Text

Exam Rept 50-424/OL-86-02 on 861021-1205.Exam results:13 of 15 Reactor operator,15 of 18 Senior Reactor Operator & 1 of 1 Instructor Certification Candidates Passed
	ML20207T403
Person / Time
Site:	Vogtle
Issue date:	02/18/1987
From:	Arildsen J, Munro J; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML20207T381	List: ML20207T380; ML20207T403; ML20207T409;
References
	50-424-OL-86-02, 50-424-OL-86-2, NUDOCS 8703230534
	Download: ML20207T403 (125)
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ENCLOSURE 1

-EXAMINATION REPORT 424/0L-86-02 Facility Licensee: Georgia Power Company P. O. Box 4545 Atlanta, GA 30302 Facility Name: Vogtle Electric Generating Plant Facility Docket No.: 50-424 Written, oral, and simulator examinations were administered at the Vogtle Electric Generating Plant near Waynesboro, Georgia.

Chief Examiner:

e A. Arildsen

[M 18 FG8 er/

Date Signed Approved by: Y [Na #

,<.A0% F. Mpu,p ction Chief 1////r 7 Date Signed Summary:

Examinations on October 21, November 12-13 and 22-26, and December 1-5, 1986.

Oral examinations were administered to 23 candidates; 23 of whom passed.

Simulator examinations were administered to 27 candidates; 26 of whom passed.

Written examinations were administered to 30 candidates; 27 of whom passed.

Based on the results described above, 13 of 15 R0s passed, 15 of 18 SR0s passed, and 1 of 1 Instructor Certifications passed.

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REPORT DETAILS

1. Facility Employees Contacted:

J. Badgett, Nuclear Training Manager

P. Rushton, Plant Training and Emergency Preparedness Manager

Attended Exit Meeting

2. Examiners:

J. Arildsen C. Casto W. Heming P. Isaksen F. Jagger N. Jensen J. Munro D. Nelson R. Picker A. Vinnola

Chief Examiner

3. Examination Review Meeting At - the conclusion of the written examinations, the examiners provided Mr. Paul Rushton, with a copy of the written examination and answer key for review. The coments made by the facility reviewers are included as Enclosure 3 to this report, and the NRC Resolutions to these coments are listed below.

a. R0 Exam (1) Question: 1.02a (5.13a)

Facility Coment:

The E0P training text has additional reasons for maintaining pressurizer level.

, NRC Resolution:

Will accept as an additional correct answer: "(Provides sufficient inventory) to maintain pressurizer pressure control."

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i-2 (2) Question: 1.10a (5.03a)

Facility Comment:

Accept alternate definitions for subcritical multiplication other than mathematical formula.

NRC Resolution:

Will accept the following additional correct answers for full credit:

" M = 1- K , " or "M = 1-Kef f" , or any descriptive answer which correctly describes subcritical multiplication and includes constant or increasing neutron levels, Keff less than unity, and the requirement to have source neutron in the vicinity of the fuel. !

(3) Question: 2.01 Facility Comment:

For each room noun name given in the answer there also exists an associated room number to identify the same room (s). These room numbers are shown on P&ID listed in the references.

NRC Resolution:

Will accept the following answers as correct:

1. Low pressure in letdown lines.

2. High temperature in letdown HX room (R-A07)

3. High temperature in CVCS valve gallery (R-A08)

4. High temperature in auxiliary buildin0 piping penetration room (R-A09)

Added VEGP P&ID 1X4DB-114 to the list of references.

(4) Question: 2.03 (6.01)

Facility Comment:

VEGP has a safety grade cold shutdown design. Going to the cold shutdown condition requires use of the reactor vessel head vent for two additional reasons.

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.NRC Resolution:

Answer key has been modified to accept as an additional correct answer: ' "Provides a safety grade letdown path (for the safety

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grade cold shutdown design - SGCSD)."

Although vessel head cooling and reduction of thermal stresses are realized by utilizing the head vent, . these are a attendant benefits to the design, and will not be included as additional correct answers in the answer key.

