Text

Exam Rept 50-335/OL-87-01 on 870420-22.Exam Results:Six Reactor Operators & Two Senior Reactor Operators Passed Oral & Written Exams.Master Exams Encl
	ML20215K895
Person / Time
Site:	Saint Lucie
Issue date:	06/17/1987
From:	Lawyer L, Munro J; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML20215K868	List: ML20215K866; ML20215K895;
References
	50-355-OL-87-01, 50-355-OL-87-1, NUDOCS 8706250481
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ENCLOSURE 1 EXAMINATION REPORT 335/0L-87-01 Faci _lity Licensee:

Florida Power and Light Company P. O. Box 14000 l

Juno' Beach, FL 33408 Facility Name:

St. Lucie Nuclear Plant I

1 Facility Docket No.:

50-335 and 50-389 I

-Written and oral. examinations were administered at St. Lucie Nuclear Plant near 1

Jensen Beach, Florida.

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Chief Examiner: @ h l D b A -

6/n /s7 Lawre L. Lawyer' Date Signed 7//7 hpproved by:

/h./

6 John F. Munro, Section Chief Date Signed i

Suninary:

Examinations on April 20-22, 1987.

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-Oral and written examinations were administered to eight candidates; all of whom passed.

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Based on the results described above, six of six R0's passed and two of two SR0's passed.

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k B706250481 870622 PDR ADOCK 05000335-V PDR l-

i REPORT DETAILS 1.

Facility Employees Contacted:

M. Shepherd, Operations Training Supervisor i

J. Spodick, License Training Coordinator

D. Sager, Plant Manager

G. Boissy, Prospective Plant Manager

'*J. Couture, Lead RC0 Instructor

K. N. Harris, Site Vice President l

T. Dillard, Plant Maintenance Superintendent

Attended Exit Meeting i

2.

Examiners:

Dr. Joe Upton

L. Lawyer R.Crlenjak(ResidentInspector) i

Chief Examiner 3.

Examination Review Meeting At the conclusion of the written examinations, the examiners provided your staff with -a copy of the written examination and answer key for review.

The comments made by the facility reviewers are included as Enclosure 3 to.

this report and the NRC Resolutions to these comments are listed below.

t Question 1.09 Agreed. The answer was changed accordingly.

Question 1.11 Disagree.

The reference material submitted is discussing a LOCA with resulting depressurization; Question 1.11 clearly asks for which one will be the first. Thus restricted, there can only be one answer,-

d, even though there may later be depressurization due to a small break LOCA.

No change required.

Question 1.17 Disagree.

The NRC does not require less knowledge of J

an R0 candidate than an R0 licensee during a requalifi-cation examination.

Also, the R0/SR0 distinction suggested is not reflected in the St. Lucie objective or KA catalogue.

See referenced learning objective and K/A 002/000 K5.15 (4.2/4.6) " reason for maintaining subcooled margin during natural circulation."

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' Question 1.20 Disagree.

Reactive load would only be decreased if

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the initial capacitive load were greater -than the "large induction motor" reactive load.

A highly unlikely and not operationally oriented assumption.

No change required.

Questian'2.17 Agreed.

The answer key was changed accordingly.

Question.2.22a Agreed. The answer key was changed accordingly.

Question 3.10 Disagree.

Logic time delays'.in some cases are too important to permit blanket prohibition of their use.

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-No change required.

Question 3.15

. Agreed. The question was deleted.

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Question 3.17 Agreed.

The additional correct answers were added to-the answer key.

Question 4.19a Agreed. The answer key was changed accordingly.

Question.4.19b Disagree.

The necessity of immediate action to implement the emergency plan and notify personnel is i

not dependent on which unit the accident occurs on.

_I However, since these two items are under the title

" subsequent action" on Unit 2, points will not be deducted if the Unit 2 assumption is stated.

-The answer key was changed accordingly.

Question 4.23 Disagree.

The suggested listing are items that are parameters (see LT 826, p. 5) as opposed to characteristics.

The wording utilized was taken from the learning objective and page 2 of the reference.

No. change required, however, equivalent answers including changes to appropriate parameters will be i

considered.

l Question 5.11 Same as 1.09.

~ Question 6.14a Agreed. The answer key was changed accordingly.

Question 7.24 Same as 4.23 4.-

Exit Meeting At the conclusion of the site visit, the examiners met with i

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' representatives of the plant staff to discuss the results of the examination.

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There were no generic weaknesses noted during the oral examination.

The material supplied for the written exam was lacking in several respects.

Specific items of note are as follows:

1.

NoLearningObjectives(L0s)forprocedures.

2.

' No index or tabs for. Plant Specific Reactor Theory (PSRT) and Plant

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Specific Thermodynamics, Heat Transfer, and Fluid Flow (PSTHTFF).

No System Description (SD)pecification sections.

No L0s on some Technical S 3.

on service water.

'4.

5.

Did not receive any lesson plans.

6.

Lecture' outline 85, Instrument Air not provided 7.

. No SD on instrument or service air.

8.

No normal operating procedures for instrument or service air.

9.

No L0s on off normal operating procedures.

10.

SD 104, Main and Extraction Steam missing.

11.

SD 117, Auxiliary Feedwater missing.

12.

SD 123,~Cire water missing.

13.

Administrative procedures such as tagging do not have L0s.

14.

AP 10122, In Plant Equipment Clearance Orders not provided.

15.

No SD on HP portable instruments.

The cooperation given to the examiners and the effort to ensure an atmosphere in the control room conducive to oral examinations was noted and appreciated.

The licensee did not identify as proprietary any of the material provided to or reviewed by the examiners.

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NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY:

_STt_LUglE_1h2___________

REACTOR TYPE:

_PWR-gE__________________

DATE ADMINISTERED _@Zl94/29________________

_ LAWYER _g@NpY___________

EXAMINER:

1 CANDIDATE:

INSIBUgIlgNg_IQ_C@NQIQ@lEl Use separate paper for th' answers.

Write answers on one side only.

Staple question sheet on tcp of the answer sheets.

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

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

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

% OF CATEGORY

% OF CANDIDATE'S CATEGORY

__YB6UE_ _IQIG6

___SCQBE___

_y@(UE__ ______________C@IEGQBy_____________

_29199__ _2Du19

________ 5.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND i

THERMODYNAMICS

_E9199__ _29199

________ 6.

PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION

_E9199__ _2Dz99

________ 7.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL

_E9A99__

29199

________ S.

ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 129199__

Totals Final Grade j

All work done on this examination is my own.

I have neither given nor received aid.

Candidate's Signature

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I fs-NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS i

During the administration of this examint, tion the f ollowing rules apply:

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 gnly to facilitate legible reproductions.

4.

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

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 Ogw page, write gnly gn gng sidg 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.

10. Skip at least thtgg lines between each answer.

11. Separate answer sheets from pad and place finished answer sheets face I

down en your desk or table.

12. Use abbreviations only if they are commonly used in facility litetatutg.

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.

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14. Show all calculations, methods, or assumptions used to obtain an answer i

to mathematical problems whether indicated in the question or not.

15. Partial credit may be given.

.nerefore, 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.

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17. You must sign the statement on the cover sheet that indicates that the I

work is your own and you have not received or been given assistance in compl etir,q the ex aminati on.

This must be done after the examination has been completed.

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18. When you complete yoar examination, you shall:

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Assemble your examination as f ollows:

(1)

Exam questions on top.

(2)

Exam aids - figures, tables, etc.

(3)

Answer pages including figures which are ~ art 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.

is 5.

THEORY OF NUCLEAR POWER PLANT OPERATION 1 FLUIDS 1_AND PAGE 2

ISERdQQyN8 digs QUESTION 5.01 (1.00)

Which one of the following statements regarding reactor pressure vessel pressure and temperature limits during heatup and cooldown is true?

c.

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility ref erence temperatures, RTNDT, at the end of the vessel's normal 40 year service life.

b.

Reactor operation and resultant fast neut.on (E >l MEV) irradiation can cause a decrease in the RTNDT.

c.

Thermal gradients during heatup produce tensile stresses at the inside of the vessel wall which add to the tensile stresses produced by internal pressure.

d.

The neutron spectra at the irr adiation samples and vessel inside radius are essentially identical.

QUESTION 5.02 (1.00)

Which of the following statements about burnable poisons is NOT true?

a.

Including burnable poison in the fuel affects the doppler coefficient.

b.

Including burnable poison in the fuel affects the moderator temperature coefficient, As core age increases burnable poison effects partially compensate c.

for fissile depletion of the fuel.

d.

As core age increases burnable poison ef f ects partially compensate for fission product buildup effects.

i QUESTION 5.03 (1.00)

Which of the following flow measuring concepts is NOT based on the square root of a measured pressure difference?

c.

Rotameter b.

Orifice f

c.

Pitot tube l

d.

Venturi L

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l QUESTION 5.04 (1.00)

Which of the following statements about temperature measurement is NOT CORRECT?

I a.

i-i. the sensing wire of an RTD breaks the instrument will read offscale high.

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b.

If a thermocouple wire breaks the instrumer+ will read off scale low, c.

RTDs respond faster to temperature changes than thermocouples.

d.

The temperature range which can be measured by an RTD is smaller.than that for a thermocouple.

I QUESTION 5.05 (1.00)

J An RCS cooldown on natural circulation should NOT exceed 50 degF/hr hot leg down to 325 degF followed by a 20.4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . soak at-325 degF. What is the purpose of this requirement?

e.

to prevent exceeding brittle fracture limits of the reactor vessel, b.

to ensure adequate mixing of HPSI injection water with RC flow into the downcomer.

t c.

to ensure that adequate heat removal through the S/Gs is possible without having to increase level above 2/3.

i d.

to prevent rapid and-erratic changes in pressurizer l evel which could 2

lead to. void formation in the vessel head..

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1 QUESTION 5.06 (1.00)

Figure.292 shows a graph of pump laws, i.e.,

the relationship between pump speed and other-pump parameters.

Which one of the following parameters is represented by curve 2?

1 a.

Power.

b.

Flow rate.

c.

Voltage.

d.

Di sch,arge pressure.

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FLUIDS _3Np PAGE 4

5.

THEORY OF NUCLEAR POWER PLANT OPERATION 2 2

ISEBd99XN8b]CS QUESTION -5.07 (1.00)

Which of the following is NOT one of the conditions necessary for brittle fracture?

a.

plastic def ormation at or below the yield point 1

b.

temperature at or below the NDTT j

c.

nominal tensile stress level

.d.

flaw such as a crack present QUESTION 5.08 (1.00) l Which of the f ollowing will NOT change over core lif e?

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a.

The TS minimum shutdown margin I

b.

The shutdown boron concentration c.

Differential boron worth d.

Doppler deficit e.

Peak Sm worth after SD from full power

-QUESTION 5.09 (1.00) j i

Which one of the following is NOT one of the ECCS acceptance criteria as j

specified in 10 CFR 50.46?

a.

Peak cladding temperature shall not exceed 2200 deg F.

j b.

Long term cooling is available to remove decay heat.

j c.

Total oxidation of cladding shall nowhere exceed 17% of the total cladding thickness before oxidation.

d.

Total amount of hydrogen generated shall not exceed 1 P. o f the maximum hypothetical amount.

s.

The DNB and linear heat rate limits shall be maintained =or<.1.28 and

=or > 13. 0 kw/f t respecti vel y.

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a is 5 __IUEgBy_9E_NUCLg86_EQWgS_E69NI_gfEggIJgN _E6UJpp,g_8Np PAGE S

1 ISE6dggyN801C@

QUESTION 5.10 (1.00)

In the StLucie reactors, the moderator temperature coefficient (MTC) varies with certain plant conditions.

The MTC: Echoose one correct answer]

c.

Becomes more negative as baron concentration is increased.

b, Varies due to temperature (Tavg) because of the non-linear density changes of water as temperature changes.

c.

Causes axial flux distribution to be tilted toward the top of the core at the beginning of life, d.

Would be expected to introduce a large negative reactivity in the event of a major steam line break.

c.

Is no permitted by Technical Specifications to be positive in any normal plant operating mode.

QUESTION S.11 (1.00)

An important consideration in plant operations is the quantity of effluent produced when changing RCS boron concentration by 100 ppm at any time in core life.

Figure 10 shows four plots of the quantity of effluent produced (system volumes) vs percent of core life.

Which plot is correct?

a.

1 b.

2 c.

3 d.

4 1

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THEORY OF NUCLEAR POWER PLANT OPERATION _FLUIQ@2_ANp' PAGE 6

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i QUESTION 5.12 (1.00)

Au Keff approaches unity, for the same increase in Keff, a greater length I

l of time is required to reach the new equilibrium level.

This is a charac-tsristic oft a.-

Moderator effects on neutron leakage.

b.

Increasing population of delayed neutrons.

j c.

Reactivity change on power increases.

d.

Subcritical multiplication.

i QUESTION 5.13 (1.00) a i

Which one of the'following sources can potentially introduce the largegt (in standard cubic f eet) amount of non-condensible gas into the RCS?

a.

Zirc-water reaction.

b.

Safety injection tanks.

c.

Pressurizer vapor space.

l d.

100% failed fuel.

f QUESTION 5.14 (1.00)

Startup of a centrifugal pump with the discharge valve shut is best characterized by which one of the following?

a.

low motor current and low discharge pressure b.

low motor current and high discharge pressure c.

high motor current and low discharge pressure d.

high motor current and high discharge pressure 1

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" THEORY OF' NUCLEAR POWER PLANT OPERATION _FLUIDg1_8Np PAGE 7

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QUESTION 5.15 (1.00)

.Concerning'the-behavior of samarium-149, in the reactor, which one of the j

following statements is CORRECT?

l a..

Most of the Sm produced comes directly from-fission.

i b.

Most of the removal of Sm is by radioactive decay, c.

Sm reactivity is in' dependent of flux once it has reached equilibrium.

d.-

Equilibrium Sm is reached about 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months after the initial startup of the reactor, e.

Sm reactivity is independent of Sm concentration.

l QUESTION 5.16 (1.00)

.One of the characteristics of water is that it will hold gases dissolved in I

solution.

Which of the following-will INCREASE the concentration of.

I dissolved gases in a quantity of water?

1 e.

