Text

Exam Rept 50-344/OL-89-01 on 890620-22.Exam Results:Four of Five Candidates Passed Exam
	ML20245G686
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	07/18/1989
From:	Meadows T, Miller L, Pereira D, Royack M, Sundsmo T; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:	;
Shared Package
ML20245G663	List: ML20245G662; ML20245G686;
References
	50-344-OL-89-01, 50-344-OL-89-1, NUDOCS 8908160158
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Examination Report No.: 50-344/0L-89-01 L Facility: Trojan Examinations adtr;inistered at Trojan Nuclear Plant, Rainier, Oregon. i i?; "

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Chief Examiner: - I--v A " -

7/ VN Thomas Meadows, Date Signed Operator Licensing Examiner Examiner: . '~

Mike Royack,

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- /1h Date Signed er tor Licensi. E miner

[P Examiner: d// #f8/t 7 Dave Pereira, Date Signed Operator Licensing Examiner m y f' ?//Wd Examiner: a[a Todd Sundsmo,

< " ~ ~ , A /pr Date Signed Operator Licensing Examiner Approved:

Le b dir- 'k/J'/N Date Signed Op/ Miller,Section erations Chi ()

Summary:

Examinations on June 20-22, 1989, (Report No. 50-344/0L-89-01)

This examination was initially scheduled per NUREG-1021, ES-201,

" Preexamination Activities," to optimize the use of examiner resources by examining no less than six(6) operator candidates per site visit. However, during the course of the preexamination administration and final facility certification two(2) operator applicants were appropriately withdrawn by the licensee, due to personnel changes and eligibility concerns. Considering that the bulk of the examination preparation had been completed, the short time remaining to the scheduled examination week, and the licensee's need for operators, the examination process was continued for the remaining five(5) operator candidates. j Operator licensing examinations were administered to one(1) Senior Reactor Operator (SRO) and four(4) Reactor Operator (RO) candidates. One(1) of the R0 t

candidates failed the written portion of his examination, and subsequently was i denied an operating license. All of the other candidates passed their {

examinations and were subsequently issued licenses.

8908160158 890727 i

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PDR ADOCK 05000344 ,

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l' REPORT DETAILS

1. Examiners Thomas' Meadows,.RV, Chief Examiner Michael Royack, RV Todd Sundsmo, RV David Pereira, RV 2.. Persons Attending t'he Exit Meeting on June 23, 1989 T. Meadows, RV, Chief Examiner R. Barr, Senior Resident Inspector

~t M. Royack, RV, Licensing Examiner T. Sundsmo, RV, Lic'ensing Examiner D. Pereira, RV, Licensing Examiner ,

R. Russell, Branch Manager, Operations S. Nichols, Branch Manager, Training ,

R. Schmitt, Manager, Operations & Maintenance :

R. Budzeck, Assistant Operations Supervisor i G. Ellis, Supervisor, Operations Training F. Bowen, Training Specialist IV, License Programs

3. Written Examinations and Facility Review The written examinations were reviewed at the Region V offices on June 13, 1989 by two persons from the licensee training and operatio>

staff. The signatures of the licensee's staff representatives are recorded in Attachment A, " Requirements For Facility Review of Written Examinations," which serves to document the security maintenance for these examinations. All original exam keys and review copies were then returned to the Chief Examiner at the conclusion of the review to ensure examin6 tion security.

The written examinations were then administered at the licensee's training i facility on June 20, 1989. At the conclusion of these examinations the Chief Examiner held a brief post examination review to compare the examination answer keys and the "as given" facility review copies. The keys were accepted without further official comments from the licensee i prior to the departure of the Chief Examiner from the site. This is attested by the licensee's documentation letter to the Region (Attachment B).

After the grading and review of these examinations it appears that a generic training weakness exists in the knowledge of " Emergency Plant Evolutions" and associated immediate operator actions. This knowledge area was tested in category 2.0 of the R0 examination and category 5.0 of the SR0 examination, respectively. Two(2) of the four(4) R0 candidates scored less than 80% in this category area (one of j these was less than 60%). Additionally, the SR0 candidate demonstrated l marginal performance on questions addressing immediate actions. These !

indicators are of concern considering the adverse safety significance l of inadequate operator knowledge and application of " Emergency Plant Evolution" precedures.

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One(1) R0 candidate failed the written portion of the examination.

-All of the other candidates passed their respective written examinations.

-4 Operating Examinations The operating portion of these examinations were administered on June'21-22, 1989. These examinations were conducted using walk-through

' scenarios that integrated events and safety systems, and included events identified in recent Licensee Event Reports. All of the candidates passed the operating portion of their examinations.

R The control room portion of these examinations identified ,three(3).

operational concerns: j

During a common scenario that encompassed the loss of a power

. range nuclear instrument (N/I) channel, the two procedures that were referenced by the operator candidates appeared to contradict _each other. Instrument test instruction, PICT-11-1, correctly indicates that the "0P-delta-T" trip is a'part of the affected instrument circuitry design. This apparently confused the' operator candidates since Off-Normal-Instruction, ONI 10.2.1, directed them to bypass all reactor trip functions associated with the failed power channel, except for the "0P-delta-T" trip.

The candidates were apparently not aware that the "0P-delta-T"-

trip input is gained to zero at Trojan, thereby having no operational effect (no functional need to bypass). This suggests that operator training and procedure clarification in the area of operational applications involving nuclear instrumentation could be improved.

The candidates had difficulty making procedural transitions, particularly when moving from "Off Normal Instructions" to specific system operating instructions in the course of their examination scenarios. Locating information in newly revised procedures became a time consuming effort. During this examination period, the licensee conducted a major procedure revision' evolution, and apparently did not provide their operating crews with adequate cross-referencing provisions during the revision period.

This handicapped the operator candidates throughout their operating examinations, and could adversely effect actual shift operations.

The operator candidates had difficulty applying procedure 01-4-3,

" Service Water System". They described various ways of transferring the service water " swing" pump between redundant system "A" and "B" trains - even with the procedure in hand. Upon further questioning, the candidates appeared to have adequate system knowledge, but could not implement the operating procedure consistently. The procedure is not of sufficient detail to accomplish the task without further training in this area.

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h The "in-plant" portion of these examinations identified two(2) generic training weaknesses in the following areas:

Candidates could only marginally describe the task for manually resetting the mechanical trip device on the unit's steam driven auxiliary feed pump. Terry Turbine. Specifically .they did not relate the proper positioning of the trip " rocker-arm" device onto the tripping base plate. Without properly seating this device,

'the-mechanical trip function could fail in an actual event.

However, all of the candidates tested in this area knew the physical location of the tripping unit, and the basics of how it operated.

' Car.didates could not adequately describe the operation of the local computer for the remote shutdown panel. They did not know;

-how to transfer control of associated safety systems from the control room to the remote shutdown panel ("decoupling operations").

The licensee representatives stated that they would look at the above stated deficiencies.

5. Exit Meeting On June 23,.1989, the NRC staff examiners met with the Senior Resident and representatives of the licensee's staff to discuss the examination.

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- . ES-201 Attachment.A '

. l REQUIREMENTS FOR' FACILITY REVIEW OF WRITTEN EXAMINATIONS At the option of the Chief Examiner, the facility may review the written examination up to two weeks prior to its administration. This review may take place at ,the facility or in the Regional office. The Chief Examiner will coordinate the details of the review with the facility. An NRC examiner will always be present during the review.

Whenever this option of examination review is utilized, the facility reviewers will sign the following statement prior to being allowed access to the examination. The examination or written notes will not by retained by the facility. - - i

a. Pre-Examination Security Agreement I . G,REGdW f . Ete uct information concerning the replacement (or initial) examinationagree th scheduled for anne 20. 1989 _ to any unauthorized persons. I understand that I am not to participate in any instruction involving those reactor operator or senior reactor operator applicants scheduled to be administered the above replacement (or initial) administered.from examination now until after the examination has been I understand that violation of this security agreement could result in the examination being voided.

hh@.Nb Signa'ttire/Date b45-83 In addition, the facility staff reviewers wi~11 sign the following I statement after the written examination has been administered.

b. Post-Examination Security Agreement I'C e tWO.EMfEM M did not, to the best of knowledge, divulge any information concerning the written examination administered on . lune 20. 1989 to any unauthorized persons. I did not participate in providing any instruction to those reactor operator and senior reactor operator applicants who were administered the examination from the time that I was allowed access to the examination.

Ihmd %b W59 SignatWe/Date Examiner Standards

_ _ _h _ _ . . _ - _ . . _ _ . . ._______.m_..__.m._

_ ___.____.____.___._____.____.___-_____m_O._________._._____..____

- ES-201 4ttachment A Rev 5 01/0p89

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I REQUIREMENTS FOR' FACILITY REVIEW OF WRITTEN EXAMINATION '

At t 1e option of the Chief Examiner, the facility may review the written examination up to two weeks prior to its administration. This review may take place at .the facility or in the Regional office. The Chief Examiner :

will coordinate the details of the review with the facility. An NRC examiner will always be present during the review.

l Whenever this option of examination review is utilized, the facility !

reviewers to will sign the following statement prior to being allowed access the examination.

by the facility. The examination or written notes will not by retained

{

a. Pre-Examination Security Agreement I /$ory u. 6/h information concerning the replacement (or initial) examinationagree t scheduled for anno 20. 1989 to any unauthorized persons. I understand that I am not to participate in any instruction involving those reactor operator or senior reactor operator applicants scheduled to be administered the above replacement (or initial) examination from now until after the examination has been administered.

I understand that violation of this security agreement could result in the examination being voided.

E24*/ 4I: 5 SignatLre/Date d E9 In addition, the facility staff reviewers will sign the following statement after the written examination has been administered.

b. Post-Examination Security Agreement I darv dd, [///$

divulge'any information concerning the written examinationdid not, to the administered on June 20, 1989 to any unauthorized persons.

did not participate in providing any instruction to those reactor I operator and senior reactor operator applicants who were administered the examination from the time that I was allowed access to the examination.

} AM4/ Wi &bC f'7 Signatiure/Date

{ Examiner Standards

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Attachment B

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Por't land General Electric Company . !

< Trojan Nuclear Plant i

UN4k 'V L L 71760 Columbia River Hwy -

Rain'ier, Oregon 97048 '

J (503) $56-3713< , ,!28 al0. 33

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' June 22, 1989

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i Mr. Dennis' Kirsch, Chief- ,'

Reactor Safety Branch, Region V United States Nuclear Regulatory Commission 1450 Maria Lane -Suite-210 .-

. Walnut Creek, California 94596 5368- ,

Dear Mr.. Kirsch:

In accordance with the O'perator Licensing Examiner S'tandards

.(NUREG-1021, ES-201) a pre-examination review was conducted in Walnut

' Creek on June 13, 1989 of the written examination administered at Trojan on June 20, 1989. The review was conducted by Mr. Tom Meadows of your staff. and Messrs. Cary Ellis 'and Greg Enterline of the Trojan-staff.

A' post-examination review was conducted after the examination. All comments by the facility staff have been fully resolved, and there are no follow-up written comment,s.

If you have any questions, please call Mr. Steve ichols at (503) 556-3713 extension 215.

Sincerely, C. P. Yundt Trojan General Manager e

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Ivl A T T E R SRO- /<E Y U. S. NUCLEAR REGULATORY COMMISSION SENIOR OPERATOR LICENSE EXAMINATION REGION V FACILITY: TROJAN NUCLEAR-PLANT REACTOR TYPE: WESTINGHOUSE. PWR. 4 LOOP DATE ADMINISTERED: JUNE 20. 1989 INSTRUCTIONS.TO CANDIDATE:

Use separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.

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CATEGORY- % OF CANDIDATE'S CATEGORY

/VALUE TOTAL SCORE VALUE CATEGORY-x v. z 24.00 W".w y, r

4. REACTOR PRINCIPLES (7%)

THERMODYNAMICS (7%) AND COMPONENTS (10%)

[ FUNDAMENTALS EXAM)

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M -=.3!MHr7- 5. EMERGENCY AND ABNORMAL PLANT m *, <x y,,. EVOLUTIONS (33%)

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43.00 4W 6. PLANT SYSTEMS (30%) AND

,+ PLANT-WIDE GENERIC y,,/f y RESPONSIBILITIES (13%)

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<A y-, FINAL GRADE All work done on this examination is my own. I have neither given nor received aid.

- - - - KEY ***** -

Candidate's Signature fi n e /w r -

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PROCEDURES FOR THE ADMINISTRATION OF WRITTEN EXAMINATIONS

1. Check identification badges.

2. Pass out examinations and all handouts. Remind applicants not to review examination until instructed to do so. '

READ THE FOLLOWING INSTRUCTIONS VERBATIM:

During the administration of.this examination the following rules apply:

~1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.

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2. After the examination has been completed, you must sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examination. This must be done after you complete the examination.

READ THE FOLLOWING INSTRUCTIONS 1.' ' Restroom trip rarb to be limited and only one applicant 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.

2. Useblaikinkordarkpencilonlytofacilitatelegiblereproductions.

3. Print your name in the blank provided in the upper right-hand corner of the examination cover sheet.

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

.. 5. You may write your answers on the examination question page or on a separate sheet of paper. USE ONLY THE PAPER PROVIDED AND DO NOT WPITE ON i THE BACK SIDE OF THE PAGE.

6. If you write your answers on the examination question page and you need more space to answer a specific question, use a separate sheet of the paper provided and insert it directly after the specific question. DO NOT WRITE ON THE BACK SIDE OF THE EXAMINATION QUESTION PAGE.

7. Print your name in the upper right hand corner of the first page of each section of your answer sheets whether you use the examination question pages or separate sheets of paper. Initial each page. -

8. Before you turn in your examination, consecutively number each answer sheet, including any additional pages inserted when writing your answers on the examination question page.

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If you are using separate sheets, number each answer as to category and number (i.e. 1.04,6.10) and skip at least 3 lines between answers to allow space for grading.

10. Write "End of Category " at the end of your answers to a category.

11. Start each category on a new page.

12. Write "Last Page" on the last answer sheet.

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

Avoid using symbols such as < or > signs to avoid a simple transposition -

error resulting in an incorrect answer. Write it out.

- 14. The point value for each question is indicated in parentheses after the question. The amount'of blank space on an examination question page is NOT an indication of the depth of answer required.

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

16. Partial credit may be given.

Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK. NOTE: partial credit will NOT be given on multiple choice questions.

17.

Proportional grading will be applied. Any additional wrong information that is provided may count against you. For example, if a question is worth one point and asks for four responses, each of which is worth 0.25 points, and you give five responses, each of your responses will be worth 0.20 points. If one of your five responses is incorrect, 0.20 will be deducted and your total credit for that question will be 0.80 instead of 1.00 even though you got the four correct answers.

18.

If the intent of a question is unclear, ask questions of the examiner only.

19. When turning in your examination, assemble the completed examination with examination questions, examination aids and answer sheets. In addition, '

turn in all scrap paper.

20. To pass the examination, you must achieve an overall grade of 80% or greater and at least 70% in each category.

21. There is a time limit of (6) hours for completion of the examination.

(or some other time if less than the full examination is taken.)

22. When you are done and have turned in your examination, leave the examination area (DEFINE THE AREA). If you are found in this area while the examination is still in progress, your license may be denied or revoked. _

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o CATEGORY 4

. REACTOR PRINCIPLES ~(7%),[ THERMODYNAMICS (7%),.

'AND COMPONENTS'(10%).

Group-I' Reactor Theory-1.50%

0UESTION.4.01 :(.75) **

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE

-During a reactor.startup, an. initial reactivityiaddition causes count rate to increase from 20 CPS'to 40 CPS subcritical. A second reactivity

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addition causes'the count rate to increase from 40 CPS to 80 CPS suberitical.

Which of the following statements is correct?

a.. The first reactivity addition was smaller.

.b. The second reactivity ~ addition was smaller.

c. The first and second reactivity additions were equal.

'd.. There'is insufficient' data given to determine relationship of reactivity values.

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.* ANSWER

b. (0.75)

REFERENCE

= License Training Program 02-H-04-LP, Neutron Sources and Subtritical Multiplication Page 17 of 20 KA 192OO8K103 3.9/4.9 1

~* QUESTION 4.02 (0.75) **

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE A reactor is initially subtritical with a Keff of 0.95 and a source range count rate of 200 counts per second (CPS). Control rods are subsequently withdrawn to add 0.025 delta k/k reactivity.

What is the subtritical steady state count rate after this reactivity addition?

a. 250 CPS

b. 300 CPS

c. 350 CPS

d. 400 CPS

ANSWER

d. (0.75)

CR2/CR1 = (1 - Keff1)/(1 - Keff2)

CR2 = 2OOCPS (1 - 0.95)/ (1 - 0.975) = 200 CPS (0,050)/(0.025)

CR2 = 400 CPS.

REFERENCE License Training Program lesson plan 02-H-04-LP, Neutron Sources and Subcritical Multiplication Page 18 of 20 KA 192OO8K103 3.9/4.0 1

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Group II Reactor Theory- 4.5%

OUESTION 4.03 (0.75) l MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE Which of the f ollowing describes the components of the power def ect ;

in an INCREASING order of significance (reactivity value) at Beginning of Life (BOL)?

a. . Void, Doppler, MTC l

b. Void, MTC, Doppler

c. MTC, Void, Doppler i

d. MTC, Doppler, Void

ANSWER

b. (0.75)

REFERENCE License Training Program lesson plan 02-H-06-SD, PWR PHYSICS Figure 32 KA 192OO4K108 3.1/3.1 l

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QUESTION 4.04 (0.75) '**

MULTIPLE. CHOICE-SELECT THE CORRECT RESPONSE Which of the following describes Xenon behavior during plant operations?

a. Peak Xenon conditions af ter a reactor trip wi~11 preclude attaining criticality near the beginning of cycle at Trojan.

.b. Equilibrium Xenon concentration' at 50% power is greater that half of-the equilibrium concentration at 100% power.

c. If the reactor' trips from 100% power, peak. Xenon concentration'will-occur in about'50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />.

d. Xenon peaking after a reactor shutdown is experienced because the reduced flux increases production about 25%.

ANSWER j

-b. (0.75) l

REFERENCE License Training Program 02-H-06-SD, PWR PHYSICS Figure 27 KA 192OO6K107 3.4/3.4 e

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QUESTION 4.05 (0.75) ,

4 MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE Which of the f ollowing describes the behavior of Xenon?

c. Xenon concentration initially increases when power is increased,

b. Xenon concentration initially decreases after a reactor trip.

c. Xenon concentration initially increases when power is decreased.

d. Shutdown equilibrium Xenon concentration after a trip is a function of the previous power level.

ANSWER

c. (0.75) l

REFERENCE License Training Programs 02-H-06-SD3, PWR PHYSICS Page 62 of 80.

KA 192OO6K105 3.1/3.1 e

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- *OUESTION 4.6 '(0.75)

. MULTIPLE CHDICE-SELECT THE CORRECT RESPONSE

.. Choose.the correct phrase to correctly' complete'the sentence.

As.the core ages, the ratio of PU239 atoms. to ' U235- atoms- increases.

This, changing ratio causes the...

- a.- reactor-period to decrease.

b. ' Void Coefficient to become less negative.

c.. Moderator Temperature Coef ficient to become less negative.

d-- delayed neutron fraction to increase.

.-* ANSWER a.- (0.75)

REFERENCE License Training Program 02-H-05-SD, REACTOR KINETICS, Page 16 of 30 KA 192OO3K106 3.2/3.3.

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QUESTION 4.7 (0.75)

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE What is the major constituent of the power coefficient at BOL? I i

2. Moderator temperature coefficient

b. Doppler coefficient

c. Doron coefficient

d. Void coefficient

ANSWER

b. (0.75)

REFERENCE License Training Program 02-H-06-SD, PWR PHYSICS, Figure 32 KA 192OO4K108 3.1/3.1 I

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QUESTION 4.8 (0.75) **'

MULTIPLE CHOICE -SELECT THELCORRECT RESPONSE l k

l~ Trojan" plant uses both soluble baron and. control rods to-control excess reactivity in the reactor.

What is the main- advantage ' of using soluble baron'to control reactivity?-

fa.. It has a minimal ef f ect on the neutron flux shape which reduces.

the maximum-to-average power density ratio.

b. It does not effect the rod worth which insures that. Rod Insertion Limits'are-reliable at different baron' concentrations..

c. Boron primarily absorbs fast neutrons; this allows more efficient' use.of thermalized neutrons by the fuel, d.- It increases reactor loading rates which allows f or rapid power.

. changes to follow customer demand.

'* ANSWER

a. (0.75)

REFERENCE License Training Program 02-H-06-SD1, Page 16

KA 192OO7K105 3.0/3.1 8

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[ Group III REACTOR THEDRY-1.00%

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DUESTION 4.9 (1.00)

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE 1

l Which of the f ollowing defines reactivity?

a. The time it takes reactor power to change by a f actor of "e",

b. It always occur indirectly f rom fission through fission fragment daughter decay.

c'. The f ractional change in neutron population per generation.

d. High energy neutrons (>1.OMev).

ANSWER

c. (1.00)

REFERENCE License Training Program 02-H-05-SD, REACTOR KINETICS Page 23 of 30 KA 192OO2K111 2.9/3.0 KA 192OO1K102 2.4/2.5 KA 192OO2K111 2.9/3.0 9

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GROUP I THERMODYNA:11CS -3%

DUESTION 4.10 (0.75) #*

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE.

10CFR50.46 establishes several ECCS acceptance criteria; the peak cladding temperature 12 mit is set at 2200 degrees F.

Why was the value of 2200 degrees F chosen?

a. It allows f or a 10% safety margin from the cladding melting point (2 ISO degrees F)

b. Any clad temperature higher than this correlates to a fuel center line temperature at the fuel's melting point.

c. The zircalloy-water reaction is accelerated at temperatures above 2200 degrees F.

d. The cladding becomes stronger, because of a =irconium phace change at tempere^ures above 2200 degrees F.

ANSWER

c. (0.75)

REFERENCE License Training Program 02-H-OB-SD2, Accident Analysis, Page 29 of 266.

KA 193OO9K105 3.1/3.5 I

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QUEETION 4.11 (0.75) **

MULTIPLE CHDICE-SELECT THE CORRECT RESPONSE The Enthalpy Rise Hot Channel Factor limit in Technical Specifications cannot be directly observed by the operator. Assurance that this limit is not exceeded is based on maintaining ceveral related plant parameters within limits.

Which limit helps ensure that the Enthalpy Rise Hot Channel Factor limit is not exceeded?

a. Control rods in group move together within +/- 12 stepc.

b. Tavg vs. Tref are kept matched to within 5 degress F.

c. Thermal power is limited to 3411 megawatts.

d. RCS minimum flow for modes 1, 2, and 0 is 90% of rated flow.

ANSWER

a. (0.75)

REFERENCE Technical Specification Bases Power Distribution Limits 3/4.2.2 and 3/4.2.3 KA 193OO9K107 2.9/3.3 i

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QUESTION 4.12 ( 0. 7 5 ) -

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE Which one of the following represents the maximum linear power density which would be expected in the core during full power operations?

a. Local Power Density multiplied by Nuclear Peaking Factor.

b. Radial Peaking Factor multiplied by the Local Peaking Factor,

c. Average Kw/ft for the core multiplied by the Nuclear Peaking Factor.

d. Nuclear Peaking Factor multiplied by the Maximum Local Power Density.

ANSWER

c. (0.75)

REFERENCE License Training Program 02-H-13-LP, Thermal Hydraulics, Page 24 of 43.

KA 193OO9K107 2.9/3.3 l

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QUESTION 4.13 (0.75)

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE The Tro.4an Boration injection system ensures that negative reactivity control is available during each mode of facility operation. The boron inventory in the RWST and boric acid storage tanks have Technical Specification limitations and conditions.

Which of the following is one those Technical Specification conditions?

' a. To maintain a constant RCS reactivity while the temperature.is 551' degrees F.

b. To compensate for an inadvertent positive reactivity addition to RCS of approxswetgly 1*/. deltaK/K while in mode 5 at 2OO degrees F.

c. The maximum boration capability requirement occurs at the Beginning of Life (BOL).

d. The required RWST volume has been decreased due to tank geometry and letdown considerations.

'

ANSWER

b. (0.75)

REFERENCE TROJAN T/S Page B 3/4 1-3 KA 193010K107 3.8/4.1 l

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GROUP II THERMODYNAMICS-2.25

QUESTION 4.14 (0.75)

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE Which ONE of the following statements is CORRECT concerning a secondary calorimetric ?

a. If feedwater temperature is read erroneously high, then the calculated reactor power will be higher than actual.

b. The calorimetric equation does NOT take into consideration the heat added by the reactor coolant pumps, or the heat lost to the containment atmosphere (ambient losses).

c. Mass flow rate of the secondary system is determined by totaling the average steam flows from each of the three steam generators.

d. The results of the secondary calorimetric may be usc-d as the basis for calibration of the power range nuclear i n st rument ati on .

ANSWER

d. (0.75)

REFERENCE

License Training Program 02-H-11-LP1, HEAT TRANSFER METHODS, Page 5 of 56.