In the future, the facility is requested to provide the most correct and complete training materials to the examiners, including any letters or other correspondence which set operational policy or further define design intent.

(5) Question: 2.17a (6.17a)

Facility Comments:

At normal operating pressure the cold overpressure mitigation system (COMS) is blocked by switches on the QMCB, panel C.

NRC Resolution:

Answer key has been modified to accept either "COMS Blocked or COMS has no effect." One-half credit for correct reference to 2185 psig PORV interlock.

This answer required change due to the inconsistency and lack of detail between different training material provided, including the figures referenced by the facility. The material provided on this

system gives no clear distinction between what constitutes part of the COMS and part of the PORV control circuit. Additionally, it was noted that material referred to the COMS in one place and COPS in another.

(6) Question: 2.17b (6.17b)

Facility Comment:

This part of the question asked how the setpoint was determined.

Only the wide range loop temperature (auctioneered low) is used to establish the setpoint. Wide range pressure is compared to the setpoint to trip bistable.

[ NRC Resolution:

l The facility comment is accepted in part. The answer is modified j to read "The auctioneered Low RCS WR Temperature inputs to j pressure programmer to produce the system setpoint."

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This answer required change due to lack of detail and consistency in training materials as in comment for 2.17a.

(7) Question: 2.20

' Facility Comment:

There are two MSIVs in series on each SG steam outlet line. If one valve fails' to shut the other MSIV will still provide isolation of the affected SG.

NRC Resolution

-The answer- has been modified to state, "The MSLIS will provide a signal to shut all SG MSIVs -or- the MSLIS will provide a signal to shut the second MSIV in the affected steamline."

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Reference:

VEGP R0-LP-21102-00 pp. 14-16, has been added. ,

(8) _ Question: 2.24 l Facility Comment:

The VEGP technical specifications provide the latest information

. on coincidences and setpoints concerning the SG hi-hi protective circuit.

NRC Resolution:

Part 4 of the answer key has been changed to reflect the most recent correct setpoint and coincidence for S/G high-high level protective circuit of 78%, 2 out of 4 coincidence. Also added VEGP T.S., Tables 3.3-3 and 3.3-4 to the list of references.

In the_ future, the facility is requested to ensure the most recent correct setpoint data is included consistently throughout all

[ training material sent to examiners, f

(9)- Question: 3.10 Facility Comment:

. The latest setpoints and coincidences for answer 2 and 6 are provided in LO-LP-27102-00 (Rod Control System)

NRC Resolution:

Parts 2 and 6 of the answer key have been changed to reflect the most recent correct setpoints and coincidence for the C-2 and C-11

interlocks as per the facility recommendation above. Also added l the following to the list of reference: VEGP LO-LP-27102-00, pp. 10 and 11.

5 In the future, the facility is requested to ensure the.most recent correct setpoint data is included consistently throughout all

. training material sent to examiners.

(10) Question: 3.12 Facility Comment:

The latest setpoints are provided in the VEGP technical specifications.

NRC Resolution:

Parts 3 and 6 of the answer key have been changed to reflect the most ;ecent correct setpoints as per facility recommendation.

Also added VEGP T.S. Tables 2.2-1 and 3.3-4 to the list of references.

In the future, the facility is requested to ensure the most recent correct setpoint data is included consistently throughout all training material sent to examiners.

(11) Question: 4.14 Facility Comment:

Procedure 19211-1, Step 2, has two RNO actions related to the turbine. Only one action is required if the turbine will not trip.

NRC Resolution:

If the turbine will not trip, alternative means must be employed to prevent an uncontrolled cooldown of the RCS due to steam flow that the turbine would require. These include:

1. Manual running back the turbine, and if the turbine will not run back then

2. Shutting the MSIVs.

Although listed separately, both are contingency actions to protect against excessive cooldown in case of failure of the turbine to trip. The answer key remains unchanged.

