

.,

-..

-. - -

-

. - -

. - -... -.

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

- -_

_ __,.

.

~*

.3 INSTRUMENTS AND CONTROLS PAGE

.

-. ANSWERS -- LIMERICK 1-86/10/20-B. K. HAJEK

-.

.

l ANSWER 3.01 (1.00)

A reactor scram could occur (0.25)

due to the stop valves being closed at < 30 percent power as indicated by first stage turbine pressure (0.75)

'[setpoint at 114 psig).

REFERENCE S01.1.A, pg. 15.

,

>2 LOT-0300,.pg.

9.

Objectives 7 and 8.

ANSWER 3.02 (2.50)

Total steam flow would indicate 75 percent while actual 3b4 La **kASI'd Etutpedb wd steam flow would remain at 100 percent (0.5].

guired. g The FWL control system would reduce FW according to the 757a moi 4*oE change in indicated steam flow [0.5).

t.

Level will begin to decrease, and a level error signal will be generated [0.5).

Feed flow will increar.e back to the 100 percent level to bring level back near normal [0.5).

'

Final level will uti$-be established slightly lower than the original level [0.5]. [ Level will be such that the level error is equal to but opposite the flow error.)

REFERENCE LOT-0550, pg. 17.

Objectives 2, 4,

7a.

l l

l

., - -.

.

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

- - -. -.

-

.,

-

- - -..

-. - - -,.

.

r

INSTRUMENTS AND CONTROLS PAGE

.

-ANSWERS -- LIMERICK 1-86/10/20-B. K. HAJEK

.

ANSWER 3.03 (2.50)

a.-

Power is interrupted between the MG set generator and the recirc pump motor.

(0.5)

This limits the coastdown time of the pump [to

.

minimize a reactivity transient on a turbine trip at EOC).

(0.5)

I b.

1.

Main turbine trip at greater than 30 % power

[as sensed at turbine first stage)~ [ Turbine trip is sensed by TCV fast closure at 500 psig EHC pressure or turbine stop valve closure at 5% from full open).

2.

Phase overcurrent

?.

Ground fault 4.

[RRCS signal at)

-38" level with a 10 see time delay OR 5.

[RRCS signal at] 1093 psig RPV pressure.

0.5 each for any three of the above.

REFERENCE LOT-0040, pg. 22.

Objectives 2.f and 6.

ANSWER 3.04 (2.50)

a.

The next rod can be pulled only to notch 8.

[0.25)

When below position 12, Group Notch Restraint is enforced [0.25), and all rods in a group may be withdrawn

one notch at a time only as far as the next Group

'

Bank Position. [0.5)

b.

Only the rod at notch 6.

c.

The next rod, or the rod at notch 28, can be pulled to notch 48. [0.5)

Above notch 12, Group Notch Restraint is not enforced. [0.5)

REFERENCE LOT-0100, ppg. 13 - 16.

Objectives 2, 6.

W red cam,G[b QA tw gwedim is h.4 pen ble d u h A M %kthaae..

,

I

/

. -

- -

-

--

- -... -. - -

-..

. -,- --

.- -.

.. - _. -

.

.3.

INSTRUMENTS AND CONTROLS PAGE

.-

.

-

ANSWERS -- LIMERICK 1-86/10/20-B.

K.

HAJEK

.

ANSWER 3.05 (2.50)

, ggWJ y a ' +uN.% ~

a.

Will be initiated if a load mismatch [0.3] of 40 ga percent (0.3] occurs between the generator stator i

amps [0;2] and cross around steam. [0.2]

b.

A runback to zero occurs until the condition clears

-[0.3],

the input to the load set summer decreases to zero [0.3], and you get a fast closure of the CVs (0.4]. Q g 9gg c.

The reactor will scram if the power is greater than 30 percent.

[What is the source of this scram?]

S}bb{f 94 W.o&JnL 4 V}Gd &

REFERENCE m h g, LOT-0590, PG. 38.

Objective 5.

ANSWER 3.06 (2.00)

a.

The RBM receives inputs from the four LPRM strings (Levels A and C) around the selected control rod.

(0.2)

It also receives a core average input from APRM C

[ Alternate E].

(0.2)

If the LPRM average is less than the APRM signal, the LPRM signal is adjusted to equal the APRM signal.

(0.4)

,

If the LPRM average is equal to or greater than the

!

APRM signal, the LPRM average is passed.

(0.4)

l l

b.

RBM A receives inputs from Flow Units A and C.

(0.2)

It passes only the signal of lowest value.

(0.3)

[It trips at 0.66W + 24, 32, 40, and 51.]

i)When the flux signal decreases below 3 percent.-Ef, M N c.

)

ils down%6lC.

REFERENCE LOT-0280, ppg. 5 - 6.

Objective 2, 5,

6.

'

pg,17 FAS

-

.

l

,

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

_

_~-

-

_

_. _ _..... _, _,...

-m

. - - _ _.

_

..

e 3-INSTRUMENTS AND CONTROLS PAGE

.

-

ANSWERS -- LIMERICK 1-86/10/20-B. K. HAJEK

.

ANSWER 3.07 (2.50)

a.

Increases b.

Decreases c.

Increases efI d ih w.g %

Y pote, M getM M 2. d e'apl,9(,re&,

..

d.

Increases f $ & !%;: s CsAl } kv<.

W e.

Increases REFERENCE LOT-0050, ppg. 16 - 17.

Objective 7.

ANSWER 3.08 (2.50)

a.

Both compressor switches are placed in AUTO. [0.3]

Pressure is sensed on the receiver tank [0.3], and the lead / lag arrangement is set by manual adjustment of pressure switches mounted on the tem receiver [0.3].

The lead compressor cycles on first (between 100 - 110 psig) [0.3], and the lag compressor cycles on if pressure drops further [to 94 - 108 psig) [0.3].

}Tu re e ed M ^^L-ADS b.

If pressure drops further [to 90 psig],

1 g g

automatic solenoid valves position to isolate the compressor discharge to the ADS header (SV-59-g pdn'd 150A/B] [0.3], and open to allow the nitrogen g gh bottles to maintain the header pressure [SV-59-152 * *gy *

A/B] [0.3).

These same actions may be taken

>

manually from the control room [on panel C626]

[0.4].

N REFERENCE

"h*

LOT-07307 ppg.

21 - 23.

Objectives 11, 12, 13.

Trovlcrw M

[fct u m nl @ oar dise d 6 A M L f.

t yANSWER 3.09 (1.50)

1.

Supply and exhaust dampers would close, f2X Erdos" I

'

2.

Standby Gas Treatment would initiate 3.

Reactor Enclosure recirc would start telv5)) p vit)e.gjhaAL}})kt.La Y k U*S W W

__

-- -. -

-

.

__

.

-.

.

3-INSTRUMENTS AND CONTROLS PAGE

r

.

_

ANSWERS -- LIMERICK 1-86/10/20-B. K.

HAJEK

.

REFERENCE LOT-0720, pg. 12.

Objective 2.

ANSWER 3.10 (3.00)

1.

Upscale 108/125 of scale 2.

INOP Low detector voltage Internal module unplugged Chan mode sw not in operate 3.

Downscale 5/125 of scale

-

4.

Detector not fully inserted 0.25 per above item (total of 9 for 2.25 pts)

Bypassed if a.

Associated channel is bypassed [0.25]

b.

Reactor mode switch is in RUN [0.25)

c.

Downscale is bypassed if in Range 1 (0.25]

REFERENCE LOT-0250, pg. 10.

Objective 10.

' ANSWER 3.11 (1.50)

[ System temperature goes up, and then)

1.

System isolates 2.

Circulating pumps trip 3.

Holding pump starts.

REFERENCE LOT-0460, ppg. 11.

Objective 9.

ANSWER 3.12 (1.00)

l l

The CONTINUOUS INSERT button can be depressed to l

accomplish this action.

REFERENCE LOT-0080, pg. 17.

Objectives 4.

5, 6.

...

_ _ _.

-

.. --.

-

-.

. -

,

.

. _,.

.

<4.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

RADIOLOGICAL CONTROL

, ' '

ANSWERS -- LIMERICK 1-86/10/20-B. K. HAJEK

-

.

ANSWEE, 4.01 (3.00)

J'5 y@ N[f Af

.

dk 1.

Verify all four ESW pumps start and recover any

,

pump that did not start.

t

}'

2.

Place both Safeguard Piping Fill, Pumps.in service.

,3.

Establish ESW flow to TECW heat exchangers.