KA 193OO7K108 3.1/3.4 193OO7K106 3.1/3.3 14

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QUESTION 4.15 (0.75) **

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE During power operations, a percentage of RCS flow through the reactor vesse1~bypassus the fuel assemblies.

Which statement describes core bypass flow?

a. Bypass flow is created by design tolerances between vessel components and serves no useful function.

b. Bypass flow cools vessel components and is greater than 4% of total core fl ow.

c. Bypass flow maintains suf ficient cooling flow in the RCS to prevent RCP cavitation.

d. Bypass flow is less that 2% of total core flow and assures that the control rod guide tubes remain clear.

ANSWER

b. (0.75)

REFERENCE Licensee Training Program 02-A-05-SD2, pages 31, 32 KA 193OOBK120 2.9/2.9 15

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QUESTION 4.16 (0.75) **

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE Departure from Nucleate Boiling (DNB) can cause rapid increase in fuel centerline and cladding temperatures by greatly reducing the heat transfer coefficient.

Which of' the f ollowing best describes the cause of DNB7

a. High heat flux produces boiling bubbles that begin to form a film on the f uel rod before they can be swept away.

b. High heat flux raises the bulk coolant temperature to greater than saturation and the coolant instantly flashes into a steam blanket.

c. Excessive local heat flux begins to rupture the fuel cladding.

d. Boiling bubbles begin to cause cavitation erosion to occur excessively on the. cladding.

ANSWER

a. (0.75)

REFERENCE License Training Program 02-H-08-SD, Page 13 KA 193OO8K104 3.1/3.3 t

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b GROUP III THERMODYNAMICS-1.75%

L

QUESTION 4.17 (0.75) **

L NULT'IPLE' CHOICE-SELECT THE CORRECT RESPONSE ~

Which one of-. the f ollowing events would cause water hammer 7

a. . Starting a centrifugal pump in a filled (vented)' system l with its discharge valve shut.

b.-Rapid depressurization during a hydrostatic test of a large

. system.

c. Rapid pressurization.of an solid' stable system.

.d. Sudden closure of a valve in which there is high water flow.

ANSWER

d. (0.75)

REFERENCE License Training Program 02-H-12-LP, FLUID FLOW, Page 39 of 43.

KA 193OO6K110 3.3/3.4-17

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QUESTION 4.18 (1.00)

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE Which one of'the following changes is likely to cause the probability of cavitation to decrease in a system containing a centrifugal pump?

a. Suction pressure decreases.

b. Suction temperature decreases.

c. Pump speed increases.

d. Suction line isolation valve is partially shut.

i

ANSWER

b. (1.00)

REFERENCE License Training Program 02-H-12-LP, FLUID FLOW, Page 27 of 43.

Exam Bank Question 02H12-4-17 KA 193OO6K111 3.1/3.3 M

1B

GROUP 1 COMPONENTS- 6%

l.

DUESTION 4.19 (0.75) **

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE A motor operated Engineered Saf eguards valve was manually shut f or surveillance testing and is being returned to service per Operations Procedure OM-3-1, Tagging.

What action must be taken bef ore the valve is declared f ully operable?

a. It must be manually returned to its original position.

b. It must be Danger Tagged until administrative approval is signed by the Shi f t Supervi sor.

c. It must be electrically cycled.

d. The operating mechanism must be lubricated with an alcohol and graphite solution.

ANSWER

c. (0.75)

REFERENCE Operations Manual OM-3-1 section 3.10.1 KA 191001K106 3.3/3.7 1

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QUESTION 4.20 (0.75) **

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE After 100% power operations for a prolonged time, an intermediate loss of coolant accident and saf ety injection occurs, complicated by a partial ATWS. The Reactor Vessel has voids occurring:

Which one of the following events would result in a larger i ncrease in the Source Range detector output?

a. Voiding occurring in the upper half of the core,

b. Voiding occurring in the upper half of the downtomer.

c. Voiding occurring in the lower half of the core.

d. Voiding occurring in the lower half of the downtomer.

ANSWER

d. (0.75)

REFERENCE License Training Program 02-B-07-LP, EXCORE NUCLEAR INSTR'UMENTATION SYSTEM, Page 14 of 56.

02-BO7-LP Obj.1.2.2.G.9 KA 191002K117 3.3/3.5 6

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QUESTION 4.21 (0.75) **

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE What type of detector is used for the source range nuclear instrumentation?

a. Compensated Ion Chamber.

b. Fission Chamber,

c. Proportional Counter.

d. . Uncompensated Ion Chamber.

ANSWER

c. (0.75)

REFERENCE License Exam Bank Question No. B-07-B2-1 KA 191002K118 2.6/2.8 21 I .

L li Ls L w-l r . .

QUESTION 4.22 (0.75) ** l MULTIPLE CHDICE-SELECT THE. CORRECT' RESPONSE An. operator drops his' Digital Alarming Dosimeter (DAD) while working in a radiation area. . Visual inspection reveals no obvious damage to the DAD 7' 1

What action should the operator take7

a. Since the DAD passed a visual inspection, no action.is necessary; his TLD will be used to determine his legal dose..

' b. The operator should leave the radiation area and report the dropped-DAD to. Radiological. Protection.

c. .The DAD is designed to take minor-impacts, however, the event should be reported to Radiological Protection whenever the Dad is: turned'in.

d. The operator should report the dropped DAD to the Shift Supervisor because the DADS cost more than $1000.00, 9

ANSWER

b. (0.75)

REFERENCE General ' Employee T; aining conducted on 2/21/89.

KA 191002K119 3.1/3.3 22

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QUESTION 4.23 (0.75) **

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE Prior to closing any alternating current (A.C.) circuit breaker,'the

-two power sources must be in phase to prevent component damage.

How is this requirement assured when closing the Main Generator output.. breakers?

a. The output ~~ breakers are interlocked with the syncroscope to prevent closing unless both lines are in phase.

b. The output breakers are always closed onto .a deenergized tids #

so that paralleling is not necessary.

c. The operator must properly. parallel across the breaker, no electrical interlocks are provided.

'd. All 230 KV breakers are the break-before-make design which allows for any breaker closing sequence.

ANSWER

a. (0.75)

REFERENCE O2-C-02-SD1, Page 9 of 27 02-C-01-LPI Objective 1.2.2.B.2.i ,

KA 191008K107 3.0/3.3 .l l

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I*DUESTION ~ 4.24- (O.75) **

MULTIPLE CHOICE-SELECT THE-CORRECT RESPONSE

. What. type (s) 'of' radiation is/are measured by a' Digital . Alarming

-Dosimeter?

a. : ' Gamma on1y.

b. . Gamma and Beta.

< 'c. Gamma'and Neutron.

~d. . Gamma, Beta. and-Neytron..

i

ANSWER

a. (0.75)

REFERENCE

, License Training Program, G 1 -F-02--H O , Page 7 (General Employee Training)

KA 191002K119 3.1/3.3 l

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QUESTION 4.25 (0.75)-

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE Which.one of the following statements about pump Net' Positive Suction !

Head- (NPSH) .is CORRECT 7 i A. NPSH can QNLY be calculated by. adding the suction pressure-and the discharge pressure.

D. When.a pump is started, the NPSH'will decrease by the amount-of the. pressure drop.in the suction piping.

C. NPSH is. essential for' operation of centrifugal pumps but not

. for positive displacement pumps.

D. NPSH can QNLY'be' calculated'by subtracting the suction pressure from the. discharge pressure.

ANSWER B.. (0.75)

REFERENCE License Training Program 02-H-12-LP, FLUID FLOW, Page 25 of 43.

KA 191004K106 3.2/3.3 25

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QUESTION.4'.26 (0.75)

U . . .

l MULTIPLE! CHOICE-SELECT THE. CORRECT RESPONSE u ::p

!- - During. Trojan. plant'cooldown and depressurization with forced.

circulation, RCS loop; flow indication becomes erratic. ..

~

&V ' What is.the most likely.cause of the erratic RCS

- loop M1ow: indication?

, A.. l R C P -' r u n o u t .

' B. : RCP cavitation.

. C.- ..RCS; hot' leg subcooled.

D.- RCS loop water hammer..

s

-

ANSWER B. CRCP cavi tation3 CO.753

-

REFERENCE

' License Training Program;02-H-12-LP, FLUID FLOW, Page 25 of 43.

KA.191004K106 -3.2/3.3 i

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QUESTION 4.27- (.75) **-

, MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE IThe numb'er of starts over aJgiven time period is limited for the Reactor Coolant Pumps.

'Wh.y is this. limit required?

a. TheLmotor start capacitors are air cooled and require significant-(30 minutes) cooldown time.

b'. High: motor starting currents can overheat and damage motor windings if. adequate cooling time is not allowed.

+

c. . High heat loads on the motor lube oil system from excessive cold h -g motor starts cen cause motor binding.

'd. .The split phase induction design motors require high starting' currents that may overload supply breakers if a rapid restar t is

. performed. ,

ANSWER

b. (.75)

REFEREldCE Operating Instruction 3-4, Rev. 19, REACTOR COOLANT SYSTEM

-NORMAL OPERATION, Page 5 of 10.

KA 191005K106 3.0/3.1 27 i I

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QUESTION 4.28 (1.75) i

.- Troj an - pl ant is operating at-100% power with RCS'Tavg at 587 F.and a steam pressure of 980 psig. Tavg must be' changed in order to maintain a

-steamfpressure of 980 psig at 100% power with 10% of the tubes in each steam generator plugged.

l'

- Using the' STEAM TABLES and assuming identical steam generator's, what;must this new TAVE'be 7 (1.50 for application, .25 value)

SHOW ALL WORK, including any applicable formulas.

ANEWER (1.50 for application, 0.25 f or value)

S/G heat transfer = Q = UA(Tavg Tstm) .(0.25F 0, U, and Tstm remain constant;-

A1(Tavgi - Tstm) = A2(Tavg2 - Tstm) (0. 25 )

Given: A2= 0.9 X A1 PSIA ='PSIG + 15 = 980 + 15 = 995 (+/ . 5) psi a (0.25)#

- From Steam Tables: Tsat for 995 psia = 544 (+1/-1) F (0.50f A1(587 - 544) = 0.9A1(Tavg2 - 544) (0.25)#

Tavg2 = 591.8 F (+1/-1) (0.25)#

.

REFERENCE License. Training Program 02-H-11-LP, HEAT TRANSFER METHODS, Page 37 of.56, and STEAM TABLES KA 191006K113 2.8/2.9 l

28

QUESTION'4.29 (O.75) **

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE Compare ' the indications that would be ' observed if ' the West RHR Pump were started and operated with its discharge flow path isolated, as compared

. to with its discharge flow path open (ASSUME RUNOUT FLOW CONDITIONS).

What are the indications f or the " isolated discharge" case?- j

a. Higher starting current and lower running amperage.

b.. Lower starting current and lower running amperage.

c. . Higher starting current and higher running' amperage.

d. Lower' starting current and higher running amperage.

ANSWER

b. (0.75)

REFERENCE License Training Program 02-H-12-LP, FLUID FLOW, Page 22 of 43 KA 191005K104 2.7/2.8 l

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QUESTION 4.30 (0.75) **

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE The B Emergency Diesel Generator is operating at 95% load in parallel-with the Normal 4.16 KV Vital Bus power supply for surveillance testing.

The operator places the GOVERNOR Raise / Lower Control Switch to the RAISE position.

How will the Emergency Diesel Generator (EDG) respond?

a. Generator Kilowatts will increase until either the operator releases the switch or the EDG trips on overcurrent.

b. Generator Frequency, RPMs, and Kilowatts will increase until either the operator releases the switch, or the EDG trips on overspeed or overcurrent.

c. Generator parameters will not change because the Governor Limit Switch will not allow EDG loading above 100% unless an ESF signal has been received.

d. Generator Frequency and RPMs will increase until either the operator releases the switch or the EDG trips on overspeed.

ANSWER

a. (0.75)

REFERENCE Lesson Plan 02-A-13-SD, Emergency Diesel Generator 02-C-04-LP-Obj.1.2.2.c KA 064000K101 4.1/4.4 END OF CATEGORY FOUR 30

p t

.le V.

CATEGORY 5 EMERGENCY AND ABNORMAL PLANT EVOLUTIONS (33%)

GROUP 1 EMERGENCY AND ABNORMAL PLANT EVOLUTIONS-19%

4 QUESTION 5.1 (3.00) **

A fire is burning out of control in the Turbine Building and has been confirmed by plant operators. The fire has not been contained and is spreading. The Control Operator announces the alarm over the plant paging system, and repeats'the announcement. You are the Shift Supervisor and have been informed. You are executing EFP-0, " Procedure in the Event of a Fire. '

a. What are the two(2) notifications that EFP-O requires the Control Opert to make upon-sure indication of a~ fire? (0.75 each)

b. What are two (2) of the three (3) conditions in EFP-O that require execution of EFP-1, Alternative Shutdown for Evacuation of Control-Room Caused by Fire? (1.5)

ANSWER

a. Notify the (Rainier) Fire Department (0.75)

Notify the Security Watch Supervisor (0.75)

b. (any two of the f ollowing; 0,75 points each)

Large fire threatens redundant equipment (AND is out of control longer than 15 minutes).

Difference exist between 2 or more CATEGORIES of instruments (AND significant loss of instrumentation).

The control room is physically uninhabitable.

REFERENCE --

EFP-0, Procedure in the Event of a Fire, Page 2a and Page 5.

02-K-09-LP Objectives 1.2.2.F and H KA OOOO67K304 3.3/4.1 31

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QUESTION 5.2 (2.00) **

The Critical Safety. Function Status Trees (CSFSTs)

.and portions of.ECA-0.0 are attached to this exam.

Station Blackout (loss of onsite and offsite AC power) has occurred and ECA-0.0, Loss of All AC Power is being executed. Step 4 was just completed when core exit thermocouple exceeded 1200 degrees F (CSFST RED path for core cooling).

a. How does ECA-0.0 inform the operator to execute Step 5 of ECA-0.0 instead of going to FR-C.1 Step-1, Response to Inadequate Core Cooling? (1.0)

^

b. 'What is the basis for NOT executing FR-C.17 (1.0)

ANSWER

a. The NOTE at beginning of ECA-0.0 states that CSFSTs are monitored for information only. (FRs should not be implemented) (1.0)

b. FR-C.1 (all FRs) assume that AC power is available to at least one 4.16 KV vital bus. (1.0) !

REFERENCE ECA-0.0, Loss of All AC Power, Page 2, and Westinghouse ERGS for ;

ECA-0.0 step 1 note 2 l 02-K-22-LP Objectives 1.2.2.c and 1 KA OOOO55G012 3.9/4.0 OOOO55K302 4.3/4.6 l

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QUESTION 5.3 (3.5) **

l CSFSTs are-attached to'this exam; ASSUME degraded containment conditions. ]

I During power operations, an accident has occurred causing a reactor trip signal and degraded core cooling. . FR-C.2, " Response to Degraded Core Cooling", is being executed as required by the Critical Safety Function Status Trees (CSFSTs). RVLIS currently indicates 50%.

The CSFSTs require departure from FR-C.2 to perform other emergency actions if certain plant. conditions (limits) are exceeded.

What are the plant conditions and setpoints requiring departure f rom FR-C.2 f or each of the f ollowing CSFST Priorities?

(Include numerical limits, when applicable)

(a) HEAT SINK (2 conditions) (0.5 each) (1.00)

(b) SUBCRITICALITY (2 conditions) (0.5 each) (1.00)

(c) CONTAINMENT (1 condition) (0.5) (0.50)

(d) RCS INTEGRITY (2 conditions) (0.5 each) (1.00)

(3.5 Total)

ANSWER (3.5 Total)

(a) Total feed capability less than (or equal to) 720 gpm/. (0.50)

AND Narrow Range SG 1evel less than (or equal to) 12%. (0.50)

(b) Power range channels greater than (or equal to) 5%/ (0.50)

OR positive intermediate range start-up rate. (0.50)

(c) Containment pressure greater than (or equal to) 60 psig. (0.5)

(d) Cooldown in all RCS cold legs greater than (or equal to) 100 degrees F in one hour (0.50)/ AND RCS pressure / temperature to the left of Limit A. (0.50)

REFERENCE FR-0, Critical Safety Function Status Trees 02-K-04-LP Objective 1.2.2.9 KA OOOO74G012 4.3/4.4 33

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.* QUESTION 5.4 . (1.00)

- MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE Off-Normal Instruction,0NI-7, Reactor. Control or Rod Position Indication Malfunction, provides guidance for the operator during a rod control or indication malfunction. An abnor mal control rod position (dropped rod) has the symptoms of rod bottom / deviation alarms, changing. primary. plant parameters, and flux tilt alarms. In additions an excessively misaligned control rod can produce significant distortions in the local core power distribution.

Which of the~following is NOT a concern / action for an excessively misaligned control rod per DNI-77

a. A change in the conditions of nearby fuel rods (Xenon concentration, pellet / clad interaction).

b. An increase in potential for localized fuel damage in '

nearby fuel rode.

c. : A large local power increase in the redded fuel assembly.

d. The recovery from the excessive misaligned control rod must be performed as fast as possible.

ANSWER

d. (1.00) i

REFERENCE ONI-7, Reactor Control or Rod Position Indication Malfunction, Page 3.

KA OOOOO3K304 3.8/4.1

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-*DUESTION 5.5 (1.50) ** ,

Troj an's Technical Specifications require ALL movable Co6 trol Assemblies to be OPERABLE and positioned within +/- 12 steps of their group step counter demand position. Rod Insertion Limits, control rod drop times, and shutdown rod positions are also limited.

What are the three (3) operational reasons for rod insertion limits?

(0.50 each) (1.5)

ANSWER

a. Acceptable power distribution limits are maintained. (0.5)

6. The minimum SHUTDOWN MARGIN is maintained. (0.5)

c. The potential ef f ects of a rod ejection accident are limited. (0.5)

REFERENCE-Trojen Technical' Specification, Bases 3/4.1.3 MOVABLE CONTROL ASSEMBLIES, Page B 3/4 1-3.

02-B-09 LP Objective 1.2.2.k KA OOOOO5K303 3.4/4.1 M

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OUESTION 5.6 (0.75)

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE Off-Normal Instruction ONI-2, Reactor Coolant Pump (RCP) Trip / Trouble, provides guidance to the operator in the event of a RCP trip or RCP l trouble. If a RCP has tripped with the power level above P-9:

Which of the following immediate operator actions is performed first7

a. Defeat the affected loop Tave.

b'. Def eat the af f ected loop Delta-T.

c. Place the rod control in MANUAL.

d. Determine and verify which RCP has tripped.

ANSWER

d. (0.75)

REFERENCE ONI-2, Page 2 KA OOOO1AA103 4.0/4.0 36

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QUESTION 5.7 (0.75)

Off-Normal' Instruction ONI-2, Reactor Coolant Pump (RdP) ' Trip / i Trouble, provides guidance to the operator in the event of a RCP trip or RCP trouble. With a RCP motor bearing temperature greater than 250 degrees F and with'the power level above P-7.

p

-: Which of the f ollowing subsequent operator actions of DNI-2 is performed ? (0.75)

a. Place rod control in MANUAL.

b. Ensure RCP supply fans are operating.

c. . Monitor.RCP motor amps.

l d. Trip the reactor.

1 .

ANSWER

d. (.75)

REFERENCE ONI-2, Page 5 KA OOOO11A103 4.0/4.0 l

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DUESTION 5.8 (1.00) r MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE Emergency Instruction EI-1, Loss of Reactor or Secondary Coolant,

, provides operator actions to recover from a loss of reactor or secondary coolant. The first operator action is to perform the immediate action steps of EI-0, Reactor Trip, Safety Injection, and diagnosis. Several action steps later in EI-1, the operator is to determine if containment spray should be stopped.

Which of.the following action steps of EI-1 is correct in determining v.hether the Containment spray should be stopped?

a. Close NADH spray additive valves if the NADH tank lo-lo level alarm actuates.

b. Spray pumps-Running > 27 minutes.

c. Containment pressure > 10 psig.

5

d. Containment radiation.> 10 R/hr.

ANSWER

b. (1.00)

REFERENCE Emergency Instruction EI-1, Loss of Reactor or Secondary Cool ant, Page 7 KA OOOO11K312 4.4/4.6 38

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. QUESTION 5.9 (1.00) e MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE Off-Normal Instruction ONI-62, Excessive Reactor Coolant Pump (RCP)

Seal Leakage / Failure, provides immediate and subsequent actions f or RCP #1 or #2 seal failures. ONI-62 provides operator actions for rapid and slow transient failures of RCP #1 seal failures.

Which of the f ollowing Immediate Operator Actions per DNI-62 differenates between a rapid #1 seal degradation and'a slow transient failure? (0.75)

a. Increase VCT temperature by 10 degrees F using letdown controller.

b. Reduce VCT temperature to 10 degrees F using letdown controller.

c. Close the affected RCP #1 seal leakoff control valve.

d. Record seal injection flows and #1 seal leakoff flows.

ANSWER

c. (1.00)

REFERENCE Off-Normal Instruction ONI-62, Excessive Reactor Coolant Pump Seal Leakage / Failure, Page 2 of 5 KA OOOO15K207 2.9/2.9 I

39

QUESTION 5.10 (1.00)

ULTIPLE CHOICE-SELECT THE CORRECT RESPONSE In f-Normal Instruction -ONI-62, Excessive Reactor Coolant Pump (RCP)

Seal L akage/ Failure, the RCP #2 Seal High Leakage / Failure procedure symptoms re as follows: RCP #1 seal leakoff indicates < 1.1 gpm, RCP standp e level trouble alarms and increased leakage to the RCDT.

Which of the f lowing immediate operator actions of DNI-62 is attempted f or RCP #2 seal high leakage /f ailure? (1.00)

a. Initiate the RER

6. DO NOT restart affect RCP until cause of the seal

~

malfunction has been de rmined.

c. Attempt to r seat #2 seal b. 'ncreasing VCT pressure and/or decreasin. VCT pressure. , _ --

rwes-w d< ' 'A '

d. Contact Trojan Plant Engineering g up f or resolution.

ANSWER

c. (1.00)

REFERENCE Off-Normal Instructi on ONI-62, Reactor Cool ant Pump - cessive Seal Leakage / Failure, Page 4.

KA OOOO15A210 3.7/3.7 "O, c,Yzs y.

((

S**eaos 44: psh-y ,'

=> Tr- se ay we c~ru<r A*fnfWf.

-M8 i n z. -

40

LP pt li,'

1 D > -* QUESTION 5.11 '(1.00) **

~

' MULTIPLE CHDICE-SELECT THE' CORRECT RESPONSE Off-Normal Instruction ONI-14, Component Cooling Water (CCW)-System,

- j- ' describes the actions necessary to mitigate the consequences of a

~

malfunction of the CCW system.

l l

Which of .the f ollowing ' probable causes would initiate a loss of

' both CCW Trains per ONI-147 (1.00)

a. Electrical fault or overload in the CCW pump motor.

h. - A malf unction of the CCW pump supply circuit breaker.

c. System rupture resulting in significant loss of CCW inventory.

d. A 1'oss of power to the 4.16 KV bus A-1 or A-2.

,3

ANSWER

c. (1.00)

REFERENCE Off-Normal Instructi on ONI-14, Component Cooling Water, Page 6 Of 19 KA OOOO26A202 2.9/3.6

~

41

i

. r.:

M., =- J l l::.

t !* QUESTION 5.12 - ( 1. 50 ) -

l During Trojan p'lant operations;-an accident occurs such that a loss . of both Component ' Cooling . Water: (CCW)' Trains:

l: occur,-andthe CCW. Trains can NOT be restored to the Reactor' Coolant: Pumps-(RCPs).-

What . two ' i mmediate operator ' actions per DNI-14.' (Loss of- CCW)

~

i; -are required after^the loss of both CCW trains? !

(two at 0.75 each) (1.50) ;

i

~

i

1'

ANSWER' (Two at 0.75 each) -l Trip the reactor :

(0.75)

Trip'the RCPs (0.75) i.

-

REFERENCE .

Off-Normal Instruction,0NI-14, Component Cooling Water, Page 4 of 19 ~

{

.KA OOOO26A106 2.9/2.9 :

1 i

) -. I ' g l

'l l

1 i

i

)

42 4

it. _ . _ _ _ - _ _ _ _ - . - - - . _ _ _ _ _ _ _ _ _ _ _ )

QUESTION 5.13 (1.00)

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE Off-Normal Instruction ONI-17, Control Room Inaccessibility, describes the actions to be taken when the control room evacuation is immi nent ,

after control room evacuation and the establishing of stable plant conditions at Hot Standby.

Which of the f ollowing personnel are directed per DNI-17, Step 15 to go to the Remote Shutdown Station Room?

a. Shift Supervisor, Assistant Shif t Supervisor, and assistant Control Operator.

b. Shift Supervisor, Control Operator, and Shift Technical Advisor (if stationed).

c. Assistant Shif t Supervisor, Control Operator, and Auxiliary Building Auxiliary Operator.

d. Control Operator, Shift Technical Advisor (if stationed), and Assistant Control Operator.

ANSWER

b. (1.00)

REFERENCE ONI-17, Page 10 of 36.