(12) Question: 4.20 Facility Comment:

The E0P Training Text instructs operators to follow E0P actions even if Tech Spec limits are exceeded.

_ - _ _ _ _ i

6 NRC Resolution:

The answer key has been modified to - accept the facility recommendation as an additional correct answer for the " explain" portion of this question. The following was added to the list of references: VEGP E0P Training Text 1-14, 15.

(13) Question: 4.22 Facility Comment:

Step 8 of Procedure 19013-1 requires realignment of containment spray when RWST level less than 16%. The caution note on page 4 of Procedure 19013-1 requires alignment upon receiving the RWST EMPTY LEVEL alarm. Also Tab 4.9 of the Plant Technical Data Book shows RWST EMPTY LEVEL alarm corresponding to 9% RWST level.

NRC Resolution:

Step 8 of 19013-1 requires realig'nment of containment spray when RWST level is 16% or less," (not less than 16%"). Answer key has been modified to read as follows:

"At RWST level of 16% or less, OR Upon receipt of RWST Empty Alarm (9%). (Either answer accepted for full credit.)"

Also, the following references are added to the list of references: VEGP E0P 19013-1, pp. 4, 9. VEGP Technical Data Book, Tab 4.9.

The following additional four changes were made to the RO exam as a result of final review and/or candidates questions during/ exam administration. 'The first three changes listed were publicly announced to all candidates by an exam proctor:

Question: 1.04 Changed the phrase "

...on the equation. . . " to ". . . to an equation..."

Question 1.15 Inserted the phrase "in the fuel" following the words "Self Shielding."

Question 3.02 Further identified the letdown valves in question as "LV-459 and

-460."

[

7-Question 3.03 Changed parts 1 and 2 of answer to NOT require specific setpoints ;

for P-8 and P-9 permissive, for candidate .to obtain full . credit.

-(Changed the specific setpoints to the phrase "...a preset value of...").

The following changes were made to the R0 answer ~ key during exam

~ grading:

1.07a Changed to accept as an additional correct answer for the second part: "Due to decrease in core delta -P

-[0.25]"

2.06 Included.the following point distribution: .

"[3 @ 0.5 each]"

In addition to change- incorporate as a result of 2.17a 4 facility comments, also changed answer key to allow half-credit if candidate correctly describes operation of the 2185 psig PORV interlock.

b. SR0 Exam (1). Question: 5.03a Same as R0 comment for 1.10a.

(2) Question: 5.13a i

Same as R0 comment for 1.02a.

(3) Question: 6.01

. Same as R0 comment for 2.03.

(4) Question: 6.05 Facility Comments:

, The ESF sequencer lesson plan provides a more detail discussion of sequencer operations. The only time e " load shed" occurs is on an i under voltage condition. A load shed will remove all loads. If an SI occurs after under voltage sequencing is compTete then SI loads will be sequenced on and an ACCW pump will be running. The ACCW pump was sequenced on with the U/V sequencer and is not a safety related load.

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'If an SI occurs while the U/V sequencer is still sequencing loads (and before the ACCW pump starts) then SI loads will sequence on; all non safety related loads will remain de-energized.

NRC Resolution:

Facility recommendation was not accepted. The reference provided does not' clearly define the condition as stated in the question, whereas, the original reference continues to support the original answer.

This answer was not modified due to the facility comments, but required additional evaluation by the examiner due to the inconsistency _ between training text and plant lesson plans. All materials should agree and clearly define the different situations under which safety equipment may be required to operate.

(5) Question: 6.08 Facility Comments:

A more complete list of temporary monitoring equipment is contained in the plant startup procedure.

NRC Resolution:

Facility recommendation was not accepted. The reference provided was not sufficiently detailed to ascertain the relationship of the equipment listed and the Temporary Core Loading Instrument System.

(6) Question: 6.09 Facility Comments:

Tech Spec Table 3.3-9 provides the list of remote shutdown panel instrumentation. Eight indications are available on the shutdown panels.

NRC Resolution:

Answer key amended to reflect the eight indications listed in the Tech Spec Table 3.3-9. Six are required for full credit as before.