4.

Establish ESW flow to RECW heat exchangers.

5.

Supply RECW to the Drywell Coolers.

6.

Start at least one CRD pump.

0.5 for each item.

REFERENCE LOT-1566. Objective 2.

Plant Procedure E-10/20, ppg. 1 - 2.

ANSWER 4.02 (2.00)

1.

Must be stamped in red " Controlled Copy".

2.

Signature and date from a.

Station Superintendent or alternate

[b.

QA Superintendent or alternate 4

-/2 g Eired h d W J*

3.

Date of use must be later than effective date N'

0.5 per item REFERENCE LOT-1570, ppg. 5 - 6.

Objective 1.

Plant Procedure: None available.

ANSWER 4.03 (1.50)

1.

Scram the reactor 2.

Trip the main turbine 3.

Close the MSIVs REFERENCE j

LOT-1563. Objective 2.

Plant Procedure SE-1, 1s.

1.

'

l l

,

. -. _. _ _. _

-

,

- - -. -

.

a 4.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

RADIOLOGICAL CONTROL

.

'

ANSWERS -- LIMERICK 1-86/10/20-B. K. HAJEK

'

ANSWER 4.04 (3.00)

Quarterly 2.5 rem /qtr if record is current and on file (0.5)

. nn

.,_ine,

<r

,,+a-

>

...

- --. _ - - -.. - - >-

g ~9'Q"y"T. ','I

' " ' ' ' ' ~ ~ ~ ~

~" L a, :__

_

.-

...,

-. - - -.

(Lifetime must remain below 5(N-18)]

Daily too meeWd Q td Wg h (900 mrtm (o,T)

,

300 mrem /dahb ir(qtrfhbalanceisgreaterthan400

-

mrem.

(0.5)

Requires a Dose Extension [per HP-103] to exceed.

(0.5)

Annual 4500 mrem (0.5)

Requires plant superintendent's approval to exceed.(0,5)

REFERENCE LOT-1760. Objective 1, 2.

Plant Procedure HP-102, ppg. 2 - 3.

,

Commt.wt M gMSWrw.

ANSWER 4.05 (2.00)

1.

Suppression Pool Temperature above 95 degrees F.

' to 24.25 8 2.

SP Level outside 22 3.

D/W Pressure above 1.68 psig.

4.

D/W Temperature above 135 degrees F.

REFERENCE LOT-1560. Objective 3c.

!

Plant Procedure T-102 Flowchart.

ANSWER 4.06 (3.00)

a.

1.

Sudden increase in Offgas System pressures.

2.

Sudden change in Offgas System flows.

3.

Sudden increase in Offgas System hydrogen concentration.

4.

Sudden increase in Offgas System temperatures, b.

To preclude or mitigate the consequences of a low condenser vacuum occurrance.

REFERENCE LOT-1550. Objectives 1, 2.

Plant Procedure ON-103.

. -

-.

_.

_ _ _ _...

-

.

. _ _ -.

- _ _ _ _ _ _. _

. - _

.

f

'4.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

RADIOLOGICAL CONTROL

-

[

. ANSWERS -- LIMERICK 1-86/10/20-B. K. HAJEK ANSWER 4.07 (1.00)

[The procedure is written primarily to handle a failure in the three-element level control system.)

If a failure in the three-element level control system has resulted in excess steam flow, this action alone may reduce steam flow (through a reduction in reactor power)

sufficiently to correct the situation.

REFERENCE LOT-1540, pg.

3. Objective 1, 2,

3.

Plant Procedure OT-100.

ANSWER 4.08 (2.00)

If fuel damage is suspected [0.5],

survey the SDV [0.5]

before releasing the water [0.5]

because it could be highly contaminated [0.5].

i REFERENCE

-

LOT-1530, pg. 38.

Objective 2c.

Plant Procedure GP-11, pg.

1.

i ANSWER 4.09 (1.50)

1.

A "I Gen Stator Coolant Trouble" alarm (on 1AC854).

2.

An auto runback of the EHC loadset.

3.

A trip of both recirc pumps.

[All three conditions are required for entry.]

REFERENCE LOT-1550, ppg. 25 - 29. Objective 1.

!

Plant Procedure ON-114.

-

...

.

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

_ -.. _

_

.

>

'4,

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAOZ

RADIOLOGICAL CONTROL

, ANSWERS -- LIMERICK 1-86/10/20-B. K. HAJEK

,

ANSWER 4.10 (2.00)

1.

To maximize Drywell cooling, put into service all available Drywell chillers (0.5] and Drywell cooler fans (0.5).

2.

To terminate Drywell inerting, close the nitrogen supply valve (1F019 on panel 10C601) (0.5] and observe that nitrogen flow has decreased to zero (on FR57 on 10C601] [0.5).

REFERENCE LOT-1540, ppg. 6 - 7. Objective 2.

Plant Procedure OT-101.

ANSWER 4.11 (1.50)

G.O g mg) ogJw a.

Anytime entry is not routine.

cap 5

&

"

b.

Any licensed RO performing licensed control

,

activities (0.5), or any licensed operator Sk QLdul'4 supervising those activities (0.5]<vs.thd W M, q

REFERENCE LOT-1570. Objectives 2, 3.

Plant Procedure AP-7, pg. 22, Section 5.5.

ANSWER 4.12 (2.50)

a.

Verify that all SDV discharge and drain valves are open.

Enter T-100 if a scram occurs.

b.

SDV Not Drained Alarm at 5 gal Rod Block at 13 gal Scram at 25.45 gal REFERENCE LOT-1540, ppg. 23 -24.

Objective 1,2.

Plant Procedure OT-105.

- -

__

.

- _ _. - _ _ _ _ _

__ _ _ _ _.

_ _ _ _.. _ _..

_.. _ _.. _ _ _

_

,,

..

.

.. -.

--

lfRCl7melyf E

..,

/

',,

{

h U.S. NUCLEAR REGULATORY COMISSION

'

'"J SENIOR REACTOR OPERATOR LICENSE EXAMINATION

.

b Facility:

LIMERICK Reactor Type:.

8WR-5

_

Date Administered:

10/20/86 Examiner: >ObARRELL

'

-

,

Candidate:

/M43?fk jjii n v

,

INSTRUCTIONS TO CANDIDATE:

-

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

.

Staple question sheet on top of the answer sheets.

Points for each question

,

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

Category

% of Candidate's

% of Value Total Score Cat. Value category 25.0 25.I, 5.

Theory of Nuclear Power Plant Operation, Fluids and Thermodynamics 25.0 25.4 6.

Plant System Design, Control and Instrumentation J3.4 25.5 2P.1I 7.

Procedures - Normal, Abnormal, Emergency, and Radiological Control 2 S*. I 24.5 2RI 8.

Administrative Procedures, Conditions, and Limitations

.

41. f

_Id TOTALS Final Grade

%

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

Candidate's Signature

.

e

.

.J

.-

.

-

--

.

..

._

__ _

. _ _ _ _ _ _.

.-

.

,

..

~

So THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 17 THEmwuYNAMICS g

--

,

ANSWERS -- LIMERICK 1-86/10/21-JARRELL, D.

t-

-x

- ANSWER 5.01 (2.00)

_

_.

a.

The bubble will collapse (condense) due to heat transfer to the surrounding subcooled liquid [+1.0].

b.

Reactivity will be increased [+0.5] due to a decreased void fmetion and the negative void reactivity coefficient -[+0.5]

_

D.~ ; *pp, (. increased moderation of fast neutrons - less core leakage).

-

REFERENCE

~

1.

Limerick Generating Station: T-LOT-1340-3, " Boiling Heat Transfer," p. 4.

.

2.

Limerick Generating Station: T-LOT-1460-1, " Void Coefficient,"

p. 4.

.

ANSWER 5.0?

(2.00)

If power were reduced from a hi the insertion of shallow rods (gh power (>90%) condition [+0.5] by rods below the core midplane) [+0.5],

the resulting positive reactivity from local decrease in void formation in the lower portion of the control cell [+0.5] would more than offset the negative reactivity of the low-worth rod insertion [+0.5].

REFERENCE

'1.

Limerick Generating Station: T-LOT-1490-5, " Control Rod Worth,"

Lesson Objective No. 9.

ANSWER 5.03 (1.50)

.

,

increase [+0.25], due to reference leg being less dense [+0.5]

a.

diaphram[+0.5]].d"$$ '+0.25, due to smaller differential pressu b.

s'

s REFERENCE 1.