KA OOOO6BK318 4.2/4.5 43

-GROUP II EMERGENCY AND ABNORMAL PLANT-EVOLUTIONS-12%

l_

DUESTION 5.14 (2.0) **

Trojan has e> perienced a reactor trip with . an inadvertent Saf ety Injection. The plant operators are performing Emergency Instruction EI-0, " Reactor Trip, Safety Injection, and Diagnosis."

l .Three immediate actions of EI-O have been perf ormed by the operators:

Verify Reactor Trip; y Verify Containment. Isolation, Phase A; and Check Containment Pressure.

What are the remaining eight (B) immediate actions of EI-O (regardless of

. order)7 (0.25 each)

ANSWER (0.25 each) (any order)

1. Verify Turbine Trip.a

2. Verify Generator Trip.Y

3. Verif y Power to AC Emergency Buses.4

4. Check if SI is Actuated.4

5. Verify Feedwater Isolation.

6. Verify AFW Pumps Running.b

7. Verify ECCS Flows.r O. Check if Main Steam Lines should be isolated.

REFERENCE Emergency. Instruction EI-0, " Reactor Trip, Safety Injection, and Diagnosis" Pages 2 thru 5 02-K-12-LP Objective 1.2.2.A.4 -

KA OOOOO7K301 4.0/4.6 44

~* QUESTION 5.'15 (1.00) **

MULTIPLE CHDICE-SELECT THE CORRECT RESPONSE Trojan-has. experienced a reactor trip.with an inadvertent Safety.

Injection. The plant operators are perf orming Emergency Instruction EI-0, " Reactor' Trip, Safety Injection, and Diagnosis". -

What is one of the conditions that must be met prior to terminating Safety Injection?

a. . RCS subcool' i ng less than 25 F.

b. Pressurizer level greater than 5%.

c. RCS pressure decreasing.

d. AFW flow less than 495 GPM.

ANSWER

b. (1.00) l

REFERENCE Emergency Instruction EI-0," Reactor Trip, Safety Injection, and Di agnosi s", Page 11 and Reference Page facing.

02-K-12-LP Objective 1.2.2.A.5 KA OOOOO7K301 4.0/4.6 45

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

l:

l,

QUESTION 5.16 (1.00)

Trojan is tripped and saf ety injection . has been initiated. The' plant' I; operators are performing EI-0, " Reactor Trip, Safety Injection,~and Diagnosis". Step.B requires the operator to verify auxiliary feed pump automatic start indication and three other immediate actions.

Besides ' the auxiliary f eedwater pump turbine auto start indication, what are the TWO(2) out of the three(3) immediate operator action-

. steps that EI-O step B uses to verify that the turbine driven auxiliary feedwater pump has started?

l, (0.50 each, maximum 1.00) l

ANSWER (Any two at 0.50 points each, maximum 1.0 points)

1. Steam supply valves open.

2. Steam stop valve open.

3. Trip Throttle valve open.

REFERENCE Emergency Instruction EI-0, " Reactor Trip, Safety Injection, and Diagnosis", Page 4 KA COOOO7A108 4.4/4.3 46

r L* QUESTION 5.17' (1.00)

The1 Troj an Reactor Cool ant : Pumps (RCP) are identical, single-speed, centrifugal. units driven by air-cooled,.three-phase induction motors.

The RCPs. motor ' has protective f uncti ons and logi cs associated . with L their operation.

Which one of f the f ollowing is a protective function of'the RCP MOTOR?

a. High. differential voltage..

b.. Bus underfrequency.

-c. Bus overvoltage.

d. Motor undercurrent.

ANSWER

' b. (1.00)

REFERENCE License' Training Program 02-A-02-LP, Reactor Cool ant Pumps, Page 31-and 32.

KA OOOOOBK304 4.2/4.6 l-l l.

c.

47 l . - _ _ _ =_ _ _ _ _ _ ~ _ _ _ - - _ . - _ _ . _-_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

r" !

l ll L -

80UESTION 5.18, _(1.00) **

1 MOLTIPLE CHOICE-SELECT THE CORRECT RESPONSE

.The Trojan. nuclear plant is in Mode 2 conducting a reactor startup'when a_ Source Range. nuclear' instrumentation (SRNI) fails.

What action is Egguited by Teghnigg1 @gggli[qstigns f or a si ngle

,SRNI~ failure in Mode 27

a. Verify:that NR-45 is not recording the defective channel.

b. Verify that the operable Source Range channel is selected as an input.to the audio count rate channel.

c. LSuspend all operations involving positive reactivity changes.

d. . Place the level. trip switch for inoperable channel to the BYPASS position.

'* ANSWER

c. (1.00)

REFERENCE Off-Normal Instruction ONI-16, Rev. 7, Excore Nuclear Instrumentation Malf unction, Page 2 of 7. Troj an T. S. 3.3.1; O2-B-07-LP Objective 1.2.2.P 3 KA OOOO32K302 3.7/4.1 48

R'

-*OUESTION 5.19 (1.00)

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE The Trojan nuclear plant is above 5 percent reactor power, when a . single Intermediate Range Nucl ear Instrument fails. The operators perf orm 'the immediate' and subsequent operator actions of- Of f-Normal Instruction-ONI-16, "Excore Nuclear Instrument Malfunction."

Which of the f ollowing operator actions is NOT a correct action in accordance with DNI-16 with reactor power above 5 percent?

a. Monitor the operable NIS channel to verify its proper operation.

b. . - Verif y that NR-45 .is not recording the def ective channel .

c. If both channels fail, refer to Technical. Specification 3.0.3 for applicable actions.

d. Place the flevel trip switch for the' inoperable' channel.in the normal position, then resume power operation.

ANSWER

d. (1.00)

REFERENCE Off-Normal Instruction,0NI-16, Excore Nuclear Instrument Malf unction Page 4 of 7 KA OOOO33A102 49 l

l

QUESTION 5.20 (1.00)

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE Trojan's Technical Specifications provide limiting conditions for operations (LCO). In the Reactor Coolant System (RCS) for operational leakage, the RCS leakage is limited.

Which one of the f ollowing correctly describes the technical specification for the RCS operational leakage limit?

a. 10 Gallon per Minute (GPM) pressure Boundary leakage.

b. 1 GPM UNIDENTIFIED leakage.

c. 700 gallons per day through any one(1) steam generator.

d. 100 GPM IDENTIFIED leakage from RCS.

ANSWER

b. (1.00)

REFERENCE Troj an Technical Specification 3.4.6.2 Page 3/4 4-14 KA OOOOO9K321 4.2/4.5 l

l l l 50 l

QUESTION 5.21 (1.00) ** 1 MULTIPLE-CHOICE-SELECT THE CORRECT RESPONSE Trojan's Emergency. Instruction-EI-3, Steam Generator Tube Rupture, provides actions to terminate. leakage of , reactor coolant' into the secondary system f ollowing a steam generator tube rupture.

Which method does EI-3 use to identify the ruptured SG7'

a. High containment radiation from PRM-1, 5 or 14.

b. , Any RCS loop. flow greater than 100% , or a 1% difference between loops.

c. Any SG pressure decreasing in an uncontrolled manner.

d. . Unexpected rise in any SG 1evel inconsistent with feed flow.

ANSWER

d. (1.00)

REFERENCE Emergency Instruction-EI-3, Steam Generator Tube Rupture, Rev.13 Page 4 of 33.

02-K-15-LP Objective 1.2.2.8 KA OOOO3BA203 4.4/4.6 51

I i

i 1

OUESTION 5.22 (1.00)

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE i

Trojan's Emergency Instruction-EI-2, Faulted Steam Generator Isolation, provides actions to identif y and isolate a faulted steam generator (SG). The emergency instruction first identifies j and then isolates the f aulted SG. One of the action / expected responses in to check Reactor Coolant System (RCS) to SG integrity.

Which one of the f ollowi ng is NQT the action used to verify i RCS to SG integrity in EI-27

a. Request periodfr activity samples on all SG's,

b. Verify secondary radiation monitors-UNISOLATED.

c. Any SG pressure decreasing in an uncontrolled manner.

d. Check secondary radiation from PRMs-6, 10, 16, 17- NORMAL

ANSWER

c. (1.00)

REFERENCE Emergency Instruction EI-2, Faulted Steam Generator Isolation Rev. 14, Page 5 of 5.

KA OOOO38A201 4.1/4.7 l

52

^

J

~

'OUESTION 5.23 (1.00) i MULTIPLE' CHOICE-SELECT THE CORRECT RESPONSE Troj an's Of f-Normal Instruction-ONI-51, Loss of Instrument Air, provides instructions / f or operation on an . imtninent or complete loss of' instrument air.. Some automatic actions in ONI-51 are the closing of service air supply valve CV-4467 at 80 psig on decreasing air pressure and the starting of the standby air compressors on decreasing air ' recei ver tank pressure. On loss of instrument air:

- Which of the following functions can the Chemical and Volume Control

System NQT' perform?

a, . - Charging flow.

b. Normal let'down.

c. Emergency baration.

d. Seal water injection.

s

ANSWER

b. (1.00)

REFERENCE Off-Normal Instruction-ONI-51, Loss of Instrument Air Page 4 and 5 of 17.

KA OOOO65K303 2.9/3.4 1

lL 53

-l 1

l

QUESTION 5.24 (1.00)

MULT1PLE CHOICE-SELECT THE CORRECT RESPONSE Trojan's Off-Normal Instruction-ONI-51, Loss of Instrument Air, provide instructions for immediate and subsequent operator actions for loss of instrument air. An immediate operator action of ONI-51 is to attempt to start an air compressor, if a compressor has tripped or failed to start. Upon a loss of instrument air:

Which of the f ollowing operator actions, per ONI-51,is necessary to minimize reactor coolant system cooldown?

a. Shutdown the air ejector.

b' . Bypass the air filtering and drying unit.

c. Isolate the rupture by closing the root valve as near the fault as possible.

d. Air pressure drops below 70 psig and critical valve position becomes uncertain, trip the reactor.

ANSWER

a. (1.00)

REFERENCE Off-Normal Instruction -ONI-51, Loss of Instrument Air, Page 2 of 17.

KA OOOO65K308 3.7/3.9 r.

1 54 l . _ _ - _ _ - _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ - _ _ _ _

(4 '

(~ g-. .

~)

- a n

R

l' f, . GROUP III EMERGENCY AND ABNORMAL PLANT EVOLUTIONS -2%

l )

QUESTION 5.25 (1.00)' -** i

. MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE-ECA O.0,1 Loss of. A11 LAC Power, states the'.: actions that are required, forla Stat. ion B1'ackout..

Which event i s an ~ Entry. Condition f or ECA O.07 a.: L A11' Emergency Diesel: Generators f ail to start after Saf ety Injection is initiated.

b. The Main: Turbine trips (load' reject) and an 'immedi ate l'oss of1Dffsite power. occurs.

c. The Control Room has' indication that. all main and emergency AC buses are'de-energized,

d. Fire .in the Cable Spreading Room causes a loss .of power to both trains of the Reactor Protection System.

ANSWER-

c. (1.00)

REFERENCE ECA O.0, Loss of All'AC Power (02-K-22-LP Objective 1.2.2.A)

KA'OOO56G011 3.5/3.8 i

1 55

. i :l

QUESTION 5.26 (1.00)

MULTIPLE' CHOICE-SELECT THE CORRECT RESPONSE The react'or is producing 100% rated thermal power at a-core delta T of.60 degrees ~and a mass flow rate of 100% when a station blackout occurs. Natural circulation is established, core delta T goes to 40 degrees, and the core mass flow is 3%.

Which of the f ollowing' is the decay heat in % of rated therma.1' power?

a. 1.3 b_ . ~ 2.0-c.- 3.0

d. 4.2

~

ANSWER

b. O = m cp delta T; (1.00) 100% = 100%cp(60 degrees) ; results in a cp =.1/60 degrees.

~

Onatural circ. = 3%(1/60 degrees) (40 degrees) = 2%

REFERENCE License Training Program 02-H-11-LP, HEAT TRANSFER METHODS, Page 27-of 56.

KA OOOO56K101 3.7/4.2 1

END OF CATEGORY FIVE Y

56

. __ _ _ _ _ _ _ _ _ _____________ _____________________________ _ J

CATEGORY 6 PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC RESPONSIBILITIES-(13%)

PLANT-WIDE GENERIC RESPONSIBILITIES-13%

~*QdESTION 6.1 (2.50)L **

Administrative Order, AD-3-26, Independent Verification, provides guidelines for determining when' independent verification is required, which components require independent verification, and how independent verification is performed. Two activities that require independent verification ares (1)' system lineups prior to startup, and-(2) hydrostatic

' t'esting::

a.' What are the two(2) other activitites that require independent-verification.per AO-3-267 (.75 each)- (1.50)

b. What are.the two (2) methods of determining the status of circuit breakers?' (0.50 each) (1.00)

ANSWER

( O.75 each)

a. 1) Clearances 2)' Surveillance testing.

( O.50 each)

b. 1) By mechanical indication'(visual inspection: closed / tripped flags, pointers).

2) Energization of downstream components.

REFERENCE Administrative Order AD-3-26, a) Page 2-3, b) Page 6 KA 194001K101 3.7/3.7 l

57

-_---_-_-_-_-_______-_-_-___a

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

l

.\

1 I

QUESTION 6.2 (2.50) l Plant Safety, PS-3-30, Trojan Holdout and Tagging Procedure delineates the steps.to be followed to obtain a clearance on equipment to allow personnel to perform work on equipment. PS-3-30 states that no person will work on any plant equipment without first obtaining properrauthorization. According to PS-3-30:

-a. Who are the THREE (3) types of personnel authorized to write a clearance? (0.50 each) (1.50)

6. Independent verification is required to ensure proper positioning of components when writing, hanging or releasing clearances involving important equipment. For these type of systems:

What are the TWO (2) cases (or evolutions) when independent verificatic is NQI required 7 (.50 each) (1.00)

ANSWER (3 at .50 each) (1.50)

a. 1) Control operators.

2) Assistant control operators.

3) . Licensed auxiliary operators.

(2 at. 50 each) (1.00) l b. 1) When replacing danger tags (with test tags).

2) When replacing test tags (with danger tags).

REFERENCE Plant Safety PS-3-30, a) Page 3, b) Page'10, Para.E KA 194001K102 4.1/4.1 58

6 (

l

(

(.

8 QUESTION 6.3 (1.25) **

Plant Safety, PS-3-10,. Confined Space Entry.' requires a confined -(

cpcce to be ventilated prior to. entry, especially if the space normally' contained fluids or gases other than air. Lines, valves,

'and electrical-equipment associated with the space are tagged

! t.o. prevent. personnel injury. Oxygen deficient atmospheres require a minimum of Q air exchanges:.

-c. Whet kind of atmospheric condition will require J.9 air changes per PS-3-107 (0.75) .

(I

b. What is ONE (1) of the THREE (3) post-Job activities that allows '

the clearance to be released and -all danger tags removed from a confined.spac67 (0.50)

! (

I

(

4

ANSWER- (

' a. - Toxic (or flammable atmospheres). (.75)

Any one of three (0.50 total):

b}

,. 1 . All. personnel are out of the space. k-

2. Tools / equipment are out of the space.

' 3. The space has been closed.

0 REFERENCE C

PS-3-10, abb)Page 6 Para.3, c)Page 9 Para.E.2 KA 194001K114 3.6/3.6 O

L L

t C

, , 59

QUESTION 6.4- (1.50)

Plant. Safety PS-3-2,' Work on High. Temperature /High Pressure Systems, describes the procedure.to be followed when working on systems that contain high temperatures and/or high pressure fluids.

' A.: How does PS-3-2. define a "high temperature"? (.50)

Bi, How does PS-3-2~ define a " high pressure" fluid 7 (.50)

C., Whose. approval must be obtained prior to starting work on:a'. cleared piping system thet'contains contaminated ~

water, but cannot be drained? (.50) n

~

ANSWER

~

A. 'High temperatures are defined greater than (or equal to) 150 degrees F. (.50)

B. High pressures are defined as greater than 100 psi.(.50)

C. Approval of the General Manager (or Duty General Manager). (.50)

REFERENCE-PS-3-2 A) Page 1, B) Page_1, C) Page 2 KA 194001K100/K109 3.4/3.4 M

60

.n. --___=__:___-__:_-_________ - _ _ _ _

n mn 4

4 n

Fi' O -

QUESTION 6.5l .( 2. 00) > **

'In order to maintain personnel radiation exposures within the limits b . established-'by'10CFR20, administrative restrictions are applied to V the rate offdose accumulation.

JA. 'What are. Trojan's OUARTERLY Administrative Exposure Limits for: (1.50) o 'Whole: Body,. head and trunk?

}: o Skin tof Whole Body 7 o Extremities?

s B. Who must. approve any request for radiation exposures in excess of 1.0 rem / quarter.7 (.50F

ANSWER-A. _ Whole Body 0.5 rem / quarter (0.25), or 1.Orem/ quarter up to 2.5 rem / quarter i f NRC-4' is on file and lif etime dose -is < 5X' (N-18) . (0.75)-

Skin of whole body 7.5 rem / quarter (.25)

Extremities 18.75 rem / quarter (.25) 4 B. The GeneralLManager (.50)

-* REFERENCE Radi'ation Protection Manual, Part II, Radiological Controls; A, Page 2-2; B, Page 2-3 KA 194001K103 3.4/3.4 9

M 61

QUESTION 6.6 (1.5)

. According to the Radiation Protection Manual, Radiation Protection personnel may immediately stop work, cancel an Radiation Work Permit (RWP), or remove personnel f rom the controlled area if certain protection. practices are not followed.

' A. What are the two (2) out of the four (4) conditions that Radiation Protection personnel may issue STOP WORK instructions? (.50)

B. What are the two (2) activities that the STOP WORK authority from

. radiation protection personnel does not apply? (.50)

C.

What kind of jobs is an extended RWP issued f or? (.25)

D. What valid time period is an extended RWP issued for? (.25)

ANSWER A. Any two at .25 each for total of .50.

1. _ Personnel in violation of the Radiation Protection Manual, Radiation Protection Procedures, or their RWP.

2. Personnel are using unsafe radiation protection practices.

3. Personnel are in danger of exceeding administrative exposure limits.

4. . Unexpected radiological conditions are discovered.

B. 1. Suspending Reactor Startup (.25)

2. Power generation (.25).

C. Jobs of a repetitive nature. (.25)

D. Valid for a maximum of 12 months (.25)

REFERENCE Radiation Protection Manual; A, Page 1-2; B, Page 1-2; C&D, Page 2-21.

KA 194001K104 3.5/3.5 62

._2_. _ _ - _ _ . _ _ _ ____-__-______ - - _ . _ _ _ _ _

OUESTION 6.7 (1.50) **

In Operations Procedure OM-2-1, "Use of Operating Procedure", it details that the procedures contained in the Plant Operating Manual shall be f o'11 owed in the performance of plant activities.

A. What should an individual do when he believes that the procedure he is perf orming cannot be accomplished as written? (.50)

B. Whom should an individual inf orm when he believes that the procedur e he is performing cannot be accomplished as written? (0.50)

C. When are reasonable departures from existing procedures, license conditions, or Technical Specifications acceptable? (.50)

ANSWER A. He shall place the system / component in a stable and safe condition. (.50)

B. Inform the Shift Supervisor. (.50)

C. When immediate action is needed to protect the public health and saf ety. (.50)

REFERENCE Operation Procedure OM-2-1, Use of Operating Procedures, A.B and C, Page 1 KA 194001A102 3.9/3.9 _

63 E-______.

l-l -.

GROUP I PLANT SYSTEMS-14%

tCUESTION 6.8 (0.75) **

I MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE The following questi on ref ers to the Full Length Rod Control System.

Assume that the plant is operating at full power with all systems operable and the Rod Control System (RCS) is in Automatic.

Which fault will cause a Rod Control System Logic Cabinet urgent failure?

a. Loss of one 120 VAC power source to the Logic Cabinet,

b. Failure of the Pulser.

c. Failure of the Power Cabinet.

d. Failure of one of the six DC Logic Cabinet power supplies.

ANSWER

b. (0.75)

REFERENCE Lesson Plan 02-B-09-SD 02-B-09-LP Objective 1.2.2.G.8 KA OO1000K403 3.5/3.8 64

r u

[ b': ;

d i

l* QUESTION 6.9. ( O . ~75')

MULTIPl.E' CHOICE-SELECT.THE CORRECT RESPONSE.

y; Assume ;that the plant is operating at f ull power with.all systems operab\e.and,the Rod Control.. System is in Automatic.

'What MODES of Control Rod. motion are inhibited by a Logic Cabinet-

. Urgent Failure?

a. Automatic . mode of ' rod motion onl y. .

b '. Automatic.and Manual modes of rod-motion only.

..r. . . .c. All modes.of. rod motion.

d. Individual mode of' rod motion.

ANSWER

b. (0.75)

.* REFERENCE Lesson Plan 02-B-09-SD Page 26.

LP O2-B-09 Objective B.7.c KA OO1000K403 3.5/3.B

~

65 l

n OL _ -- -__ _ - _ _ . , _ - _ _ _ _ _ _ _ _ , , _ . _ _ , _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _

i.{..-

DUESTION 6.10 (0.75) **

^

. MULTIPLE CHOICE-SELECT THE; CORRECT-RESPONSE

- Assume that the plant is operating at rated thermal power.

If the RCS Loop 3 cold leg temperature instrument f ails hight o .What-will<be.the demanded'(Automatic mode) control rod-motion?

a. Outward rod motion at 72 steps / minute.

b. Outward rod motion at 48 steps / minute.

c. Inward red mot 2on at 48 steps / minute.

d..Inward rod motion at 72 steps / minute.

l l

l 1.

l' l

1 4 ANSWER

d. (0.75)

REFERENCE Lesson Plan 02-B-09-SD 02-B-09-LP-Objective-1.2.2.H KA- OO1000K403 3.5/3.B c ., 66 L-=_-_,-___-_-_-__-_____:---_-___-___-_-___,___________-_______,__-__-__._____-___--_

l p

QUESTION 6.11 (0.75) **

MULilPLE CHOICE-SELECT THE CORRECT RESPONSE While the reactor is operating at 30% power, the "B" RCP trips on overturrent. The plant continues to operate at about~30% power and no operator action is taken.

Why does the initial rapid change in the "B" Steam Generator occur?

a. Level increases because feed flow i s greater than steam flow.

b. Level increases because of swell.

c. Level decreases because of shrink.

d. Level decreases because steam flow is greater than feed flow, l,

ANSWER

c. (0.75)

REFERENCE Lesson Plan 02-A-02-SD 02-A-02-LP Objective 1.2.2.H.2 KA OO3OOOK201 3.1/3.1 OO3OOOK302 3.5/3.8 67

QUESTION 6.12 (0.75) **

MOLTIPLE' CHOICE-SELECT THE CORRECT RESPONSE

'I The plant'is operating at full power 1 and all systems are operable. The

- Chemical, and Volume Control System (CVCS) is aligned f or normal makeup and letdown.

Why'is_the Volume Control Tank isolated . on . a Saf ety. Injection signal?,

a.;-To prevent the charging pumps from becoming gas bound and lcavitating.

b. To maximize the boron concentration of the injected water.

~

c. To prevent pump' cavitation by increasing the net positive suction head.

'd. To maintain the water volume of the VCT in reserve for makeup

,.. during plant cooldown.

l l'

l

ANSWER H a. (0.75)

REFERENCE Trojan EDB O2-A-06 and Lesson Plan 02-A-06-SD.

02 -A-06-LP Obj ecti ve B. 2. h. 4 KA OO4000K407 3.O/3.3 s

68

m e

QUESTION'6.13 (0.75) **

MULTIPLE-CHDICE-SELECT.THE CDRRECT RESPONSE The plant is. operating at.ful1~ power.and all systems are operable. The:

Chemical and: Volume Control System-(CVCS) is aligned for normal makeup and letdown.

iWhich conditions must be met to satisfy the interlocks for opening a q

. letdown isolation valve?-

-a. Pressurizer level must be less than 17 */. .

'b. At least one orifice isolation valve must be open.

~

c. At least one charging pump must be running.

d..All. orifice. isolation valves must-be' shut.

ANSWER

d. (0.75)

.* REFERENCE

, Troj an EQB O2-A-06 and Lesson Pl an 02-A-06-SD.

02-A-06-LP Objective 1.2.2.E.3 KA OO4000K405 3.3/3.2 I

69 l

1. , m - _ _ _ _ _- " ^ - " - ^ - - - - - - _ _ _ _ _ _ _ _ _

i... <

x: ,

7@! ;n' < .<

^

' lll- ,

l 1":i:.; , _ _

ll 1* QUESTION l 6.14 (0.75) **

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE J' .

' L The . pl ant' is - operating' at f ull power and all: systems-are operable. The-

<! . Engineered Safety Features Actuation System (ESFAS) is in its normal-'

.1 i n eup . .

F o- . What.is the power. supply to ESFAS Train A7 a.'4.16=KV.ESF Bus A1.

3: :b. 120.VAC Preierred' Instrument Bus Y11.

c. 480 VAC MCC B-25.-

d. . 125 VDC: Bus D-30.'

'* ANSWER

b. (0.75)

REFERENCE-Lesson Plan 02-C-06-SD page 24 KA 013OOOK201 3.6/3.8

?