9 (7) Question: 6.10 Facility Comments The interlocks specified in the Vogtle Training Text were based upon a generic SIGMA refueling machine before Vogtle was purchased. Subsequently, a machine has been purchased with the interlocks listed in the attached lesson plan.

NRC Resolution:

Answer key was changed to use new SIGMA interlocks in reference.

The answer required changing due to inconsistency between different training material provided to examiner.

All training material should reflect any changes such that if equipment or design references are compared there will be no conflict between them. This also ensures that the students receive correct information. Additionally, the lesson plan of the utility is not adequate in' the detail concerning the refueling

-machine interlocks.

(8) Question: 6.12 Facility Comments:

Additional signals will stop blowdown flow.

NRC Resolution:

Two of the facility's recommended answers are already' contained in the answer key as "high energy line break monitors. Therefore, the answer key has been changed to accept "High energy line break monitors" as one of the three required answers, OR (1) High Pressure, (2) High flow, as two answers of the required three, and add fourth and fifth answers of High Temp from Auxiliary Building area monitors and High Temp to the demineralizer.

(9) Question: 6.16 Facility Comments:

Power range high negative rate and high positive rate trips are two separate trips. Each has its own basis and each is a separate line for required operability in Table 3.3-1 of the VEGP technical specifications.

NRC Resolution:

Answer key amended to list the rate trip as separate functions.

Eight of nine listed required for full credit.

10 (10) Question: 6.17a Same as R0 comment for 2.17a.

(11) Question: 6.17b Same as R0 comment for 2.17b.

(12) Question: 7.06 Facility Coments:

The foldout page, which includes SI actuation criteria, is applicable throughout E-0. In addition step 4 RNO contains 4 items requiring manual actuation of SI. Two sets of indications relating to SI are in 19000-1.

NRC Resolution:

-Answer key has been changed to accept either of the following sets of answers:

1. Subcooling monitor less than 28 F.

2. Pzr level cannot be maintained greater than 4%.

OR

1. Pzr pressure less than or equal to 1850 psig.

2. Steam line pressure less than or equal to 585 psig or steam line pressurization rate greater than or equal to 110 psig/2 sec.

3. Ctmt press greater than or equal to 4 psig.

4. Any automatic alignment of ECCS equipment to the injection phase.

(13) Question: 7.08D Facility Comments:

Both 2.2.3 and 2.2.4 in 12000-1 discuss RHR operations in mode 5.

NRC Resolution:

Answer key modified to the following "2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months OR no time limit Dependent on S/G Level. Both required for full Tredit."

(14) Question: 7.19 Facility Comments:

Step 2 of 18038-1 has a RNO column with different actions than the action / expected response column based on the presence of a fire.

11 Thus the presence of a fire also dictates the number of actions completed.

NRC Resolution:

Facility recommendation was not accepted. The actions done on a control room fire is at the discretion of the Shift Supervisor, according to 18038-1 Step 2. The Shift Supervisor must use the guidance of the note prior to Step 2. Answer unchanged.

(15) Question: 7.20 Facility Comments:

Procedure 19200-1 has been revised and changed setpoints referenced in this question (19000-1 foldout page supports 19200-1).

NRC Resolution:

The facility recommendations were accepted and the answer key changed to reflect setpoint changes. However, this question was taken directly from the Facility Exam Question Bank which required modification due to recent changes. This question should have been modified to reflect the changes, since it could lead to misinformation, misunderstanding and safety errors by operators if the values in. the original question were learned and used instead of the actual values. Additionally,- it was noted that the facility stressed to the examiners, who prepared the examinations, that this issue of the Facility Exam Question Bank forwarded was an up-to-date issue.

(16) Question: 7.21 Facility Comments:

Cooldown rate is a technical specification LCO. The E0P training text instructs operators to follow the E0P actions even if tech spec LC0 actions are entered.

NRC Resolution:

Answer key has been amended to accept either the original answer OR "cooldown should not be limited by Tech Spec."