Increasing the pressure and/or lowering the temperature.

l b.

Decreasing the pressure and/or lowering the temperature.

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c.

Increasing the pressure and/or raising the temperature.

d.

Decreasing the pressure and/or raising the temperature.

QUESTION 5.17 (1.00) 1 You are plotting an inverse multiplication plot (1/m) during fuel loading.

The normal _ shape of the 1/m plot seen at PSL will look like ______ in Figure 298.

3.

b..

C.

d.

c

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QUESTION 5.18 (1.00) j The unit i linear power range safety channels all use boron lined detectors. Which of'the'following is the CORRECT reason for use of boron?

l a..

It reduces the critical volume (size) of the detector.

Detectors which rely solely on gas ionization by neutrons are much larger.

b.

Neutrons do not carry a net electric charge.

Neutron detection must i

depend upon their interaction with. target nuclei.

c._

Ionization of the boron by neutrons is much more responsive end

. accurate than other ionizations such as neutron-rhodium used in the in-core detectors.

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.d.

The neutron-boron reaction produces beta particles which.have a much higher specific ionization than neutrons alone.

QUESTION S.19.

(1.00)

Salect the CORRECT statement concerning pump cavitation.

a.

Vapor bubbles are formed when the enthalpy difference between the l

pump discharge and pump suction exceeds the latent heat of i

vaporization.

b.

When the vapor bubbles enter a higher pressure region, the bubbles collapse which produces high pressure pulses or shock waves.

c.

Vapor cavities (bubbles) are produced when the localized pressure exceeds the vapor pressure at the existing temperature.

d.

As the vapor bubbles are discharged from the pump they impinge on downstream piping and valves causing water hammer..

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- THEORY OF NUCLEAR POWER _ PLANT _gPgRATIQN _FLUIgg2_AN9 PAGE 9

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4 QUESTION 5.20 (1.00)

When synchronizing the generator to the grid, OP2-OO30124, " Turbine' start up zero to full load" directs the operator to regulate generator speed to slowly rotate the synchroscope in the fast (cl oc kwi se) direction.

Which choice below CORRECTLY gives the two parameters that the synchroscope is indicating?

a.

Current and voltage differences b.

Voltage and frequency differences c.

Frequency and phase differences

?

d.

Phase and resistance differences o.

Resistance and current differences QUESTION 5.21 (1.00)

Which of the following is a true statement concerning radioactive decay?

R2 member the atomic number is the number of protons and the mass number is the number of neutrons plus protons.

.a.

When an element decays by beta emission, the new element will have increased in atomic number by one and the mass number will remain the same as the original element.

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b.

When an element decays by alpha emission, the new element will have decreased in atomic number and mass number by two, from the original element.

c.

When.an element decays by neutron emissien, the new element will have increased in atomic number by one and. decreased in mass number by one, j

from the original element.

d.

When an element decays by gamma emission, the new element will have increased in atomic number by one and the mass number will remain the same as the original.

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I r 's.5." THEORY OF NUCLEAR POWER _ PLANT _gPgRATJgN _ FLU 19 2_AND PAGE. 10 1

ISEBU99XN8bigS QUESTION 5.22 (1.00)

Which one of the f ollowing correctly describes the response of core deltaT if.the RCS coolant flow rate is reduced while the steam system is controlling to maintain 80/. electrical output?

Unit i has been operating at 80% of full power, steady state, for 10 days.

All control rods.are fully withdrawn (ARO).

I a.

Delta T will remain the same since the turbine / generator output power j

I has not changed.

b.

Delta T will remain the same but nuclear power will increase to provide the same megawatt thermal output.

c.

Delta T will increase but nuclear power remains the same to provide the same megawatt thermal output.

d.

Delta T will increase and nuclear power must increase to provide more steam flow to the turbine..

QUESTION 5.23 (1.00)

One isotope of plutonium, Pu239, can be found in the nuclear reactor core of St Lucie unit 1 because a.

of the non fission absorption of a thermal or epithermal neutron by U238 nuclei.

b.

of the doppler broadening of neutron absorption peaks for U235 nuclei.

c.

of the fissioning of U238 nuclei by fast neutrons.

d.

of the non fission absorption of a thermal neutron by U235 nuclei, e.

of the prompt neutron emission of Pu240 nuclei.

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5a_'_ISE96Y_9E_NUg6E88_Egyg6_E6@NI_gEEBSIJgN _ELUlpg2_9ND PAGE 11 1

ISE5099XU8dlg@

QUESTION 5.24 (1.00) a The reactor is critical at 10E-3 */. power when a S/G Atmospheric dump valve-fails open.

. Assuming BOL conditions, no rod motion, and no reactor trip, J

choose the answer below that best describes the values of Tavg and nuclear power for the resulting new steady state.

(POAH = point of adding haat).

a.

Final Tavg greater than initial Tavg, Final power above POAH.

4 b.

Final Tavg greater than initial Tavg, Final power at POAH, i

i c.

Final Tavg less-than initial Tavg, Final power at POAH.

d.

Final Tavg less than initial Tavg, Final power above POAH.

j QUESTION 5.25 (1.00)

-Initially, one centrifugal charging pump is in operation when a second centrifugal charging pump in parallel with the first pump is also put into operation.

Which statement below correctly describes the effect on system volumetric flow rate and system head loss?

i a.

Higher flow rate, higher head loss b.

Same flow rate, higher head loss c.

Higher flow rate, same head loss d.

Same flow rate, same head loss QUESTION 5.26-(1.00)

Which one of the curves on the attached figure 1172 correctly represents the xenon concentration for the given power history?

t f

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1 QUESTION 5.27-(1.00) 3 Which of the following will cause plant efficiency to increase?

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e.

' Total S/G blowdown is changed from 30 gpm to 40 gpm.

b.

Steam quality changes from 99.7% to 99.9%.

i c.

Level increase to higher than normal in a feedwater heater.

d.

Absolute condenser pressure changes from 1.0 psi to 1.5 psi.

QUESTION 5.28 (1.00) l Which one of the following must the main condenser remove the most heat from to condense?.(assume steam is of equal quality) 1

a. one' pound of steam at O psia b.

one pound of steam at 300 psia c.

two pounds of steam at 600 psia d.

two pounds of steam at 1200 psia QUESTION 5.29 (1.00)

Figure 1390A depicts flux and xenon concentration for varying lengths of time af ter control rods are inserted a significant distance into f ull power equilibrium flux at your reactor. Which one of the folowing plots of flux and xenon concentration (1-4 in B) would you expect at 28 hours3.240741e-4 days 0.00778 hours 4.62963e-5 weeks 1.0654e-5 months after the rod insertion?

j e.

1 b.

2 c.

3 d.

4 l

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THEORY-OF NUCLEAR POWER PLANT OPERATION,FLUIDg1_AND PAGE. 13

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-QUESTION 5.30

( 1. 00 ) '

-Match the correct letter (sL with the correct number.

1.

Quality A.

. Fuel rod power per unit length.

2.. 'Enthalpy

-B.

Energy' conter.t in' BTU /lbm.

3.-

Latent Heat of C.

Ratio of. steam mass to total mass Vaporization of steam and water.

4.

DNBR D.

Ratio of steam volume to total volume of steam and water.

5.

Linear Heat Rate E.

Heat. required'to raise' temperature of water. 1 deg F.

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F.

Heat required to convert i lbm-of water to.

1 lbm of steam.

G.

Ratio of actual heat flux to' local heat flux required to keep DNB greater than 1.30.

H.

Ratio of heat flux required to reach DNB to actual local heat f l ux.-

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6:__E69NI_SYgIgdS_Qgg]GN _CgNIB962_@Np_JNgIBUdgNI@IlgN PAGE 14 1

QUESTION 6.01 (1.00)

'Which one of the following reactor trips DOES NOT receive an input from the core protection calculators?

l a.

Thermal margin / low pressure b.

Local power density c.

High rate of change of power d.

Variable high-power QUESTION 6.02 (1.00)

Which one of the following statements correctly describes how to reset the MECHANICAL overspeed on the turbine driven AFW pump?

a.

It must be reset locally, after driving the limitorque to the open position, to relatch the linkage.

b.

It must be reset locally by using a lever to relatch'the linkage.

i c.-

It will reset automatically as turbine speed decreases below a pre-determined setpoint.

d.

It may be reset locally, but can also be reset from a switch on RTGB 102'in the control room.

i QUESTION 6.03 (1.00)

Which one of the following statements concerning the LPSI/ shutdown cooling system is correct?

a.

The hot leg suction valves to the LPSI pumps automatically close at a l

decreasing RCS pressure of 500 psig, b.

The LPSI pumps can be used in fill and makeup to the SITS with RCS pressure less than 1500 psig.

c.

When using the system for shutdown cooling, purification of the RCS coolant using the CVCS demineralizers is not available.

d.

The LPSI pumps start automatically on SIAS and are stopped i

automatically when RCS temperature reaches 325 degrees.

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.. st__Pb8NI_gl@Igdg_QgSJgN _CgNISQ61,8NQ,JNgIBUDEUISI]90 PAGE 15 2

QUESTION 6.04 (1.00)

The reactor cavity cooling system consists of two full capacity fans with the fan that is _______. (choose the correct phrase below) a.

in standby started automatically: if the reactor cavity ambient temperature, as measured by either of the two TCs located near the excore neutron detectors, alarms at 150 degF.

l b.

in standby started automatically on a CIAS.

c.

in standby started automatically after a ten second delay on LO flow.

d.

in operation restarted automatically after a loss of offsite power as demanded by the D/G shutdown sequencer.

)

QUESTION 6.05 (1.00)

The shutdown cooling heat exchangers are used to remove heat during cooldowns a.

if the pressure is =or<

1500 psia and the temperature is =or< SOOdegF.

b.

if the CCW inlet temperature is =or< 55 degF.

c.'

with the cooldown rate controlled by thrcttling the SDC return valves to the LPSI headers.

d.

with the cooldown rate controlled by throttling the CCW flow control valves to the SDC heat exchangers.

QUESTION 6.06 (1.00)

The A and B trains of the 125 VDC system are normally lined up such thats a.

the C train is powered from either of the A or B buses via the swing bus AB in unit 2.

b.

the swing bus AB is powered from the A side train in unit 1.

c.

the battery charger AB is connected to the A side train in unit 2.

d.

the swing bus AB is powered from the A side train in unit 2.

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QUESTION 6.07 (1.00)

The blowdown line of steam generator.1A hast a.

a temperature measuring element which, together with a CIAS or blowdown high radiation signal, closes an air actuated containment isolation valve inside containment.

b.

a temperature measuring element downstream of the containment isolation valve which can provide inf ormation to determine S/G temperatures for RCP starting requirements.

c.

a flowrate sensor which controls both the inside and outside of containment isolation valves to provide protection in the event of a high energy'line break.

d.

A flowrate sensor which controls the pressure control valve located Just upstream of the flowrate sensor.

QUESTION 6.08 (2.00)

Indicate whether each of the following statements concerning the containment. spray system apply to unit 1,

unit 2 or both, a.

When the RWT decreases to its low level setpoint, a RAS signal is generated which shifts the containment spray pump's suction from RWT

)

to the containment sump.

j b.

The containment spray system can be lined up to supply all or part of the suction to the HPSI pumps.

c.

The containment spray system utilizes hydrazine in the spray as an iodine removal agent.

l d.

The containment spray pumps are provided cooling water by the CCW l

system.

i t

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QUESTION 6.09 (1.50) i Indicate whether each of the following statements regarding the i

overpressure mitigation system (OMS) applies to UNIT 1, UNIT 2 or to BOTH UNITS.

a)

There are FOUR pressure comparators (PC's) that are used as inputs to Channel A and Channel B for low pressure protection, one channel per PORV.

b)

The PORV low pressure-protection system is designed to prevent i

exceeding the Tech Spec' PTS limits if an inadvertant Safety Injection f

by two HPSI pumps and three charging pumps.

l l

c)-

A temperature interlock prevents PORV actuation due to a low pressure relief open signal if RCS temperature is > 320 degrees F.

l l

QUESTION 6.10

(.75)

Fill in the blanks below to correctly complete the statement regarding SBCS operationt A Quick Opening Signal is generated subsequent to a ____________

if both secondary pressure is > ______ psia and Tavg > _____ deg F.

I l

QUESTION 6.11 (1.00)

What happens to the Main Steam Isolation Valves on a loss of electric power and a loss of air supply (treat each case individually)?

QUESTION 6.12 (1.00)

Describe how many fans and at what speed they will run for BOTH UNITS when f.

an SIAS starts the Containment Cooling System.

i

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i

QUESTION 6.13 (2.00)

What automatic actions (other than alarms) occur, if any, when the f ollowing process radiation monitors exceed their setpoints?

Consider each of the four monitors separately.

a.

Liquid discharge monitor b.

Steam generator blowdown monitor

-c.

Letdown process monitor d..

Component cooling water monitor QUESTION 6.14 (3.00)

Consider the excore neutron detectors and their instrumentation for unit 1.

a.

What are two indications of a failed section of a dual section UIC safety channel?

b.

LIST three control / protective functions of the wide range channels.

c.

-What are two functions (inputs to other systems) of the control channels?

QUESTION 6.15-(1.50)

List the five CEA motion inhibiting signals that are bypassed when the UNIT 1 Motion Inhibit Bypass button is depressed.

l QUESTION 6.16 (1.75)

Describe the two operational modes of the UNIT 2 Main Feedwater Isolation Valves.

Include a description of the motive force for the valves, number of valves per feedline, their response in both normal and accident conditions and the manner or method in which speed is controlled.

QUESTION 6.17 (1.00)

Dascribe how the Heated Junction Thermocouple (HJTC) System detects a collapsed liquid level above the core.

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QUESTION 6.18 (1.00)

The AFAS actuates auxiliary feedwater to a steam generator on low level unless the steam generator or its associated auxiliary. f eedwater supply header have been identified as being ruptured.

What criteria determines whether such a ' rupture exists?

(i. e.

What auto feature prevents AFW supply to the ruptured steam generator)

QUESTION 6.19-(2.00)

Explain in general the method utilized in fuel dry-sipping.

Include flow paths of both of the two main fluids (gases or liquids) involved, name of these fluids and how a cladding defect is detected.