Limerick Generating Station: LOT-0050, " Reactor Vessel Instrumentation, p. 17.

.

ik, g e#[

..

m e'ht d =*

.-.

-.

_-,

- - - -

..

.

-___

_____

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

.

..

.,,

THE0RY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 18

'

-

THERMODYNAMICS g

-

'

' ANSWERS -- LINERICK 1-86/10/21-JARRELL, D.

-

.,

ANSWER 5.04 (3.00)l

_

_ _,

a.

- Doppler, Moderator Temperature, Void [+0.75]

'

b.

As the rods are withdrawn, power level <incNasesF 8ut the

_ -additional heat generated in the fuel is not immediately Ty, gg'trensported 'to.t he coolant, aThe fuel ~ time constant [of A

,a Ja 1 approximately 8 seconds) slows the rate-that' the' heat yeger 4

"i erated in the fuel is conducted into the coolant [+0.75].. l

~

' t takes about 3 time constants or 30 seconds for the total increase in heat generated to be transported to the coolant.

/

So the fuel temp erature rises first, causing the doppler to be the first effect. [+0.5] The next effect would be the g' /\\hig moderator temperature, as the coolant is heated to satura-p tion. [+0.5] Finally comes the effects of voids, as the

/

lieated generated in the fuel boils the water flowing through

4 the core. [+0.5]

.

,

REFERENCE 1.

Limerick Generating Station: LOT-1440, " Reactivity Coefficients and Defects," Lesson Objective No. 2.

...

g.

ANSWER 5.05 (1.50)

a.

Increase b.

Decrease tc \\lDecrease l

.

e,

c,

-.

, _.

[+0[5] each l REFERENCE

.

1.

Limerick Genernting Station: T-LOT-1260, " Fluid Properties,"

Lesson Objective No. 6.

2.

Limerick Generating Station: T-LOT-1270,"HeadLoss,"Leisson

~

Objective No. 7.

..

.

* C k ' ',

-

ts.,

me i te.Wm.

. _.....

.

_

.

__ _

.

.,,

..

,-

5.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 19 THE.0 DYNAMICS p.

.

'

ANSWERS -- LIMERICK 1-86/10/21-JARRELL, D.

.

e ANSWER 5.06 (3.00)1

,

a.-

Decrease [+0.25]. Lower steam flow results in a lower pressure

'-drop [+0.75]l

'-e m

....

m ng

,

..

b.

(Increase [+0.251. Less heat input to condenser with same cooling itions'resultszin lower condensate' temperature;[+0.75](-

" -'

'

(s. j.%,p)GDecrease [+0.25]. ' Extraction steam energy to7feedwater

'

57"prg/gn!;-,-

"#,. 'y.3

... u

,D

"~

"

^

W

'

=

.

decreases faster than feed flow [+0.75]\\

-

REFERENCE 1.

Limerick Generating Station: LOT-1270, " Fluid Friction, Head Loss," Lesson Objective No. 7, 2.

Limerick Generating Station: LOT-1180, "First Law," Lesson Objective No. 3.

3.

Limerick Generating Station: LOT-1190, " Application of Energy l

Eq.", Lesson Objective No. 3.

l

,

i e

e

.

I

.

,, w

% * 17

--..m

... -.

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

_,

m.w.,_-._,-,

,

_.,,

-.. _,..

.,m

_

,,... _

-

.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 20 THEi@i0 DYNAMICS

.

,.,v

-

ANSWERS -- LIMERICK 1-86/10/21-JARRELL, D.

z.S*

,

ANSWER 5.07 (2.r00)

_ _ _ _ _ _.

a.

. With one feed pump inoperative power operation will be limited

- --will be taken as an ac[ceptable starting point.)

to approximately 70%

+0.5].

(Any answer between 50% and ?d%

To Wmvwz ;hT 4.ggcm u.g,gt

-

a

,v,.g.,: t g...;g

...,,

-

.I

<

.

,4-

.

'

..

., - -

..

,..,

.c.. c.. a.

...

..

a0 -,-

D-

~ :~.

-

..

-: g 4,

7.:.

.,

., -

".. :

,*

.

s

-

.

60 '

-

.

&

se

,

m o

8

16

24

32

40

--

_

TimW__ _____ __

3- --

2-

!

.

=

.

g A

'

d 4-

12

20

28 3'2

40 Tim (H)

1e

. _ _...

Grading Notes: 1. Initial power between 50 and 16%, initial Xe reactivity between -2.1 and -2.7 [+0.5]

2. Xe burnout adds between.0.3 and 0.5% reactivity (+) at approximately 6 hrs [+0.5]

3. Power increases 10-20% (3% min void increase-not required) at approximately 6 hrs [+0.5]

4. Final Xe equilibrium at approximately 3% delta K/K

~

(clearly above initial level) [+0.5]

,

5. Final power level approximately 90% (clearly below initial level) [+0.5]

REFERENCE 1.

Limerick Generating Station: LOT-1510 (Training Objectives 5 and 6), " Xenon."

2.

Limerick Generating Station: LOT-1460-2, " Void Coefficient."

.

...

-e-*.emP..

-

- - -

--

-. -

-

.

,

-- 5.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 21 THERMODYNAMICS

.

ANSWERS -- LIMERICK 1-86/10/21-JARRELL, D.

3.

Limerick _ Generating Station: LOT-0540, "Feedwater System."

ANSWER 5.08

- J(2l00,)[jl',

~

t (

!

' Low tempir'ature '[/625]!becTausYthe~l1tactivity feedback effects that.,.

-

^

turn the power rise '(doppler,' mod,(41.5] l ~ ' 45 voi.d)1are : lower (or possibly positive)

>

in magnitude at low't peratuFes.'

~~

Joes

w ' e'

porus e (fiele : nJun prose,

,

' i Q 4 M n Alt. o u t a % o p a

^

REFERENCE inp.=,)

i 1.

Limerick Generat. ion Station: T-LOT-1440-1, " Reactivity Coefficients and Defects".

'

ANSWER 5.09l (2.00)

UsingP=Poe**(-t/T)

[+A0fr][o,G3

= 15 e**(-120/80)

= 3.35 on Range 4 [10d5][o,5}

= 33.5 on Range 1 Therefore, Range 1 is the lowest Range [+0.5]

or (Range 2 if administrative limit is considered.)

Assumptions: On a down power transient, with large negative reactivity insertiors, the stable decay period is determined by the longest lived half-life. For this example, it is assumed to be -80 seconds.

[+0.5]

,

l

.

i e

,

ese et

.e E * * /

d b=ee,=

= *

. hoe *M Aasmew.

.,.

-

- -.. -

,

.

- :

5.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 22

..

THEiM0 DYNAMICS

'

' ANSWERS -- LIMERICK 1-86/10/21-JARRELL, D.

REFERENCE 1.

Limerick Generating Station. T-LOT-1430-1, pp. 4-14.-

_

ANSWER 5.10l-(3.00)

.

Assumptions:T*~" ~- - WTT"

... *

,.

'Wu4ecctor steam pressure'is 1000 psig

,...

,

.

2.

T-102 scram temperature = 110 deg F

-

3. 'Cp = 1.0 Btu /lbm-deg F

[+0.25 each]

Heat Rate Input into the Suppression Pool:

0 = m delta (h)

[+0.25]

h(steam) = 1193 Btu /lbm! [+0.25]

h(sup. pocl) = 48 Btu /lbm [+0.25]

Q = 800,000(1192-48)

.Q = 9.152 x 10**8 Btu /hr heat input [+0.25]

Heat increase necessary to scram per T-102:

,

Q = aCp delta (T)

[+0.25]

. 80 Q = 122,220(62.4 lbm/cu. ft.)(110-

= (122,200)(62.4 lbm/cu. ft.)(1 Btu /lbm-deg F)(110-80)

Q = 2.29 x 10**8 Btu [+0.25]

Time required:

T = (heat necessary)/(heatup rate) [+0.25]

<T = 2.29 x 10**8 Btu /9.15 x 10**8 Btu /hr = 0.25 hr T = 15 min [+0.25]

[+0.25 for math 14 ( T ( 16]-

(AleI8.* Precedae m a T-st4 rep.'e.g.s seresee a Mn effer ope.,;ng o f SAV. ).

REFERENCE 1.

Limerick Generating Station: Technical Specifications, pp. 3/4 6-22.

~

2.

Limerick Generating Station: T-LOT-0330, " ADS", p. 3 of 19.