70

y ,, ' ,

.pu;

...,"^1' c,

u W' s .

t ,,

QUESTION 6.15 - ( 1. 0) . ;**-

MULTIPLE ~ CHOICE-SELECT THE. CORRECT. RESPONSE f The plant.was operating?at full; power.with all systems operable

' when the Engineered-Safety Features' Actuation System.'-(ESFAS) was- '

.a inadvertently activated. LAfter ESF has been resets-Wh'chifunction i must be manually reset?--

a.. Containment ventilation isolation. signal.

' b. Centrif uga1' Charging Pump (CVCS) ' start. signal .

c. Saf ety:. Inj ection Pump Start signal .

d. Containment air coolers start signal.

L

!* ANSWER-

a. - (0. 75) .

-

REFERENCE Lesson Plan 02-B-02-SD pages 21 &22. Figure 40 02-F-04-LP Objective 1.2.2.E.3 KA 013OOOK401 3.9/4.3

-4 6

71

QUESTION 6.16 (0.75) **

1 MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE ]

The plant is operating at full power and all systems are operable.

Maintenance is in progress on the Engineered Safety Features Actuation ,

System (ESFAS) when all bistables for HI-HI containment pressure are inadvertently actuated.

How will the Containment Spray System (CSS) respond?

l a. CSS will be actuated and spray the containment,

b. CSS pumps will start, but CS isolation valves will stay shut.

c. CSS isolation valves will open, but CS pumps will not start.

d. CSS will not be effected by this incident.

ANSWER

c. (0.75)

REFERENCE Lesson Plan 02-B-02-SD page 24.

02-A-09-LP Dbjective 1.2.2.C.2.g KA 013OOOK105 4.1/4.4 j 72

i l

QUESTION 6.17 (0.75) **

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE The plant is operating at full power and all systems are operable. The Auxiliary Feedwater System (AFW) is in its normal lineup.

~ What is the power' supply to the AFW Turbine Trip and Throttle Valve 7

a. 4.16 KV ESF Bus A2.

b. 120 VAC Preferred Instrument Bus Y22.

c. 480 VAC ESF Bus B-23.

d. 125 VDC Bus D-10.

ANSWER

d. (0.75)

REFERENCE Lesson Plan 02-A-12-SD page 37 02-A-12-LP Objective 1.2.2.B.1.j KA 061000K201 3.2/3.3 4

i 73

QUESTION 6.18 (0.75) **

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE The plant is operating at full power and all systems are operable. The Auxiliary Feedwater System (AFW) is in its normal lineup.

When does the Diesel Driven AFW pump trip on low CST level?

a. 35%

b. 30%

c. 9%

d. 60%

ANSWER

a. (0.75)

REFERENCE Lesson Plan 02-A-12-SD page 25.

02-A-12-LP Obj ective 1.2.2.C.3 KA 061000K401 3.9/4.2 74

s:

QUESTION 6.19 (0.75) **

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE To. assure adequate decay heat removal and RCS heat transfer, Red Path actions are required in some cases when AFW flow is less than a given value. The E-O (Reactor Trip, Safety Injection, and Diagnosis) Red Path Summary for Heat Sink uses less than 495 gpm AFW flow, but E-1 (Loss of Reactor or Secondary Coolant) Red Path Summary uses -less than 720 gpm.

Why is the Red Path value f or AFW flow' required by E-1 more than E-O?

a. E-11 assumes an adverse-environment flow measurement and allows for

. larger flow errors,

b. More decay heat would be generated under the conditions when E-1

.would be used.

c. E-1 ' allows f or allowable AFW pump speed variations and flow measurement errors.

d. More AFW flow is required to prevent water hammer in the SG f eed rings under Loss of Coolant conditions.

ANSWER

a. (0.75) l

REFERENCE Lesson Pl an 02-A-12-SD page 60.

02-K-13-LP Objective 1.1.2.2 KA 061000K501/502 3.2/3.6 75

y.-. 1

.A

QUESTION.6.20 (1.0)'

(

MULTIPLE CHDICE-SELECT THE CORRECT RESPONSE The primary function of the Post Accident Range Instrumentation-

-(N48)- is to: provide a environmentally and. seismically qualified neutron detection. system.

What is the secondary design function of this system?

a. To provide qualified shutdown power level instrumentation for' station blackout as required by;10 CFR 55.

b. To provide qualified and independent instrumentation f or remote shutdown as required by 10 CFR 50, Appendix R.

c. To. provide' qualified instrumentation for post-accident assessment of containment' radiation levels for 10 CFR 100.

d. To provide an indirect indication for post accident reactor vessel. water' level as required by 10 CFR 50, Appendix K.

ANSWER

b. (1.00)

REFERENCE Lesson Plan 02-B-07-SD, page 34.

KA 015000K601 2.9/3.2 76 l-

DUESTION 6.21 (1.00) **

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE The plant is operating at rated thermal power with rod control in automatic when one power range nuclear instrument (N43) fails high.

What is the resulting control rod motion without operator action?

a. Control rods drive in until the Turbine First Stage Pressure is less than 15% power, then stop because C-5 will not allow further insertion.

b. Control rods drive in until the Over Temperature Delta T (C-3) stops rod motion within 3% of the trip setpoint.

c. Control rods drive in until Tref - Tave compensates for the failed instrument (rate input only), then out to restore Tave.

d. Control rods drive in until temperature mismatch balances power (rate only) mismatch, then stop at the lowest position reached.

I l

ANSWER

d. (1.00)

REFERENCE Lesson Plan 02-B-09-SD l 02-B-09-LP Objective 1.2.2.H l KA 015000K302 l

l l

i

) 77 1

QUESTION 6.22 (0.75) **

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE The plant is operating at full power and all systems are operable.

The Auxiliary Feedwater (AFW) and Main Feedwater (MFW) Systems are in their normal system lineups.

Where does the AFW feed line penetrate the Main Feedwater System?

a. Between the MFW Isolation Valves and the MFW flow venturi.

I

b. Between the MFW flow venturi and the containment.

c. Between the containment and the Steam Generator.

d. Between the MFW seismic category I check valve and MFW Isolation Valves.

i l

ANSWER

b. (0.75)

REFERENCE Drawing M-213, sheet 2; Lesson Plan 02-A-12-LP O2-A-12-LP Obj ecti ve 1.2.2.B.2 KA 059000K102 3.4/3.4 78

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _i

I

QUESTION 6.23 (0.75) **

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE The plant is operating at. full power and all systems are operable.

The Au>:iliary Feedwater (AFW) and Main Feedwater (MFW) Systems are in their normal system lineups.

Which MFW parameter is used by the Steam Generator Water Level Control System to control MFP speed?

a. Main Steam line Pressure.

b. MFW Flow Control Valve position.

c. MFW flow.

d. Steam Generator water level.

l

ANSWER

a. (0.75)

REFERENCE Lesson Plan 02-B-05-SD, Figure B.

02-B-05-LP Dbjective 1.2.2.B.2.k KA 059000K104 3.4/3.4 79

l l

QUESTION 6.24 (0.75) **

MULTIPLE CHDICE-SELECT THE CORRECT RESPONSE The plant is operating at full power and all systems are operable.

All Reactor Coolant Pumps (RCP) are in their normal system lineup.

What is the power supply to the "C" RCP7

a. 12.47 KV bus H1.

b. 230 KV bus V 81.

c. 4.16 KV bus AS.

d. MCC B25 1

ANSWER

a. (0.75)

REFERENCE Lesson Plans 02-A-02-SD, Page 17 02-A-02-LP Objective 1.2.2.D.1 KA OO3OOOK201 3.1/3.1 OO3OOOK302 3.5/3.8 1

f l

80

i.

QUESTION'6.25 (0.75) **

- MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE The plant' is operating at' f ull power'and all systems are operable.

All Reactor Coolant Pumps (RCP) are in their normal system lineup.

Which signal will cause an RCP trip during normal. full power operations?

a. Loss of power from the Startup Transformer due to~

a phase to phase fault within the transformer.

b. The LOCAL PANEL SELECTOR SWITCH is placed in LOCAL /RUN while

. the REMOTE CONTROL SW' I TCH is lef t in RUN..

c. Steady state Under Frequency occurs on a RCP power supply bus due to an instrument malfunction.

d. Steady state Under Voltage occurs on a RCP power supply bus and the Unit Auxiliary Transformer breaker is open.

9

ANSWER

d. (0.75)

REFERENCE Lesson Plans 02-A-02-SD, page 17 02-A-02-LP Objective 1.2.2.F.5 KA OO3OOOK302 3.5/3.8 1

4 81

GROUP II PLANT SYSTEMS-13%

QUESTION 6.26 (2.00) **

Appendix A to ES-0.1, Reactor Trip Response, provides six (6) plant conditions that are used to verify natural circulation cooling in the Reactor Coolant System (RCS). One condition is S/G pressures-STABLE OR DECREASING.

What are four (4) of the remaining five (5) plant conditions?

(0.50 each for any four; 2.0 maximum) i 1 ANSWER (O.5 each for any four; 2.O maximum)

1. RCS subcooling > 25 degrees F.

/

2. RCS hot leg thermocouple stable or decreasing.

3. Core exit thermocouple stable or decreasing.

4. y NCS cold leg temperatures near saturation temperature f or SG pressure.'

5. ,RCS delta T < 66 degrees F and stable.

REFERENCE ES-0.1, Appendix A. ._

O2-K-12-LP Objective 1.2.2.D.21 KA 002020K510 3.5/3.9

)

82

QUESTION 6.27 (2.5) **

Trojan Technical Specifications requires that RCS average temperature be maintained above a given limit in Modes 1 and 2.

A. What is the minimum temperature f or criticality (MTC)? (0.50)

B. What are the f our (4) bases for this limit? (2.00)

ANSWER (.50 each for five parts)

A. 551 degrees F- minimum temperature for criticality. <

B. The minimum temperature for criticality is based upon:

1. That the MTC is within it's analyzed temperature range.

2. The pressurizer is capable of being in an operable status with a steam bubble.

3. The reactor pressure vessel is above it's minimum NDT temperature.

4. The protective instrumentation is within it's normal operating range.

REFERENCE Troj an Technical Specification Bases 3/4.1.1.5, Page B 3/4 1-2 02-A-01-LP Objective 1.2.2.A.12.c KA 000002G006 2.6/3.8 83

QUESTION 6.28 (1.50) **

What are the Technical Specification leakage limitations for the following types of Reactor Coolant System leakage? (0.50 each)

A. Identified B. Unidentified C. Controlled (All RCP's)

ANSWER A. 10 gpm. (0.50)

B. 1 gpm. (0.50)

C. 20 gpm total from all RCPs (0.50)

REFERENCE Reactor Coolant System Description, 02-A-01-SD, Page 46 of 54.

02-A-01-LP Objective 1.2.2.A.13.f KA 02OOO2K401 3.6/3.8 l

84

p.

l f i

'

QUESTION 6.29. .(1.50)

The Emergency Core Cooling System - (ECCS) is designed - to provide-core cooling'and additional shutdown capability during'iour (4) types of accidents.

' Wha't-are three (3) of these types of accidents?

l

ANSWER- (Any 3 at 0.50 each ) .

l

.I

.1. . Engahs_i D_BGS(up to and including the design basis accident).

2. Bugtytg_gf_GBDd(pressure housing and RCCA Ejection) .

l 3. .Stgghg_tn_ Mein _Etggm_@ygtem( up.to and. including a double-ended guillotine sheer of largest pipe).

'4 . Sweigte_91_SZ9_Iwbs.

(Underlined portions constitute essence of answer.)

REFERENCE Emergency Core Cooling System Description, 02-A-07-LP, Page 7 of 62.

02-A-07-LP Objective B.2.a KA OO6000 GOO 4 3.5/3.8 85 L-__-________________-_______-_-_--__

k

.e

OUESTION 6.30 (2.50) **

Trojan's. Technical. Specifications Limiting Conditions f or Operation 3.5.5.11mits the. volume and Baron concentrations in the Ref ueling Water-Storage Tank (RWST).

1. What.are the two '(2) bases for these limits? (1.OO:each) (2.00)

2. What is the~ minimum temperature for operability'of the RWST7 (.50)

ANSWER'

. 1. a.fTo ensure that sufficient water is available within containment to permit recirculation cooling flow to the core in the event of a LOCA. (1.00)

b. To ensure that the reactor will remain subcritical (in the cold condition following mixing of the RWST and the RCS water

volumes) following a LOCA (with all control rods inserted except f or the most reactive control assembly). (1.00)

2. 37 degrees F (0.5) l

.

REFERENCE Trojan's Technical Specifications 3/4.5.5 RWST Pg B 3/4 5-2.

02-A-07-LP Obj eci ve B. 2. m. 9 KA OO6000 GOO 6 2.9/4.0 86

1 w.

1. '

. i l'

l 1

1

QUESTION 6.31 (2.25) **

"The pressurizer Pressure and' Level Control System is. designed to' accommodate three. analyzed transients without causing a reactor trip.

-What'are these three (3) transients (include magnitudes /setpoints)?

(0.75 each ) (2.25)-

l

*ANSWER ('O.75 each)

1. Leading or unloading at a rate of 5%/ min. (with automatic rod control).

2. Instantaneous load changes of +/- 10% (with automatic rod control).

3. Step load reduction of 50% (with automatic-rod control and steam dumps).

REFERENCE Pressurizer Pressure and Level Control System Description, 02-B-04-SD, Page 3 of 38.

02-B-04-LP. Objective 1.2.2.A KA 010000 GOO 4 3.1/3.3

- )

1 57

____2___1______ _ _ _ _ _ _ . _ _ .

)

QUESTION 6.32 (0.75)

MULTIPLE CHDICE-SELECT THE CORRECT RESPONSE Which of the f ollowing is the Pressurizer Low Pressure Reactor trip? (0.75)

A. 2385 psig. !

B. 1865 psig.

C. 1777 psig.

D. 2235 psig.

ANSWER B. (0.75)

REFERENCE Pressurizer Pressure and Level Control System Description 02-B-04-SD, Page 36 of 38 KA 010000K101 3.9/4.1 88

c __

GROUP III PLANT SYSTEMS-3%

QUESTION 6.33 (1.00) **

The Residual Heat Removal (RHR) system suction isolation valves (MO-8701, MD-8702) have two interlocks to prevent overpressurization of the RHR system.

What are these two (2) interlocks (include setpoints)?

(0.50 each) (1.00)

ANSWER

1. The valves cannot be opened when Reactor Coolant System wide range pressure (0.25) is greater than 425 psig (0.25).

2. The valves will go shut automatically if pressure increases (0.25) to 600 psig (0.25).

REFERENCE License Training Program 02-A-08-SD, Residual Heat Removal System Page 6 of 28. i 02-A-OB-LP Objective 1.2.2.E.5 KA OO5000K407 3.2/3.5 89

l

OUESTIOl1 6.34 (1.00) **

MULTIPLE CHOICE-SELECT THE CORRECT RESPONSE Precautions in Operating Instruction 01-4-1, Residual Heat Removal (RHR),

prohibit personnel entry or presence in the RHR rooms when an Reactor Coolant Pump is started.

Which of the following is the reason for this precaution?

a. The potential for an over pressure surge is increased with possible rupture of RHR piping.

b. Flow increase in the RCS may starve the RHR suction and cause major pump demage with potential personnel safety concerns.

c. Radiation levels may increase as crud is swept from the RCS.

d. System safeties (relief valves) may lift returning radioactive water directly to the room floor drains thereby increasing personnel exposure.

ANSWER

c. (1.00)

REFERENCE Operating Instruction 01-4-1, Residual Heat Removal, Page 3 of 7.

02-A-08-LP Objective 1.2.2.I KA OO5000G010 3.3/3.5 90

I .i (J

I

_ f I: ,

f

-

QUESTION 6.35 (1.00) **

What'are.four (4) out' of the six -(6) Engineered Safeguards related: components' cooled by the Seismic Category I Component Cooling ' Water . System loop? . (f our at O'

. 25 each)

L* ANSWER .(Any four.at 0.25 each) a.- Positive. Displacement (CVCS) Charging Pump.

b. Saf ety1Inj ection. Pumps. /

c. Containment. Spray Pumps.

d. Residual Heat Removal Pumps,

e. Residual Heat Removal Heat Exchangers./

f. Containment. Air Coolers.V

-* REFERENCE.

Component Cooling Water System Description, 02-A-11-SD, Page 5 of 31.

02-A-11-LP Objective 1.2.2.G 4

KA OOBOOOGOO7 3.3/3.4 END OF CATEGORY SIX END OF WRITTEN EXAM 91

_____=____---__:--- - - _ . . - _ - . -

p Critien1SafdyFunctionStetuiTrass(CSFST)

FIGURE FR-0.1 SUBCRITICALITY l

Red I. -

GO TO FR-S.1 POWER NO R

4

<ANGE5%

orange YES GO TO

l. _ _ .

FR-S.1 i

Yellow GO TO FR-S.2 IR SUR NO ZERO OR NEGATIVE IR SUR NO YES MORE NEGATIVE THAN -0.2 YES DPM Green

( a

. NO SR ENERGIZED YES Yellow

} ,

Go to FR-S.2 SR SUR NO ZERO OR NEGATIVE YES Green CSF Sat FR-0.1 FR-0 Page 5 of 12 Page 1 of 1 Revision 4 ;

I 0

_ _ _ _ _ _ _ _ _ _ - - _ _ _ - _ _ _ _ _ _ - - _ _ _ _ _ - _ ~

y', '

Critieni S-fc\ty Functirn Stntus Trces (CSFST)

I FIGURE FR-0.2 I

( LORE COOLING Red CORE e GO TO FR-C.1 NO EKIT Red

T/cs

< 1200*F ygg FR-C 1 RVLIS NO FULL RANCE

> 39% YES.

Grange .

GO TO CORE FR-C.2 NO EXIT

"~~"

T/Cs

< 700*F YES Orange GO TO FR-C.2 AT LEAST NO RYLIS NO ONE RCP FULL RUNNING RANCE YES > 39% YES Yellow ~

!. GO TO FR-C.3 SUB- Oringe NO COOLING GO N

> 25'F FR-C.2 (75'F) YES RVLIS DYNAMIC NO HEAD RANGE !

> 44% 4 RCPs

> 30% 3 RCPs

> 20% 2 RCPs

> 13% 1 RCP YES I i

Yr.llow i l

~

,. GO TO l FR-C.3 i Grien f CSF {

SAT ]

i FR-0.2 Page 1 of 1 FR-0 Page 6 of 12 Revision 4 l

l ,

r e

, Oritical

.;_; m . a -- .Shfatv. Tunction Status Trems '05FST)

~ .. . . . _. . .

FIGURE FR-0 3 C EEAT S!h*K Red

,..... ..... .- CO TO 9TR-H.1

. . . . - - . ~ - . . . . - . - -

. um . _ , - -.

TOTAL FEED FLOW TO NO S/Cs

>495 TES

~

(720) gpia

~

AVAIIABLE

.1,.... .,.

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, CO TO FE-H.3 e

II N m g

li.of rr a

- . - 1 .

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< 1170 PSIC gg Yellow CO TO FR-E.5 ER LEVEL II ALL M S/C

> 5%

(12%) TES Creen

~

O . CSF SAT ra-o.3 Page 1 of 1 TR-0 Page 7 of 12

\

Revision 6

Critical Safety Function Status Treas (CSFST)

. \. . .. , ..

FIGURE FR-0.4 INTEGRITY C- g , , , .

256C !

' Tedkra e 75*

Red LLL RCS l

9 GO FR-P.1 TO '

PRESSURE- 2 Orange ECOLD POINTS - TO TO RIGHT OF

I

~IMIT A COOLDOWN (SEE ATTACH-IN ALL NO ALL RCS Mmv o NO TCOLD TCOLD TEMPS

--> < 100*F IN yg fJLST 60 YES > 245*F

- . ~ - ...

(yg) ES - M TO MINUTES'- ' ~ ~ ' -

g '

ALL RCS 33 TCOLD l(L TEMPS

> 275*F YES (T2)

Green l CSF SAT Orange GO TO l FR-P.1 ALL RCS NO TCOLD TEMPS

> 245'F YES (TI)

RCS Yellow NO l PRESS .

GO TO

< 440 FR-P.2 PSIG gg Green RCS TEMP NO I CSF

> 290*F -

SAT YES Green G CSF FR-0.4 SAT Page 1 of 1 FR-0 Page 8 of 12 Revision 4

. +

. ..-m C. FR-0.4 ATTACHMENT A TROJAN OPERATIONAL I.IMITS CURVE 3000

_., _.._. i ,

=. . . . . ... .

' 22.90 2flS

* .: 1 2iF JJfF

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1 0 100 190 245 275 400 200 500 300 1 TEMPERATURE (*F) - FR-0.4 Attachment FR-O Page 9 of 12 Page 1 of 1 Revision 4 l l

_ Critical Safety Function Status Trees (CSFST) - FIGUE FR-0.5 C- . . . ... .. . CONTAINMENT Red

GO TO _ FR-2.1 CONT NO . 4 PRESS

                                    < 60     YES PSIG Orange               l GO TO FR-Z.1 CONT       NO PRESS                                                                                    -
                                          < 30     _TES. . . _ _ _ _ , .

PSIG Grange j k/ - 2 1 CONT NO SUMP LEVIL TES

                                                  <),2d*

e Tellow g'9 M TD (Q V FR-Z.3 i con NO RADIATION

                                                           < ARM 15A/B                 TES ALEE Creen
                                                                               -                                                           CSF SAT

_) 1L-0.5 age 1 of 1 FR-0 Page 10 of 12 Revision A

3' $ _y

                        .                                                  ,                                                    \                             ,

CriticalSafet[FunetionStatusTrees(CSFST)- p

                     -q(..

c FIGURE FR-0.6 c i

                                                                                   . - - - - ~ ~ ~ .

INVENTORY 4..._.. ...._,.... ,, , Yellow GO TO FR-I.3 I RVLIS INDICATES NO UPPER HEAD FULL YES

                                                                                                  . . . . . . . . .       ,             (> 94%)-                                                                         l Yellow GO TO FR-I.1 PER       NO
                                                     --+            LEVEL-
                                                                    < 92%

YES

                 . {f ..                                                              .. . - . . . .,

Yellow c0 To FR-I.2 h c NO Pza-LEVEL

                                                                                                 > 17%-

YES I Yellow rg GO TO 2 FR-1.3

RVLIS INDICATE 3 NO UPPER HEAD FULL (> 94%) YES g Green CSF SAT FR-0.6 Page 1 of 1 FR-0 Page 11 of 12 Revision 4 l' l

1 +

1 l

l PORTLAND GENERAL ELECTRIC COMPANY UPDATED M

                                              - TROJAN NUCLEAR PLANT                                                                   NED BY COPY HOLDER April 19, 1988 Revision 3                                                               QUALITY            RELATED EMERGENCY CONTINGENCY ACTION - ECA-0.0 LOSS OF ALL AC POWER APPROVED BY                                                                    t to                                DATE       l4 8>9./

I~ n A. PURPOSE This procedure provides actions to respond to a loss of all AC power. B. ENTRY CONDITIONS

1. Indication that all main and emergency AC busses are de-energized.

2. This procedure is entered from EI-0. REACTOR TRIP, SAFETY INJECTION AND DIAGNOSIS, Step 4, on the indication that all AC emergency busses C' are de-energized.

i i T0001/1R - ECA-0.0 Page 1 of 21 Revision 3 i

s . I, V*.. 1 LOSS OF ALL AC POWER Step Action / Expected Response Response Not Obtained CAUTION: A loss of all AC power . requires initiation of the RERP. Carry 6dt~ " "~~'----"-~~"- -

                                                                  . RERP actions in parallel'with this procedure I

NOTE: Steps 1-4 are IMMEDIATE _ . . .....p. ACTION steps.

                                                                           ,I NOTE: CSF statur trees should be monitored for informa-tion only. FRs should NOT be implemented.

I NOTE: Many chart recorders will be de-energized during a loss of all AC power. Use associated meters if

recorders are not C~. t 1 Verify Reactor Trip: available. l a. Reactor trip and bypass a. Manually trip reactor. breakers - OPEN

b. Neutron flux - DECREASING b. Continue with procedure.

WHEN a source of emergency power is restored and a CCP is running. THEN initiate emergency boration. 2 Verify Turbine / Generator Trip

a. All stop valves - CLOSED a. Manually trip turbine.

(verify CLOSED locally)

b. Generator output breakers - b.

OPEN Manually open output breakers 30 seconds AFTER turbine trip.

c. Generator exciter fleid c. Manually open.

breaker - OPEN ,, ECA-0.0 Page 2 of 21 Revision 3

                                                                                                                        ~-   ~ ~ - - - ~                  ~~ ' ~ ~ ~ ~ ~                                            ~ ~ ~ ~

o- ..

                     .,s                                                                                                                                                                                                                         1 q

l . i

LOSS OF ALL AC POWER I

                                                                  ~ ~ ~ '

Step 'Ac tion / Expe'c te,d' Response

                                                                                                                                                                '                                        ~

Response Not Obtained "7' * " Check ~1'f' RCS~iit Tsdlated "~~' ~ ~" ' ~ ~ ~ ~ "' ' ' a. i Pressurizer PORVs - CLOSED a. IF pressurizer pressure

                                                                                                                                                                                                         < 2.335 psig. THEN manually l-close PORVs.,

b. Letdown isolation valves: b.

l Manually close letdown

LCV-459 - CLOSED containment isolation CV-8152.