QUESTION 6.20 (1.00) l During initial condensate pump startup with the system depressurized, the procedure requires that the main condensate recirc regulator controller setpoint be reduced to 1000 gpm. (see step 9 on figure 1275 attached)

Explain why this is done by stating the two reasons given in the procedure.

QUESTION 6.21 (1.00)

When performing a-fill and vent operation on the RCS in accordance with OP 0120020, why must the fill rate be maintained below 500_gpm?

Indicate the limiting component (s) that is/are the basis for this precaution.

QUESTION 6.22 (2.50) a)

What is the purpose of the Power Trip Test Interlock?

b)

What trips are automatically bypassed by the RPS?

c)

How is the Zero Power Mode Bypass initiated and when is this bypass automatically removed?

Note any Ul/U2 differences.

i

(***** END OF CATEGORY 06

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1 PAGE 20 PROCEDU8gg NORMAL _8pNQRMAL3_EMg89ENgy_AND.

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2 689196901GOL_GQNIBQL

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I

.'QUESTIOl' 7.01 (1.00).

r Which one of the f ollowing is NOT a symptom of Sondenser-tube leak as centained.in off-normal OP-1-610030, " Secondary Chemistry -Off Normal"?

a.

' Increase'in cation conductivity'in the condensate header b. =.

1 A S/G specific conductivity increasing 1

c.

Increase in.hotwell' chloride monitor d.

Hotwell quadre.rtt cation conductivity increasing or alarming QUESTION -7.02 (1.00) l i

IWhich one of the f ollowing' is NOT 'a' correct caution during fi1~1ing and v'enting of the RCS?

a.

Do not attempt to vent if the RCS temperature is above 200-degF.

b.

Do not allow RCS pressure-to decrease below atmoss,heric, c.

Do not install the head on the reactor vessel unless two pressuricar l

safety valves are operable.

d..

Do not operate a RCP unti1 ~ the RCS has been pressurized and major air pockets removed.'

i

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. PROCEDURES - NORMAL _ABNQRMAL,EMER@EN9Y,AN9 PAGE 21 2

1 88DI96991986_CgNI696 QUESTION

'7.03

'(1.00)

Which one of the f ollowing is NOT a requirement for increasing RCS pressure to greater than 1750 psi ~g on a unit 2 reactor plant hettup from cold to hot standby?

I a.

Two independent containment spray systems shall be operable with each spray capable of taking suction from the RWT on a CSAS and automatically transferring' suction to the containment sump onLa RAS.

b.

At least three safety injection tanks operable wi'th each tank

.containing 74-94% borated water with a concentration between

-1720-2100 ppm, having nitrogen cover pressure between'235-650 psig and its isolation valve open.

Two' independent ECCS subsystems shall be operable consisting of.at c.

s least one-HPSI' pump, one LPSI. pump and one charging pump, d.-

RCS dissolved oxygen concentration shall be <0.1 ppm.

QUESTION 7.04 (1.00)

The refueling machine bridge drive has interlocks that make it inoperative in 'ei ther direct Jond Which one of the following is NOT a correct interlock?

3 a.

The mastibumper is actuated.

b '.

The grapple is in or below' the upper grapple operate zone.

'l c.

The hoist is being operated, d.

The spreader is retracted.

l J

QUESTION 7.05 (1.00) 1 IfftheJHPSI pumps were running during an excess steam demand event at

-i unit 1, the 'HPSI pumps snould NOT be throttled or stopped if:

)

J a.-

the RC.is 40 degF subcooled.

i

.b.

the pressurizer level is 20%.

both 'Eeum-generator levels are 50%.

s c..

d.

~the reactor vessel level is 60%.

]

l

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PROCEDUAES ~ NORMALn !) nog Alc ETAERGENCY AND e'PAGE 22 A

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E QUESTION 7.06 (1.00)

/

)

Whic'h one ni tha tc) lowing' is correctly stated regarding a permissible dose to ar, operator lw/7 form NRC-4) in +. restricted area as specified in

-100FR207,

- h.. ' Under non-accident conditionr, the operator is permitted to receive'no~

I more than 7 1/2 rens per calencer quarter to each hand and to each i

foot.

b.

Under non'-accident conditions, the operator is permitted to receive no

.I more than.1 1/4 rents of beta per calendar qvpcter to the lens of the

'Y

I c.

Under accident or emergency conditions, the operator is, permitted to receive up to 23 r'em once in h, lifetime er.posure.

.j

/

emE;fgency conditions, the operator is permitted to

.d.

Under accident or receive up to 190 rem once 5,n a. lifetime opposure.

' l l

/.

}

i

)

/

QUESTION 7.07 (1.005 LJhich one of the follo.ving "Radi ati on Ar ea" ceH ni tions" - means die same as thht ejiven in HP-2, "FP&L health physics manual"?

a.

any. area where the dose rate exceeds five mrnq/hr or where, in any five consecutive day period,/ exceeds 100 mrv.q/hr at any time.

i I

b.' ' any accessible area where a major portion of th'e body could exceed a dose of five mrem in any one hour, or 'in any five consecuti ve days a 6

dose in excess of '100 mee.n.

t cs any accessible area where any portion'of the body could exceed a dose

.I

/

rate of five mrem /hr or where, in any five consecutive day period,

[

could f,xceed a dose rate'of 100 mrem /hr at any time.

y y f

d-any area where the Gose rate to any por tion of the body)could exceed a F

dose of five mren in any one hour, or in any f ; /e consecuti.ve days a

' dose in excess of 100 mrem.

/-

1, c

i f

/

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1 5891969G1906 99NIBQL I

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QUESTION 7.08 (1.00) if a steam generator tube rupture occurs on Unit 1 and if all RCPs are f stopped, RCP. restart criteria CANNOT be met ifs (choose one) a.

the CCW has been lost for four minutes.

l b.

the RCS is 30 degF subcooled.

?, \\

f.

c.

.the pressurizer level is 40%.

j d.

the' unaffected steam generator level is 30% wide range.

' QUESTION 7.09 (1.00)

'The unit.1 hydrogen recombiner should always be plared in service when the j

hydrogen concentration in containment is in which one of the following ranges?

a.-

0.5% to 3.5%

b.

0.5% to 4.5%

c.

1.5% to 3.5%

d.

1.5% to 4.5%

e.

0.0% to 4.0%

QUESTION 7.10

. (1.00)

}

i Which one of the following locctions is where the "B" RCO reports to on a Con t r o'l Room Evacuation due'to inhabitability?

a.

Remote Shutdown Panel b.

El ec t r i c a'l Equipment Room, Peactor Auxiliary Building c.

Turbine Operating Level i

d.

LPSI Pump Room

)

i I

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689196991986 99NIBQL l

QUESTION 7.11 (1.00)

Which'one of the following is the RCP. trip strategy in EOP-5(Excess Steam Dsmand),' assuming CCW is maintained to the pumps, RCP operating limits are maintained and. Pressurizer pressure f alls to < 1300 psia?

c.

Trip all RCPs b.

Trip one RCP in each loop c.

Trip the two RCPs associated with the faulted S/G d.-

Trip one RCP in.the loop associated with the faulted S/G o.

Leave all four RCPs running I

QUESTION 7.12 (1.00)

While operating at-100% of full power on. unit 2, an increase in component cooling water (CCW) activi ty is observed on the indicators PLP-101 and PLP-

-102.. This is.followea by an indicated rise in the level of the CCW. surge tank.

Which one of the f ollowing must be eliminated as a potential source of the. activity in the CCW system?

I a.

Letdown heat exchanger b.-

Sample heat exchangers I

c.

LPSI pump seal cool ers j

d.

Reactor coolant pump seal coolers j

e..

HPSI pump seal coolers QUESTION 7.13 (1.00)

Which of the following is an indication that Natural Circulation is being established?-

a.

Loop Delta'T is equal to Full Power Delta T b.

That and Tcold are increasing slowly at the same rate I

c.

That is equal to Tsat associated with S/G pressure d.

Thot is stable and Tcold is slowly decreasing

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609196991C86_CgNIBQL

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QUESTION 7.14 (1.00) i A plant calorimetric was completed and other power level indications are to bn adjusted for agreement.

Which one of the following statements is correct concerning this adjustment?

a.

Thermal energy balance and delta T power are being adjusted to agree with nuclear power.

b.

Nuclear power and thermal energy balance are being adjusted to agree with delta T power.

c.

Delta T power and nuclear power are being adjusted to agree with thermal energy balance.

d.

Nuclear power is being adjusted to agree with thermal energy balance and delta T power.

QUESTION 7.15 (1.00)

Indicate whether each of the following statements regarding cooldown of the reactor plant are applicable to UNIT 1,

UNIT 2 or BOTH UNITS.

a)

The shutdown cooling systeci:SDC) shall NOT be placed in service until 4

the RCS pressure is < 265 psia and the temperature is < 325 degF b)

When the RCS temperature is < 500 degF and RCS pressure is < 1500 psig rack in the breakers and close the SI Tank discharge valves.

Rack the breakers out again once the valves are closed.

QUESTION 7.16 (1.50)

List all the conditions that must have been completed, so that a system that had been tagged out for maintenance, can be considered to be " ready for service".

QUESTION 7.17 (1.50)

List three conditions for which an inverse count rate ratio (ICRR) shall be plotted during an approach to criticality at St Lucie unit 2.

)

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PROCEDURES'- NORMAL _ABNgRMAL _EMERGENgY_ANQ 1

2 8991969G1986_ggNI696 QUESTION 7.18 (2.00) 4 Wh'ile operating at 100% offfull power, St Lucie unit 2 experiences a' slow.

1

~

i

'dacrease in the condenser vacuum.

List eight things that could be checked l

to determine the cause of the' decrease in the' condenser-vacuum.

l QUESTION 7.19 (2.00) l l

In off normal operating procedure 2-0120037, " Reactor coolant gas vent j

system off normal operation" there is listed the following symptom,...

'" Plant events have occurredtsuch as safety injection tank discharge,:

rapid RCS cooldown, or core uncovery events] that may result in-the presence of a gaseous void in the vessel head."

Regarding'this symptom, answer the followilig questions.or' fill in the l

blanks.as appropriate.

a.

What'is the primary source of gas'for the void following SIT di'scharge?

i b.

What is the primary source of gas for the void following rapid RCS cooldown?'

c..

What is the primary source of gas for the void following a-core.

uncovery event?

i d.

The procedure.contains steps to determine the " venting time period" in order to limit the concentration of what gas within containment?

QUESTION 7.20

(.50)

What safety function, in relation to RCS heat removal, can be impacted severly upon loss of FW if some type of feedwater cannot be restored?

l I

I QUESTION 7.21 (2.00) i What are the EXIT CONDITIONS from the Standard Post Trip Actions (SPTA)?

l 1

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QUESTION 7.22 (2.00)

. Answer the.following questions regarding RCP Trip Criteria:

a)

Which location is worse for a LOCA assuming that the RCPs are still operating, Hot Leg or Cold Leg?

(0.5) b)

Assuming that RCP operating limits are met, CCW is flowing to the RCPs and pressurizer pressure falls below 13OO psia, why do the SGTR (EOP-4) and Excess Steam Demand (EOP-5) procedures have you maintain 1 RCP i

operable in each loop?

Give three reasons.

(1.5)

QUESTION 7.23 (2.00) a)

-What condition is indicated if, while lowereing a fuel element into i

the core, a Dillon cell UNDERLOAD condition occurs?

i I

b)

List two situations where the Bridge Speed is NOT limited to slow speed.

i i

- QUESTION 7.24 (1.50) l Define an excess steam demand event (ESDE) by stating, in one or two words l

cach, its three main characteristics.

i j

1 1

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QUESTION 8.01 (1.00)

Which one of the f ollowing is NOT in an LCO at Unit 2?

f e.

The reactor coolant system lowest operating loop temperature (Tave) shall be =or> 515 degF when the reactor is in modes 1 or 2.

1 1

b.

The calculated value of TFR (total integrated radial peaking factor) f shall be limited to =or<

l.70.

j c.

Primary containment internal pressure shall be maintained between -0.7 and +2.4 psig.

j 1

4 d.

The pressurizer shall be operable with a minimun, water level of

=or>

d 27% indicated level and a maximum water l evel of

=or< 68% indicated level.

QUESTION 8.02 (1.00)

Which one of the following items is NOT required to be in the control room storage locker?

A.

SCBA b.

Coveralls c.

Neutron TLDs d.

Dosimeter charger QUESTION 8.03 (1.00)

Prior to review by the FRG, jumpers / lifted leads may be installed in safety related systems provided that three criteria are met.

Which one of the following is NOT one of those three criteria?

a.

The intent of the system operation would not be altered.

b.

The request has cleared the QC department.

c.

The request has been approved by two members of the plant staff at least one of whom holds an SRO license on the affected unit.

d.

The request will be reviewed by the FRG and approved by the plant manager within 14 days of the installation.

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' QUESTION 8.04 (1.00)

During a Unit i startup with the reactor at 2% power, one power range j

ncutron flux monitoring channel is found to be inoperable.

Which of the i

following statements is correct?

Refer to the attached Tech Specs. (figs.

981A-E) a.

Operation above 5% rated thermal power is not allowed until the inoperable channel is repaired and declared operable.

b.

If the inoperable channel is placed in a tripped condition and the other three channels are operable, you must verify compliance with the shutdown margin requiremen*:s of Tech Specs.

c.

If the inoperable channel is placed in a tripped condition and the j

other three channels are operable, operation to 100% rated thermal power may proceed only if all functioning units receiving an input from it are tripped (may not be bypassed).

i d.

The only restriction on proceeding to 100% rated thermal power are that the inoperable channel be placed in a bypassed or tripped j

condition; however, if it is bypassed it must be only f or surveillance testing and must be tripped within an hour.

QUESTION 8.05

( 1. 00')

i Which one of the following is correct regarding EP-310029E, " Duties of an Individual Who Discovers an_ Emergency Condition"?

l a.

All significant information, events and actions taken during the emergency period must be recorded by the emergency coordinator.

l 4

b.

Primary examples of emergency conditions covered by this procedure are l

tornado, acid and caustic spills, security threats and most other j

emergencies not involving radioactive material.

i c.