3.

Limerick Generating Station: T-LOT-1180, T-LOT-1190.

.

$

.

-

'

,.

..i.+-

.-.,.m.

.

_m.

. _. -..

,. _.

.

-

.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 23 THE@i0 DYNAMICS

-

.;

'

AN.iWERS -- LIMERICK 1-86/10/21-JARRELL, D.

ANSWER 5.11l (2.50)

a.'.The MAPRAT of 1.02 is NOT conservative [+0.5].

With a MAPRAT l

'

of greater than 1, it means that the MAPLHGR has been exceeded

--[+.0.5]. p e t n =tthir L ;;.$within 15 minutes [+0.5].

-

Tske corred:ve able n e e.riep

.

. pt.. +,gg$ h,, r.y[.~MAPRATn"MAPLHGR(actual)/MAPLHGRLCO v.y

,.

<-

..

-:[+T.5 ' Total] l

"'~ W

.

'b.

1.

_ false [+0.5]

gj 2.

true

.(+0.5] f

-

..

[+1.0 Total]

j

'

REFERENCE

-

1.

Limerick Generating Station: T-LOT-1410, p. 4.

2.

Limerick Generating Station: Technical Specifications, Section 3/4 2-1.

-

I l

s

.

F

!

l

.

. _.,

--

-

--

9.,. e

- - - - -

<

.

..

.

6.

PLAN 1 SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 24 E

ANSWERS -- LINERICK 1-86/10/21-JARRELL, D.

-

,

ANSWER 6.01 (2.50)

[a.8]

w._

..

E."h

'

Yes,4.evel would indicate zero {$0:S]( Jere***f e./.ccerf*M

l-1 253 l

b.

Acceptable effects are:

Sci.

.all rods remain out

- w

'

2.

xenon fece Anf.'c-of

~

'

~ i@G4.';ivoid collapse A?7temperaturedeficit

-

"

>

,. >

dN

.

,,

.,

.

<

-

s

"~ 5 * moderator deficit

<-

doppler deficit 6.

7.

control rod worth decrease (moderator temp. decrease)

~

Part b answers worth [+0.5] each, +2.0 maximum.

REFERENCE

-

1.

Linerick Generating Station: LOT-0310, " Standby Liquid Control

'

System."

ANSWER 6.02 (3.00)

a.

Indicated level will increase [+0.25] to maximum because both sides of the cell see the same pressure [tD X5].

2. 5

b.

Feed flow will decrease [tod5 Tin response to an apparent high level [+.045].

5 c.

The Main Turbine trip will be unaffected [+0.25' because its signal comes from the D level transmitter [+0.5l. (Trip will leastueseks c4 :. )

.,.

eaca re latw 14,As=(

,

ao d.

The recirculation pump will be unaffected [+0.25] unless vessel level is allowed to reach 27.5 in and the RFP Low Flow alam will result) [+0.5]perator action (75% recirculation runback is not cleared by o

~

.

'(Answers may vary if valid assumptions are presented.) l

,

REFERENCE 1.

Limerick Generating Station: Hot License Training Quiz - Week 1.

2.

Limerick Generating Station: LOT-0050, " Reactor Vessel Instrumentation", p. 7.

3.

Limerick Generating Station: LOT-0040, " Recirculation Flow Control", p. 8.

4.

Limerick Generating Station: LOT-0540, "Feedwater System", p.18.

.

2,

e a

no s *.h w ed esA

'

.,mem. -

-

-.

-

.

..

-

6.-

PLANT SYSTEMS DESIGN.-CONTROL. AND INSTRUMENTATION PAGE 25

~~

ANSWERS -- LIMERICK 1-86/10/21-JARRELL, D.

,

.

ANSWER 6.03 (3.00)

.

,

if a.--control valves close 5% [+0.35] lbypass valves open.5% ~[+0.40]

.)~

b.

' control valves close 5% [+0'.35] freactorscramprobabledueto increasing pressure because) bypass valves >will.not open [+0.40]l

.

~

2- ' ~Regulatorh8" controls at 100% power but with CV. closed.a small.

- N'.

unt [+0.35]\\(to compensate f ssure bypass * "

i@. 20: remains closed [+0.40] ~ (sleadipr highem reacgor pgJavi[) 6 s )

farf Q.Jss if'

d.

control valves close to approximately 94% [+0.35J l(to maintain

~ Rx pressure at 920 psig) while approximately 6% of flow goes through bypass [+0.40]i REFERENCE

'1.

Limerick Generating Station: LOT-0590, " Electro-Hydraulic Control Logic", Sections III.E.4, III.E.12, and IV.A.

2.

Limerick Generating Station:

LOT-0120-7, " Main Steam System".

3.

Limerick Generating Station: Simulator Malfunction No.105b.

ANSWER 6.04 (2.00)

a.

No ADS response [+0.5] because RHR/CS pump discharge is not above set point (RHR/CS pump not operating [+0.5].

,

i-b.

No [+0.5]

l l

c.

Yes

[+0.5]

i I

REFERENCE 1.

Limerick Generating Station: T-LOT-03'30-1, " ADS", p. 3.

'

2.

Limerick Generating Station: Lesson Objective No. 6.

s

,

.

., ~,

.

..-...-

- - - - - -

._

-

"

.

. _.

.

-_

-

.

y

.

.-

PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 26 I

ANSWERS -- LIMERICK 1-86/10/21-JARRELL D.

,

.

F

, ANSWER 6.05 (2.00)

,

'

"a.

.YES - [+0.5] ----

--

-

..

.+

,

!b. : The condenser will fill with condensate [+0.5] resulting in radioactive wirorsteam leaking from the turbine seals [+1.0].

"I V

i avemee%.cl-Yffl-.

' * '

-

--

~+

=

S si;a w

n 1.

Limerick Generating Station: LOT-0340, III.D.

.

.ANSWEk 6.06 (3.00)

g An ATWS RPT is initiatedjREACTOR LOW LEVEL (-38 in.) [+0.3] or a.

REACTOR HIGH PRESSURE (MEO psig) [+0.3], and an EOC RPT initiates on a TURBINE TRIP [+0.3] when power is GREATER THAN 30% [+0.3] as

,

sensed by FIRST STAGE PRESSURE [+0.3].

b.

Both'ltrips open breakers between the MG SET and the' RECIRC

/'

'

PUMP (MOTOR)

(, g, c g p p )g,,(yd

-

'

%' c. l At the End-of-Cycle, a pressure increase transient from the turbine trip could add positive reactivity faster than the control rods can insert negative reactivity. Tripping the recirculation pumps helps by adding negative reactivity.

[+1.0] ( M cF R W u * e..,% 1fc l

I$ +;<d 4o T.S. %.1.s )

REFERENCE

, 1.

Limerick Generating Station: LOT-0040, " Recirculation Flow Control", pp. 16, 22, and 23, Typical Question 5.

.

l l

.

,'

,,

.

6.

PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 27 ANSWERS -- LIMERICK 1-86/10/21-JARRELL, D.

,

f

. ANSWER 6.07 (2.00)

d:'..

.

IID12/oads:

.

708 ESW pump OB RHR service water pump 18 RHR pump

~~'~i18, turbine building; equipment compartment exhaust fan

>

.

'18 tore spray pump w

,

'0124' load center

'

,

~[+0.33]l for each load.

f'

,

REFERENCE 1.

Limerick Generating Station: LOT-0660, "4.16 kV AC Power Distribution", Figure T-LOT-0660-1 and Section III.B.3.d

__

(Training Objective 4.b).

ANSWER 6.08 (1.50)

1.

Pump would trip off (to unload the D11 bus).

[+0.5]

55-second time delay would 2.

diesel to come up to speed) prevent motor restart (to allow

[+0.5]

.

l 3.

ESW pump would restart following the time delay.

[+0.5]

REFERENCE 1.

Limerick Generating Station: LOT-0680, " Emergency Service Water", Lesson Objective No. 5.

.

%

.

-

.

--

m.--

m.

_ _,, _ _ _ _ _

y

--- -- -

$

,

--

g-

.. - - - -

- - -

_.

PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 28

-

ANSWERS -- LIMERICK 1-86/10/21-JARRELL, D.

-

,

ANSWER 6.09 (2.50)

-

a.- 1.

low reactor water level - 38 in.

2.

high ventilation differential temperature - 32 deg F 3.

high equipment area temperature - 122/135 deg F 4.

high differential flow - 54.9 gpm 5.