                                                                               + LCV-460 - CLOSED'                                                                                                                                               I

c. Excess letdown isolation c.

l Manually close valves. valves:

                                                                              + CV-8153 - CLOSED

CV-8154 - CLOSED.. .. .

d. RCS and pre =surizer sample d. Manually close sample isolation valves CLOSED: containment isolations.

CV-5655, CV-5657, and

                                                                              + MO-5653 - CLOSED                                                                                                        CV-5659.
                                                                              + MO-5654 - CLOSED C!           .

MO-5656 - CLOSED
MO-5658 - CLOSED l

4 Verify AFW Flow - > 720 gpm perfor1a the_following:

a. Stop the diesel APW pump if

Verify turbine-driven AFW runninr, pump running. IF NOT, THEN manually open steam supply valves.

b. Line up emergency cooling to the diesel AFW pump per

Verify proper alignment of DI-8-2

                                                                                                                        -                                                                           AFW valves. IF NOT. THEN

c. manually align valves as Restart the diesel AFW pump necessary, if necessary ECA-0.0 Page 3 of 21 Revision 3

p' 4 e LOSS OF ALL AC POWER

                    -- Step                Action / Expected Response                                                   Response Not obtained
                   ---  -  -Try to- Restore Power--to Any AC             .--

Emerrency Bus: a... Restore power to A1/A2 from the EDCs:

1) Start the EDCs 1) Start the EDG locally.

a) Place local / remote switch in LOCAL. . b) Ensure all lockouts are reset. t c) Start EDG. IF EDG * * " does NOT start THEN C, r2.as.u .4 refer___,,,,,,,,,,,,,t

,m.

to ONI-4T EL . [Hb h$bb b[ bDC, and g.g[ continue with step 6 of this precedure.

2) Bring EDG to rated speed and voltage C 3) Synchronize the EDG to A1/A2 and close A108/A208

3) Locally close A108/A208.

CAUTION: Cooling to the EDCs must be es-tablished within 3 minutes after

                                                          'they are started.

b. Verify automatic loading on b. Manually load the EDCs by the S/D sequencer:

manually closing the fol- - lowing A1/A2 load breakers.

CCPc - START
B01/B02 breakers.
CCW Pumps - START
B03/B04 breakers.

SW Pump - START; verify a SW pump breakers (ones SWBP starts when SW pump lined up for service). starts ' ECA-0.0 Page 4 of 21 Revision 3 i

                   , ,                                                                                                                                                                                                    i d

__ .. ... .. . c

                                                                        " LOSS OF ALL AC POWER C                   Step                 Action / Expected Responge.

Response Not ObtainEd ! 5 ' ' - * - ~ ^ '

SWBP breakers.

CCp breakers.
CCW pump breakers (ones lined up'for service).

c. Check A1/A2 buses - AT LEAST c. ONE ENERGIZED IF the EDG cannot be sta ted u y Q: "Mg g r to ONI-k ,$${.Id.e. .b i.

_ rn EM, and centinue 4M; with Step 6 of this procedure. 5 ' p.

d. Return to procedure and step e

in effect I LAST PAGE OF EcA-o o l RCA-0.0 Pete 5 of 21 Revision 3 _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ - - - _ - - - - - - - - - - - - - - - - - - - - - - - - - ~ - - -

l~ L AO

                                                                                                           /cr Y U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION REGION V FACILITY:               TROJAN NUCLEAR PLANT REACTOR TYPE:           WESTINGHOUSE. PWR. 4 LOOP DATE ADMINISTERED:      JUNE 20. 1989 INSTRUCTIONS TO CANDIDATE:

Use separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80L Examination papers will be picked up six (6) hours after the examination starts.

                                                                                                              % OF CATEGORY % OF         CANDIDATE'S          CATEGORY
                                                                 /VALUE      TOTAL          SCORE            VALUE                    CATEGORY '

273 27.0 M 1. e-- REACTOR PRINCIPLES (7%)

                                                                               '# "A 7                                           THERMODYNAMICS (7%) AND COMPONENTS (11%)

(FUNDAMENTALS EXAM)

                                                                  , ,, o     arz/7~Y" M-           E 0 0-                                      2. EMERGENCY AND ABNORMAL PLANT
                                                                        <n ur een y                                              EVOLUTIONS (27%)

wr 47.0 4W 3. PLANT SYSTEMS (38%) AND

                                                                                  < A ,.7 , ,                                    PLANT-WIDE GENERIC RESPONSIBILITIES (10%)

v v. o

                                                                 %0 00 'n                                            %           TOTALS c,,,,,,,     FINAL GRADE All work dea r on this examination is my own.             I have neither given nor received aic.

t

                                                                                                                             ***** KEY *****

Candidate's Signature

y,Q t /f4'Y

7

                                                                                                                                                ,,, ;< ;r y     e--c- e:

me e -

                                                                                                                                                      /                 # ##                  ;
                                                                                                                                                 < wa:r Erw d en                              j

,/

l PROCEDURES FOR THE ADMINISTRATION OF WRITTEN EXAMINATIONS I

1. Check identification badges.

2. Pass out examinations and all handouts. Remind applicants not to review examination until instructed to do so. ;

READ THE FOLLOWING INSTRUCTIONS VERBATIM: During the administration of this examination the following rules apply:

1. Cheating on the examination means an automatic denial of your application 4 and could result in more severe penalties.

2. After the examination has been completed, you must sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examination. This must be i done after you complete the examination.

READ THE FOLLOWING INSTRUCTIONS

1. Restroom trips are to be limited and only one applicant 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.

2. Usebladkinkordarkpencilonlytofacilitatelegiblereproductions.

3. Print your name in the blank provided in the upper right-hand corner of the examination cover sheet.

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

 . 5. You may write your answers on the examination question page or on a separate sheet of paper. USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAGE.

6. If you write your answers on the examination question page and you need more space to answer a specific question, use a separate sheet of the paper provided and insert it directly after the specific question. DO NOT WRITE ON THE BACK SIDE OF THE EXAMINATION QUESTION PAGE.

7. Print your name in the upper right hand corner of the first page of each section of your answer sheets whether you use the examination question pages or separate sheets of paper. Initial each page.

8. Before you turn in your examination, consecutively number each answer sheet, including any additional pages inserted when writing your answers I on the examination question page.

l '

                                           . I 9.

If you are using separate sheets, number each answer as to category and i number (i.e. 1.04,6.10) and skip at least 3 lines between answers to allow space for grading.

10. Write "End of Category " at the end of your answers to a category.

11. Start each category on a new page.

12. Write "Last Page" on the last answer sheet.

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

Avoid using symbols such as < or > signs to avoid a simple transposition error resulting in an incorrect answer. Write it out.

14. The point value for each question is indicated in parentheses after the question. The amount of blank space on an examination question page is NOT an indication of the depth of answer required.

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

16. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND 00 NOT LEAVE ANY ANSWER BLANK. NOTE: partial credit will NOT be given on multiple choice questions.

17. Proportional grading will be applied. Any additional wrong information that is provided may count against you. For example, if a question is worth one point and asks for four responses, each of which is worth 0.25 points, and you give five responses, each of your responses will be worth l 0.20 points. If one of your five responses is incorrect, 0.20 will be deducted and your total credit for that question will be 0.80 instead of 1.00 even though you got the four correct answers.

18. If the intent of a question is unclear, ask questions of the examiner only.

19. When turning in your examination, assemble the completed examination with examination questions, examination aids and answer sheets. In addition,

      ,               turn in all scrap paper.

20. To pass the examination, you must achieve an overall grade of 80% or greater and at least 70% in each category.

21. There is a time limit of (6) hours for completion of the examination.

(or some other time if .less than the full examination is taken.)

22. When you are done and have turned in your examination, leave the examination i

area (DEFINE THE AREA). If you are found in this area while the examination i is still in progress, your license may be denied or revoked. - 1

+,

i. (EOUATSON SHEET (Pere 1 of ?)- [n ' x 1* N- 19. PE - mgz W AU ' , gJ- c

2. : - cN:

20. D'

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4; - P - P,10(SUR t) -21. in - p A U '

5. P -'.P e e(t/ )- 22. h-mcoT p 3

6. Xeff 23. Q - mah' 1.... p 1

7. Keff 24 .Q - UALT

                                                                                                                                                                                                        ~ ~ ~ ~ ~

3.

8. F - Keff -l 25.

                                                    .g,gg                                                Q - Gwr:L h --h                                                    m f*                                                                                                                              "

L 9 .- P-g,ff 7 +1 + AT-Beff 26. Xh fE - m S

                                                                                                                                                                                            +m g
                                  -10.'T - 3eff'-                    o                                                             ""

27*

                                                                  -                                       r; - Qin AP
                                                                           .                       28. PE3 + KE3-.+ V                                                  3 r    ~E 3    +Q-
                                 .11. . SUR .'" f                                                         PE, + KE2 + V, + FE, + U

12. ng,- Ms o 29. pv - nRT

CR,: 3 'FV 2 P,V,

13. ICRR 30. _

CR 3 's 7a CR 2 31. CV+. 2 2 .+CV -CV

14. M -

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                                                                                                                                       - At-

15. M.-

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16. CR (1

Keff ) - CR,(1 - Keff,)

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                                                                                                                                                                                               ',)-o                     .

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                                                                                                         .._._-_....____m.-s__           ____.__________________.________m.__________m_

v- - g s. , [1- 'EOUATION SHEET'(Sare 2'of ?)' y lC j.: :36. I . I,e-Mx - I,10*(x/TYL). (

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                              ' 3 8. ' I'idi^-'I;di' O.5 CE 39..R/hr              d2 -

(*' *## <

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40. R/hr .6;CE/d2-(feet)

CONSTANTS 6 CONVERSIONS N 6 023 X 1028 atoms / mole-1 vart -.3.12 X 102* fissions /sec'ond . 4 1 Curie - 3.7 X 3020 disintegrations per second

p

                             'l amu.- 931 Mev                                                                                                                                                             '
                              =g      ,32 ft/sec2' g,: - 32. f t-lbm/sec2-1bf-
                                                ,.w,
                   ' J - 778 -[3 g "
                                                           =
                                 *F - (9/5'C) + 32 L'c -L(*F-32)'5/9
                                *R         *F + 460
                               *K          'C + 273
                             .c      - 1.0 ETU/lbn *F @ STP E:D'- 62.4 lb=/f: 8                        -

1 g=/c=8 @ STP 4

                 .- i_____         m        _         _      __m-    . . _ .         ___    _.m__m._ m -.m..--____m -_ __mm.-mm_--

CATEGORY I REACTOR THEORY (7%), THERMODYNAMICS (7%) COMPONENTS (11%) GROUP 1' REACTOR THEORY oQUESTION 1.01 (1.0) MULTIPLE CHOICE (Select the correct answer) A startup is to be perf ormed 4 hours af ter a trip from 100% power, and the Estimated Critical Position (ECP) has already been calculated. Which event would cause the ACTUAL critical control rod position to 'be LOWER than the calculated ECP?

a. The startup is delayed until 8 hours after the trip.

b. The condenser steam dump pressure setpoint is increased 25 psig.

c. Baron concentration is increased from 1140 ppm to 1150 ppm.

d. All steam generator levels are raised 5% just prior to startup.

OANSWER D

REFERENCE GOI-8, Estimated Critical Position KA 192OOOK101 3.4/3.5 i

_ _ _ _ - _ _ . _ _ . ___-______-__________-__a

cQUESTION 1.02 (1.0) MULTIPLE CHOICE- (Select the correct answer) A reactor is subcritical with a Keff of 0.95 and a source range count rate of 200 counts per second (CPS). Control rods are subsequently withdrawn to add 0.017 delta k/k reactivity, and the reactor remains subcritical. What is the subtritical steady state count rate after this reactivity addition?

a. 250 CPS b.,

300 CPS

c. 450 CPS

d. 600 CPS oANSWER B

S/ (1 - 0.95) = 200 (given information); S= 10 10 / (1 - 0.967) =X X = 300 CPS

 + REFERENCE Lesson Plan 02-H-4-LP, page 18 KA     192OO8K103    3.9/4.0 l

l 2 l

                                                              . - - - - - - _ - - -             _a

P 19 QUESTION' ** 1.03'~(O.75) MULTIPLE CHOICE (Select 1the correct answer)

                 ,During a reactor startup, an initial reactivity addition causes count rate

to increase from 20 CPS to 40 CPS'(stable)... A second reactivity-

F addition causes-the count rate to increase from 40 CPS to 80 CPS-(stable).

                 .Which reactivity addition was smaller?-

a. The first reactivity addition was' smaller.

L b. The second reactivity addition was smaller. c[ 'The first and second reactivity additions were equal.

d. .There is 'insuf ficient data given to determine relationship of reactivity values.

               # ANSWER B
             '# REFERENCE Lesson Plan 02-H-04-LP, page 17 KA    192OOOK103    -3.9/4.0 M

t 3

,m____-.__.-___ - - - - - . _ - - - . _ - - __ _ - - - - . - 6

              'OQUESTION l'.' 04    (0.75)
              ; MULTIPLE CHOICE               (Select.the correct answer)

The plant is prepared for startup after refueling. .Keff = 0.95 and count rate is 10 cps with all rods inserted. The operator pulls the shutdown banks fully out, adding + 3.5% dK/K of reactivity.- No other operator-action is taken, and the reactor remains subcritical. How will nuclear power respond?

a. Continue increasing until the fuel or moderator. temperature-

                           ' increases.

b ., ' Increase to a new equilibrium count rate depending on Keff and source neutron level..

                     .c. Increase until the positive reactivity addition is offset by the Isothermal Temperature Coefficient.
                     ,d. Increase while rods are moving but will decrease by an observable amount as soon as rod motion stops.

ANSWER

                        'B
               -OREFERENCE Lesson Plan 02-H-04-SD, page 17 KA          192OO8K104 3.8/3.8 1

I 4

                                                +                                   -     -   .               _.                                --- - - _ - _ - - - - _ _ _ _ - - _ - - - _ - - -

1;

     , ~OQUESTION 1.05~ (1.0)
       -MULTIPLE CHOICE . (Select the correct answer)

Which parameter (s) : determines if the moderator temperature coefficient

         ' i s . positive or negative?

l l a.. Both the moderator-to-fuel. ratio and the boron concentration because as moderator temperature. increases, the moderator-to-fuel ratio decreases,

b. 'Only the boron concentration because raising -the. boron concentration makes the moderator temperature coefficient more negative.

ci Only the moderator temperature because the magnitude of'the moderator temperature coefficient is greater at higher temperatures.

d. Both.the boron concentration and moderator temperature because although both the moderator pressure and void coefficients are positive they are neglected.

oANSWER A OREFERENCE Lesson Plan 02-H-06-SDi, pages 23-28

       ' KA    192OO4K101'      3.1/3.2 5

n - y f.- GROUP II REACTOR ~ THEORY

         .* QUESTION-1.06                                                   (1.0)

MULTIPLE CHOICE (Select the correct answer)

Which statement best describes' Xenon characteristics during plant.' operations? t a.- Peak Xenon conditions after a reactor trip-will preclude attaining. criticality now the beginning of cycle at Trojan.

b. Equilibrium xenon concentration at 50% power is greater than half of the equilibrium concentration at 100% power.

c. If the reactor trips from 100% power, peak xenon concentration will-occur in about 50 hours.

d. Xenon peaking after a reactor shutdown is experienced because the reduced flux increases production about 25%.

          # ANSWER' B

REFERENCE Lesson Plan 02-H-06-SD, figure 27

            'KA                                                  192OO6K105 3.1/3.1,                                                                               192OO6K107  3.4/3.4
                                                                                                                                                                                            )
                                                                                                                                                                                            )

6

oOUESTION ** 1.07 (1.0) MULTIPLE CHOICE (Select the correct answer) Which statement best describes Xenon 135 as a fission product poison?

a. It is a stable fission product; the largest source is from direct fission and secondary source f rom Iodine 135 decay.

b. It has a 9 hour half-life; the largest source is decay of Iodine 135 and secondary source from direct fission.

c. It has a 6 hour hal f-lif e; the largest source is direct fission and secondary source from Iodine 135 decay,

d. It has a 3 hour half-life; the largest source is decay of Iodine 135 and secondary source from direct fission.

eANSWER B OREFERENCE Lesson Plan 02-H-06-SD and Objectives X through BB of 02-H-06-SD KA 192OO6K102 3.0/3.1 7

                                                  .-______-_---_--________-__-_-_-_-_a

\:

GROUP 3 REACTOR-THEORY cQUESTION. ** 1.08- ( 1. 0) - MULTIPLE.CHDICE (Select the correct answer)- Trojan uses both soluble boron and control rods to control excess:- reacti vi ty i n the. reactor. What is the matn advantage of using soluble boron to control reactivity?

a. It has. a minimal effect.on.the neutron flux shape which reduces the maximum-to-average power density ratio.

i

b. It'does'not effectLthe rod worth which' insures that Rod Insertion Limits are reli'able at different boron concentrations.

fc. Baron.primarily absorbs fast neutrons; this allows more efficient use of thermalized neutrons by the fuel.

d. It increases reactor loading rates which allows for rapid power-changes'to; follow customer demand.

ANSWER A
REFERENCE

        " Lesson Plan 02-H-06-SD1, page 16 KA'         192OO7K105                        3.0/3.2
                                                                                                                                                                                                                         ~

8

 ;a                    .

l GROUP 1.THERMO-LOQUESTION' 1.09 (0.75) . MULTIPLE CHOICE -(Select the~ correct' answer) 10CFR50.46 establishes several ECCS acceptance criteria; the peak cladding temperature limit is set. at 22OO degrees F. Why was the value of 2200 degrees F chosen?~

a. 'It allows for a 10% safety margin from the cladding melting point (2450 degrees F).

b ., .Any clad temperature higher than this correlates to a f uel center line temperature at .the f uel 's melting point. c.. The zircalloy-water reaction is accelerated at temperatures above. 2200 degrees F.

d. The' cladding becomes weaker, because of a rirconium phase change at temperatures.above 2200 degrees F.

                  .OANSWER
                                  .C
                   'OREFERENCE Lesson Plan 02-H-OB-SD2, page 29

KA 193OO9K105 3.1/3.5 l

l L 1 9

           ,4
                             )                                                                                                                       I jy "GQUESTION' 2 1 ^.10 ~ -(0.75):

MULTIPLE.CHDICE (Select the: correct ~ answer)- - C Brittle' f racture is a f ailure mechanism of concern for low 1 alloy steel' pressure vessels._ Several OPERATIONAL LIMITATIONS; help reduce.the-

possibility of brittle. fracture ~ occurring.

J- - Which OPERATIONAL LIMIT effects brittle fracture potential?

a. ,RCS/ Pressure-Safety Limit.-

                          .b.                                                       RCS cooldown rate limit.
                         .c., Minimum' temperature for criticality limit.

d. RCS ' chemi stry; 11 mi ts.

j oANSWER' B ,_ cREFERENCE' l Not provided.

KA 193010K104 -3.3/3.7 Gum

                   - - - _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _           ._,._1., _ _ _ . _ _

                                                                                                                          .1 q

oOUESTION- **. 1.11 (0.75) MULTIPLE'CHDICE (Select-the correct answer) The Enthalpy Rise Hot Channel Factor limit in Technical Specifications ccnnot beldirectly observed by the operator. Assurance that this. limit

 ;is~not exceeded 1s -based on maintaining -several related plant parameters within' limits.

Which limit helps ensure.that the Enthalpy Rise Hot Channel Factor 1-i mi t

 .is not exceeded?
     ~n.    -Control rods in a group move together within +/- 12 steps.

b., Tavg vs. Tref are kept matched to within 5 degrees F.

c. Thermal power is limited to 3411 megawatts.

    .d.      RCS minimum flow for. modes 1,          2,  and 3 is 90% of rated flow.

oANSWER A

 .* REFERENCE
   . Technical' Specification Bases 3/4.2.2 and 3/4.2.3 KA     193OO9K107               2.9/3.31 1

1 I

~~

12

                                                                  '  --  - - _ _ _ . _ _ _ _ _ _ _        ---'-%~_.._,__,

    ',, ?I r

1 l:,' ! -# GROUP'2 THERMO-l

               'eQUESTION                                                                                                                                                                    ~ **

1 '.' 12 .(0.75).

               '. MULTIPLE CHOICE..(Select the correct' answer)                           -

Whichi statement'best describes the Departure from Nucleate ~ Boiling

                   -Ratio (DNBR)7

a. DNBR increases with decreasing RCS flow.

b. DNBR decreases with decreasing reactor power.

c. DNBR' decreases with increasing pressurizer pressure.

d. DNBR increases with decreasing Tavg.

                  # ANSWER D.
               -* REFERENCE Lesson Plan 02-H-13-LP, page 13 KA                 '193OOOK105 3.4/3.6' W

12 _ _ _j

L* QUESTION; .

              " 1.' 13 .= .1( 0. 7 5 )
             !.MULTIP'LE CHDICE -(Select the correct-answer)-
    ,g            1 Reactor power;is about to be lowered-from 95"/. to 75% by the method
             ~ , indicated?below.

S . . . . How will the Reactor Coolant'Subcooling Margin (RCSM) change? a.. The RCSM is not a - f unction of reactor power and.will remain constant. 1b.--The RCSM will' decrease.because inward rod motion will increase' control rod density.

                       - c .' The'RCSM will increase because RCS hot leg temperatures will decrease as power'is reduced.

d. The RCSM~will decrease because RCS pressure and. Pressurizer Level will increase on a down-power transient.

                 # ANSWER C

REFERENCE Lesson. Plan 02-H-07-SD, Transient Analysis KA' 193OO8K115 3.6/3.8 6

13

cOUESTION 1.14 (0.75) MULTIPLE CHDICE (Select the correct answer) During power operations, a per centage of RCS flow through the reattor vessel bypasses the fuel assemblies. Which statement describes core bypass flow?

a. Bypass flow is created by design tolerances between vessel components and serves no useful function.

b. Bypass flow cools vessel components and is greater than 4% of total core flow.

c. ' Bypass flow maintains suf ficient cooling flow in the RCS to prevent RCP cavitation.

d. Bypass flow is less than 2% of total core flow and assures that the control rod guide tubes remain clear.

OANSWER B OREFERENCE Lesson Plan 02-A-05-SD2, pages 31, 32 KA 193OO8K120 2.9/2.9 14

  -A-    ;y .
               .* QUESTION 1.15 :(1.0)

MULTIPLE CHOICE (Select the correct answer) Departure from Nucleate Boiling.(DNB) can cause_ rapid. increases in fuel centerline and clidding temperatures by greatly reducing the_ heat

transfer coefficient.. t Which of. the following. best describes the cause of DNB7 a.. High . heat . flux produces boiling bubbles that begin to f orm a film : on the fuel' rod before they can be swept away,

b. High heat flux raises the bulk coolant temperature .to greater than

                         . saturation .and the coolant instantly flashes into a steam blanket.

c. Excessive local heat flux begins to rupture the fuel cladding.

                    'd. Boiling bubbles begin to cause cavitation erosion to occur excessively on the cladding.

ANSWER i

A

                '# REFERENCE Lesson Plan-02-H-OB-SD page 13 KA     193OOOK104' 3.1/3.3 15

9 ,

                                                                                                              = ,s
 's         ,

4s t f IGROUP.E3:.THERMO OQUESTION **

                        ' 1 '.16        (O. 75).

E_ MULi2PLE' CHOICE '(Select the correct answer)-

          +                -Which one of the following events would'cause water hammer?
                            *La.-        Starting a centrif ugal . pump in a filled '(vented) system-with:its discharge, valve shut..
                                                                           ~

16. Rapid: depressurization during ' a hydrostatic, test of a large system.

                              .c.        Rapid pressurization of a solid, stable' system.

1*

d. Sudden' closure of a valve in which'there is high water flow, h

                      =OANSWER D:

REFERENCE Lesson Plan 02-H-12-LP, page 39 KA .193OO6K110 3.3/3.4

wm v

6 4 f 16

i' COUESTION 1.17 (0.75) MULTIPLE CHOICE (Select the correct answer) Which condition is likely to cause cavitation in a closed system containing a centrifugal pump:

a. Suction pressure increases.

b. Pump speed decreases.

c. Suction temperature increases.

d. Discharge line. throttle valve is partially shut.

       ' OANSWER C

cREFERENCE Lesson Plan 02-H-12-LP, page 27p and Exam Bank Question 02H12-4-17 KA 193OO6K111 3.1/3.3 4 17

(V 1-GROUP:1-COMPONENTS.

         . OQUESTION 1.18                 (0.75)
      . MULTIPLE CHOICE- (Select the correct answer)

A motor operated Engineered Saf eguards' valve was manually shut . f or surveill ance testing and is being returned to ' service per - Operations Procedure OM-3-1, Tagging. What. action must be taken bef ore the valve is declared f ully operable 7

a. It must be manually: returned to its original position.

b. It must.'be Danger Tagged until administrative approval :is. signed by the Shift Supervisor.

c. It must be electrically cycled.