An individual who discovers an emergency condition shall notify the nuclear plant supervisor and the nuclear watch engineer, d.-

The individual should take any immediate action that will minimize the emergency (eg, close an upstream valve or l oc al l y stop machinery)

{

i except extinguishing fires; fires shall only be fought by the fire brigade members.

l I

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d

. QUESTION 8.06 (1.00)

On unit 2, with the reactor head closure bolts less than f ully tensioned, i

the boron concentration of'the RCS and refueling canal shall 'De maintained j

uniform and of sufficient concentration to ensure that the more restrictive of which of the following reactivity conditions is met?

l

_e.

Keff of 0.95 or baron concentration =or> 1720 ppm.

b.

Keff of 0.95 or baron concentration =or) 2100 ppm.

c.

Keff of 0.97 or. boron concentration =or) 1720 ppm.

d.

Keff of 0.97 or baron concentration =or> 2100 ppm.

4 QUESTION 8.07 (1.00)

)

l Immediate notification of NRC is required if the release of radioactive j

materials in concentrations which, if averaged over a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

{

would exceed ______ times the limits specified for such materials in Appendix "B",

Table II of 10CFR 20; ______ damage to property in excess of i,

Fill in the blanks with the correct information by choosing one J

of.the below combinations.

a.

500; or; 2,000 b.

500; and; 2,000 c.

5,000; or; 200,000 d.

5,000; and; 200,000 l

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-QUESTION O.08 (1.00)

Diesel Generator 1A, which supplies 4KV Bus 1A3 is INOPERABLE.

LPSIP B supplied by 4KV Bus 183 is INOPERABLE.

The Tech Specs for ECCS and AC' sources are attached as figures'1259a thru d.

Which statement is CORRECT concerning continued operation in mode 1?

a.

The Action Statements for both the LPSIP and the DG are applied indepen-dently, each must be restored to operable in 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

b.

Since the DG is required in Mode 4 and the LPSIP is not, the Unit must l

f be taken to Mode 4 within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

I c.

LCO 3.0.3 applies.

'd.

LCO 3.0.5 applies.

)

i I

i

. QUESTION 8.09 (1.00)

According to unit 1 Tech Spec 3.

7.1.5, the MSIVs are required to be operable in what modes?

(select one) a.

1,2,3 and 4 only I

b.

1,2, and 3 only j

c.

.1 and 2 only d.

All modes i

r i

s 1

l l

l

(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)

i

i 4.

8___epd]NJSIB8IJyE_E89CEgg6ES _CgNp]IJgN32_9NQ_LJdJI9IJgNg PAGE 32 2

1 i

QUE' TION 8.10 (1.00)

S Which one of the following actions is correct if unit 2 is in mode 1 with one: safety injection tank inoperable-due to the discharge valve being closed?

i t

s..

Open the valve within one hour AND be in at least hot standby within l'

the next six hours.

b.

Open the valve within.one hour OR be in at least hot standby within j

the newt six hours.

c.

Open the valve immediately AND be-in at least hot standby within one f

. hour.

d.

Open the valve immediately OR be in at least hot standby within one f

hour.

e.

Continue operating.

(Three accumulators operable meet the requirements of Tech Specs) l QUESTION 8.11 (1.00) i According.to EPIP 3100033E, which one of the following methods of release

)

rate determination' requires a HP backup survey 7 l

i a.

Post-LOCA monitors 1

b.

Chemistry grab sample c.

Effluent monitor d.

Containment hi range monitor i

l

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94_'_09dIU19I68IIME_P699EQU6gg1_ggyp]IJ9N!1_80p_6JUJI@l19N@

PAGE 33 l

QUESTION 8.12 (1.00)

Which one of the f ollowing is a correct definition of' " site -area" emergency?

a.

Major failure of systems that are needed to protect the public and potential for significant uncontrolled releases..

[

4 b.-

Major failure of systems that are needed to protect the public with potential ~f or limited uncontrolled releases.

c.

Substantial plant degradation and potential for significant uncontrolled releases.

j d.

Substantial plant degradation with potential for limited uncontrolled releases.

QUESTION 8.13 (1.00) 1 l

During plant emergencies which create the potential for release of radioactive material to-the environment, the-emergency coordinator or'his 3

designee is responsible for: (select one of the following) a.

ensuring that off site monitoring is performed.

l i

b.

informing the appropriate off site agencies concerning off site field monitoring data.

i c.

the direction of the off site monitoring team to ensure that their exposures are within 10 CFR 20 limits.

t d.

comparing estimated dose projections obtained from field measurements l

to those based on plant release conditions.

i QUESTION 8.14 (1.00)

The person directly responsible for maintaining the disconnected lead and temporary jumper log is thet(select one of the following) a.

NPS b.

ANPS c.

STA d.

RCO I

(***** CATEGORY 08 CONTINUED ON NEXT PAGE

- - - - )

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x

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e et[_epdINJSIBgIJVg_P8QggpOBEg1_ggNp))]QNS _6NQ_LJd]IBIJgN5 PAGE 34 1

QUESTION 8.15 (1.00)

In the event of_a medical emergency at St. Lucie, the duties and responsibilities of the emergency coordinator includes (select one) i 1

a.

-the noti ication during business hours of the duty call supervisor who should notify health physics, the operations superitendent and the emergency control officer.

b.

the notification during business hours of the plant manager who should notify health physics, the operations superintendent and the emergency control officer.

c.

ensuring that the victim's TLD and self reading dosimeter are transported with the victim to the Lawnwood Medical Center.

d.

ensuring that the victim is decontaminated onsite and then transported

-j to the Lawnwood Medical Center.

QUESTION O.16 (1.00)

Unit 2 is in Mode 3 during a Reactor startup with the following deficiencies:

One Main Steam Isolati on Valve is inoperable and closed f

One Motor Driven Aux. Feedwater Pump is inoperable l

l

-Which one of the following actions most. accurately details the allowances and/or limitations imposed by the Tech Specs in this instance?

i

a. Mode 3 must be maintained (Entry into Mode 4 acceptable)

{

b. Startup activities may continue; Mode 2 may be entered but not exceeded.

c. Startup and power operation into Mode 1 may be accomplished provided i

Mode 1 action statement for MSIV met.

i t

i

d. Startup activities may continue into Mode 2 provided subsequent restoration of the MDAFW pump to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

i

(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)

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9I_090]NJgIB8IJVE,P69CggUSEg1_C9N9]IJgNg2_8N9_LJd))SIJgNg PAGE 35 l

l

[

QUESTION 8.'17 (2.00) i

)

' Indicate whether the following statements are TRUE or FALSE with respect to i

)

plant ~ work orders (PWO).

a a.

If, in the opinion of the Nuclear Plant Supervisor or Assistant Nuclear Plant Supervisor, emergency work is required to restore equipment or. systems to normal operating conditions; the work can be f

performed under these emergency conditions without a complete PWO package.

b.

All modification, preventative and corrective maintenance not covered by blanket' work orders which is to be performed by Florida Power and Light" personnel at the St. Lucie plant, shall be authorized through the plant work order, form 1784.

c.

When a condition requiring modification or maintenance is known, it must be reported and documented by requesting that the Operations Supervisor initiate a PWO.

d.

If an AO observes that a relief valve in the CVCS requires replacement or. maintenance, he should report the situation by completing Section A l

of.the PWO and by checking the "yes box" in Section B under " Rad work permit required".

.i i

QUESTION 8.18 (1.00)

]

i Match one of the plant operating staff members given in column A to one emergency function given in column B.

j column A column B 1.

Watch engineer a.

emergency coordinator j

j 2.

Health physics supervisor b.

primary radiation team leader 3.

Nuclear plant supervisor c.

fire team leader 4.

Shift technical advisor d.

Team leader, first aid /decon 5.

Chemistry supervisor e.

Accident assessment and initial technical support necessary for repair or corrective action.

j

(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)

Sc io r ai_IOQd1NISI6@IlyE_P6QCEDU6ES_QQNDlIlgNS_@NQ_61dlI@IlgNS PAGE 36 z

t

. QUESTION 8.19 (2.00)

List-the accident / condition which the following reactor trips are designed to protect against.

a)

Pzr Pressure High (accident) g b)

S/G Pressure Low (accident) l c )-

Thermal Margin / Low Pressure (c ondi t i on )

d)

Local Power Density High (condi ti on )

t QUESTION 8.20 (1.00)

Fill-in the BLANK.with the appropriate Tech Spec definition.

~

" A ______ sh al l be the injection of a simulated signal into the channel as close to the primary sensor as practicable to verify operability including I

.al arm and or trip functions".

e l

I f

QUESTION 8.21

(.50)

Fill-in. the BLANK f or the f ollowing:

1 In accordance with 10 CFR 55 "if a licensee has.not been actively perform-ing the function of an Operator or Senior Operator for a period of months, or longer, he shall, prior to resumir.g activities licensed pursuant to this part, demonstrate to the Commission that.his knowledge l

and. understanding of f acility operations ar;d administration are j

satisfactory."

QUESTION 8.22 (1.50) l St Lucie Unit 2 TSs allow temporary changes to the fire protection program implementing procedures if three conditions are met.

State the three conditions.

I

(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)

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t QUESTICN 8.23' (2.00)

With_the. plant critical, the reactor operator reports to you'that Tavg is 510 degF.

(For a and b below note any unit 1/ unit 2 differences) a.'

What are the Tech Spec requirements?

(Include time limit)

{

b.-

What required action must you take if you cannot meet the time limit in "a" above?

(Include this time limit also) l QUESTION 8.24 (1.00) 1 Tech Specs defines _ Shutdown Margin as..." Shutdown Margin shall be the instantaneous amount of reactivity by which the Reactor is subcritical or l

would be subcritical from its present condition assuming l

STATE the assumptions made for the plant conditions which complete the definition of Shutdown Margin.

QUESTION 8.25 (3.00) i Define IDENTIFIED LEAKAGE by listing in detail the three types.

In this listing, state where the leakage is f rom, where it goes to, and its intervening path to the extent that this is done in the unit 2 Tech Specs.

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(***** END OF CATEGORY 08

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

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8.2 Initini Condensate Pump Start With System Depressurised.

hiitn Section 8.2 can be performed using the IC Condensate Pump as the initial pump.

However, Section 8.4 or 8.5 must be ~ performed prior to pump s tart to supply the 1C l

pump with electrical power.

1. Ensure normal system valve alignment has been completed as per Section 8.1.

2. Check hotwell level to be equal to or greater than normal opera ting level.

3. Check the ha twell makeup regula tor and sprays to be in service and an adequate supply of water available from the Condensate Storage Tank and/or the SGBTF Monitor Storage Tanks.

4. Ensure the TCW System is in seevice and lined up to the Condensa te Pump vith greetar than 12 GPM flow.

5.1.ine up ' the backup seal water supply to the Condensa te Pumps by I

perforning the following:

1 A.

Ensure transfer pump suction and discharge valves are open.

B.

Ensure Y-12558 (Makeup spray / seal wa ter supply) is open.

C.

Ensure the individual seal wa ter supply valves are open at each pumps V-12242 (IA Condensate Pump) l V-12245 (IB Condensate Pump)

V-12590 (1C Condensata Pump)

6. Ensure V-9218 (Condensa ta Pump / Steam Genera tor fill) is open to allow filling and pressurization of the feedwater sections of

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piping.

1

7. Close the discharge valve on the condensate pump to be started, !

then reopen the discharge valve just enough to crack it off the sea t.

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8. Verify the pump casing and suction strainer vents are open to i

the Condenser.

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9. Reduce the main condensa te recire regulator controller se tpoint to 1000 GPM.

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p 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1.1 As a minimum, the reactor protective instrumentation channels and bypasses of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in Table.3.3-2.

~

APPLICABILITY:' As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

I SURVEILLANCE REQUIREMENTS i

4. 3.1.1.1 Each reactor protective instrumentation channel shall be.

'/. '.(,

l demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL I

k CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-1.

i

4. 3.1.1. 2 The logic for the bypasses shall be demonstrated OPERABLE during the at power CHANNEL FUNCT.IONAL TEST of channels affected by bypass operation.

The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation, j

4.3.1.1.3 The REACT 0n TRIP SYSTEM RESPONSE TIME of each reactor trip i

function shall be demonstrated. to be within its limit at least once per

]

18 months. Each test shall include at least one channel per function 1

such that all channels are tested at least once everf ' times 18 months where N is the total number of redundant channels in a specific reactor

. trip function as shown in the " Total No. of Channels" column of Table

3. 3-1.'

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0 TABLE 3.3-1 (Continued)

TABLE NOTATION h'

j With the protective system trip breakers in the closed position and l

the CEA drive system capabh of CEA withdrawal.

The provisions of Specification 3.0.4 are not applicable.

(a) Trip may be bypassed below 1% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 1 1% of RATED l.

THERMAL POWER.

(b) Trip may be manually bypassed below 685 psig; bypass shall.be.;.,,

e automatically removed at or above 685 psig.

(c) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass 1

shall be automatically removed when THERMAL POWER is 1 5% of RATED THERMAL POWER.

Trip may be bypassed below 10'# and above 15% of RATED THERMAL (d)

POWER; hpass shall be automatically removed when THERMAL power-

~

is g 10 % or < 15% of RATED THERMAL POWER.

(e) Trip may be bypassed during testing pursuant to Special Test Excep-tion 3.10.3.

l (f) There shall be at least two decades of overlap between the Wide Range l

i Logarithmic Neutron Flux Monitoring Channels and the Power Range Neutron Flux Moni.toring Channels.

ACTION STATEMENTS ACTION l With the number of channels OPERASLE one less than l

required by the Minimum Channels OPERABLE requiregjertt, l

restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the protective system trip breakers.

With the number of OPERABLE channels one less than the ACTION 2 I

Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

The inoperable channel is placed in either the bypassed a.

or tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

For the purposes of testing and maintenar.ce, the inoperable channel may be bypassed for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from time of initial loss of OPERABILITY; however, the inoperable channel shall

,_. W g T h / h-then be either restored to OPERABLE status or placed in the tripped condition'.*

U ST. LUCIE - UNIT 1 3/4 3-4 Amendment No. 75, 27, 45 t

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' i.

6 TABLE 3.?-1 (Continued)

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ACTIONJTATEMENTS 1

?

b.