SBLC initiation

" ~, ; j. J. ~ ~ ~ "'

L_

-

9g)FMch.' -

'

%

-

,

.

'

+..... w.

b.

Low flow through the demineralizer bed would allow " sloughing

. off" of the resin precoat [+0.5].

The pump auto starts at 60 gpa (decreasing flow through the on-line demineralizer) [+0.5].

REFERENCE 1.

Limerick Generating Station: LOT-0110. " Reactor Water Cleanup System", Sections III.D.5 and iV.A.3.

ANSWER 6.10 (2.00)\\

Gama rays from nuclide decay and other background sources entering SRM or IRM detectors cause ionization in the detector. This gama-induced ionization is not a direct function of neutron power level.

!

l[+1.0] \\ In the APRM region gama flux is proportional to power (also I

neutron flux is much greater than gama flux).

[+1.0] l REFERENCE.

1.

Limerick. Generating Station: LOT-0240, " Source Range Monitors",

i

.,

e;

?*

ANSWER 6.11 (1.50.)

I a.

half-scram b.

hal f-scram

~

c.

rod blockl

~( s.... 4 p a a,a ae+;o. ;; sb++ad ana f.)

I ( 6..n.-s i.

,. > set )

I

[+0.5each]

'

REFERENCE 1.

Limerick Generating Station: LOT-300, "RPS", Learning Objective No. 8.

2.

Limerick Generating Station: LOT-270, "APRM", Learning Objective No. 2.

3.

Limerick Generating Station: LOT-070, "CRDH", fearning Objective No. 5.

.

,

f-S 1.

e

'^

~-

_

. _ _

___ _

-. _ _

.

c..

7.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 29

.)

.

RADIOLOGICAL CONTROL

'

ANSWERS -- LIMERICK 1-86/10/21-JARRELL, D.

ANSWER 7.01

... --

(2.00)

a.

site emergency M l 0-damage of fuel element [+1.0]

y._._

.

-

-- REFERENCE

-

, t~NWSt;ce.titi:

"%.

' "

~

.

,g,

-

~-

1.

' Limerick Generating Station:

pp. 11, 13.

EP-101, Classification of Emergencies,

'

.

.g

,

,,-

--

g

-

I ANSWER 7.02

.(2.00) I l

Toavoidexcessivethermalstress(onthick(section) metal a.

components) [+1.0]

I b.

To prevent shortening seal life [+1.0] /

,

REFERENCE 1.

Limerick Generating Station: LOT-1530, " General Plant Procedures," p. 23.

.

O

%

l I

.-.

.

_ -

-

- -

.. -

.

..

__ _ _.

.

. _ _

-

.

,

.

,

7.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 30 RADIOLOGICAL CONTROL

)

-

,

.

ANSWERS -- LIMERICK 1-86/10/21-JARRELL, D.

ANS 7.03_

(2.00)

'

Adequate co cooling is assured [+0.5]

-9 Y

.Even though the e may be uncovered, since at 1 one ECCS pump

is injecting into t re (RHR 'A'), then efinition of adequate f

core cooling is met. [41.

M

"

k' bE %~ ' '

5,

.. *[5

"'

M 1.

Limerick rating Station: Mitiga on of Core Damage, ot License

. tion, given at Limerick 01/14/

ANSWER 7.04 (1.50)

!

Group 1 [+0.5]; this minimizes the total work force exposure [+1.0]

,l REFERENCE 1.

Limerick Generation Station:

LOT-1770, ALARA Program, pp. 2

,

and 10.

l

.

-

_..

g

.

l l

I

.

,,

,

. dn %

e

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

__.-_..._.___,___._.____..m._-~,%,.....,,w-

.. _,, _..... _..,,., _ _

..

.

'*s*6L

.. -. _,

y

,

-

.

.

7.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 31

-

RADIOLOGICAL CONTROL j

.

'

ANSWERS -- LIMERICK 1-86/10/21-JARRELL, D.

ANSWER 7.05 j(2.50)

,,

a.

1. Sudden drop in primary containment pressure.

2. Abnormally high N2 makeup to primary containment.

3. The failure of a containment penetration:. leak rate test

.u

-

resulting in contaminant leakage exceeding 0.6 times the

"f,g %w d aaximum allowable leakage.

'

.--

.2

-

..

?0bservation of a primary containment breach.

"

qv.y p

.

[3 of 4 require at +0.5 each]

-

b.

1. Visual observation of damage to or unauthorized opening of the secondary containment boundary.

(be *. s a++s )

2. SBGTS is unable to maintain secondary containment at -0.25

inc es of water pressure w6ct a flow rate of less than 57000-e SC (Wh

15*

.

_

[+0.5 points each]

REFERENCE

.

1.

Limerick Generating Station: ON-110 (Loss of' Primary Containment)

2.

Limerick Generating Station: ON-111 (Loss of Secondary Containment)

/

ANSWER 7.06-(3.00) /

a.

1.

suppression pool temperature > 95 deg F [+0.5]i 2.

suppression pool level ( 22 ft or > 24.25 ft [+0.5] l l

3.

drywell pressure > 1.68 psig [+0.5)J ]

l 4.

drywell temperature > 135 deg F

[+0.5 b.

All paths must be pursued in parallel.

[+0.5]

,

c.

When all entry conditions have been cleared.

[+0.5]

REFERENCE 1.

Limerick Generating Station: Trip Procedure T-102, " Containment Control".

.

he..

-

.

-

..

'

PROCEDUP.ES - NORMAL. ABNORMAL, EMERGENCY AND PAGE 32

RADIOLOGICAL CONTROL

-

.

-86/10/21-JARRELL, D.

ANSWERS -- LIMERICK 1 I

ANSWER "'t7.07 f'(2.50)\\

.

,

-

,- mu

-.I

'

1.

Runback recirculation flow [+0.5] to maintain pressure below

'

1020 psig [M,,5]

)

j- -

l

[

,.

,,

2; Control reactor pressure below 1020 psig with bypass valves l

'

gh.;,a[nP:s4v. jg:.3,.g z,...

'.l.., \\

,

+0.5]_using the jack

..,

s

,

,4,

,%. e g, %

v.

3.-

If a scram occurs, enter Procedure T-160' [+0.5],-

-,~,. s,.

,,.

j-

-

.

- ' "

.

REFERENCE 1.

Limerick Generating Station: OT-102 (Reactor High Pressure) pg.1.

l t

!

,

e S

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

.--

r

-

-.

m

-

_

,-.

- ' ' '

.

-

.

7.

PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE 33

mRADIOLOGICAL CONTROL

-

'

ANSWERS -- LIMERICK 1-86/10/21-JARRELL, D.

\\

ANSWER 7.08

!;(2.00)

g p,

,

_

. 1.

Turbine trip 22.2 inches

,

2.

Feed p trip

inches

'=

"^ S,

-

it. N itSIV c osure

.

10.5 psia

> '

r-r

4.4 '8ypass Valve closure 7 inches l:

def*KIk,"aeagdeAlt o'f ssipeiof.h yi"*)

-

'

.

_

..

.

1.

Limerick Generating Station: OT-116,(LossofCondenser' Vacuum)

pg. 2.

I ANSWER 7.09

'(1.'50)l

~

Only the APRMs are environmentally qualified to be reliable during a

~

LOCA.

[+1.5] /

REFERENCE 1.

Limerick Generating Station: T-101, Caution No. 8.

,

ANSWER 7.10-(2.00)i

)

'

7*)

a.

1.

'l GEN STATOR COOLANT TROUBLE' alarm [+0.33]

,

I 2.

automatic EHC runback [+0.33] !

trip of both recirculation pumpsf+0.3 3.

b.

1.

to remain within the capacity of the bypass valves [+0.33] l and maintain an adequate scram margin. [+0.33],

2.

to minimize Generator Stator Amps. [+0.33]

l REFERENCE

,

1.

Limerick Generating Station: ON-114 (Loss of Stator Water Cooling Runback) pg. 1.

.

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PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE 34 y..

-IIADIOLOGICAL CONTROL

'

ANSWERS -- LIMERICK 1-86/10/21-JARRELL, D.

..

I ANSWER 7.11 (2.00)

b.

This setpiont ensures that a RFP will be operating prior to

< reactor pressure exceeding the capacity of the condensate pump. [+1.0]

-..3x

,.

c.'

'. Establishes a continuous steaming rate allowing ] continuing

>

"

ggy flow through the feedwater inlet nozzles.

[+1.0

.

qvges,o 4.

.

,

.