(17) Question: 8.038 Facility Comments:

The answer mentions a five (5) hour cooldown to RHR. The bases of the Vogtle Tech Specs has no time associated with the cooldown,

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- 'O NRC Resolution:

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. i N The facility recommendations was accepted and the "five. hours" '

removed from. answer key. -

This answer required modification due to inconsistent material provided to the examiners. This questica was taken from the Facility Exam Question Bank ar.d the same comment about it applies as that of question 7.20 reg'ar' ding the, question bank.

t (18) Question 8.08(a) s'.

Facility Comments:

The reactor trip occurred because of I&C shorting RTB U.V. coil.

A reactor. trip signal is then produced dee to a turbine trip and a negative rate due to rods dropping into the core wtlich should have ,

tripped RTA. If RTA remained closed then thtt becomes a Condition 2I event. In addition, the failure oO'.his intercept valve is a tech spec item which provides protection to safety related components. ,

T NRC Resolution: 1 Answer changed,to Condition II. s

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f (19) Question: 8.08(b)

I Facility Comments: w The response to the question 'is a double jeopardy question.

t Depending'on how the examinee answered part "a" determined his t' l response to part "b".

l- NRC' Resolution: '*- .

I a .! s Part b is deleted since it is a double jeopardy question. " N '?

(20) Question: 8.14 l

Facility Comments: T

! There is or.ly one condition that allows the OSOS or the department i supervisce to release a subclearance when the subTlearance holder i

is not on. site. Tbst is when the subclearance holder is contacted l by phone and'gives his permission.

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, 5 NRC Resolution:

? ic s Facility recommendation was not accepted. Item 4 of reference page provided allows the second part of answer key. Answer unchanged.

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(21) Question: 8.16 (Facility Comments:

Procedure 00052-C list six (6) items that constitute a procedure s intent change.

. ' 's NRC Resolution:

J' The power key has been changed to read as the facility recommended

, except that item 4 has parentheses enclosing all words after I safety.

l The following~ changes were made to the SRO answer key during exam grading:

5.02 Answer changed from a. to c.

Reason: Answer was incorrect due to distractor being rearranged during exam formatting and answer not being cbrrected.

6.14 Added to answer "All Intermediate stop valves remain 6 pan [0.25]" and inserted the word " intercept" between i

iy' the words six and valves in the first line of answer.

Changed point breakdown in line one to [0.25] instead of

[0.5].

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Reason: Intermediate stop valves were not addressed in original answer as required by question. Intercept added to clarify answer. Point breakdown changed due to

addition of answer about Intermediate Stop Valves.

'6.18 Parenthesized setpoints in answer.

! Reason: Setpoints not required.

4. Exit Meeting ,

l \ At the conclusion of the site visit the examiners met with representatives l

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

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There were generic weaknesses noted during the oral examination. These,'as welf. las areas of below normal performance included:

1. ' Candidates were uncertain of the approval status of the available " pen and ink" copy of the plant Technical Specifications.

2. B0P operators did not know the required coincidence for a valid OT runback.

3. One group of operators failed to express any immediate concern for a

't possible plant shutdown when given a known 10 gpm steam generator tube leak.

' 4. All candidates - tended to ignore SPDS alarms due to the system's

, , excessive alarm frequency.

5. Several SR0s were reluctant to call for assistance when needed.

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6. _Several candidates were uncertain of the distinction between Cold

' Overpressure Mitigation System (COMS) and Cold Overpressure Protection f - System (COPS).

g 7. Candidates were uncertain of the phase continuity light indications' response to phase grounding.

8. Candidates were uncertain of who was specifically authorized to sign for the release radioactive material.

9. Candidates' were uncertain of the specific, procedural response to a control rod being stuck off the bottom.

The examiners discussed specific simulator limitations and planned enhancements with< representatives of the plant staff.