'Within ene hour, pil functional units receiving an inputtinn the iruperable channel are also pir.ced in the sam condition (either bypassed or trippwd, as applicable) as that required by a. above for the inoperable channel.

c.

The Minimum Channels OPERABLE requirement is met';

however, one additional channel may be bypassed for up to 48 shours while performing tests and maintenance

{

on than channet provided the other inoperable channel is placed in the tripped condition, j

?

ACTION 3 With the number of channa'is OPERABLE one less than reyuired l

J by the' Minimum Channels OPERABLE requirenent, verify I

compliance with the SliUTDOWN MARGIN requirements of 1

Specif.ication 3.1.1.1 or 3.1.1.2, as applicable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

ACTION 4 With'the number of channels OPERABLE one less than required l

)

by the Minimum Channels OPERABLE requirement, be in HOT

(

STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ;%2 wever, one channel may be bypassed i

for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for surye111ance testing per Specification 4.3.1.1.1.

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I 961 5 rop ST. LUCIE - UNIT 1 3/4'3-5 AmendNtntNo.

27 L,

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3/4 LIMITfNG CON _DITTONS FOR OPERAT10N AND SURVEf LLANCE REQUIREMENTS 3/4.0 APPLICABILITY, LIMITING CONDITION FOR OPERATION 1

3.0.1 Limiting Conditions for Operation and ACTION requirements shall be applicable during the OPERATIONAL MODES or other conditions specified for each specification.

3.0.2 Adherence to the requirements of the Limiting Condition for Operat' and/or associated ACTION within the specified time interval shall constitu compliance with the speci fication.

In the event the Limiting Condition fo-Operation is restored prior to expiration of the specified time interval, completion of the

" ION statement is not required.

3.0.3 In the event a Limiting Condition for Operation and/or associated ACTION requirements cannot be satisfied because of circumstances in excess of those addressed in the specification, the facility shall be placed in at least HOT STANOBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , at least HOT SHUTDOWN within the following 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months and at least COLD SHUT 00WN within the sLbsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless

corrective measures are completed that pemit operation under the permissible ACTION statements for the specified time N.m val as measured f rom initial discovery or until the reactor is placed in a :.00E in which the specification is not applicable.

Exceptions to these requirements shall be stated in the individual specifications.

3.0.4 Entry into an OPERA 110ltAL fiODE or other specified applicability condi-tion r.N11 not be made unless the conditions of the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION state-ments unless othenise er,eptnd.

This provision shall not prevent passage through OPERATIONAL HOUES as re. quired to comply with ACTION statements.

3.0.5 When a system, Jubsystem, train, component or device is determined to bo inoperable solely because its emergency power source is inoperable, or tolely because as norm 31 power source is inoperable, it may oe considered OPERABLE for the purpose of satisfying tha requirements of its applicable Limiting Condition for Operation, provided:

(1) its corresponding n6rmal or emergency (power source is OPERABLE; and (2) all of its redundant system (s),

subsystem s), train (s), component (s) and device (s) are OPERABLE, or likewise satisfy the requirements of this $pecification.

Unless both conditions (1)

I and (2) are satisfied, the unit shall ce placed in at least HOT STANDBY l

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , HOT SHUT 00WN within Oe following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and at least 1

COLO SHU100WN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

I l

This specification is not applicable in MODE 5 or 6.

I SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be applicable during the OPERATIONAL MODES cr other conditions specified for individual Limiting Conditions for I

11 Operation unless otherwise stated in an Individual Surveillance Requirement.

ST. LUCIE - UNIT 1 3/4 0-1 Amendment No. 40 Q

l1.b b

f

{l3

>jy I,

~jc --

,/,.i

,,, y

~n'

?

.r,,

a

,; ; s t s.,

<.7 i'. L. ( (; jjf '

'l

}

-f}/

,f l

(1

]t.; 1 <

l Ql%,.,,'jh!/d,,.5 EMERGENCYCORE_gGLIN'GSYSTEMS(ECC$,).

t t s lI '

jgg,Y_ INJECTION TI,\\fh d

I

' y J_,,,

f.

7

. LIM 1YItiR____ ONDITION FOR DPEREION C

t 3r Each reactor ct. dant c9 stem safety injection tanii'shall be OPERABLE 3.5.1 witp:

o The isolatten vrTE/open,

/'a' e

a.

b.

Between 10'70 and 1170 cubic feet of borated water,

,I A minimum boron concentration of(1720 PPM, and c.

d.

A nitrogen cover-pressure of ba'tweett,200' and 250 psig.-

j

,s APM.ICABILITY : MODES 1, 2 and 3.*

c ACTION:

9'With one, jafety injection tank inoperable; except as a result c.

.of 'a. closed isolation valve, restore the inoperable tank to

.0FIRABLE r.tatus within one ' hour or.be 'in HOT SHUT 00WN within J the next B hours.

b.

With one safety injection tank inocerable due~to thE isolation valve being closed, either immediately. open the it.olation valve or be in h0T STANDBY within cne hour at:d be in HOT SHUT 00WN l

I within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

i.

e SURVEILLANCE REQUIR!f g S'

_,'L_,_,

4.5.1 Each safety injection tank shall be demonstrated OPERABLE:

r a.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:

1.

Verifying the water level and nitrogen cover-pressure in the tanks, and 2.

Verifying that each safety injection tank isolation valve is open.

With pressurizer pressure 2,1750 psia.

. f d ' ~*

S ib ST. LUCIE<- UNIT-1 3/4 5-1

e

./ a 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION l

3.8.1.1 As a minimum, the following. A.C. electrical power sources shall be OPERABLE:

a.

Two physically independent circuits between the offsite trans-I mission network and the onsite Class 1E distribution system.

and b.

Two separate and independent' diesel generator sets each with:

1.

Engine-mounted fuel tanks containing a minimum of 152

{

gallons of fuel, j

2.

A separate fuel 3torage system containing a minimum of i

16,450 gallons of fuel, and 3.

A se arate fuel transfer pump.

f r

APPLICABILITY:

MODES 1, 2, 3 and 4.

ACTION:

a.

With either an offsite circuit or diesel generator set of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by I

performing Surveillance Requirements 4.8.1.1.1.a and 4.8.1.1.2.a.4 within one hour and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereaf ter; L

restore at least two offsite circuits and two diesel ger.erator j

sets to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With one offsite circuit and one diesel generator set of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.1.a and 4.8.1.1.2.a.4 within one hour and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereaf ter; restore at least one of the inoperable sources to OPERABLE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT 00WN within the following 30

(* } r j

i l1f9 1

.ST. LUCIE - UNIT 1 3/4 8-1 e

l

y.+

ELECTRICAL POWER SYSTEMS ACTION'(Continued) hours.

Restore at least'two offsite circuits and two diesel

-generator. sets to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from the time of initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , c.

With'two of the above required offsite A.C. circuits inoperable, demonstrate.the OPERABILITf of two diesel. generator sets by perfonning Surveillance Requirement 4.8.1.1.2.a.4 within one hour and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter, unless the aiesel generator sets are-already operating; restore at least one of the inoperable offsite sources to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With only one offsite source restored, restore at least two offsite circuits to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from time of initial loss or be in at least HOT STAN0BY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following.

30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , d.

With two of the above required diesel generator sets in-

' operable; demonstrate tha OPERABILITY of two offsite A.C.

1 circuits by performing Surveillance Requirement 4.8.1.1.1.a

(..

within one hour and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter;.

restore at least one of the inoperable diesel generator sets to OPERABLE status within 2 bours or be in at least HOT-STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Restore at least two diesel generator sets to OPERA 5LE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from time of initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months l

}..

and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

i SURVEILLANCE REQUIREMENTS 4.8.1.1.1 Two physically independent circuits between the offsite transmission network and the'onsite Class 1E distribution system shall be determined OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying l

correct breaker alignments and indicated power availability.

~

TQ

\\

(2. T*\\ D ST. LUCIE - UNIT 1

, 3/4 8-2 s

F\\ h S16 S 4

' 91_1IUEg8y_QE_Ngg6ggB_EgWgB_g69NI_gEEBSIJgN _E691gS _8Ng PAGE 38 2

2

':ISEBd99YN9dJgS ANSWERS -- ST. LUCIE 1&2

-87/04/20-LAWYER, SANDY ANSWER 5.01 (1.00)

(d)

REFERENCE TS Bases p 3/4 4.10

'SL,TS bases,p4-8.LO-none 002/000 K5.18 (3.3/3.6) 367 01/23/84 03 01 00 01 05

+0.600

+0.600 ANSWER 5.02 (1.00)

(a).- Doppler is a'U238 cross section effect.

' REFERENCE NETRO, 8.2,11.3 SL,PSRT,pp5-7,8,91LO 5.5.

192/7K1.04(3.1/3.4) 302 03/05/85 02' 00 03 02 07

+0.143

+0.143 ANSWER 5.03 (1.00)

(e) - Rotameter.is not based on Bernouli's principle.

REFERENCE CR fluids & mechanics, pp 50-53.

SL,PSTHTFF,p 2-10 l

191/2K1.01(2.2/2.4)

SL,LO1,PSTHTFF,p 2-1

302 03/05/85 02 00' 05 00 07

-0.429

-0.429 l

i

i 50 " THEORY OF ' NUCLEAR POWER _ PLANT _gPgRATIgN _ FLUIDS _ANg PAGE-39 1

1

-1656099X000199 ANSWERS -- ST. LUCIE 1&2

-87/04/20-LAWYER, SANDY j

ANSWER 5.04 (1.00) j (c) -thermocouples are smaller and faster REFERENCE.

.CR,STM-7, pp2-7 SL,SD-407.

'A-2,K1.14(2.8/2.9)

SL,LO407-4

'SL,LO201-15 302 03/05/85 00- 02 02 0'3 07

-0.143

-0.143 ANSWER 5.05 (1.00)

(d)

-l REFERENCE CR,AP-530, p4 SL,0NP2-012OO39 r 1, p 3. -

SL,LO(none for procedures).

'O/74EK3.01(3.4/4.2) 302 03/05/85- 03 00 03 04 10

+0.400

+0.400 I

' ANSWER 5.06 (1.00)

(d)

REFERENCE-NETPP, p 6.2-5.

GP,.HT&FF, P323.

SL,PSFF,p2-20,LO11.

A9-8'(2.6/2.6) 269:-8/19/85 12 04 00 16

+0.S00

+0.500 71.0001 1R

...(,A'S)

(

)

i 1

J

S __IHgg6%_QE_UQCLE86_EQW56_E68UI_QEg6811gN _E(ylQS _@dQ PAGE 40 t

t t

ISEBdQQYUed1CS ANSWERS -- ST. LUCIE 1&2

-87/04/20-LAWYER, SANDY f

i ANSWER 5.07 (1.00)'

(a)

REFERENCE DPC, FNRE, pp 200 & 221.

SL,SD-201,p41. SL,LO,SD-201-7,10.

O/9 EK3.08 (3.6/4.1); 4/O K5.11 (3.6/3.9) 269 8/19/05 09 03 02 02 16

+0.375

+0.375 ANSWER 5.08 (1.00)

(a)

REFERENCE l

OC, OP 1103/15, enci 8.0, 5.1, 2.0.

OC,TS 3.1-23, 3.5-10.

SL,PSRT,PH-7,LO-1c.

SL,TS 3.1.1.1.

1/10 K5.35 (3.3/3.6) 269 8/19/85 10 02 03 01 16

+0.375

+0.375 ANSWER 5.09 (1.00) a (e) l REFERENCE GP, HTT&FF, p271.

SL,SD207,LOl6,p7.

6/30 K5.01 (2.7/3.0) 269 8/19/85 12 00 04 00 16

+0.500

+0.500 l

ANSWER 5.10 (1.00)

(b)

REFERENCE McGuire System Manual MC-CP-RCF-3 SL,PSRT,LOl6, sect 6.0,p7-19.

1/OKS.16(3.4/4.0) a

5. THEORY OF NUCLEAR POWER PLANT OPERATION _FLUlD@t_AND-PAGE 41 t

ISE8dODYN@dlC@,

LANSWERS -- ST. LUCIE 1&2

-87/04/20-LAWYS'*t, SANDY

.369 01/23/84 3

_1 1

0 5

+.200

+.200 ANSWER 5.11 (1.00) x (d)

_ REFERENCE Catawba CN-SYS-NC-5 SL,PSRT,LO9;9;p5-13 4/2K6.22(2.4/3.2)

.413 03/20/84 2

1 5

1 9

-0.333

-0.333 413 03/20/84 13 1

3-1 18

+0.555

+0.259

' ANSWER 5.12 (1.00)

(d)

REFERENCE PWR Reactor Theory Handout, General Physics Corp. p 33 SL,PSRT,LO5b,p9-1 413.03/20/84 3 0 6 0 9 -0.333 -0.333 l

' ANSWER 5.13 (1.00) 1 1

1 CR,(a) - 420,000sef; 26,000 scf (b); 140 scf (c) ; 1,320 scf (d) i SL,(a) - 440,000scf ; 52,500 scf (b ) : 1100sc f (c ) ; 1,166 scf (d).

l REFERENCE CR lesson'"FLndamentals of Natural Circulation";

CR, ROT 3-14, p2 j

SL,0NOP 120037, p 9.

l K O/74 EA1.01(4.2/4.4) i 1

.302 03/05/85 02 00 05 00 07

- 429

- 429 302 08/18/86 02 00 00 00 02

+1000

- 111 a

335 11/17/86 04 02 00 00 06

+ 333

+ C67 335 11/17/86 02 01 00 00 03

+ 333

+ 111 I

l

4

' g v'

,v.>

5.

' THEORY OF' NUCLEAR POWER PLANT OPERATION _FLUlpS _6Np PAGE 42 l

1 2

ISEBd9QXN8dlCE I

ANSWERS -- ST. LUCIE 1&2

-87/04/20-LAWYER, SANDY i

ANSWER 5.14 (1.00)

]

i (b)

REFERENCE.

NETPP, p 6.4-5.

SLiPSFF,p2-19,LO19.

l A9-11 (3.0/3.2) 269 8/19/85 11- 01 00 04 16

+0.875

+0.875 i

ANSWER.