ANSWER 7.12

/

(2.00)

l l

a.

Scram by placing MS radiation monitors A, B, C, and D in the tripped condition [+0.5] from the auxiliary equipment room

,

[+0.5].

'

b.

Scram by tripping the RPS [+0.5] breakers from inverter room 1

[+0.5].

!

.

l t T.,p% +,9 r., p.. > wr n

.,..,fou. )

REFERENCE 1.

Limerick Generating Station: Special Event Procedure SE-1,

" Remote Shutdown", Section 4.2 and 4.3.

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.

4 e

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.-

-

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-

.

ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 35

ANSWERS -- LIMERICK 1'

-86/10/21-JARRELL, D.

\\

ANSWER 8.01

.(3.00) I j

_

'....:~..

Mince 1 APRM channel is inoperable and the pmtection system setpoint on

' M;.a second channel is less conservative than Table 2.2.1-1 [+0.5], the RPS

-

' channel must be declared inoperable [+1.0] within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of.the

'i.dFPMnitiation of the surveillance [+0.5].dFollowingt its declaration as

'M inoperative, the deficient RPS channel shall be placed,in a tripped 1l ition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . [+0.5] - Per Tech. Spec. Section 3.3.1.a.[+0.5].

'

&

.3 3 \\

,

~-

44,, REFERENCE

' ' ' ~

~

-

.

..

E ' " 1.

Limerick Generating Station: Technical Specifications, Table 3.3.1-1(a),Section3.3.1.

"

Note: Technical Specifications must be provided for this question.

ANSWER 8.02 (2.50)

1.

Themal power shall not exceed 25% of rated [+0.5] when vessel pressure (steam dome) is less than 785 psig or core flow ( 10%

of rated. [+0.5]

2.

The Minimum Critical Power Ratio shall not be ( 1.06 [+0.5]

with reactor pressure > 785 psig and core flow > 10% of. rated.

[+0.5]

l The GEXL correlation (power / flow map, which establishes safe cladding amperatures at high power levels) is only valid above 785 psig and I.

10% of rated core flow. [+0.5]

REFERENCE 1.

. Limerick Generating Station: Technical Specifications, Section 2.1.1, 2.1.2, and B2-1.

.

ANSWER 8.03 (2.00))

a.

no [+0.5]

b.

no - [+0.5]

\\

c.

no

[+0.5]I d.

yes [+0.5] i REFERENCE 1.

Limerick Generating Station: Technical Specifications, Section

,

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-

.

8.

ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIM 79TIONS PAGE 36

-

ANSWERS -- LIMERICK 1-86/10/21-JARRELL, D.

.

1-7.

.

ANSWER 8.04 (3.00)l

-

W f.e

-

m-Mas:. [MMJ] Surveillance requirement 4.8.1.1.2.a.7.cannot be met (air start receiver pressure > 225 psig) [*Vf]. l

-

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REFERENCE

_

,

.kkk$ brick Generating Station: Technical Specifications,' Sectio'nf

Y

'

~ 1.2.5 and 4.8.1.1.2.

.

Note: Technical Specification Section 3/4.8 must be provided.

ANSWER 8.05

' - (2.00)

Minimum volume limit to provide a sufficient supply]of water a.

to emergency systems in the event of a LOCA.

[+1.0 b.

Maximum temperature limit provides complete quenching of blowdown steam [+0.5] and provides ECC pump NPSH [+0.5].

REFERENCE 1.

Limerick Generating Station: Technical Speq1fications, Sections 3.6.2.1.a.1, B 3/4.5.3, and 8 3/4.6.2.A 5 % 4 -3 ( m.os,volu.,w lo eker b ioc A u eske.as r } -

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ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 37

-

. ANSWERS -- LIMERICK 1-86/10/21-JARRELL, D.

,

ANSWER

~8.06 (2.00)

[0perab111ty-

'

'

'

either recirculation loop flow differs (by > 10h from given a.-

. g.; speed /f1w characteristic.

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-

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tal

,:

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cP-diffuser-to-lower plenum delta P of any individual jet pump differsfromnorm(by>10%),

Any two (2) [+0.5] each, maximum +1.0.

Bases -

a.

Increases blowdown area during LOCA [+0.5]

b.

reduces capability of reflooding core following a LOCA [+0.5]

REFERENCE 1.

Limerick Generating Station: Technical Specifications, Sections 3.4.1.2, 4.4.1.2, a; i ; 3/4 4-1.

.

ANSWER 8.07 (1.50)

a.

shift supervisor [+0.5:1 b.

control room operator [+0.5]

c.

the "as found" position [+0.5]

i REFERENCE 1.

Limerick Generating Station: Administrative Procedure A-41, Sections 5.2.5 and 5.3.1.

.

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ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 38

-

Y ANSWERS -- LIMERICK 1-86/10/21-JARRELL, D.

-

,

, ANSWER.

8.08 (2.00)

r. -

.s.-1.-

complex or extensive jobs where reliance on memory cannot be trusted

.

,

.

g 2.

tasks that are infrequently performed "

""' x,

J'*couplex tasks ~that must be performed in a specific order.

V ny two (2) g o.dia,, yad

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A Y+0.5] each, +1.0 maximum.

,

b.

Plant personnel shall take action so as to minimize personnel injury and damage to the facility and to protect the health and safety of the public [+1.0].

REFERENCE 1.

Limerick Generating Station: Administrative Procedure A-7, SHIFT OPERATION, Section 5.2.1, "Use of Procedures" and Section 5.2.5,

" Response to Transients and Emergencies."

.

f

.

,

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-

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

-

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'.

ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 39 i

ANSWERS -- LIMERICK 1-86/10/21-JARRELL, D.

,

i ANSWER.

8.09 (2.50) \\

3.b '

i Operable [+0.5], apply action for 3.7.0."

1.e.,closethevalve)

~

[fl.0]

Valve closure makes HPCI then action 3.5.E.2 applies

"

[+1.0]

(1/2 credit for giving only HPCI action.)

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1.

Limerick Generating Station: Technical Specifications 374.5.1

"HPCI) and 3/4.6.3 " Primary Containment).

Provide Technical Specification pages 3/4 5-1 to 3/4 5-7 and 3/4 6-17 to 3/4 6-18.

ANSWER 8.10 (2.00)

' 0.5'

a.

false

+

b.

false,'+0.5

.

c.

true

+0.5 d.

true l+0.5:

REFERENCE 1.

Limerick Generating Station:

A-41, " Procedure for Control of Plant Equipment", pp. 10, 12.

2.

Limerick Generating Station: A-26, " Procedure for Corrective Maintenance", pp. 14, 16.

3.

Limerick Generating Station:

A-3, " Temporary Changes", pp.1, y

3.

.

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T..

.

'

8-ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 40

'

ANSWERS -- LIMERICK 1-86/10/21-JARRELL, D.

,

t.oo ANSWER 8.11 _. I

{2<S0)

-_..i.~7 would~not - -'

.

'

,

b.

would

' c.

would d.

.would

.

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[+0.5each]

-}

vt. iNY?' ' :' ;~ '

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REFERENCE

-

1.

Limerick Generating Station:

A-31, " Notification of NRC", pp.

3-4.

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.

_

_. _ _ _

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ATTACHMENT 3 Facility camments 'and NRC Resolutions of Comments on Written Examinations The following represents the NRC resolution to those comments made by the facility as a result of the current exam review policy.

Only those comments resulting in significant changes to the master answer key, or were "not accepted" by the NRC, are listed and explained below. Comments made that were insignificant in nature and resolved to the satisfaction of both the examiner and the licensee during the post exam review are not listed.

ie:

typo errors, relative acceptable terms, minor set point changes.

SRO Examination 5.10 Facility comment that OT-114 requires a scram if a SRV is open for two minutes.

NRC resolution: Comment accepted as an alternate answer.

5.11 Facility comment that action to correct must be started within 15 minutes.

'

NRC resolution: Comment accepted as this is the correct Tech Spec interpretation.

6.01 A.

Facility Comment: Alternate correct answer could be that the SBLC tank level indication would not necessarily drop to zero, but it certainly would decrease depending upon extent of loss of instrument

-

air.

NRC Resolution: Since it was not explicitly stated that the instru-ment air supply pressure failed to zero, an alternate correct answer for part a) is a decreasing indication.

6.03 C.

Facility Comment:

Answer stated indicates CV closes a small amount. Analysis of EHC logic diagram indicates that CV would return

,

l to original position (i.e., open to support 100% rated steam flow).