                ~d.              The. operating' mechanism must be lubricated with an alcohol and
                                     ~

graphite solution. OANSWER C-

          =0 REFERENCE Operations Manual OM-3-1 section 3.10.1 KA           191001K106'         3.3/3.7 4

18

E

QUESTION **

1.19 (0.75) MULTIPLE CHDICE (Select the correct answer) What type (s) of radiation is/are measured by a Digital Alarming Dosimeter? l

a. Gamma only.

b. Gamma and Beta.

c. Gamma and Neutron.

d. Gamma, Beta, and Neutron.

OANSWER A

REFERENCE Lesson Plan G1-F-02-HO, page 7 (General Employee Training)

KA 191002K119 3.1/3.3 i L -- , l 1 2'

  - _ _ - -                    -    -                                                        l

N' 1OGUESTION 1.20 (0.75) MULTIPLE CHDICE (Select the correct answer)- An, operator' drops his Digital Alarming Dosimeter (DAD) while working in a radiation area.- Visual inspection reveals no obvious damage to the DAD. (.. What action sh'ould the operator take?

a.- Since the DAD passed a visual inspection, no action is.necessary; his TLD will be used to determine his' legal dose.

                  'b. .             The-operator.should leave the radiation area and report'the
                      -             dropped ' DAD to- Radiological Protection,

c. The DAD is designed to take minor impacts, however, the event should be-reported to Radiological Protection whenever the DAD is turned in.

d. The' operator sho"* d report the dropped DAD to the Shif t Supervisor because the DADt rust more than-$1000.00.

DANSWER B

REFERENCE

             ,' General Employee Training conducted on 2/21/89.

KA 191002K119 3.1/3.3 t ( 20 _ _ - _ - _ _ -

i ! .OQUESTION u 1.21 (1.0) MULTIPLE CHOICE (Select the correct answer) During full power operations, an RCS piping failure caused a large LOCA. Reactor Trip and Safety Injection activated, but voiding in the core is still occuring. Which event would result in a larger increase in the Source Range detector output?

a. Voiding occurring in the upper half of the core.

b. Voiding occurring in the upper half of the downtomer.

c. Voiding occurring in the lower half of the core.

d. Voiding occurring in the lower half of the downcomer.

OANSWER D OREFERENCE Lesson Plan 02-B-07-LP, page 14 (Objective 1.2.2.G.9) KA 191002K117 3.3/3.5 21

8 . ;.g k; - ? ...

                    ,-i
    ,p.,

3. q...

                     .. OQUESTION!

1.22 -(O.75)- :

                      . MULTIPLE" CHOICE                .(Select'the correct answer)-

What type of detector is used.for the' source range nuclear

                            .: instrumentati ora?                                                                                            .

a. : Compensated (Ion Chamber.

b.' Fi'ssi on' Chamber. i . c.. Proportional Counter.

...g d. . Geiger-Muller Counter.

                    'OANSWER C ..

OREFERENCE License Exam Bank ' Question No. B-07-B2-1 KA 191002K118 2.6/2.8-O 4 22

l i s eQUESTION 1.23 (0.75) 1 MULTIPLE CHOICE (Select the correct answer) Indicate which of the following DECREASES the available NPSH of a centrif ugal pumps

a. The pump discharge valve is throttled shut.

b. The temperature of the pump suction fluid decreases.

c. The nitrogen blanket pressure on the suction side supply tank increases.

d._ The pump discharge valve is throttled open.

ANSWER-D cREFERENCE Lesson Plan 02-H-12-LP, page 25 KA 191004K106 3.2/3.3 l

1 23

c'

              . *OUESTION                                                                                              **

1.24 -(0.75)

MULTIPLE CHOICE. (Select the' correct answer)

-f         ,     .The North Centrifugal' Charging Pump 'i s running with the' flow control

valve,~FCV-121,..in mid position.

                  'If FCV-121-was fully. opened, which parameter'would INCREASE 7
                      .a. . Pump discharge pressure upstream of FCV-121.

b. Motor amps.-

c. Avail able NPSH to ' pump.

                     'd. Pump suction pressure.
             ' OANSWER B
              -OREFERENCE Lesson Plans'02-H-12-LP, page 25; and 02-A-06-SD (CVCS)
                 'KA 191004K106 3.2/3;3
  . es
       .e i.

M 24

                                                                .__ _ _ _ _ - - _ _ _                                                 ~

          ,  f6                     1
                           )-
          .y.     :
                ..7
                '

QUESTION i.;. r.; 1.25 .(0.75)

                . MULTIPLE CHDICE                   (Select the correct answer)

The - CVCS' Positive Displacement Pump i s operating: with insuf ficient net -'

positive suction head.

                         '.Which one of the'fol1owing is 1ikely to occur?
                            . a.      . Slip.-

b. Decreased pump speed.

ck Viscosity' loss.

d. Vapor binding.

                ' cANSWER
                               -D
   's

REFERENCE

Lesson Plan 02-H-12-LP, page 25 KA'191004K106 3.2/3.3

  '.m.

0 25

oOUESTION ** 1.26 (1.0) MULTIPLE CHOICE (Select the correct answer) Prior to closing any alternating current (A.C.) circuit breaker, the two power sources must be in phase to prevent component damage. How is this requirement assured when closing the Main Generator output breakers?

a. The output breakers are interlocked with the syncroscope (" sync check relays") to prevent closing unless both lines are in phase.

b. The output breakers are always closed onto a deenergized bus so

     ,that paralleling is not necessary.

c. The operator must properly parallel across the breaker, no electrical . interlocks are provicled.

d. All 230 KV breaker s are the break-bef ore-make design which allows for any breaker closing sequence.

OANSWER A

REFERENCE' Operating Instruction 1-2, page 2 (02-C-01-LP1 Objective 1.2.2.B.2.1),

and 02-C-02-SD1 page 9 of 27. 'KA 191008K107 3.0/3.3 { 26 i

                                                               ._-___-_____-_____-_A

v - oQUESTION 1.27- -(0.'75) MULTIPLE. CHOICE ~(Select the correct answer) During power operations, one 4.16 KV ESF bus (A1) has been deenergized because of'a spurious feeder breaker trip (152-101). The Emergency Diesel Generator failed to start.. The operator is ready to reenergize

          'the ESF. bus tnr reclosing the feeder. breaker (152-101).

I- What should [the syncroscope needle motion be just prior to closing the feeder breaker onto a deenergized bus?-

a. It will be stationary, no rotation will be' observed.

b. It,will rotate slowly in the FAST (clockwise) direction.

c. It will ' rotate slowly in the SLOW (counter- clockwi se) direction.

d.- The syncroscope would not be selected if the bus was.deenergized. OANSWER

                 'A OREFERENCE Lesson Plan 02-C-04-SD, 4.16 KV System (Objective 1.2.2.C.8).

KA 191008K107- 3.0/3.3 i 27

1 i OQUESTION , 1

           . 1.28    .(1.0)..                                                                   !

MULTIPLE CHOICE (Select the correct answer) During power operations, one 4.16 KV ESF bus (A1) has been deenergized-because of a spurious feeder breaker trip (152-101). The Emergency

           . Diesel Generator failed to start.       The operator is ready to reenergize lthe ESF bus by reclosing the feeder breaker (152-101).               The operator places the Control . Room switch to close WITHOUT turning on the -

syncroscope. Will the breaker close?

a. Yes,.but the.syncroscope circuit will delay actual breaker closure i until both lines are in phase. l

b. Yes, because closure interlocks are bypassed when the oncoming bus -

is deenergized. j c.- No, because.the two lines may be out of phase and breaker closure could damage the breaker,

d. No, because the sync selector switch is interlocked with the breaker closure circuitry.

CANSWER

                 .D OREFERENCE Lesson Plan 02-C-04-SD, 4.16 KV System (Objective 1.2.2.E.2).

KA '191008K107 3.0/3.3 9 28 , 1

l;; 1' 1 L . GROUP.2 COMPONENTS 1

          .' OQUEST I ON                                                                                                                                                                 **
            .1.29    (1.0) l MULTIPLE' CHOICE- (Select the correct answer) l ..

L .The number of starts over ' a given time period . is limited f or the' Reactor ll Coolant-Pumps, per Trojan Operating Instuction (OI) 3-4: 1 l Why is- this limit required?- a.' The' motor start capacitors.are. air cooled-and require significant'(30-minutes) cooldown time. b,: ;High' motor starting currents can overheat and damage; motor windings if' adequate -cooling time is not allowed.

c. High:' heat loads on'the'_ motor 1ube oil. system from excessive cold motor starts.can cause' motor binding.

d.- The split' phase induction design motors require high starting l', currents.that.may overload supply breakers if a rapid restart is. performed. 1 OANSWER ! B OREFERENCE

             -OperatingLInstruction 3-4, page 5 KA     191005K106    3.0/3.1 1                                                                                                                                                                                                               i 1

l i _ _ _ _ _ . _ . _ . .____M._____.._ _ _ _ . _ . _ _ . _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

OOUESTION 1.30 (0.75) MULTIPLE CHOICE (Select the correct answer) Compare the indications that would be observed if the West RHR Pump were started and operated with its discharge flow path isolated, as compared to with its discharge flow path open (ASSUME RUNOUT FLOW CONDITIONS). What are the indications f or the " isolated discharge" case?

a. Higher starting current and lower running amperage.

b. Lower starting current and lower running amperage.

c. Higher starting current and higher running amperage.

d. Lower starting current and higher running amperage.

OANSWER , B-

REFERENCE Lesson Plan 02-H-12-LP, page 22 KA 191005K104 2.7/2.8 l

l I 30

p p. oQUESTION 1.31 (1.0) MULTIPLE. CHOICE (Select the correct answer) The B Emergency Diesel Generator is. operating at'95% load'in parallel with the Normal 4.16 KV-Vital' Bus power supply for surveillance testing. The operator places the GOVERNOR Raise / Lower Control Switch to the RAISE f: position. How will.the Emergency Diesel Generator,(EDG) respond?

a. Generator kilowatts will increase until either the operator _

                        . releases the switch or the EDG trips on overcurrent.

b. Generator Frequency, RPMs,-and Kilowatts will increase until either the operator releases the switch, or the EDG trips on overspeed or.overcurrent..

c. Generator parameters will not change because the Governor Limit-Switch will not allow EDG loading above 100% unless an ESF signal has been received.

d. Generator. Frequency and RPMs will increase until either the operator releases the switch or the EDG. trips on overspeed.

              # ANSWER A

eREFERENCE Lesson. Plan 02-A-13-SD, Emergency Diesel Generator (02-C-04-LP Dbjective 1.2.2.C) KA 064000K101 4.1/4.4 i

31 j

A p, LOQUESTION

         ;1.32 ' (0. 75) -

MULTIPLE CHOICE' .(Select the correct answer)

           ?The B. Emergency Diesel Generator is operating at 95% load.in parallel with the Normal 4.16 KV Vital Bus power' supply f or surveillance testing.

The operator places the VOLTAGE Raise / Lower' Control Switch-to the RAISE position for three (3) seconds. How wil1 ~ the Emergency Diesel Generator (EDG) respond? l

a. Generator. Voltage will. increase.

b. Generator Kilovars will increase.

c. Generator Kilowatts will: increase.

d. 'EDG Frequency and RPMs.will increase.

CANSWER B. OREFERENCE Lesson ~ Plan-02-A-13-SD, Emergency Diesel Generator (02-C-04-LP Objective-1.2.2.C) KA 064000K101 4.1/4.4' l l E. END of CATEGORY ONE

32 3

CATEGORY II EMERGENCY PLANT EVOLUTIONS (27%) GROUP I EMERGENCY PLANT EVOLUTIONS QUESTION ** 2.01 (3.0) A fire is burning out of control in the Turbine Building and has been confirmed by plant operators. The fire has not been contained and is cpreading. The Control Operator announces the alarn. over the plant paging system, and repeats the announcement. The Shift Supervisor has been i informed and is executing EFP-0, Procedure in the Event of a Fire:

a. What are the two(2) notifications that EFP-O requires the Control 1 -Operator to make upon sure indication of a fire? (0.75 each)

6. What are two (2) of the three (3) conditions in EFP-O that require execution of EFP-1, Alternative Shutdown for Evacuation of Control Room Caused by Fire? (0.75 each) 1 ANSWER

a. Notify the (Rainier) Fire Department (0.75)

Notify the Security Watch Supervisor (0.75)

b. (any 2 of the following; O.75 pts each)

Large fire threatens redundant equipment (and is out of control longer than 15 minutes). Differences exist between 2 or more CATEGORIES of instruments (and significant loss of instrumentation). The control room is physically uninhabitable. l REFERENCE EFP-0, Procedure in the Event of a Fire (02-K-09-LP Objective 1.2.2.F and H) KA OOOO67K304 3.3/4.1 33 _ _ - _ _ - _ _ _ _ _ _ _ - _ - - - _ _ -

QUESTION 2.02. (3.O)

CSFSTs are attached to this exam ASSUME degraded containment conditions. During power operations, an accident has occurred causing a reactor. trip signal and degraded core cooling. FR-C.2, " Response to Degraded Core Cooling," is being executed as required by the Critical Safety Function Status' Trees (CSFSTs). RVLIS currently indicates 50%. The CSFSTs require departure from FR-C.2 to' perform other emergency actions if. certain plant conditions (limits) are exceeded. One condition would be if Containment pressure exceeded 60 psig. What' are the other plant conditions requiring departure f rom FR-C.2 f or each of the following CSFST Priorities (INCLUDE SETPOINTS)? (a) HEAT SINK (two parameters, 0.5 each)- (b) SUBCRITICALITY (two parameters, 0.5 each) (c) RCS INTEGRITY (two parameters, 0.5 each) (3.0)

ANSWER (0.5 each parameter, 1.0 each) (3.0)

(a) Total feed capability less than (or equal to) 720 gpm, AND Narrow Range SG 1evel less than (or equal to) 12%. (b) Power range channels greater than (or equal to) 5% OR positive intermediate range start-up rate. (c) Cooldown in all RCS cold legs greater than (or equal to) 100 F in one hour AND RCS pressure / temperature to the lef t of Limit A.

Reference FR-0, Critical Safety Function Status Trees (02-K-04-LP Objective 1.2.2.A.9)

KA OOOO74G012 4.3/4.4 4 34

7 l l 1 cQUESTION 2.03 (3.0) . i The reactor is initially operating at rated power with all systems in i cutomatic. The INSTRUMENT AC BUS UNDER VOLTAGE annunciator alarm and control board indications confirm that power was lost to Instrument Bus YO2. Power is also lost to MCC-B26. The Main Feed Pumps (MFP's) trip on low suction pressure, and then the reactor trips.

a. What THREE (3) component failure responses (NOT electrical systems) contributed to the MFP trips?

(0.5 each)

6. What are two (2) of the three (3) immediate operator actions required by ONI-46, " Loss of 120 VAC Instrument Bus," IF plant conditions did UgT cause the MFP's and reactor to trip? (0.75 each)

CANSWER

e. The recirculation valves (f ailed on loss of power) for (0.5 each)

Condensate Pumps, Heater Drain Pump, and MFPs (1.5)

      .b.     (any 2 of the following; O.75 pts each)                      (1.5)

Reduce power to less than (70% at the maximum rate). j (Determine cause of power loss and) restore power to the bus. (If MCC-B26 power was lost,) open input breaker from MCC-B26 and close input breaker from MCC-B25. i

REFERENCE ONI-46, page 13 (02-C-06-LP Objectives B.2.h and j)

KA OOOO57K301 4.1/4.4 l 35 ; i

L l

            -GROUP II EMERGENCY AND ABNORMAL PLANT EVOLUTIONS
            'OQUESTION                                                                                                            **

2.04 (2.0) Trojan han experienced a reactor trip with an inadvertent Safety Injection.- The plant operators are performing Emergency

            ' Instruction EI-0, " Reactor Trip, Safety Injection,.and Diagn'osis."

Three immediate. actions of EI-O have been performed by the operators: Verify ~ Reactor Trips Verify Containment Isolation, Phase A; and Check Containment Pressure. i i What -are the remaining eight -(0) immediate actions-of EI-O (regardless of' order)? (0.25 each) OANSWER (0. 25 ' each') 1.. Verify Turbine Trip.

2. Verify Generator Trip.

3. Verify Power to AC Emergency Buses.

            ,4. . Check if SI is Actuated.

5.-Verify Feedwater Isolation. l

6. Verify AFW Pumps Runr.ing.

7. Verify ECCS Flows.

B. Check if Main Steam Lines should be isolated. i. f OREFERENCE-Emergency Instruction EI-0, " Reactor Trip, Safety Injection, and Diagnosis" Pages 2 thru 5 (02-K-12-LP Objectives 1.2.2.A.14) KA OOOOO7K301 4.0/4.6 4 36

l i.

h. <

                      ;*QOESTION                                                                                                                                                                                                           **

l , 12.05 (1.0) MULTIPLE CHOICE (Select the correct answer) Trojan has experienced.a reactor trip and Safety-Injection. The plant. operators'are performing Emergency Instruction EI-0, " Reactor Trip,. Safety;' Injection, and Diagnosis".

                      . !4 hat is one of the conditions.that must be met prior to terminating Safety Injection?.

a. RCS subcooling less than 25 F.

                                 .b.        Pressurizer level greater than 5%.

c. .RCS-pressure decreasing.

d.- .AFW flow-less than'495 GPM. OANSWER (0.75)

1. B OREFERENCE Emergency' Instruction EI-0," Reactor Trip, Safety Injection, and. Diagnosis", Page 11 (02-K-12-LP Objectives 1.2.2.A.3 and 5)

KA .OOOOO7K301 4.0/4.6 up 6 37

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

r id cQUESTION 2.06' (1.0) MULTIPLE. CHOICE- (Select'the correct answer), Trojan.has experienced a. reactor trip and Safety Injection. The. plant operators are performing ~ Emergency Instruction.EI-0,." Reactor Trip, Safety Injection, and Diagnosis". Which of'.the following-would require use of " adverse containment"

             - parameters?

a.. Containment. pressure greater than 3.5 psig.

b. Containment radiation levels greater than.10,000 R/ hour.

c. Containment temperature / pressure greater than saturation.

d. . Containment pressure AND radiation levels greater than.their alarm setpoints.

             # ANSWER     (0.75)

A-OREFERENCE Emergency Instruction EI-0," Reactor.' Trip, Safety Injection, and. Diagnosis", Page 11 (02-K-12-LP Objectives 1.2.2.A.3 and 5) KA- OOOOO7K301 -4.0/4.6 38 - _ _ _ - - - _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ -

                                   . s:-
 '1 g

M

                          \   .'

x

QUESTION **

i . 2.07 (1.0) MULTIPLE. CHOICE (Select the correct answer) The: Trojan nuclear' plant is in Mode 2 conducting a reactor'startup when

                 'o Source Range. Nuclear: Instrument.(SRNI) fails.
              'What is'a probable cause'of the Source' Range malfunction?

6

a. Source range counts greater: than 1xiO ,

o

b. Depletion of boron trifloride gas'in the detector.

c. Excessive discriminator voltage.

                        -d. Source range drawer crowbar circuit is actuated.

cANSWER

.D

REFERENCE

                    .Off-Normal' Instruction ONI-16', Rev. 7, Excore Nuclear Instrumentation Malfunction, Page 2 of 7-(02-B-07-LP Objective 1.2.2.G.6).

KA OOOO32K302- 3.7/4.1

_____1_____..____

1

   '* QUESTION-2.08           (1.0)

MULTIPLE'CHDICE f(Select the correct' answer)

  'The Trojan nuclear plant is in Mode 2 conducting a reactor startup when
   .a:: Source Range Nuclear Instrument (SRNI) f ail s.

What action is.tgguited by Igghnigal_ Sagg[fi_gatigas for a single SRNI failure'in Mode'27 t a.- Verif y that NR-45 is not recording the .def ective channel . b.- Verify.that'the operable Source Range channel is selected as an input' to the audio count rate channel. I. .- L :c. Suspend all operations involving positive reactivity changes. !

d. Place .the level trip switch for inoperable channel to the BYPASS position.

oANSWER C.

   'aREFERENCE Trojan Technical Specification 3.3.1                                                                           (02-B-07-LP1 Objective 1.2.2.K.3)

KA OOOO32K302 3.7/4.1 m 40 _ _ = . - -__-_:___ _ - _ _ , _ - - - - _ - _ ___ _ -- _ _____--________- _ _ ___ _ _ _ _ _ -

?)hi + 1, ' U e H Jc-QUESTION. , ~2.09: (1.0)- MULTIPLE CHOICE' '(Select the correct answer)

           . Emergency;I'nstruction EI-3, Steam' Generator Tube Rupture, provides

actions to . terminate leakage of ' reactor coolant- into the secondary system f ollowing a steam. generator tube rupture.

O Which method does EI-3 use to identify the ruptured SG7 p c .' High containment-radiation from PRM-1, 5 or 14.

            'b. Any RCS loop flow greater than 100%, or a 1%' difference between
                  ' loops.

c. .bnySGpressuredecreasinginanuncontrolledmanner.

s

d. Unexpected rise in SG 1evel inconsistent with feed flow.

            .oANSWER D

eREFERENCE , Emergency Instruction-EI-3, Steam Generator Tube Rupture Rev.13

              'Page 4 of 33 (02-K 15-LP Objective 1.2.2.8).

KA- OOOO38A203 4.4/4.6 41 j

QUESTION 12.10. (3.0) Trojan has been. in Mode 6 f or three weeks and ref ueling operations are in progress. -HIGH radiation alarms were just received on ARM-12 (Maintenance Shop) and ARM-13 (New Fuel Storage). Alarm. validity has been verified-ai ' and the radiation level in these. areas is 250 mR/ hour.

             '1..What are the three (3) immediate actions required by ON1-12, High Activity Radiation Monitoring? (0.5 each-)

2. What is- the' maximum stay time f or a worker with no previous exposure

                  .(documented on NRC Form 4) without exceeding 10CFR2O limits?

SHOW YOUR WORK!

                   -                                                                                                               (1.25 f or application)

(0.25 for answer)- ANSWER

1. (0.5 each)

Evacuate personnel (as necessary to the controlled access point) Determine if - a f uel handling accident has occurred (refer to FHP-13) l. l Start'SFP exhaust fan l~-

2. 10CFR2O quarterly limit is 3,000 mR (this may be inferred by the problem solution, 0.75) 3000 mR / 250 mR/hr = ~ (0.5 f or f ormul a)

                                                  = 12 hours                                                          (0.25 f or answer)
   ,          REFERENCE-ONI-12, page 3 of 12, and 10CFR20.101 (02-J-04-LP Objective 1.2.2.C and 02-J-05-LP Objective 1.2.2.B)

KA 000061G010 3.3/3.3 194001G103 2.8/3.4 42 __- __ .-__ - - - - . _ _ - _ _ - _ - - -

o ESTION

2. (1.0)

MULTI _E CHOICE (Select the correct answer) Trojan i operating at rated thermal power when an RCS pressure transient causes a essurizer Safety Relief Valve to lift (below its setpoint) , and fail to reseat. The SRV leak rate (LOW FLOW) estimate is being calculated. Which annunciat would be the First (most probable) indication that the SRV was leaking? HI TEMP would alarm and not clear. ({)PZRRELIEFLI AKAGE HIGH would alarm and will not (f)PZRSAFETYVALVE clear.

c. PZR SAFETY VALVE OPEN ould alarm and not clear.

d. PZR PROTECTION LO PRESSU - would alarm and the reactor would trip.

cANSWER A gj? O

REFERENCE ONI-36, Pressurizer Relief or Safety V ve Actuation or Leakage; and Lesson Plan 02-A-03-SD pages 7-9 (Object've 1.2.2.B.2.e)

KA 000008G005 3.8/3.9

                                                                  )e & rrf 4       ,
                                                                             //% *//f sr esLc         /'# ' f Gm A c- w w w

("z p csrsddd" s e f *=- eve f

6* * .C Sysmsep ( sT V

\ \

43

GOUESTION 2,12 (1.0) MULTIPLE CHOICE (Select the correct answer) Trojan is operating at rated thermal power when an RCS pressure transient causes a pressurizer Safety Relief Valve to lif t (bel ow its .setpoint) , and fail to close. The SRV flow rate is HIGH. Which annunciato- would be the First (most probable) indication that the SRV was passing a high flow rate?

a. PZR SAFETY RELIEF VALVE OPEN would alarm and clear.

b. PZR PROTECTION LO PRESSURE would alarm and the reactor would trip.

c. PZR SAFETY RELIEF VALVE LEAKAGE HIGH would alarm and not clear.

d. PZR RELIEF LINE LOW TEMP would alarm and high flow indicated by temperature porportional to flow.

OANSWER C OREFERENCE ONI-36, Pressurizer Relief or Safety Valve Actuation or Leakage; and Lesson Plan 02-A-03-SD pages 7-9 (Objective 1.2.2.B.2.e) KA OOOOOBGOO5 3.8/3.9 l l 44

oQUESTION 2.13 (1.0) MULTIPLE CHOICE (Select the correct answer) The plant is in Mode 5 with the RCS partially drained for maintenance. The operating RHR pump (chutdown cooling) trips and it is estimated that boiling will occur in the core in about i hour. What immediate action should the operator take?

a. Investigate the cause of the RHR pump trip.

b. Attempt to restart the RHR pump that tripped.

c. Start the other RHR pump.

d. Initiate Safety Injection.