5.15-(1.00)

(c)

REFERENCE r

DPC, FNRE, p 169.

l i

NETRO, 10.5-2.

Westinghouse NTO, p I-5.77.

.SL,PSRT,p.4-19,LO-5 1/0 KS.13-(3.7/4.0) 269 8/19/85 12 00 04 00 16

+0.500

+0.500 l

t ANSWER 5.16 (1.00)

(a)

REFERENCE GP, HT&FF, chapter 1.

DPC, T,

FF&HT, Chapter 1A.

Cr3, ANAO-29, p2.

OC, CH-PC, p 13.

'2/O KS.06 (2.3/2.6) i 269 8/19/85 09' 03 04 00 16

+0.125

+0.12 f

i

-]

1

}

50 THEORY OF NUCLEAR _PQWER_ PLANT _gPEdATIgN _FLUIgS _AND PAGE 43 l

1 1

.i ISERdQQyy@ digs ANSWERS -- ST. LUCIE 1&2

-87/04/20-LAWYER, GANDY j

l I

ANSWER 5.17

.(1.00)

(d)

REFERENCE DPC,FNRE, pp 124-128.

SL,PSRT.p.9-22,LO11.

15/O KS.06 (3.4/3.7)

I 269 -8/19/85 no history 2 ans.

l

-)

ANSWER 5.18 (1.00)

(b)

REFERENCE n

DPC, FNRE, pp 253-257.

SL,SD403,LO3.

15/O K5.01 (2.9/3.2) l 269 8/19/85 10 02 04 00 16.+0.250

+0.250 L

ANSWER-5.19 (1.00)

{

(b)

REFERENCE FPC, CR3, HTFF/Thermo, p 82.; SL, PSTHTFF, Ch2, p45, LO12.

DPC, OC, T,FF&HT, p 150.

SL,PSFF,p2-2,LO12 3/O A1.01 (2.9/2.9) 269 8/19/85 12 00 01 03 16

+ 875

+ 875 335 11/17/86 03 00 00 00 03

+1000

+ 895 I

i

^

i

_---_-_---.-.--._--_--_-_---__-_.____--.----.__--__._____._----____-.--..__.-__)

r-5 _IISE96Y_9E_NWGLE@B_EQWEB_ELQNI_QEEB@llgN _E(91DS _@ND PAGE 44 t

t t

IUEBdQDYUQUlGQ ANSWERS -- ST. LUCIE 1&2

-87/04/20-LAWYER, SANDY ANSWER 5.20 (1.00) l (c)

REFERENCE OC, GP 1106/01,-p10

BK, GP-03, p 19.

OC, EL-MPD, p 32.

1 BK, 20-2-C4, p37.

SL, OP2-OO30124, r 18, p l8 (no LO on proc. )

AS-16 (2.1/2.2);

62/O K5.03 (2.4/2.6) 269 8/19/85 10 02 03 01 16

+0.375

+0.375 ANSWER 5.21 (1.00)

(a)

REFERENCE l

NETRO, S.

7.2, 7.3, 7.4.

Nuclear Reactor Analysis, Dunderstadt & Hamilton, p 13.

SL,PSRT,p4-1,LO2 l

269 8/19/85 12 00 03 01 16

+0.625

+0.625 302 12/17/84 04 00 00 00 04

+1.000

+0.700 i

I ANSWER 5.22 (1.00)

(c)

REFERENCE 3L, PSTHTFF, HT p 43 & LO 4.

335 1.1/17/06 05 01 00 00 06

+ 666

+ 666 I

335 11/17/86 03 00 00 00 03

+1000

+ 777 ANSWER 5.23 (1.00)

(a)

REFERENCE SL, PS4T, p 8-11,LO1.

335 11/17/86 03 00 00 00 03

+1000

+1000 I

i 1

-5.- THEORY OF NUCLEAR POWER' PLANT OPERATION _ FLUIDS _8ND PAGE 45 t

t ISEBdQQXUedigS LANSWERS -

ST~.

LUCIE 1&2

-87/04/20-LAWYER, SANDY ANSWER 5.24' (1.00)

(d)

' REFERENCE.

.l Wnstinghouse Reactor Physcis, Sect.

1-5, MTC and' power defect.

DPC, FNRE;4 SL, RP, Sect.

7.6, 7.7.; SL,.PSRT, p 12-2, LO4.

335 11/17/86 02 01 00 00 03

+ 333. + 333 I

i ANSWER 5.25 (1.00) 1 (a)-

l 1

REFERENCE CNTO, " Thermal / Hydraulic Principles and Applications, I I ',', pp 10-45/48 l

SL,PSFF,p2-2,LO19.-

006/050; K3.01(2.9/3'1) i 335 '11/17/86 05 01 -00 00 06

+ 666

+ 666 ANSWER 5.26 (1.001 (c) d W

i REFERENCE-1 SL,_CNTO, pp 4-16/27.

q

.SL,PSFF,p2-2,LO19.

d KOO1/OOO K5.13(3.7/4.0) 335 11/17/86 06 00 00 00 06

+1000

+1000 l

ANSWER 5.27-(1.00)

-(b)

REFERENCE Ganeral Physics, Heat Transfer Thermodynamics and Fluid Flow, pp. 145 - 160.

ST Lucie Thermo Handbook, Chapter 2d SL, PSTHTFF, Ch 2, LO 5 OO2/OOO-K5.01 (3.1/3.4) j

Q : _ _'IS E98 Y _9 E _Nyg(E 66_EQW E g _E(g NI _QEE 6 @IlgNt_E(glQS _@NQ PAGE 46 t

IdE609910001GS ANSWERS -- ST. LUCIE 1&2

-87/04/20-LAWYER, SANDY 1

)

ANSWER 5.28 (1.00)

(c)

REFERENCE

]

steam tables 1

SL,PST,p1-1,LO2,6.

335 11/17/86 06 00 00 00 06

+1000

+1000 t

ANSWER 5.29 (1.00)

(a) i REFERENCE SL,PSRT, p4-14, LO13.

1/0 K5.10 (3.9/4.1)

ANSWER 5.30 (1.00) 1.

C 2.

B 3.

F 4.

H 5.

A REFERENCE Steam - Its generation and use, Babcock & Wilcox Co.

SL,PST,LO1,2.

pA17#20(2.0/2.3) 369 01/23/84 4 0

1 0

5

+0.600

+0.600 1

]

j

6 _1E60NI_SygIEdg_gEgIgyg_CgNIRgL,$Np_JNgI69dENIgIIgN PAGE 47 2

ANSWERS -- ST. LUCIE 1&2

-87/04/20-LAWYER, SANDY ANSWER 6.01 (1.00)

)

(C )'

-REFERENCE-SL,~ Lecture outline RPS, figure 7.2.21; SL, LT404, p32, LO2.

1 335 11/17/86 06 00 00 00 06

+1000

+1000 335 11/17/86 03 00 00 '00 03

+1000

+1000 li.

t I

ANSWER 6.02 (1.00) l (b)

REFERENCE ST Lucie SD117 "AFW",

pp 91 CWD 87708-326; LT 412, p9, LO7.

061/000; K4.07(3.1/3.3) l 335 11/17/86 04 02 00 00 06

+ 333

+ 333 335 11/17/06 03 00 00 00 03

+1000

+ 556' ant?WER 6.03.

(1.00)

(b)-

i REFERENCE SL, lecture outline #46, ppi,2,4.

5/O K1.13(3.3/3.5) 335-11/17/06 03 00 00 00 03

+1000- +1000 ANSWER 6.04 (1.00) l (d)

REFERENCE-SL, SD2, ppi-7,1-8; SL,LT602,p11,LO3.

6 __'ELONI_gySIgdg_pggigNg_ggNIBg(2_9NQ_JNgIBydgNIBIIgN PAGE 49 t

ANSWERS -- ST. LUCIE 1&2

-87/04/20-LAWYER, SANDY ANSWER

'6.05 (1.00)

{

(c)

REFERENCE o

SL, SD24, pi-20 ;SL,SD207,p19,LO10.

l i

i ANSWER 6.06 (1.00)

(d) l REFERENCE I

SL,.SD145, p-0-7=and fig.

1.

SL,LY503/504,LO2.

j I

ANSWER 6.07 (1.00)

(b)

. REFERENCE SL, 50102, pp 1-6,1-9.

SL, LT408, LO4 i

SL, LT410,o19,LO6.

ANSWER.

6.08 (2.00) a.

both t

b~.

unit i

=c.Junit 2 d.

unit 1 REFERENCE SL, 1ecture 47, pp2&3, difference analysis, p34.

' SL, LT207, LO3.~

'335 11/17/86 03 00 00 00 03

+1000

.J

E c

6,

PLANT' SYSTEMS DESIGN _ggNIRgl _@Np_JNSIByMENI@IlgN PAGE 49

)

2 i

ANSWERS -- ST'. LUCIE 1&2'

-87/04/20-LAWYER, SANDY

' ANSWER 6.09 (1.50)-

t j a)

Unit 2

(+.5 ea) b).

Both c)

Unit 2 REFERENCE

.ST Lucie SD9 "PZR Pressure, Relief and Level Control", pp 33-36 SL,LT201,p41,LO10.

010/000; K4.03(3.8/4,1)

]

1 i

l ANSWER 6.10-(

.75).

i Reactor Trip, 806 psia, 560 deg F

(+.25 ea)

REFERENCE ST Lucie SD108 "SBCS",.pp 17; SL,LT406, LO2b.

041/020; K4.17(3.7/3.9) l l

ANSWER 6.11

() 00) electric-fails open

(+.5 ea) air-fails close REFERENCE ST Lucie SD13, pp 3; SL,LT304, LO3.

039/000;-K4.05(3.7/3.7) l

~

ANSWER 6.12 (1.00) l UNIT 1-1(or 4) fans at Normal speed

(+.5 ea)

UNIT 2 Slow 1

-)

REFERENCE ST Lucie 508, pp 3;SL,LT207, L0s,14,15.

103/000; K 1. 05'( 3. 6 / 3. 8 ) -

j

e o

6t__E60NI_@y@IEd@_DE@lGN _CQNI6Q(t_QND_lN@l6QUENI@IlgN PAGE 50 t

" ANSWERS -- ST. LUCIE 1&2

-87/04/20-LAWYER, SANDY

]

ANSWER 6.13 (2.00)

a. The liquid discharge valve is closed b.

The blowdown valves and the sample valves are closed

c. None

d. The. surge tank vent is valved from the atmosphere to the chemical drain tank REFERENCE SL, SD 37; SL, LT410, pp12,18,20, LO 6.

335 11/17/86 06 00 00 00 06

+1000

+1000 335 11/17/86 03 00 00 00 03

+1000

+1000 1

1 ANSWER' 6.14

-(3.00) a.

Subchannel deviation alarm high-r ch nnel trip, TM/LP alarm,LOE ZbH3 Dhef4f64LC5%(h w p N

%) Ay W67'/Weh t$, W, DNCr a :

_b. SUR trip to the.RPS, Sul anable 10E-4%,zero mode bypass.1% powe-7 c.

Input to RRS /m eactor vs turbine power, input to CPCs for calculation core barrel movement monitor. 4 of ASI, input

-3 REFERENCE SL,LT403,LO11.

335-11/17/86 2.17 0.83 0.00 0.00 03

+ 447

+ 447 ANSWER 6.15 (1.50)

]

1)

PDIL

(+.3 ea) 2)

Group out of sequence 3)'

Group deviation 4)

ISH (reg grp wthdwl perm) 5)

IRG (sd grp wthdw; perm)

REFERENCE ST Lucie SDS "CEDS", pp 31;SL,LT402,LO1.

i i

001/000; K4.07(3.7/3.8) i 1

I i

L6__'ELQNI_@ySIEdg_QEgigN_CgNISQLt_@NQ_lN@I6UdENI@IlgN PAGE 51 t

t ANSWERS -- ST. LUCIE 1&2

-87/04/20-LAWYER, SANDY ANSWER 6.16 (1.75)

Their are two electro-hydraulic operated valves per feed line

(+.5)

They are operated in fast speed in accident conditions by two Hydraulic (N2) accumulators

(+.5).

In normal ops (slow speed) they use a pneumatic pump (+.5)

A spool valve, solenoid air operated determines speed (+.25)

REFERENCE j

ST Lucie SD112 " Condensate and Feedwater", pp 18; SL, LT301, pp8,9, LOS.

{

059/0003 K4.19(3.2/3.4) l 1l 335 11/17/86 5.43 0.57 0.00 0.00 06

+ 810

+ 810 335 11/17/86 2.71 0.29 0.00 0.00 03

+ C07

+ 809 i

i ANSWER 6.17 (1.00)

A heated thermoccuple's output is compared with an unheated thermoccuple's output.(+.3)

With no liquid to remove the heat, the temperature difference becomes large, creating a large voltage cutput

(+.7)

REFERENCE ST Lucie QSPDS Handout, pp 16; SL, LT407, p21, LO5.

l 017/020;K4.01(3.4/3.7) 335 11/17/06 6.0 0.0 0.0 0.0 06

+1000

+1000 335 11/17/86 3.0 0.0 0.0 0.0 03

+1000

+1000 i

l ANSWER 6.18 (1.00)

Header pressure is 100 psi below the opposite steam generator or auxiliary feedwater header pressure.

REFERENCE SL, LT 412 r3, p12, LO9.

I 335 11/17/8A 1.5 1.5 0.0 0.0 03

+ 000

+ 000 l

l l

F 6.

PLANT SYSTEMS DESIGN CONTROL _AND_lNSTRUMENTATION PAGE 52 t

t ANSWERS -- ST..LUCIE 1&2

-87/04/20-LAWYER, SANDY 1

ANSWER 6.19 (2.00)

Pressurized nitrogen is inserted into the hoist box by hose, replacing the cooling water.

The cooling water flows out the bottom into the fuel pool.

This gas can then be vented through a scintillation counter after a soak period has been completed.

Cflow paths (0.5 ea); fluids (0.25 ea); detection (0.5)]

REFERENCE SL, LT 200 r2, p59, LOS.

O/6 EA204 (2.6/3.4) 335 11/17/86 2.00 1.00 0.00 0'.00 03

+ 333

+ 333 ANSWER 6.20 (1.00)

1. Allow f or system pressurization to be controlled from the pump discharge l

valve.