NRC Resolution: Answer will remain as is. Simulator malfunction No.

105b states that "this malfunction results in a transient identical to a step 10 psi increase in the EHC pressure setpoint." Further,

" turbine load will be esse-tially the same as before, but with pres-sure 10 psi higher." With higher inlet pressure, the only means of maintaining constant turbine load is to provide additional throttling at the turbine control valves. Credit will be given if a candidate recognizes that the actual bias on the "B" regulator is set l

. _.. _ _,. _ _ _

_, _ _ _

. - _ _

_.

~

_

-

_

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M

.Attacnment 3~

approximately 2 psi above the "A" set point (just enough to prevent

" hunting" oscillations) and therefore will not produce a noticeable reactor pressure shift.

6.06'C.

Facility Comment:

Alternate correct answer for why EOC/RPT trip is

"

necessary could be MCPR thermal' limit consideration. Tech Spec 3/4.2.3-indicates that MCPR LC0 Tech Spec Action required if,E0C/RPT inoperable.

-NRC Resolution: -Acceptable alternate answers are that a MCPR consid-eration based on LCO action statement is required per T.S. 3/4.2.3,

'and to assist in prevention of a double low reactor water level as a result of " shrink".

6.11 Facility Comment: Alternate answer for part a) is no action if

. shifted manually. Alternate answer for part b should be based on

. candidate's assumption of reactor. power.

NRC Resolution: Part a) No action would be produced if shifted manu-ally (credit will be given if this answer is supported).

Part b) Acceptable alternates:

Power > 42% => ROD BLOCK Power > 51% => 1/2 SCRAM 7.03 Facility Comment: Answer key states that adequate core cooling is assured. The reference for this answer was-given as LGS:MOCD, Hot-License Examination, given at Limerick 1/14/85.

As was indicated in the utility response to this same question in Jan. 1985, there are several alternate correct answers to this ques-tion depending upon how the SRO justifies his answer:

!

(1) Adequate core cooling is assured (as per T-116, step RF-14)

because core submergence can be assured when the operator has increase level indication due to ECCS pump injection

,

!

into the vessel until at least 3 SRV's are open and Rx (

pressure is stabilized at or increasing through 72 psig.

!

(2) If the G.E. Emergency Procedure Guidelines G.E. EPG Document

8308-2, Section 3, Definition of Adequate Core Cooling) are j

used then the answer to this question is no, adequate

core cooling cannot be assured. This is because the LGS I

reference to Spray Cooling, as this phenomenon is discussed in Enclosure 3, is covered in LGS TRIP Procedure T-114,

.

step SPK-3 which states that "with one core spray subsystem i

operating as described in the General Section of the bases

!

adequate core cooling is assured".

!

f

,

.

O Attachment 3

Thus, depending upon logic that SRO candidate used in answering question various answers may be considered accurate.

NRC Resolution: Due to the ambiguity concerning the question state-ment and responses, this question was deleted from the exam.

7.05 B.

Facility Comment:

Answer key actually states two answers within the first given answer. The SRO candidate may state visual observation of damage to secondary containment as one symptom and the unauthor-ized opening of secondary containment as the second symptom. The question was not specific to say list the symptoms but just asked

"what are two (2) symptoms".

NRC Resolution: The first answer listed actually contains two acceptable answers - credit will be given for either or both. The flow rate figure given in answer 2. should be corrected to 1250 SCFM, and enclosed in parentheses.

7.09_

Facility Comment:

This question places the burden on the SRO candidate of having to memorize a caution in a TRIP procedure. The TRIP procedure Lesson Plan objective page for LOT-1560 specifically indicates that "Given a caution statement contained in one or more TRIP procedures, state the bases behind that caution." Since the question does not state the entire caution statement, this question is confusing to the SRO candidate because he is asked to remember an item which he is only required to explain once it is given in its entirety.

Although the answer key response indicates that SRM's are not used because APRM;s are environmentally qualified, an alternate correct answer could also be given. This answer would be that the SRM's can only monitor the % power range of 10 * to 10 3% and would never be able to tell if reactor power is greater than 4%.

NRC Resolution: Not accepted. The candidate was only asked to state the bases.or the use of APRMs following a LOCA. A full statement of the caution note would have provided the candidate with the approp-riate answer.

Future usage of this question will utilize more exact wording of the cautionary statement so as to specifically reference only the APRMs, but will still require an understanding of this important ATWS recognition principal. Credit for logically defended answers will be considered where appropriate.

7.12 Facility Comment: Alternate correct answers to this question are as follows:

(1) Manually trip the Main Turbine form the Front Standard.

This will cause a resultant generator lockout and reactor

.

scram due to Turbine Stop Valve closure (see LOT LP-0300 l

for RPS); and i

l

...

. - _ _ _ _. -

-

._

-

.

._.. -.

.

Attachment 3

'4 (2)

In Aux. Equip. Room open breakers for RPS Trip System 1AY160 and 18Y160 (see SE-6 procedure, page 2). These breakers are located just inside the door of the Aux.

Equip. Room.

NRC Resolution:

Comments accepted as viable alternatives.

8.04 Facility Comment: The answer agreed to during the post-exam 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months review was that the Division 1 D/G is not operable according to Tech-nical Specifications.

Looking at Tech Spec Surveillance requirement 4.8.1.1.2.a.7., it would appear that the action of declaring the C/G inoperable due to air start receiver pressure <225 psig would be interpreted as the most conservative action. However, in discussion with both the LGS Superintendent-Operations and Operations Engineer they interpreted this condition as the 0/G still being operable.

This is based on the assertion that the D/G is connected to the 4KV bus during this surveillance and it is performing its intended func-tion.

If one assumes that the air pressure was indicating 200# after the D/G was operating and supplying load then the D/G is not inoper-able until it is shutdown after the surveillance test has been completed.

The answers by the individual SR0 candidates should be graded based upon what interpretative Tech Spec logic they used to come to a conclusion in their answer.

NRC Resolution:

Comment Accepted.

Logically supported answers will be carefully considered during grading.

8.05 Facility Comment: An alternate answer is a minimum suppression pool volume required to absorb LOCA Energy release.

Reference Tech Spec B

3/4 6-3.

NRC Resolution:

Comment accepted. Answer key will be corrected.

8.06 Facility Comment: This question asks the SR0 to memorize Tech Spec Surveillance Requirements 4.4.1.2.a, b., and c.

These items are not required for SR0's to memorize as part of their Hot License training program. An acceptable, alternative answer for this question would be to refer to Off-Normal (ON) procedure and Jet pump operability ST to check for operability. This ON procedure (0N-100) is what SR0's are expected to know and effectively utilize.

NRC Resolution:

Not accepted. The question only requires that sufficient knowledge of the indications of a jet pump failure be demonstrated. Without the ability to recognized these symptoms it is impossible for the candidate to utilize either the OT or the tech spec (see LOT-1550 lesson objective NO.1).

Since the symp' ~ recog-nition of the LOT and the operability checks given in the tech spec are functionally identical, simply referring to the ON or jet pump ST l

Attachment 3

is not acceptable. The quantifier "by 10%" should be placed in parentheses (as agreed upon at the post-test review) in all three answers.

8.08 A.

Facility Comment:

Another acceptable answer would be as indicated on page 13 of A-7 paragraph 5.2.1.

This additional condition for

-

requiring direct reference to procedures is if there is any doubt to the procedural action required.

NRC Resolution: Corr:nent accepted as alternate answer.

8.10 Facility Comment: The answers listed for parts a.,

b.,

c., and d.

are not correct as indicated on the answer key. The following justi-fication applies:

a.

Equipment Trouble Tag would have generated a local permit as per A-26.

Local permit cleared (E.T.T.) removed - individual verified permit cleared would then remove E.T.T Correct answer is true.

b.

Tag would not be reapplied, valve identification not on local permit.

Correct answer is true or false.

c.

As per A-3, two members of unit management staff at least one of the individuals shall be the Shift Superintendent.

Correct answer is false.

d.

As per part A of question E.T.T. would be lifted and oper-ability of minimum flow would have been verified

previously.

Correct answer is f alse.

NRC Resolution: Alternate answers will be accepted if justified by the candidate.

8.11 C.

Facility Comment:

Alternate answer by SR0 candidate to this event as 1-hcur reportable may be that it is not due to the fact that a loss of one 125VDC bus may not remove a significant portion of i

I control room indication. Annunciator power is supplied by all four (4) 125 VDC panels (buses). Therefore loss of any one bus may not mean a "significant portion of control room indication" is lost.