CANSWER A OREFERENCE ONI-13, Page 6 of 10 (02-A-OB-LP Objective 1.2.2.J) KA OOOO25G010 3.9/3.9 l 45 __ -_-____-_-____________- _ -

w e s. L 'oOUESTION. 2.14 (1.0) MULTIPLE' CHOICE (Select the correctranswer) l< ?The..RCS isisolid. The plant is in' Mode 5 with' the EsCS in a normal solid plant: condition. 'The operating RHR pump- (single pump operation) trips and the' loop suction isolation valve (MO-8701) inadvertently isolates. What:immediate operator action should be taken?-

a. Start the second RHR pump.

b.: Attempt'to reopen MD-8701.

               .c. Start ~a Reactor Coolant Pump.

d. Stop all' Charging Pumps.

               #ANEWER D.

OREFERENCE ONI-13,-page 4 of 10 (02-A-OB-LP Objective 1.2.2.J). KA OOOO25G010 3.9/3.9 L. l 1-l 46

p l l GROUP III EMERGENCY PLANT EVOLUTIONS oDUESTION 2.15 (1.0) MULTIPLE CHOICE (Select the correct answer) ECA O.0, Loss of All AC Power, states the actions that are required for a Station Blackout. Which event is an Entry Condition for ECA O.07

a. All Emt,rgency Diesel Generators fail to start after Safety Injection is initiated.

b. The Main Turbine trips (load reject) and an immedi ate loss of Offsite power occurs.

c. The Control Room has indication that all main and emergency AC busses are de-energized.

d. Fire in the Cable Spreading Room causes a loss of power to both trains of the Reactor Protection System.

oANSWER C OREFERENCE ECA O.0, Loss of All AC Power (02-K-22-LP Objective 1.2.2.A) KA OOOO56GO11 3.5/3.8 1 47

i

        / oOUESTION j'            -

2.161 (1.0) h , NULTIPLE CHOICE (Select"the-correct answer) During Of f Normal Instruction ONI-50, . " Plant Operations af terz a loss of Off-SiteLPower", several loads.are sequenced-onto the Diesel Generators

         ' by the-loadLsequencers.

Which~ component (s)'must be MANUALLY;1oaded onto the Diesel. Generators?-

a. Instrument and' service air ~ compressors.

b. Component cooling water pumps.

c.. Service water pumps.

d. . Turbine Generator. emergency bearing. oil pump.

l

         -OANSWER.

A

        . OREFERENCE
             .ONI-50,. Plant Operations After A Loss'of Off-Site Power, Pages 1
             .and 2 of'7 (02-C-04-LP-Objective 1.2.2.G.2 and H).
            'KA         OOOO56K301                                     3.5/3.9-w em 48

s LoDUESTION 2.17. (1.0)

         ' MULTIPLE CHOICE: (Select the correct answer)

The controlling pressurizer pressure channel has failed. Switch PS-455F (Pressurizer Pressure Control Selector Switch).was used to select an

        ' operabl e ' channel .
         . Which condition may .have been caused by .operati on of ' PS-455F7
         'a. An automatic reactor. trip.

b.- Lifting or seating the Power Operated Relief Valves. c.;. Lifting.or. seating the code Safety. Valves.

        .d.      Loss'of pressurizer level control.
       - oANSWER' B

oREFERENCE Operating' Instruction 01-2-B, " Transfer of Pressurizer Pressure Control Channels", Page 1.of'2 (02-B-04-LP Objective.1.2.2.G). KA OOOO28K305 3.7/4.1 END of CATEGORY TWO 1; 49

,.3 f :- b CATEGORY 3-PLANT SYSTEMS (38%) .AND PLANT-WIDE GENERIC RESPONSIBILITIES (10%) PLANT-WIDE GENERIC RESPONSIBILITIES

QUESTION 3.01 .( 2.5)

Administrative Order, AO-3-26, Independent Verification, provides guidelines on independent verification. Two(2); ectivities that require independent verification ares-(1) system lineups pricr to startup.after an outage, and (2) hydrostatic. testing.

 ,         .a. What are two(2) OTHER activitites (or evolutions) that require independent verification per AO-3-267         (0.75 each) (1.5) b.-  What are the two(2)' methods of. determining the status of circuit breakers per AO-3-267       (0.5 each) (1.0)
          'cANSWER

a. (0.75'each)

                 -1) Clearances.

2) Surveillance testing.

           .b . By mechanical indication (visual inspections closed / tripped flags, pointers) (0.5) or energization of downstream components-(0.5)

REFERENCE Administrative Order AO-3-26, a) Page 2-3, b) Page 5, c) Page 6 KA 194001K101 3.7/3.7 50

i, 4 ,

                                                                                                                                                )
                 *OUESTION 3.02-       (2.25)-

LPS-3-10, Confined Space Entry, requires a confined space to be ventilated

             -; prior to entry, especially if the space normally contained fluids or gases other than. air. Lines, valves,'and electrical equipment associated with!the space are tagged to prevent personnel injury. Oxygen. deficient atmospheres require a minimum of 5 air changes:  .
  ~

a. What kind of.-atmospheric condition-would require'J.O air changes, per PS-3-107 (0.75) b..What are.the THREE~(3) . conditions required.to release the'-confined space protective clearance after all work is complete? (0.5-each)

ANSWER

a. . Toxic or (flammable atmospheres)'. (0.75) b' . All i personnel- (0. 5)', tools / equipment are out of the space (0.5),,

and the space has been closed (0.5). eREFERENCE-PS-3-10, a)Page 6 Para.3, b)Page 9 Para.E.2, c)Page 10 Para.F.2

               ,KA-194001K114        3.6/3.6 b

51

i fcQUESTION

3.03 (1.50)

Plant' Safety PS-3-2, Work on.High Temperature /High Pressure Systems,

describescthe. procedure to be followed when working on systems that

contain high. temperature and/or high pressure fluids.

c'. . How does PS--3-2 define a "high temperature" ' fluid? (0.5) (0.5)

t :b. How does-PS-3-2 define "high. pressure" fluid?

c. Whose approval must be'obtained prior to starting work on a cleared piping _ system that contains contaminated water, but cannot be drained?

                                                                                                                      .(0. 5 )
         -* ANSWER

m. High temperatures are defined greater than (or equal to) 150 F.-(0.5) b._High pressures are defined as greater than 100 psig. (0.5)

c. Approval of the General Manager (also accept Duty General Manager).

                                                                                                                        .(0. 5 )

9 i 1i aREFERENCE PS-3-2, a) Page 1, b) Page 2 -

     . KA 194001K108/K109                3.4/3.4 L '

52

m - - a, - r

-]

i, j '.;,, r t' L-( oQUESTION- **

                .3.04'          (2.0)
        ,J      . In forder ~ to ; maintain personnel r adiation- exposures within the limits
    -           ' established,by:10CFR20, administrative restrictions are applied to the rate of dose. accumulation.

l oi What' are l Trojan 's- QUARTERLY Administrative Exposure Limits ' f or : a b -Radiation Worker'with a documented NRC-4 dose history:. I o- Whole Body, head and' trunk? ~ ( 1. 0 ) o Skin of Whole Body?. (0.25)- I o : Extremities 7: (0.25)

b. . Who must -approve any request f or whole body radiation exposures in -

                        -excess of 1.0. rem / quarter?        (0,5) eANSWER a.'     Whole Body-       (1.0 rem / quarter up to) 2.5 rem / quarter if lifetime dose is less than 5 * (N - 18).                                                                        (1.0)
                          ' Skin ' of whole body         7.5 rem / quarter          '(.25)

Extremities 18.75 rem / quarter (.25)

b. The General Mt1ager (0.5)

                 ' *' REFERENCE r

Radiation Protection Manual, Part II, Radi ol ogi cal Controisi a) page 2-2; b) page 2-3.

                     '.KA 194001K103 3.4/3.4 1

53 p - - - . ._ ._ - _ - _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -

oOUESTION: 3.05 (2.00) OM-2-1, "Use of' Operating Procedure", requires the procedures. contained-in the Plant'. Operating Manual to be followed while performing plant. activities. c. What two (2)' actions'are required when an operator believes that the

                                  . procedure he is performing cannot be accomplished as written in a safe mannor?                                                                                                                             (1.0)

b.. When are reasonable departures from existing procedures, license conditions, or-Technical Specifications acceptable? (1.0) eANSWER m' . - He shall place the system / component in a (stable and) safe condition (0.5) and inf orm the Shif t Supervisor - (0.5) .: b. When-immediate action is needed to protect the public health and safety. (1.0)

REFERENCE Operation Procedure OM-2-1, Use of Operating Procedures, A,'Page 1; B, Page 1 KA 194001A102 3.9/3.9 54

e e t ,- GROUP I SYSTEMS ** QUESTION 3.06. (1.0) MULTIPLE. CHOICE .(Select the correct answer):

                               'The: plant.is operating at full. power with all systems' operable'and the Rod Control System (RCS) is in Automatic.

Which fault will:cause a Rod Control System Logic Cabinet. Urgent failure? I

                                           'a. Loss of one 120 VAC power source to the Logic Cabinet.

I

b. Failure of the Pulser.

c.' Failure of.the Power Cabinet. d.' Failure:of one of the six DC Logic-Cabinet power supplies. ANSWER B-REFERENCE Lesson-Plan 02-B-09-SD (Objective 1.2.2.G.8)

                                      'KA      OO1000K403                                                                 3.5/3.8 O

m M 55

b - ,

  ;                    l       .,

s k

                           'QUESTI'ON                                            GROUP I S STEMS 3.07' (0.75) p
                                               ~
                          .MUL.TIPLE CHOICE                    (Select'the correct answer)'

I.~ sThe. plant is operating at full power.with all systems operable and the Rod . Control: System-(RCS) is.in Automatic.

       ~
                                  .What MODES of Control Rod motion are inhabited by.a Logic Cabinet-Urgent-Failure?
                                  .a.,    .

Automatic mode ofirod motion only..

b. Automatic and Manual-modes of rod motion only.

c.eIndividual Bank mode of rod' motion only.-

d. All- modes of rod motion.

                         ' ANSWER B

i. REFERENCE. Lesson Plan 02-B-09-SD (Objective 1.2.2.G.8) KA OO1000K403 3.5/3.8 1

     'e i..

i, I l M

I i 56 I

h:. . b' . QUESTION- ** [L.

              -3.08             ( 0. 75 ) .
        ,   nMULTIPLE. CHOICE -(Select the correct answer) 7 Thelp1~ ant is operating at full power with all systems' operable and the
            . Rod Control .' System - (RCS) is in Automatic.                                                       The RCS' Loop.3' cold leg-temperature instrument fails high.
   ; n-      'What will-be the' demanded control' rod motien7 a.. Outward rod motion.at 72 steps / minute.

b. ' Outward rod motion at 48 steps / minute.

c. Inward rod motion.at 48 steps / minute.

d..Inward-rod motion.at 72 steps / minute. ANSWER d. .r

             ' REFERENCE-Lesson. Plan 02-B-09-SD (Objective 1.2.2.H)

KA 'OO1000K403 3.5/3.8-

57

it

             ' _I f' l QUESTION-             ..
                    ?. 3. 09.       ( 0. 75)'

MULTIPLE' CHOICE.-(Select.the correct answer) LThe plant is'operatin'g at. full power and all systems.'are operable. The Chemical.L and Volume Control System (CVCS) is aligned f or nor. mal makeup

                      'and' letdown.

Why:is the Volume Control Tank'. isolated on a Safety' Injection si gnal '??

a. To prevent.the charging pumps from becoming gas bound and cavitating.

                           .b. To maximize the boron concentration of the injected water..

c..To prevent pump cavitation by increasing the net positive suction

head.

To maintain the water volume of the VCT in reserve.for makeup

d. during-plant cooldown. ANSWER A s. REFERENCE Trojan EQB'.02-A-06'and Lesson Plan 02-A-06-SD (Objective B.2.h.4).

                       . MA       OO4000K407 3.O/3.3 l'.

O [ 58 l

QUESTION l 3.10 (0.75) MULTIPLE CHOICE (Select the correct answer) -The plant is operating at full power and all systems are operable. The Chemical and Volume Control System (CVCS) is aligned for normal makeup and letdown. Which conditions must be met to satisfy the interlocks for opening a letdown isolation valve?

a. Pressurizer level must be less than 17%.

b. At least one orifice isolation valve must be open.

c. At least one charging pump must be running.

d. All orifice isolation valves must be shut.

ANSWER D REFERENCE Trojan EQB O2-A-06 and Lesson Plan 02-A-06-SD (Objective 1.2.2.E.3). KA OO4000K405 3.3/3.2 i i l 59

                                                           . _ _ _ _ _ _ _ ---____-_________-_a

R f ,c A.,

     .f i

u' .) LGOESTIONL ** 33.11:- '(0.75) . 1. E 'MULTIPLELCHOICE~ (Select the correct-answer)

,The plant'is; operating at'. full power and all systems are operable. The Engineeredf Saf ety Features Actuation System -(ESFAS)' is in its normal .

lineup. What'is the power.. supply to ESFAS Train'A?' a.s 4.16 KV ESF Bus'A1. Ib. 120 VAC; Preferred Instrument Bus Y11. c.:480 VAC MCC B-25.

                ~ d..125 VDC Bus D-30.

ANSWER B I LREFERENCE Lesson P1an.02-C-06-SD page.24. KA013OOOK201-3.6/3.8 J 60 -=__:-___:-_____-_--___-_-_-__-______-_________--.____-________

QUESTION **

  .3.12           (1.0)

MULTIPLE CHOICE (Select the correct answer) The plant was operating at full power with all systems operable when the Engineered Safety Features Actuation System (ESFAS) was inadvertently actuated. AFTER ESF has been reset Which function must be manually reset?

a. Containment Ventilation isolation signal.

b. Centrifugal Charging Pump (CVCS) start signal.

c. Safety Injection Pump start signal,

d. Containment Air Coolers start signal.

ANSWER A REFERENCE Lesson Plan 02-B-02-SD pages 21 & 22 (Objective 1.2.2.E.3). KA 013000K401 3.9/4.3

  -t SIDE 61

       ,L
  , r5 Li QUESTION'

. 3.13 (0.'75)

MULTIPLE CHOICE (Select the correct: answer)

          'The. plant;is operating at full power and allEsystems are operable.
          '. Maintenance is ~in progress' on the Engineered Saf ety Features Actuation System L(ESFAS) when all bistables f or HI-HI. containment pressure are inadvertently actuated.

How will.the. Containment Spray System (CSS); respond?

a. . CSS will be actuated and spray the containment.

E b.': CSS pumps will start, but CS isolation valves will stay shut. i- ! -c;' CSS isolation valves'will open, but CS pumps will not start.

d. ~ CSS will not be' effected by.this incident.

ANSWER-C REFERENCE Lesson' Plan 02-B-02-SD page 24 (Objective 1.2.2.C.2.g).

             .KA   013OOOK105 4.1/4.4 1

1 62 ,

                                                ,                                       i

t - t, [Q'UESTION ** [. [- 13.'14* (0.75) b MULTIPLE, CHOICE' -(Select the correct' answer) Thelp1' ant.is operating at full power and all systems are operable. The Auxiliary Feedwater System ( AFW)~ . is i n its normal lineup. What is.the power supply to the AFW Turbine Trip and Throttle Valve? a.'-4.16 KV ESF~ Bus A2.

b. 120 VAC Preferred Instrument Bus Y22.

c. 480 VAC ESF Bus B-23.

d.^.125 VDC Bus D-10. ANSWER D REFERENCE Lesson Plan'02-A-12-SD page 37 (Objective 1.2.2.B.1.j). KA 061000K201 3.2/3.3 I 63

'^

Dz -. , .; gya s

                                                                                                                                                     ..u
                                                                                                                                                      .j d,
             , .                                                                                                                                          1 I

m )1 lOUESTIOct: ** .' m. e

             +:! 3.15 - (0. 75)'-                                                                                                                      .j
                   . MULTIPLE CHDICE       (Belect the correct ~ answer)

+ 1The plant isioperating at: full. power and..all systems are' operable.. The. l Aukili ary Feedwater System '-( AFW) :i s Din its . normal-- lineup. When does the' Diesel Driven AFW pump-trip on low CST level?

a. 35%'

                           .b.:30%.

N . c'. ; 9% r ... dI 60%' ANSWER

                           'A
                  . REFERENCE Lesson Pl an 02-A-12-ED page 25. '(Objecti ve 1.2.2.C.3).

KA- 061000K401 3.9/4.2 l

l M

l 64

C-QUESTION- ** 3.16.:(0.75)

                                                     ~
                               ' MULTIPLE. CHOICE   (Select the correct answer)

To assure" adequate decay heat removal and RCS heat transfer,. Red Path

                               - octions are required in some cases when - AFW flow is less than a given
                               - value. .The E-O (Reactor Trip, Safety. Injection, and Diagnosis) Red Path Summary f or Heat Sink uses less than 495 gpm AFW flow, but E-1 (Loss-of Reactor or-Secondary Coolant) Red Path Summary uses less than 720 gpm.

Why is the: Red Path value f or AFW flow required by E-1 more than E-O?

a. E-1 assumes an adverse environment flow measurement and allows for larger flow errors.

b. .More decay heat would be generated under the conditions when E-1 would be used.

c. E-1 allows for allowable AFW pump speed v'ariations and flow measurement errors.

d.,More AFW flow is required to prevent water hammer in the SG feed rings under Loss of Coolant conditions. ANSWER A

                                . REFERENCE Lesson Pl an 02-A-12-SD page 60 (02-K-13-LP Objective 1.2.2.2).
                                  -KA 061000K501/502 3.2/3.6 q                      i 65

L-q OUESTION 3.17 (1.0) . MULTIPLE CHOICE (Select the correct answer) ; Reactor shutdown is in progress with Intermediate Range power level at 1 x 10 E-8. The Control Operator accidentally actuates one Source Range Reset Switch. What are the consequences of this operator error?

a. The Source Range channel energizes early and excessive anode burnup occurs until the operator drives rods in far enough to energize the other channel,

b. The Source Range channel will energize when reactor power is lowered below P-6 and may cause a reactor trip if power is greater than 1 X 10 E6 cps.

c. The Source Range channel will energize and fail high. The reactor must be manually tripped to reduce further damage.

d. The Source Range channel will energize and will trip the reactor if indicated power is greater than 1 X 10 ES cps.

ANSWER D REFERENCE Lesson Plan 02-B-07-SD, page 46 (Objective 1.2.2.B.2). KA 015000K604 3.1/3.2 66 _ _ _ _ . - . ._ _ _ _ __ - _ L

[ b' ;

    . QUESTION                                                                                                   **

3.18. ( 1. O) MULTIPLE CHOICE (Select the correct answer) The plant is' operating at rated thermal power.with rod control.in automatic when one power range nuclear. instrument (N43) fails high. What is the resulting control rod motion without operator action?

          -a. Control rods drive in until the Turbine First Stage Pressure is less than'15% power, then stop because C-5 will not allow further insertion.
         -b. Control rods drive in until the Over Temperature Delta T (C3) stops rod motion within,3% of the trip setpoint.

c. Control rods drive in ' until Tref - Tave compensates for the f ailed instrument (rate input only), then out to restore Tave.

d. Control rods drive in until temperature mismatch balances power (rate only)' mismatch, then stop at the lowest position reached.

ANSWER D REFERENCES Lesson Plan 02-B-09-SD (Objective 1.2.2.H). KA. 015000K302 :3.3/3.5

67

(- N. p i. i

     .. QUESTION
     .3;19 !(1.0)

MULTIPLE CHDICE (Select the correct answer) LThe initia1 post-refueling plant startup is in progress with reactor-power at 35%.- Power Range Nuclear Instruments N41 and N44 fail low. (Before immediate Actions are initiated, channel N43 also fails low

       -(common mode detector failure).

's What.is the effect on the Reactor Protection System (RPS) without operator' action?

a. RPS is inoperable and cannot initiate any protective f unctions without two channels operable or one channel manually tripped.

be RPS-high flux trips (108%) are inoperable, but all of the Automatic and Manual mode rod stops (C1 through C4) will actuate if needed.

c. RPS cannot initiate trip functions with the power range

                  . instruments, but intermediate range instruments will' initiate a reactor trip.

d. RPS will initiate a reactor trip because the General Warning self check fails anytime both RPS trains cannot trip the reactor.

L ANSWER

      -      C REFERENCES Lesson Pl an 02-B-03-SD, page.28 (Objective 1.2.2.A.2).

KA 015000K301.3.9/4.3 l 68

QUESTION. L3.20: (0. 75)' MULTIPLE CHOICE- (Select the correct answer). The. plant is operating at full power and all systems are operable. The

           . Auxiliary-Feedwater (AFW)'and Main Feedwater (MFW) Systems are'in their.

normal system lineups.- Where does. the AFW 'f eed line penetrate the Main Feedwater System?

a. Between the'MFW Isolation Valves and the MFW flow venturi.

!b.'Between the MFW flow venturi and the containment.

c. Between_the containment and the Steam Generator.

d. Between the MFW seismic category I. check valve and MFW Isolation Valves..

           ' ANSWER
                -B-                                                                                                                                                      !
          REFERENCE

, Drawing M-213, sheet 2; Lesson Plan 02-A-12-SD (Objective 1.2.2.B.2). KA '059000K102 3.4/3.4 aum 69

l

 *I i QUESTION.                                                                                                                                                                                               **

E 3. 21 - .(0.75)

                .MOLTIPLE CHOICE                         (Select the correct answer)

The p'lant:is operating'.at full power and'all systems are operable. The Auxiliary Feedwater-(AFW) and Main Feedwater (MFW)' Systeins are in . their normal. system. lineups. .

                ;Which MFW parameter is used by the Steam Generator Water Level-Control
                 , System to control MFP speed?

a. Main' Steam line' Pressure..

b. MFW Flow Control Valve position.

                      .c.-       MFW flow,

d. Steam Generator water. level.

ANSWER A REFERENCE Lesson Plan' 02-B-05-SD, Figure:B-(Objective 1.2.2.B.2.k). KA 059000K104 3.4/3.4-m O 4 70

QUESTION 3.22 (0.75) MULTIPLE CHOICE (Select the correct answer) All The~ plant is operating at f ull power and all systems are operable. Reactor Coolant-Pumps (RCP) are in their normal system lineup. What is the power supply to the "C" RCP?

a. 12.47 KV bus H1.

b. 230 KV bus V 81.

c. 4.16 KV bus A5.

d .' MCC B25 ANSWER A REFERENCE Lesson Plan 02-A-02-SD,.page 17 (Objective 1.2.2.D.1). KA OO3OOOK201 3.1/3.1 OO3OOOK302 3.5/3.8 M l 71 l ___ _ _ - - - __ - _-___ - _ _ _ _ _ _ -

p-1! iOUESTION' **

                  -3;23              (0.75)

F. MULTIPLE CHOICE (Select the correct answer) E The. plant'is operating at full power'and all systems are operable. ' All Reactor Coolant Pumps (RCP)'are in their normal system lineup.

Which signal"will cause an RCP ~ trip during normal full power operations? n:

                         'a . Loss of power from the Start Up Transformer due to a phase to phase' fault within the transformer.

b..TheLLOCAL PANEL SELECTOR SWITCH-is placed in LOCAL /RUN while the REMOTE CONTROL SWITCH is left in RUN. c.' Steady state Under Frequency occurs on a RCP power supply bus due F to an-instrument malfunction.

d. Steady. state Under Voltage occurs on'a RCP. power supply. bus and the Unit Auxiliary Transformer breaker is open.

ANSWER D REFERENCE Lesson Plan 02-A-02-SD, page 17 (Objective 1.2.2.F.5). KA 'OO3OOOK201 3.1/3.1 OO3OOOK302 3.5/3.B 4 M 7 72

       .s i_________________________________. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ . _ _ _ _ _ . . _ _    _ _ _ _ _

's 1 DUESTION

          '3.24        (0.75)
         ' MULTIPLE CHDICE       (Select the correct answer)
         . While'the' reactor-is operating at'30% power, the "B" RCP trips on overcurrent. The plant continues to operate at about 30% power and ' no
         ,. operator' action is taken.

Why' does the ' initial rapid change in the "B" Steam Generator occur?

a. Level increases because feed flow is greater than steam flow.

S. Level' increases because of swell.

c. Level decreases-because of shrink.

             ' 'd'. Level decreases because steam flow is greater than f eed flow.'.

ANSWER

            '.C REFERENCE Lesson P1an'02-A-02-SD (Objective 1.2.2.H.2).

KA OO3OOOK201- 3.1/3.1 OO3OOOK302. 3.5/3.O 1 M S 73

i l OUESTION 4 3.25 (0.75) MULTIPLE CHOICE (Select the correct answer) 3 Hydrogen Recombination Units X-318 A and B are designed to reduce post ] LOCA hydrogen levels in Containment, if needed. J

How is the hydrogen recombined in these units? i

a. Containment air is heated to about 600 degrees F and passed through a catalytic filter which causes hydrogen / oxygen recombination.

b. Containment air is filtered and sont into an Ignition Chamber that uses igniters to recombine (burn) any hydrogen / oxygen present.

c .' Containment air is heated to about 1200 degrees F which is hot enough to recombine any hydrogen / oxygen present.

d. Containment air is filtered and heated to about 3,000 degrees F which causes hydrogen to disassociate and combine with oxygen to form water.