2. Aid in valve opening ease once the pump is started.

REFERENCE SL, OP 1-700020 r16, p 33; SL, LT301, LO2a.

335 11/17/86- 0.00 -3.00 0.00 0.00 03

-1000

-1000 ANSWER 6.21 (1.00)

To prevent uncontrolled pressurization

(+.7)'due to the sizing of the RCGVS orifices in the vessel head and PZR vent lines (+.3) i REFERENCE ST Lucie SD7 "RCS",

pp 52;SL,LT201, LO6.; SL,OP2-0120020,r19, p.5.

002/020; PWG-7(3.7/4.3) i ANSWER 6.22 (2.50) a)

Ensure the operator does not deliberately or inadvertantly defeat parts of the protection system by switch misalignment (+0.5) b)

High SUR; Loss of Load; Local Power Density

(+.33 ea) c)

Must turn 4 keylock switches located on each RPS panel

(+.5) and is bypassed above.1% (+.25) C.5% Unit 2]

(+.25)

\\

r i

h__'EL9NI_py@ Igd @_gggjgh_CgNI69L _8Np_JN@I6UdENI@IIgN PAGE 53 2

ANSWERS -- ST. LUCIE 1&2

-87/04/20-LAWYER, SANDY REFERENCE St. Lucie SD 62, pp 31/32; SL,LT404, pp28,29, LO3,8.

1 l

l L

e Z:_ _'P 69 C E gyB E g_ ;_ d ggd@61_ @ @ Ng8d 862_ E d E 69E NC Y_ ON 9 PAGE 54 699196991996_G9dI696 ANSWERS -- ST. LUCIE 1&2

-87/04/20-LAWYER, SANDY ANSWER 7.01 (1.00)

(No chloride monitor listed among symptoms)

(c)

REFERENCE SL, OP-1-610030 l

39/0 SG8(2.7/2.9)

)

335 11/17/86 03 00 00 00 03

+1000

+1000 ANSWER 7.02 (1.00)

(c)

REFERENCE SL, OP-1-120020, ppi,2.; SL,LT201,LO9.

335 11/17/86 03 00 00 00 03

+1000

+1000 1

ANSWER 7.03 (1.00)

(b)

You need 4 to go over 1750 psig.

This is a requirement for going above 275 psig.

(Note - d is referenced to going >250 degF, but you can't go over 1725 psig without Tavg >250 degF without exceeding the 350 degF deltaT limit between the PZR and Tavg.

REFERENCE SL, OP2-30121 pp1-5.;SL,LT201,LO7.

335 11/17/06 02 01 00 00 03

+ 333

+ 333 I

ANSWER 7.04 (1.00) 1 (d)

REFERENCE SL, OP-2-1630024, p4.; SL,LT208,LO3a.

333 11/17/86 03 00 00 00 03

+1000

7 P

70

' PROCEDURES - NORMAL t_A@NQRMAL _EMERQENgY_AND PAGE 55 t

60D196901G06_GQN16QL ANSWERS -- ST. LUCIE 1&2

-87/04/20-LAWYER, SANDY ANSWER 7.05 (1.00)

(b)

REFERENCE SL, 1-EOP-05, p6.

006/050 K4.02 (4.1/4.3)

ANSWER 7.06 1.00)

(b)

REFERENCE 10CRF20.101 R 1/1/86, p 258.

PWG-K1.03 (2.8/3.0) 302 08/18/86 00 02 00 00 02

-1000

-1000 33b 11/17/86 04 02 00 00 06

+ 333

+0000 ANSWER 7.07 (1.00)

(b)

REFERENCE SL, HP-2 r5, p 16.

072/000 A1.01 (3.4/3,6) 1 335 11/17/86 02 04 00 00 06

- 333

/

i 335 11/17/86 03 00 00 00 03

+1000

+ 111 ANSWER 7.08 (1.00)

(d)

)

REFERENCE

)

SL, 1-EOP-04, p6.;LT823,LO11.

i EPE-038 EA2.17 (3.8./4.4) 335 11/17/86 06 00 00 00 06

+1000

+1000 1

335 11/17/86 03 00 00 00 03

+1000

+1000 J

I C

)

sPROQEQUREg_,_NQBM@L _A@NQRM@L _EMgRQgNQy_@NQ.

PAGE 56

~J 7.-

t t

B00196991G06_QQNIBQL

' 87/04/20-LAWYER, SANDY ANSWERS -- ST. LUCIE 1&2 f

1 i

ANSWER 7.09 (1.00)

(a) i REFERENCE SL, 1-EOP203, p21.

028/000 K1.01 (3.4/3.8) 335 11/17/86 05 01 00 00 06

+ 667

+ 667 335 11/17/86 03 00 00 00 03

+1000

+ 778 1

ANSWER 7.10 (1.00) 1 (c)

REFERENCE ST Lucie EP 00.*0.41; SL, ONOP 30135,-p6.

PWG-11 (EOP Immaalate Actions) (4.3/4.4) i 335 11/17/80 06 00 00 00 06

+1000

+1000 i

ANSWER 7.11 (1.00)

-(b)

REFERENCE i

'ST Lucie EOP-5, Steps 7-9 EPE-074 'EK3.04(3.9/4.2) 3 f

335 11/17/86 06 00 00 00 06

+1000

+1000 I

I ANSWER 7.12 (1.00)

(c)

J REFERENCE

.5L, ONOP 2-310031 r3, pl. SL,LT411,LO6.

t 333 11/17/86 03 00 00 00 03

+1s00

+1000

t 7.

' PROCEDURES - NORMAL t_ABNQRMAl _ EMERGENCY _9NQ PAGE 57 t

609196991ce6_CQNIBQL ANSWERS -- ST. LUCIE 1&2

-87/04/20-LAWYER, SANDY ANSWER 7.13 (1.00)

(d)

REFERENCE

)

ST Lucie EOP-1, pp 7 i

EPE-015; EK1.01(4.4/4.6)

)

ANSWER 7.14 (1.00)

(c)

REFERENCE SL, OP " Nuclear and delta T power calibration".

015/000 A1.01 (3.5/3.8)

ANSWER 7.15 (1.00) a)

Unit 1

(+.5 ea) b)

Unit 1 REFERENCE STLucie OP 1/2-0030127, pp 1,

7; pp2,10.;SL,LT201,lo9c.

002/020; PWG-12(3.7/3.7) 335 11/17/86 5.5 0.5 0.0 0.0 06

+ 833

+ 833 335 11/17/86 2.5 0.5 0.0 0.0 03

+ 667

+ 778 ANSWER 7.16 (1.50) 1.

All clearances have been properly released.

2. All tags have been removed and valves, switches, etc. are in their required positions.

3. The system has been tested to the extent that it is evaluated as safe to be returned to service.

REFERENCE SL, OP 10122 r21, p2.; SL, (no LT,no LO) 194/1 K1.07 (3.6/3.7) 335 11/17/86 2.33 0.67 0.00 0.00 03

+557

+ 557 t

]

f t

7.

DROCEDURES - NORMAL _ABNgRMAL _ EMERGENCY _AND PAGE 58 2

2 589196991996 99NISQL ANSWERS -- ST..LUCIE 1&2

-87/04/20-LAWYER, SANDY s

ANSWER 7.17 (1.50) 1.

After each refueling.

2.

The conditions used in the ECC calculation are questionable and/or unreliable due to a lack of steady operation as determined by the NPS.

3.

An excessive reactivity anomaly, which could affect this approach to criticality, has existed since the last time at which critica111ty was achieved.

(An excessive reactivity anomaly is defined in Reg. Guide 1.16 and STS 6.9.1.8.d).

i REFERENCE SL, OP 2-30126 rl, p2.; SL, OP 30221. SL,PSRT,p9-2, LO12.

(

335 11/17/86 2.07 0.93 0.00 0.00 03

+ 380

' ANSWER 7.18 (2.00) 1.

the steam supply to the turbine seals 2.

air in-leakage to the condenser 3.

the deltaT of the circulating water across the condenser

4..the status of the circulating water pump 5.

the circulating water box prime 60 the steam Jet air ejector operation 70 'the SJAE loop seal 8.

the condition of feedwater heater alternate dump valves REFERENCE i

SL, ONOP 2-610031 ro, p2.; SL,LT304, LO1.

335 11/17/86-5.38 0.62 0.00 0.00 06

+ 793

+ 793 335, 11/.17/86 1.75 1.25 0.00 0.00 03

+ 167

+ 584 ANSWER 7.19 (2.00) l a..the nitrogen cover. gas i

.b..RCS coolant flashing to steam c.

hydrogen from oxidation of the cladding d.

hydrogen REFERENCE SL, ONOP 2-0120037, r4, pp 1,2,9.; SL,LT,201, LO9c.