,

i i

_

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.. - -

-

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-

... -. -

,

..

......

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

.

4-Attachment 3

NRC Resolution: 'If justified as indicated, the alternate answer will be accepted. Future use of this question will specify a loss of two of the 125 VDC busses.

.

b

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R0 EXAMINATION.

1.02 Facility Comment: Alternate answer could include reference to all of the reactivity coefficients (i.e., doppler, moderator, or voids) due to operator interpretation of shutdown /cooldown of the reactor.

NRC Resolution: Written comment was not accepted because the altern-ate reactivity coefficients have no effect under the conditions stated in the question.

1.03 NRC Resolution of Verbal Comment:

Verbal comment that some candi-dates might indicate that condenser vacuum would draw down the pres-sure in the heaters was accepted as a valid explanation of the physi-cal phenomena.

1.04 Facility Comment: An alternate, correct answer to this IRM detector response may be that it is difficult to determine approximate water level because of how the steam / moderator mixture during the LOCA affects neutron thermalization. The C.E. MOCD test (specifically pages 5-3 and 5-4) indicates that the use of SRM's (and hence IRM's since they are very similar in response to answer should be graded appropriately in light of this additional informatien.

NRC Resolution: Written comment was accepted that results could be erratic at the surface, but other parts of the comments were not accepted because the physics described in the lesson plan should be true independent of what happens near the surface.

1.08 NRC Response to Verbal Comment:

Verbal comment that the operator can only affect the level over the core was considered during the grading.

1.09 a.

Facility Comment: Alternate answer for power decrease due to Recir-culation System runback to 28% to loss of all feedwater.

NRC Resolution: Written comment accepted as a valid answer if given.

1.10 Facility Comment: Alternate answers for INITIAL effects on power and pressure due to HPCI initiation at 50% power.

Reactor pressure may also decrease due to HPCI steam flow (approximately 100,000 lbm/hr)

being drawn off the "C" Main Steam line. Reactor power for this transient is taught to operators as changed due to a positive reactivity insertion (see OT-104 Bases). Additionally, the LGS FSAR for HPCI inadvertent initiation indicates that Rx power increases about 3-5 sec. after actual initiation.

In reality, this is the parameter change the operator would see and react t. _ - -

?V

,$

R0 Examination

NRC Resolution: Written comments were accepted and considered during the grading, noting in particular the context of the candidate's answer.

1.12.C.

Facility Comment:

Statement made by examiner that "The Doppler effect always increases..." could mislead licensed candidates to provide alternate answer to the question. Answer to this question

~

may be stated as FALSE due to two considerations:

.

-(1) doppler effect substantially increases as fuel temperature increases; however, at higher temperatures the broadening of the peaks is less for higher fuel temperature changes.

(2) As fuel temperatures increase the doppler coefficient becomes less negative vice " smaller" as indicated in the question.

NRC' Resolution: Comment was not accepted. The question was clearly stated,and required the candidate to consider the statement care-fully.

Limerick comment only reaffirms the statement in the question

usirg alternate wording.

2.01 Facility Comment: Alternate answer may include other responses such

>

as:

Main Steam Bypass Valves Various Moisture Drain ecnnections Safety / Relief Valves

'

Steam Seal Evaporator

NRC Resolution: Additional loads accepted except for the drains.

2.02 B.

Facility Comment:

Alternate answers to "what advantage does this have'on changes in core parameters when a control rod is moved?"

,

could be:

(1) the peak nodal LHGR adjacent to control rods is reduced and the change in LHGR due to control rod withdrawal is reduced; and l

(2) local peaking factors are reduced due to the fact that partially i

inserted control rods are never inserted next to high power (

bundles.

NRC Resolution: The written alternate answers were accepted.

2.03 NRC Resolution of Verbal Comment: Verbal comment that alternate wording could be " damage to containment" was accepted.

2.06 B Facility Comment:

Answer key reflects components in the flow path (i.e., breakdown bushing which question does not ask for. Answer should just be "the reactor".

l l

-

_.. _,

.....

_ _ _ _,

._

-

a I. -

R0 Examination

NRC Resolution: Written comment that question did not request flow path was accepted.

2.08 NRC Resolution of Verbal Comment: Verbal comment that the suction is actually upstream of the pumps, and that some might say the cooling tower basin was accepted. This is also stated in the lesson plan.

3.02 NRC Resolution of Verbal Comment: Verbal comment that exact propor-tions were not asked in the question and therefore should not be required was accepted as long as the candidate indicated an under-standing of what was occurring in the system.

3.04 Facility Comment: Just for information purposes, these rod configur-ations discussed in question could not occur at LGS.

Reactor Maneuvering Instructions (RWM Sequences for S/U and S/D) at LGS are specific on this point.

NRC Resolution: This comment was noted.

3.05 b.

Facility Comment:

Effect on turbine operation should also include bypass valve opening.

NRC Resolution: Comment on bypass valves opening was accepted.

3,06 Facility Comment: Additionally, an LPRM signal is automatically bypassed if:

1.

Edge rod selected 2.

RBM bypassed 3.

Ref. APRM less than 30%

NRC Resolution: This written comment was withdrawn by Ed Firth (after its submittal) after discussion of how these bypasses were generated relative to the question asked.

3.07 Facility Comment:

RO candidate alternate answer may be that indicated level may be unaffected (remain the same) due to actual location of all LGS water level transmitter reference legs outside the containment.

NRC Resolution: This written comment was accepted, and it was noted that the lesson plan was incorrect. Plant stated that the lesson

,

'

plan would be corrected.

3.07 B.

Facili'.y Comment:

Other alternate answer to actions that assure sufficient pressure to ADS valves may be AOS accumulators.

!

__

o

R0 Examination

NRC Resolution: Both the verbal and written comments were accepted that the provisions discussed in the lesson plan were for long term cooling, and that the accumulators were all that were necessary.

It was noted that the lesson plan needed updating, and that it should cross reference the ADS lesson plan.

3.09 Facility Comment: Answer (1) as indicated by answer key states that supply and exhaust dampers would close. Candidates may state that this answer is equivalent to "Rx Enclosure isolation".

NRC Resolution: Written alternate wording was accepted.

3.11 Facili ty Comment: Alternate answer to three actions required would also be automatic annunciation.

NRC Resolution: Written comment not accepted because question clearly requested actions rather than alarms.

4.01 NRC Resolution of Verbal Comment: Verbal comment that the requested actions are initial rather than immediate actions was noted.

4.02 Facility Comment: Answer key indicates answer for 2.b. as QA Supt.

or alternate. This requirement is only applicable to "Adminstrative Procedures" as is stated in the lesson plan.

Plant procedures (not including "A" Procedures) only require answer key responses 1, 2a, and 3.

NRC Resolution: This verbal and written comment was accepted and no deduction was taken for this item.

4.04 Facility Comment: Answer key indicating quarterly exposure require-ments for whole body rad. exposure guides should not include "1.25 mr/qtr if estimated current qtr exposure is on file". This part of answer is not applicable to R0.

A LGS R0 qtrly dose record is always known, current, and on file.

As to the question statement "Also include any special requirements for exceeding these guidelines", an alternate answer to this portion of the question may be the Emergency Plan dose extensions up to 75 rem, 25 rem, and 5 rem for life saving, equipment, and protection of the public safety, respectively.

NRC Resolution: Written comment accepted and points adjusted for grading.

4.08 Facility Comment: Question as stated may mislead the candidates.

Regardless of the outcome of water surveyed in SDV as either release-able or not, Steps 1, 2, and 3 of this procedure (GP-11) are always performed. However, Step 4 must not be completed prior to the SDV survey.

LGS operators are consistently trained in this manne.-

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.

v

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b R0 Examination

NRC Resolution: Writen comment noted.

Candidates were not confused by the wording of the question.

v t e Inspection Report - Limerick - 1986024
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 \|
1986 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 7(0) 8(0) 9(0) 10(0) 11(0) 12(0) 13(0) 14(0) 15(0) 16(0) 17(0) 18(0) 19(0) 20(0) 21(0) 22(0) 23(0) 24(0) 25(0) 26(0) 27(0) 99(0) 7(0) 9(0) 10(0) 11(0) 12(0) 13(0) 14(0) 15(0) 16(0) 17(0) 18(0) 19(0) 20(0) 21(0) 99(0) Security Operator Exams Emergency Planning Other

IR 05000352/1986024

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