ANSWER C REFERENCE Lesson Plans 02-F-01-SD page 25 (Objective 1): 02-A-11-SD figures 2 & 3. KA .022OOOGOO4 3.1/3.3 022OOOG010 3.2/3.4 l 74 ; i

i: .

-)

i l

                                                                                                                                                                 'i QUESTION.

3.26 .(0.75) ) MULTIPLE CHDICE (Select the correct answer) Safety GuideL7: establishes a maximum allowable hydrogen concentration limit for the Containment to' prevent' uncontrolled hydrogen burning. This limit'is also'used by..FR-C.1 Response to Inadequate' Core Cooling to prevent operation,of the hydrogen,recombiners.- What is the maximum Containment' hydrogen concentration in which the Hy'drogen Recombiners can be~used? E a. 4% b,.8%

c. 12%

d.-20% l ANSWER'

                        .A REFERENCE Lesson Plan 02-F-01-SD page 23 (Objective 1.2.2.F).

KA 022OOOGOO4 3.1/3.3 022OOOG010 3.2/3.4 i I 75

OUEST10N 3.27 (0.75) MULTIPLE CHOICE (Select the correct answer) The plant has just tripped f rom f ull power because of.a large LOCA. Safety, Injection and Containment Isolation Phases A and B have actuated. All systems are operable. What changes occur in the Containment Air Cooler System (CACS)? o .' CACS f ans in NORMAL are started by the DBA sequencer, and the cooling water supply increases,

b. Phase B CIS trips the CACS fans and isolates the cooling water supply.

c. CACS fans in NORMAL or STOP are automatically started, and cooling water flow increases as additional cooling pumps are started.

d. -CACS fans in NORMAL are automatically started, and Phase A CIS isolates CACS cooling water.

ANSWER A REFERENCE Lesson Plan 02-F-01-SD page 22 (Objective 3). KA 022OOOGOO7 3.3/3.5 l s 76

                                                           - - _ _ _ _ __-_____-_-._-.___-__D

b l l GROUP 11 PLANT SYSTEMS QUESTION 3.29 (0.75) MULTIPLE CHOICE (Select the correct answer) The plant is operating at full power and all systems are operable and are in thier normal system lineups. Anich of the following will cause the Overpower Delta T Calculator to !

           - reduce its setpoint7

a. Tave above rated.

b.- Primary pressure below 2235 psig.

c. Rate of change of Tave in a decreasing direction.

d. Del ta flu >: exceeding the deadband.

ANSWER A REFERENCE Reactor Protection System Description 02-B-03-SD, Page 34 (Objective 1.2.2.A.1). KA 012OOOK611 2.9/2.9 P O G 77 _ __ ____________j

n . TQUESTION **- 3.29~ (2.O)

                          ' Appendix'A to ES-0.1',. Reactor Trip Response, . provides six (6) plant conditions that are used to verify' natural circulation cooling in the Reactor Coolant-System'(RCS).. One condition is S/G pressures - STABLE.OR

DECREASING.

                         ' What:- are f our - (4) of the' remaining five.(5) plant conditions?

ANSWER (0.5 each for any fourg 2.0, maximum)

                                  .RCS subcooling . > 25 degrees F.

RCS hot leg temperatures stable or. decreasing. Core exit thermocouple stable or decreasing. RCS cold leg temperatures near saturation temperature for SG' pressure. RCS delta T < 66 degrees F'and stable. REFERENCE. ES-0.1, Appendix A (O2-K-12-LP Dbjective 1.2.2.D.21). KA, OO2020K510 3.5/3.9 78

N .,I.

        .g
                  )

QUESTION 3.30 .(2. 0) [--

                    /What are the -Technical- Specification'. leakage limitations f or the -

ifo11owing. types of'RCS leakage?: : (0.5'each)'

F if ~a') Identified b). Unidentified-c) Controlled. (all '~ RCPs)-

                                  ~
                    ' d)I'Controlledf(one RCP) .

J. -

i -; ANSWER' -(0.5 each)

a. 10'gpm.

b. 1.gpm.

c. 20 gpm'from'all RCPs ld. 6.gpm from a' single RCP REFERENCE Reactor Coolant System Description, 02-A-01-SD, page 46 (Objective 1.2.2.A.13.f)-

                        'KA OO2020K401                                                                         3.6/3.0 79
                   =_-               _ _ - - _ _ _ - - _ _ - - - _ - _ _ - - _ - - _ _ _ _ - _ - - . _ . - _ .

QUESTION-3.31 (1.50) The Emergency Core Cooling System (ECCS) is designed to provide core cooling and additional shutdown capability during four (4) types of accidents. What are three (3) of these types of accidents? ANSWER (Any 3 at 0.50 each )

1. Breaks in RCS (up to and including the design basis accident).

2. Rupture of CRDM (pressure housing and RCCA Ejection).

3. Breaks in Main Steam System (up to and including a double-ended guillotine sheer of largest pipe).

4. Rupture of S/G Tube.

REFERENCE Emergency Core Cooling System Description, 02-A-07-LP, Page 7 of 62. (Objective B.2.a) KA OO6000 GOO 4 3.5/3.8 80

         ;         L        .
                  -QUESTION 3.32      -(2.5)

Trojan Technical ~ Specifications Limiting Condition for Operation 3.5.5 r , limits'the volumefand boron concentrations in the Refueling Water. Storage. Tank (RWST). a (1;O each) 7 A) Whatiare the?two (2)Lbases for these limits? B) What11s the minimum temperature for operability of the RWST7- (0.5)

                  ' ANSWER      (2.5)

A) To ensure that. sufficient water is availab*le (within containment)~ to permit recirculation cooling flow to the: core af ter a LOCA. (1.0) To ensure'that the reactor will remain subcritical following a LOCA (in the -cold condition af ter mixing of the RWST and RCS water volumes; and with all control rods inserted except for H the most reactive control assembly). (1.0) B). 37 degrees F (0.5) J RECERENCE Trojan 's Technical Specifications 3/4.5.5 RWST page B 3/4 5-2 (02-A-07-LP Objective B.2.m.9). KA'OO6000 GOO 6 2.9/4.0 l t b > - _@8. . - - . _ - _ _ _ _ - _ _ _ _ _ . _ _ - _ _ _ _ _ _ _

QUESTION 3.33 (2.25) The pressurizer Pressure and Level Control System is designed to accommodate three design transients without causing a reactor trip. What are these three (3) transients (include magnitudes /setpoints)? i (0.75 each) ANSWER (0.75 each) Loading or unloading at a rate of 5%/ min. (with automatic rod control). Instantaneous load changes of +/- 10% (with automatic rod control). Step load reduction of 50% (with automatic rod control and steam dumps). I i i l REFERENCE Lesson Pl an 02-B-04-SD, Page 3 (Objective 1.2.2.A). KA 010000 GOO 4 3.1/3.3 l l l I

OUESTION 3.34 (2.00)

                          'Four (4) of the six (6) control rod block signals block BOTH Manual cnd Automatic modes of rod withdrawal.

What are these f our (4) control rod block signals? (0.25 each setpoint, 0.25 each signal) ANSWER (0.25 for each signal; O.25 for each setpoint)

1. Intermediate Range HI Flux at 20% Current Equivalent

2. Power Range Hi Flux at 103%

3. Overtemperature Delta-T at 3% less than Trip setpoint

4. Overpower Delta-T at 3% less than Trip setpoint REFERENCE Lesson Pl an 02-B-03-SD, pages 54-55 (02-B-09-LP Objective 1.2.2.E.2).

KA Ot2OOOK604 3.3/3.6 83 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ l

y , f , 1

                  .GROUPJIII PLANT. SYSTEMS
     . j:
                   ' QUESTION

3. 35, ' (1.5)

                  ' The Residual Heat Removal . (RHR) system suction - i sol ation valves, (MO-8701, 7-MO-8702)'have two interlocks to' prevent overpressurization of the RHR L                      cystem.
                     'What' are th'ese two (2) interlocks (include setpoints)?                            (0.75 each) p

ANSWER The valves cannot be opened when RCS pressure (0.5) is: greater ~than 425 psig (0.25).

The valves.will go shut' automatically if pressure increases (0.5)

         '              to 600 psig.(0.25).

l-REFERENCE Lesson Plan 02-A-08-SD, Residual Heat Removal System, Page 6 of 28 (Objective 1.2.2.E.5).

                         'KA OO5000K407 3.2/3.5 1

84 j ___ _________m_________________ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

J ' QUESTION

        .3.36'    (1.00)'

MULTIPLE CHOICE -(Select-the correct answer) Precautions in_ Operating Instruction 01-4-1, Residual Heat Removal (RHR), prohibit personnel entry or. presence in the RHR' rooms when an Reactor Coolant' Pump is1 started, l What 'is :the reason f or this precaution?

a. The over pressure surge when a RCP is started could possibly rupture the RHR piping and cause personnel injury and contamination'.

b. FlowLincrease in thefRCS may starve the RHR suction and cause major

             . pump damage. with ' potential - personnel safety concerns.

c. Radiation levels may increase af ter the RCP is started because radioactive crud may be swept into the RHR system from the'RCS.

d. RHR' Safety Relief Valves may' lift, returning radioactive water

to the room floor drains and. increase personnel exposure.

ANSWER C-

        . REFERENCE-Operating Instruction 01-4-1, Residual Heat Removal, Page 3 of 7 (O2-A-08-LP Objective 1.2.2.I).

KA OO5000G010 3.3/3.5 9 mm 85

t

1:

QUESTION 3.37.. (2.0).

          = What ' are the 'f our' .(4). automatic actions that occur in the Component -

' ' A-Cooling Water (CCW) System during a' Safety Injection Signal? (0.5 each)

           '[DD'NOT-LIST the: actions caused by Containment Isolation Signal!]
          ' ANSWER                         (0.5 each )

a. . Auto' start of-the stand'by CCW train and associated Service-Water system train (by the DBA. sequencer).

ib'. Separation of'.(the Seismic Category I) trains and isolation of the Seismic Category II sections.. ic. Flow control valves f or the Containment Air coolers

                                                               ~
                      . return header open to increase. flow.

d. Surge tank nitrogen supply and vent valves close.

4 REFERENCE

              . Lesson Plan 02-A-11-SD, pages 9 and 23 (Objective 1.2.2.E.4).

KA OOOOOOA304 3.6/3.7 M _ _ . _ _ . _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _

l QUESTION 3.38 (1.5) What are the si>t (6) Engineered Safeguards related components cooled by the Seismic Category I Component Cooling Water System loop? (0.25 each) ANSWER (0.25 each) Positive Displacen ent Pump. Safety Injection Pumps. Containment Spray Pumps. Residual Heat Removal Pumps. Residual Heat Removal Heat Exchangers. Containment Air Coolers. REFERENCE Lesson Plan 02-A-11-SD, page 5 (Objective 1.2.2.G). KA OO9000 GOO 7 3.3/3.4 END OF CATEGDRY THREE END OF WRITTEN EXAM 87 _ _ - - _ . ____-_________-___w i

-4 .

                            . .~                    ..

4 t PORTLAND GENERAL ELECTRIC COMPANY ()P D A TED l

                                                                                       .                '                                  N   N TROJAN NUCLEAR PLANT                     BY COPY HOLDER j

April 19,1988 Revision 3 QUAllTY RELATED EMERGENCY CONTINGENCY ACTION - ECA-0.0 LOSS OF ALL AC POWER APPROVED BY m t De_ DATE b h Pa?> I_ . A. PURPOSE

                                                                            . This procedure provides actions to respond to a loss of all AC power.

B. ENTRY CONDITIONS

1. Indication that all main and emergency AC busses are de-energized.

                           ,                                                 2. This procedura is entered front EI-0, REACTOR TRIP, SAFETY INJECTION AND DIAGNOSIS, Step 4. on the indication that all AC emergency busses are de-. energized.

T0001/1R ECA-0.0 Page 1 of 21 Revision 3 6m _ _ . _ _ _ . _ - _ _ _ _ _ _ . _ - . . _ _ . _ - . _ _ _ _ -

                                                                                        .o-a
                        .~

t

                                                                                 -          LOSS OF ALL AC POWER
                                  -.j                                                                                                                                                                              -

Step Action / Expected Response Response Not Obtained

                                                                                                                                                                                      ~.             . . .

CAUTION: A loss of all AC power requires initiation of

                                        ~                                                             the RERP. Carry cidt~                                                                    ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ' ~~

J

                                                                                                 ,    RERP actions in parallel with this procedure l

NOTE: Steps 1-4 are IMMEDIATE ACTION steps. NOTE: CSF status trees should be monitored for informa-tion only. FRs should NOT be implemented. I NOTE: Many chart recorders will be de-energized during a loss of all AC power. Use associated meters if

                                                                             ,                     recorders are not available.

1' 1 Verify Reactor Trio: l a. Reactor trip and bypass a. Manually trip reactor. breakers - OPEN

b. Neutron flux - DECREASING b. Continue with procedure.

WHEN a source of emergency power is restored and a CCP is running. THEN initiate emergency boration. 2 Verify Turbine /Geherator Trio:

a. All stop valves - CLOSED a. Manually trip turbine.

(verify CLOSED locally)

b. Generator output breakers - b. Manually open output OPEN breakers 30 seconds AFTER turbine trip.

c. Generator exciter field c. Manually open.

breaker.- OPEN 1 .. l .. ECA-0.0 Page 2 of 21 Revision 3

I LOSS OF ALL AC POWER Step - Action / Expected Ret, mse Response Not Obtained

                    '3     Check if RCS is Tsolated:                                                          ~            "'                               ~

a. Pressurizer PORVs - CLOSED a. IF pressurizer pressure

                                                                                                        < 2.335 psig. THEN manually close PORVs.

b. Letdown isolation valves: b. Manually close letdown c9ntainment isolation

LCV-459 - CLOSED CV-8152.
LCV-460 - CLOSED

c. Excess letdown isolation c. Manually close valves.

valves:

CV-81$3 - CLOSED
CV-8154 - CLOSED

d. RCS and pressurizer sample d. Manually close sample isolation valves CLOSED: containment isolations, CV-5655 CV-5657, and

MO-5653 - CLOSED CV-5659.
MO-5654 - CLOSED

                                = MO-5656 - CLOSED 4
                                + MO-5658 - CLOSED 4       Verify AFW Flow - > 720 gpm                                  Perform the following:
                        ~

a. Stop the diesel AFW pump if

Verify turbine-driven AFW running pump running. IF NOT, THEN manually open steam supply valves,

b. Line up emergency cooling to the diesel AFW pump per

Verify proper alignment of 01-8-2 AFW valves. IF NOT, THEN manually align valves as

' c. Restart the diesel AFW pump necessary. if necessary i i ( ECA-0.0 Page 3 of 21 Revision 3 L __ __ - --

                                                                                                                                                     .__--------_~.------3 t                                                                                                                                                                      I I
         .                                         LOSS OF ALL AC POWER                                                                                                     '

Step - Action / Expected Response Response Not Obtained

                   ~
                  ~5-   Try to Restore Power-to Any AC            - - - -

Emeraency Bus

a. Restore power to A1/A2 from the EDCs:

1) Start the EDGs 1) Start the EDG locally, a) Place local / remote switch in LOCAL.

b) Ensure all lockouts are reset, c) Start EDG. IF EDG

does NOT start THEN "C,

j refer to ONI-4rf R' O'#*" h .htie.; T: " :ter:

                                                                  %It+                   *:nr te ;;DG, and                                                       g, p' continue with step 6 of this procedure.
              ,               2) Bring EDG to rated speed and voltage t                 3) Synchronize the EDG to                   3) Locally close A108/A208.

A1/A2 and close A108/A208 l l CAUTION: Cooling to the EDCs must be es-

                                    ,            tablished within 3 minutes after l            they are started.
                ~

b. Verify automatic loading on b. Manually load the EDCs by the S/D sequencer: manually closing the fol -

lowing A1/A2 load breakers.

CCPs - START + B01/B02 breakers.
CCW Pumps - START = B03/B04 breakers.
SW Pump - START; verify
SW pump breakers (ones SWBP starts when SW pump lined up for service).

starts ECA-0.0 Page 4 of 21 Revision 3 l l 1

ya ..

$b; :...

LOSS OF ALL AC POWER Step Action / Expected Responge 'tesponse Not Obtained 5 -

SWBP breakers.

t

CCP breakers.

CCW pump breakers (ones lined up for service).

c. Check A1/A2 buses - AT LEAST c.

ONE ENERCIZED IF the EDG cannot '>g started. .b. i 4,g,ggNgr to ONI-jrr,3:21Ie... t; ^_ _ _cre "M. and continue with Step 6 of this i .. procedure. %

p.

d. Return to procedure and step in effect

ECA-0.0 Page 5 of 21 Revision t

i CriticalS2fbyFunctionStatp2 Trces (CSFST) FIGURE FR-0.1 SUMRITICALITY Red GO TO ! FR-5.1 POWER NO R 4

        <ANGE 5%                                                            orange YES GO TO I

FR-S.1 Yellow GO TO FR-S.2 IR SUR NO ZERO OR NEGATIVE IR SUR NO YES MORE lECATIVE THAN -0.2 YES DPM l Green O CSF Sat

 .                                            NO SR ENERGIZED YES Yellow
                                                                             },

Go to IR-S.2 SR SUR NO ZERO OR NEGATIVE YES Green CSF Sat FR-0.1 FR-0 Page 5 of 12 Page 1 of 1 Revision 4

Critien1 S"f y Functicn Status Trm s (CSFST) FIGURE FR-0.2 CORE COOLING Red O GO TO FR-C.1 CORE. . _ . NO Red EKIT

                   "+   T/cs                                                                                                  GO TO
                        < 1200*F                                                                                              FR-C.1 YES RVLIS     NO FULL RANGE
                                                                         > 39%     YES Crange GO TO FR-C.2 CORE      NO EXIT T/Cs
                                                        < 700*F   YES                                   Orange GO TO FR-C.2 AT LEAST NO                      RYLIS     NO ONE RCP                          FULL RUNNING                          RANGE
                                     ~

YES > 39% YES Yellow CO 'TO TJ,V FR-C.3

~~

Orange SUB-

               ,                         NO                                                                                  GO TO COOLING FR-C.2
                               > 25*F (75*F)    YES RVLIS DYNAMIC NO HEAD RANGE
                                                                          > 44% 4 RCPs
                                                                          > 30% 3 RCPs
                                                                          > 20% 2 RCPs
                                                                          > 13% 1 RCP   YES Yellow l                    ,j            GO TO
                                                                                                 ._        ._,'             FR-C.3 Green O

CSF SAT FR-0.2 Page 1 of 1 FR-0 Page 6 of 12 Revision 4

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

1 1 1

                                      'ri:1ce.1 Shfe:v Fune: ion Sta:as Trees '05757',

t FIGURE FR-0.3 HEAT SINK Red eGOTO FR-H.1 TOTAL FEED FLOW TO N S/Cs

                                    >495 (720) spa      D AVAILABLE
            .      . .                                                                                                                                Yellow GO TO FR-B.2 NR LEVEL                                  PRESSURE IN ONE        #
      - 4 S/G IN ALL S/Cs
              > 5%

YES < 1230 PSIC (12%) TES Yellow

                                                                                                                                                           , CO TO FR-H.3 e

c WR LEVEL IN ALL

                                                                              ,,3                                                                                     }gg l
 '                                                               SICS,                                                                                                ; Ejg'
                                                                 *M           TES 7D*/e                                                                              Yellow GO TO FR-H.4 PRESSURE                    g IN ALL 3/Cs
                                                                         < 1170 PSIC                gg Yellow GO TO TR-H.5 Et LEVEL IN ALL                                                    14 0 S/C
                                                                                    > 5%

(12%) YES H Green O CSF SAT rR-o.3 Page 1 of 1 FR-0 Page 7 of 12

   's Revision 4

Critien1 Safety Function Status Trees (CSFST)

                                              \                   .

FIGURE FR-0.4 INTEGRITY 8 . --2560 l E lC

e 1 .

Temkra e 75* Red ALL RCS I 9GOTO FR-P.1 PRESSURE- M Orange l* COLD POINTS GO TO 10 RIGHT OF FR-P.1 I LIMIT A C00LDOW (SEE ATTACH- YES NO ALL RCS IN ALL grnr 43 TCOLD NO , TCOLD TEMPS

   --> < 100*F IN                                          > 245*F                                                                      Yellow LAST 60        YES                                             YES
                                                                                                                                           , g', GO TO MINUTES                                                                                                                                y FR-P.2 ALL RCS NO TCOLD TEMPS
                                                                      > 275*F                                     ygg (T2)

Green l l CSF I SAT Orange GO TO I FR-P.1 ALL RCS NO TCOLD TEMPS

                                                                      > 245*F                                    gg
                                      .                               (T1)

RCS I' ** NO l GO TO PRESS

                                            < 440                                                                                                                         P.2 PSIG g

Green RCS TEMP I

               > 290*F                                                                                                            -

SAT YES l Green l l CSF SAT FR-0.4 Page 1 of 1 FR-0 Page 8 of 12 Revision 4

                                                                                           \                                                                            .

FR-0.4' ATTACHMENT A TROJAN OPERATIONAL LIMITS CURVE 3000 7..... i ,

                                                                                       .,I                                                         ,                                                      ,
                                     ,; ,255Q,,,5iL';

i

               - - - - -2500                                                                                                   245'F_27F  J' 4 -..
                                                - a +._._ _. . . . _. . . . .

a c 2 7 , t j,.-. . _........ .

4. . . . .

                                                  .s. _ -
                                                     '                             ....s-...                                                                               ..
                                                                              . . 2 20 *F- - -*

2000 _.. 2050 ps:.g ; .

                                                           . -.             _ . . .      ..s j..                                                   .

n M

                                           . .                        .    ...                .. .                                              t.

m _ ___ . 4 ..- m. ;. su .. .. , w w { __ .. ...

p. _. .

1500 . . . . _ . . .. ., m .

                                                                -_        ..                ..               .         . --                   i.                       . .                .        . .                                   ,
                                                  '                                                                                           g                                                     .

{.... k. a. 4... .. . . .

3. . . .. . . . .

_ _ . . . .. .. _ _ _ _ . .[. _. . .. ... ... 9....,.. _ . t , 4.. 1000 4 v- -- - 4- . . . . . _ . _ . __

                                         .t.._._                                                                                                                                             _.7 .. .

_ a. . . .

6. .

                                                                                                                                                                   ._......t......                                                     ,

_ .7 -- -_4_.... . . . . 3 .. .. _ ... .i._. y _..___.4._..._ _ . ,

                                                                                                                                                                                  ..._..}_....                                      ...

500 ~~M C ~ -" ' - ' ' ' ' ' ' ~ ~

                                                                                                                                                                                                                                   'f
                                                                                                                  .._" .'._M                                     .
                                                                                                                                                                                    . . ._d. . .. . .
                                             .,                                       ..-_.... - ..                                          -. _ . _ ~ -. . . . -                                  . . _                ....

O I

                                                                                                                                           ~' - ~ ~ ~ ' '                                               ~ ~ ~ ~ ~
                                                                                                                                                                                                                      ~~---'t~~

0 100 190 245 275 400 500 200 300 TEMPERATURE (*F) FR-0.4 Attachment FR-0 Page 9 of 12 Page 1 of 1 Revision 4 _-_________-_-----D

p

Critical Safety Function Status Trees (CSFST) - T neure rR-0.5 Coh7AINMEh7 Red

GO TO _ FR-Z.1 C0h7 NO - 4 PRESS

          < 60     YES PSIG 1

Orange GO TO FR-Z.1 C0h7 NO f PRESS

                < 30        YES PSIG 1
                                               '                                                                                              Orange
      ,/                                                                                                                                                    GO TO FR-Z.2
                                      \

Coh7 NO SUMP

        ,                 LEVEL
   ,C                 -

TES

                          <J26*

fM" Tellow g<Y, . 70 ( FR-Z.3 CONT NO RADIATION

                                  < ARM 15A/B      YES ALERT Creen O                    CSF SAT 1-0.5                                                                                                               -.

age 1 of 1 FR-0 Page 10 of 12 Revision A

                                               \                  .

Critical Safety Function Status Trees (CSFST) FIGURE FR-0.6 INVENTORY

         . . _ _ _ _ . . .                                                             Yellow GO TO FR-1.3 RVLIS INDICA *fES NO UPPER HEAD FULL

(> 94%) YES [ Yellow FR-I.1 PZR N0 f 4 LEVEL

            < 92%

YES Yello.i GO TO FR-I.2

PZR N f LEVEL

                                > 17%

i YES Yellow

                                                                                                               's GO TO FR-I.3 RVLIS INDICATES NO UPPER HEAD FULL YES

(> 94%) l Green OCSF SAT m-0.6 Page 1 of 1 FR-0 Page 11 of 12 Revision 4 a}}

ML20245G686

Text

Dear Mr.. Kirsch:

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