..~,.._.J

(1

~~~

7

' PROCEDURES - NORMAL _ABNgRMAL _gMgR@gNgY_ANQ PAGE 59 i

1 2

509196991996_99NIBQL ANSWERS -- ST. LUCIE 1&2

-87/04/20-LAWYER, SANDY l

1 l

ANSWER 7.20

(.50) i RCS inventory control.

l i

REFERENCE j

SL,LTB27,p.25,LO3.

1 l

ANSWER 7.21 (2.00)-

l 1)

Any of the Rx Trip Safety Functions status check acceptance criteria not met

(+.5)

OR (+.25) 2)

All of the Safety Functions being. maintained

(+.5)

AND-(+.25)

RCS conditions are being controlled / maintained in Mode 3

(+.5) l REFERENCE

.ST Lucie EOP-1 (SPTA)

EPE-OO7; PWG-10(4.1/4.2) 1 ANSWER 7.22 (2.00) a)

Hot Leg (as you are keeping level above break)

(+.5) i b)

Faster Plant Response

(+.5 ea for any 3) l No Hot spots in RCS j

Main Spray Available More cooling of isolated S/G l

REFERENCE CE EPG Guidelines on RCP Trip Strategy EPE-074; EK3.08(4.1/4.2) j

/T i

?

'Zg__ PROCEDURES'- NORMAL _@@NQBM@L _EMESQENQY_@NQ PAGE 60 t

t 680196991G06 QQNIBQL ANSWERS'-- ST. LUCIE 1&2

-87/04/20-LAWYER,. SANDY i

ANSWER-7.23 (2.00) a)

Warns of possible fuel bundle binding during insertion into core (+1.0) b)-

Bridge in CORE CLEAR ZONE

(+.5 ea for any two)

In' transfer zone AND hoist is at UP limit

~

In'CEA zone'AND. hoist is at UP Limit

. Trolley indexed REFERENCE ST Lucie Refueling Equipment Handbook, pp I-13, I-29 034/000 K4.01(2.6/3.4) & K4.02(2.5/3.3)

ANSWER 7.24 (1.50) 1.

size-2.

location

3. whether i sol abl e.

(30 0.5 each)

-REFERENCE SL,LT826,LO1.

)

l l

l i

i

F.,

1, J

u s

n,,

... t.

s q

t i

r:

.l3

/

\\

'1, s

U h'I..

/ (,1 i

s

,.t

.i Es._690ldl@l6611Mg_66QCEQUBE_gt_ CONDITIONS M NQ LIMITATIONS PAGE 61

.- j ' '

l.s

's

. ANSWERS -- ST. LUCIE 1&2

' s'( p-87/04/20"LAWYEp EANDY 4

( j ;

a

).,

f, d :-

T f

t N

s\\

q s

\\ 1. \\ -

ANSWER

'O.01 (1.00) t M

(c)

,I

}

.s l D~

'\\

y REFERENCE SL,.TS-2, p 1-7,2-9,6-12,4-9.

PWG-5(TS) (2.9/3.9) 335 11/17/86 02 01 00 '00 '03

+333

+333 s

ANSWER 8.02 (1.00)-

a i

s.

A c

% }

}

gg).

,(

a t

REFERENCE i

s i

M i

i.-

SL, EPIP 3100023E, r7;

<N c

i s

-; si

'C s

194/001 A1.16 (3.1/4.4).

l c-

?

N : 4 g

is i.

i y

4 i.

4 ANSWER-8.03

( l '.' O O )

(b)

\\!N REFERENCE

/

.SL,.APOO10124, r12, p5.

1

194/001~K1.02 (3.7/4.1) v a

l V\\

4 f,

i i $'

ANSWER 8.04 (1.00)

'\\

(c) f REFERENCE.

SL,'TS, pp 3-2,4,S.

15/201.PWG-5(2.8/3.9) 335 11/17/86 03 00 00 00 03

+1000

+1000 I

(

l L-

F t

r

9. _ _'69 q.;NI SI58I I V E_ P BgC E QW65$1_ C g Np]IJ gyp 1_9Np_( { MJ I@IJ g Ng PAGE 62 ANSWERS -- 9,T.

LUCIE 1&2

-87/04/20-LAWYER, SANDY r }., -

1 I

T g

'4 i

i (1.00)

ANSWER 8.05 i

(a) i REFERENCE SL, EP-310029E, p 2.

194/L K1.16 (3,5/4.2) 6

/

ANSWER 8.06 (1.00) '

(a) 1 REFERENCE i ~

SL, TS2, p 9-1.

34/0; PWG-5(2.8/3.7)

~336 11/17/86 03 00 00 00 03

+1000

+1000 "ANSWEN 8.07 (1.00)

Ji (c) t REFERENCE C(<, EM-207 Rev.

16,'p 5; VG, 10CFR20.403, (2) and (4).

194/1 K1.03'(2.8/3.4) 335 11/17/86 03 00 C;0 00 03

+1000

+1000 l

l ANSWER R. 08, (1.00) l (di s

d

' REFERENCE 1

SL, TS1, p 3/4 0-1 1

64/50; PWG-5(3.1/4.1) 335 11/17/86 03 00 00 00 03

+1000

f

~

8 __'990]N]gI68Ilyg,66pCEDU65g1_C9dp]IJgNg2_8Np_bjd]I9IJgNg

-PAGE 63 2

AASWERS.-- ST.' LUCIE ik2.

-87/04/20-LAWYER, SANDY

ANSWER 8.09

( 1. 00)-

n (b) l l

REFERENCE' 1

SL,fTS1, p 3/4 7-9 3 SL,LT304,-LO3.

l 335 11/17/86 03 00 00 00 03

+1000 '+1000

.j ANSWER B.10 (1.00) i (d)

.l REFERENCE 1

LSL, TS2, p 3/4 5-1.;SL,LT207,L0s 15,17.

335 11/17/06 03'.00 00 00 03

+1000

+1000 1

' ANSWER B.11 (1.00)

(A)

' REFERENCE SL, EPIP-3100033E, r9 l

194/1'A4.16 (3;1/4.4)

' ANSWER 8.12 (1.00) 1 (a)

REFERENCE St, EP!P-3100021E, 17s/1 A1.16 (3.1/4.4) l ANSWER 8.13 (1.00) i (a) l i

REFERENCE

SL, EPIP-310003!'

t.

i

-19a/1 A1.16'(3.

e'

)

3

f*!

1

'at__'00d1N1@l8@Ilyg_ESQQEQU6ES _QQNQlIlQNS _@NQ_(ld11@IlgNS PAGE 64 t

t ANSWERS -- ST. LUCIE 1&2

-87/04/20-LAWYER, SANDY ANSWER 8.14 (1.00)

(d) j REFERENCE SL, AP 0010124, r12, p4.

191/1 K1.07 ( 3. 6 / 3. 7 ')

ANSWER 8.15 (1.00)

(b) l REFERENCE SL, EPIP-3100021E, r19, pp16,17.

i 194/1 A1.16 (3.1/4.4)

ANSWER 8.'6 (1.00)

(a)

REFERENCE St Lucie Tech Spec l

061/000; PWG-5(3.3/4.1) & 039/000; PWG-5(3.1/3.7) i ANSWER 8,17 (2.00) a.

true b.

true c.

false d.

false REFERENCE SL, AP 10432 r24.

194/1 A1.12(3.1/4.1) 335 11/17/06 03 00 00 00 03

+1000

+1000 l

l I

L

l l

~

9:__0901015IBBI1YE_BB9CEQUBgS _CgNpillgNS _@ND_61dlI@IlgN5 PAGE 65 1

2 l

l ANSWERS -- ST. LUCIE 1&2

-97/04/20-LAWYER, SANDY I

1 ANSWER 8.18 (1.00)

(5 0 0.2 ea) 1 c

2 b

l 3

a 4

e 5

d l

REFERENCE SL, EPIP-3100023E, r29, ppi-4.

194/1 A1.16 (3.1/4.4) l ANSWER 8.19 (2.00) a)

Lors of Load without a reactor trip

(+.5 ea) b).

Steam Line break c)

Power ops when DNBR < 1.22 C1.283 d)

Prevent Peak Local Power Density < 21 KW/ft C22KW/ft) l REFERENCE SL,LT404,p14, LO2 3

l l

l ANSWER 8.20 (1.00)

Chananel Functional Test l

REFERENCE St. Lucie TS definitions l

l PWG-5(TS Knowledge) (2.9/3.9)

I l

l ANSWER 8.21

(.50) 4 months l

REFEFi2NCE j

10 CFR 55 i

l PWG-23(Shift Staf f i ng/ Ac t '. vi ti es) (2.8/3.5) l 1

i

7 Iai__0901NISI68IJYE_EBgCEQUBES,CQNQJIJgySx_6NQ_6Jd]I611gNS PAGE 66 2

ANSWERS -

ST..LUCIE 1&2

-87/04/20-LAWYER, SANDY LANSWER' 8.22

-(1.50) 1 1.

The intent of the original procedure is not altered.

2.,The change-is approved by two members of the plant management staff,-at.

l

-least one of whom holds an SRO license on the unit affected.

3.

The chance isfdocumented, reviewed by the FRG and approved by the Plant-

. Manager.within 14. days of' implementation.

REFERENCE SL, TS2, p 6-14.

194/1 K1.16 (3.5/4.2) i ANSWER-8.23 (2.00) 1

a. Restore'Tavg to =or> 515 degF within 15 minutes, b.

Be in hot standby within the next 15 minutes.

REFERENCE-SL, TS1,2, pp.3/4 1-7,6 1_SL,LT201,LO7.

1 335 11/17/86 2.25 0.75 0.00 0.00 03

+ 500

+ 500 1

ANSWER 8.24-(1.00)

All full-length _ control element assemblies shutdown and reg. are fully I

inserted CO.53 except for the single assembly of highest reactivity worth which is assumed to be fully withdrawn.CO.53 REFERENCE St. Lucie Tech Spec def. 1.29 001/050; PWG-5(2.9/4.3)

ANSWER 8.25 (3.00) i i

-1. Leakage (except controlled leakage)CO.23 into closed systemsCO.23 from pump seals'or valve packingCO.23 that are capturedCO.23 and conducted to

_ a sump or collecting tankC0.23

.2._ Leakage into containment atmosphereCO.333 f rom specifically known sourcesCO.333 but that is NOT pressure boundary leakageCO.343.

3. RCS leakageCO.333 through a S/GCO.333 to the secondary systemCO.343.

1

.a 4 i 9___09dit!~.@ISOIlyE_P6QCEQUSES _CQUQlligN@t_@NQ_Lidll@llgN@

PAGE 67 t

ANSWERS -- ST. LUCIE 1&2

-87/04/20-LAWYER, SANDY REFERENCE

.i SL, U2TS, p1-3 1SL,LT-201,LO8.

l l

l I

l l

I i

V'

~3 c-M h 5 f5R-Th' s-4 U.

S.

NUCLEAR' REGULATORY COMMISSION l

w REACTOR OPERATOR LICENSE' EXAMINATION.

l FACILITY:

_@ls_(Ug1E_1%2___________

l 1

/

i RE' ACTOR TYPE:

PWR-CE

)

DATE ADMINISTERED: _07f94f?g________________

_ LAWYER _gANQy___________

)

EXAMINER:

3 CANDIDATE:

v i

1 INDI6gCI1ggg_Ig_COND198IE1 I

Use separate paper for the answers.

Write answers on one, side only.

j Staple question sheet on top of the answer sheets.

Points. f or ~ each j

question are indicated in parentheses after the question.

.The passing

~

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

Examination papers will be picked up six (6).

hours after the examination starts.

% OF i

==

' '""/

f

=== =

-N

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A:

D: _

B:

C:

E:

F:

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H:

ft) Reciruelating Impeller (2) Auxiliary Impeller D) Integral Heat Exchanger M) Mechanical Seals 1

45) Cooling Water Outlet (6) Seal Injection Inlet (7) Control Bleadoff

$4t4RE //hO (8) Cooling Water Inlet re&W C-

r._

ENCLOSURE 3 p o sox i4000 JUNO BE ACH. F' 33408 NUCgs BCENSLNG 0FFLCgg C0?Y g

FLoAIDA POWER & LIGHT COMPANY j7'-

All : 36 APRll es igu af;'

f L 87-187 cLosc 'T"*

87 ~ U 3 Mr. John F. Munro, Chief Operator Licensing Section, Region 11 U. S. Nuclear Regulatory Commission 101 Marietta Street, N.W., Suite 2900 Atlanta, CA 30323

Dear Mr. Munro:

Re: St. Lucie Units I and 2 Docket Nos. 50-335, 50-389 Comments on License Examination i

of April 20,1987 Attached are the Florida Power & Light Company comments on the St. Lucie License Examination administered on April 20,1987.

Should you have any questions, please contact us.

Very truly yours, h Y' C. O. Woo Group Vi President Nuclear Energy COW /GRM/gp Attachment

)

f GRM4/045/l PEOPLE... SERVING PEOPLE

c.

COMMENTS ON ST. LUCIE LICENSE EXAMINATION APRIL 20, 1987 Comment #1 Question 1.09 Based on the' equation for determining the volume of l

1 makeup necessary to change boron concentration,. the curve which most correctly depicts the waste water volume is (d).

l i

Recommendation #1

]

correct the answer key. The answer should be (d)

Reference 1

i See Attachment #1 j

comment #2

-Question 1.11 The rapid depressurization and rapid void formation would cause the void coefficient to insert negative reactivity. Both doppler coefficient and void coefficient will add negative reactivity simultaneously, i

Recommendation #2 correct the answer key. Accept c or d as.the correct answer.

Reference SL,PSRT,p 7-9 (See attachment #2) f

)

S

' ^ ' ' ' ' ' ' ' '

e J

Comment #3 Question 1.17 This question tests for knowledge of a little known phenomena,

" Transient superheating of a

steam-water mixture".

This phenomena was included as an " Industry Events" lecture topic during PSL's 1987 Licensed Operator Requalification Program.

The PSL Training staff believes that this type of question would be more appropriate for SRO exams and Licensed Operator Requalification Exams.

Recommendation #3 Delete tihe question.

Comment #4 Question 1.20 does not state the initial reactive load condition, (in the lead or lag). If the student assumed MVARS in the lead then reactive load would be decreased.

Recommendation #4 J

Correct the answer key. Accept c or d as the correct answer.

Comment #5 Question 2.17 part 6.

The setpoint on unit ' 2 for SIAS is 1736 psi. part 10 does not specify the unit. SIAs 91ock Permissive on unit 1 is-1700 psia, for unit 2 1836 psia.

Recommendation #5 Correct answer key. Part 6 the most correct answer is (b). Part 10 accept (b) or (c).

Reference Unit 1 and 2 Technical Specifications 4

. _ r-

7 ;- -- --------- -

1 l

p i

Comment #5 Question 2.22 part a. There are no relief valves parallel to the interr.ediate stop valves.

i.

Recommendation #6 Correct answer key. Accept the following responses:

~

Governor & intercept valves close.

MSR block valves automatically close on unit 1 MSR block valves are closed by the operator on unit 2 post trip Shell side safeties lift Reference St Lucie Plant Main Steam Diagram sheet 3 (see attachment #3)

Commsnt #7 i

Question 3.10 To differentiate between the possible answers requires detailed knowledge of specific logic time delays.

Recommendation #7 Avoid the use of time delays in distractors. g

.s.

~

Reference' Examiners standards NUREG 1021 section ES-202 page.3 of 6 item 6.

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d.

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i comment #8 i

Question 3.15 part a and b. The question asks about a conditio l

breaker not to diesel ' generator l

close.

The question references a

Westinghouse

- that would ~ cause the i

I automatically design and is not (priicable to PSL.

Recommendation #8 Delete the question.

Reference PSL schematic diagram low pressure safety injection pump 1A, 4160v switchgear 1A3 load shed relays, sheet 949.

l sheet 251.

(ces attachments'#4 & #5) 1 Comment #9

)

Question 3.17 part b and c.

Other design indications are available to the operator.

i Recommendation #9 Correct answer key. Include for part b, CCW rad monito Reference SL ONOP 2-0310031, 1-0310031, 2-0120036, 1-0120036 f

I i

7-Comment #10 i

Question 4.19 part a and b.

The question states a spent fuel l

assembly in the spent fuel pool and does not state what unit.

{'

1 1

l Recommendation #10

)

I Correct answer key.

Part a.

Delete CIAS and Containment evacuation, due to the accident not being in containment.

Part b.

Delete Immediate operator actions #4 & #5 as they do not appear in the unit 2 Off Normal procedure.

Reference SL,0NOP 1-1600030 and 2-1600030 i

Comment #11

~

Question 4.23 The question asked to define an ESDE in terms of its characteristics. The answer key lists the characteristics that define an ESDE " type".

Recommendation #11 Correct answer key. Accept ESDE characteristics:

Rx power, Pressurizer level & pressure,RCS temperature, S/G pressure & level, Rx vessel level, loud noise.(as referenced by EOP).

Reference SL,LT 826 pg 6 (See attachment #6) y SL, EOP-5 Comment #12 Question 5.11 Refer to comment #1 i

I 1

l

I Comment #13 f

Question 6.14 part a. A DNBR alarm does not exist at unit 1 or 2.

Recommendation #13 Correct answer key. Delete DNBR alarm. Add LPD trip, dropped CEA (RPS), N.I. channel inoperable, Reactor nuc/ delta T power channel deviation.

i Reference SL ONOP 1-1220030 Comment #14 Question 7.24 Refer to comment #11 1

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4.2 Void Coefficient 5.0 FUE;.

TDtFERATURE (00FF1.ER)

COEFFICIDff/ DEFECT In a reactor, voids are the see'as bubbles caused by nucleate boiling.

The fuel temperature or Doppler reactivity is one of the principle The void coefficient is defined as means by which the stability of the the reactivity change (pes) per reactor is maintained.

It is giggy.g i,

change in void content (5 voids).

negative and is the fastest acting of all the reactivity feedbacks.

Typical PWR operation is about 0.255

$1nce it originates in the fuel it r-void or less.

The effects of an is the most important reactivity increase in void fraction are the effect in limiting any rapid power same as an increase in moderator level transier.t such as a

rod temperature.

It causes a decrease ejection event.

In moderator density (which results in the effects on the factors of Ken 5.1 Fuel Temperature Coefficient listed previously).

Thus, a,

is (FTC) negative in an undermoderated reactor.

The fuel temperature coefficienc (FTC) is defined as follows:

A typical value of.

the void coefficient is about, FTC - The change in reactivity (PCM) per unit change is fuel

- 100 pcm/ percent void temperature (*F).

1 Since-PSL operates with at least It has units of PCM/*F.

?'

50*F subcooling, voids should not exist void co tent is In order to understand how FTC works ins teanc during ermal to affect control of the reactor we i

estration.

must discuss in detail the absorption of neutrons by resonant The pressure coefficient and th poisons.

void coefficient play an Lacortant i

role in maintaining the inherent 5.1.1 Resonance stability and safety of the reactor.

Considee what would happen if a

As discussed in Chapter 4,

transient such as a Loss Of Coolant resonances occur at energy levels Accident were to result in where the target nucleus is allowed depressurization and loss of coolant to exist other than the ground

)

from the reactor.

The pressure state.

The two nuclides of interest decrease would insert' negative when describing the fuel temperature i

reactivity as would tite velds formed coefficient are U-238 and Pu-240.

from the increased boiling tha Figures 5-1 and 5-2 show the would result from the pressure dro.

probability of assorption vs neutron

^ he reactor would shut Lt if energy for these two isotopes.

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Dear Mr. Munro:

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