It was noted that'41.4% of the facility comments on the October 21, 1986 written examination involved errors or changes to the material provided by the-facility to the examiners for use in exam preparation.

i The, cooperation given by the VEGP training staff to the examiners was noted S

. and: appreciated.

- The licensee did not identify as proprietary any of the material provided to

' or reviewed by the examiners.

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_t O U. S. NUCLEAR REGULATORY COMMISSION

.j SENtoA. REACTOR OPERATOR LICENSE EXAMINATION-FACILITY: VOGTLE 1 REACTOR TYPE: PWR-WEC4 I=l=

\

DATE ADMINSTERED: 86/10/21 EXAMINER: PICKER. B.

CANDIDATE _,..

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.

% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 28.00 24 cb 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS,AND THERMODYNAMICS PLANT SYSTEMS DESIGN, CONTROL,

, 29.50

_ L_25.hb 6.

AND INSTRUMENTATION LL 24.46- __7. PROCEDURES - NORMAL, ABNORMAL, E .50 EMERGENCY AND RADIOLOGICAL CONTROL 29 SD 2s. 5%

ADMINISTRATIVE PROCEDURES, fM*4W} So.18 8.

CONDITIONS, AND LIMITATIONS II.S. f Totals

-11F.5 All work done on this examination is my own. I have neither given nor received aid.

Candidate's Signature

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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:

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

2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.

3. Use black ink or dark pencil only to facilitate legible reproductions.

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

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

6. Use only the paper provided for answers.

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

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

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

10. Skip at least three lines between each answer.

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

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

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

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

15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

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

M 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in l

completing the examination. This must be done after the examination has

( been completed.

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18. When you complete your examination, you shall:

a. Assemble your examination as follows:

(1) Exam questions on top.

(2) Exam aids - figures, tables, etc.

-(3) Answer pages including figures which are part of the answer.

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

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

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

J

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. 5.. THEQBY OF NUCLEAR POWER _ELANT OPERATION.

Pcga 4 FLUIDS.AND THERMODYH&MICS QUESTION 5.01 (1.00)

Concerning equilibrium Samarium-149 (Sm) reactivity, which of the following statements is correct?

a. 50% equilibrium Sm reactivity is one-quarter of 100% equilibrium Sm reactivity.

b. 50% equilibrium Sm reactivity is one-half of 100% equilibrium Sm reactivity.

c. 50% equilibrium Sm reactivity is three quarters of 100% equilibrium Sm reactivity,

d. 50% equilibrium Sm reactivity is equal to 100% equilibrium Sm reactivity.

QUESTION 5.02 (1.00)

When performing a reactor S/U to full power that commenced five hours after a trip from full power equilibrium conditions, a 0.5%/ min ramp was used. How would the resulting xenon transient vary if instead a 2%/ min ramp was used?

a. The Xenon dip for the 2%/ min ramp would occur sooner and the magnitude of the dip would be smaller.

b. The Xenon dip for the 2%/ min ramp would occur later and the magnitude of the dip would be smaller.

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c. The Xenon dip for the 2%/ min ramp would occur sooner and the magnitude of the dip would be larger.

d. The Xenon dip for the 2%/ min ramp would occur later and the magnitude of the dip would be larger.

QUESTION 5.03 (1.50)

a. Define suberitical multiplication (M).

b. During a reactor startup, count rate is 250 CPS with a corresponding K-eff of 0.95. The count rate increases to 500 CPS. What is the re-sultant K-eff? (Show all calculations.)

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

e

. 5.. THEORY OF NUCLEAB POWER PLANT OPERATION. Pegs 5 r FLUIDS.AND THERMODYNAMICS QUESTION 5.04 (1.50) ,&(kl What is "Self Shielding"^and how does it affect reactor operations as power changes?

QUESTION 5.05 (2.00)

Indicate whether the following will cause the differential rod worth to INCREASE, DECREASE or have NO EFFECT.

a. An adjacent rod is inserted to the same height

b. Moderator temperature is INCREASED

c. Boron concentration is DECREASED

d. An adjacent burnable poison rod depletes QUESTION 5.06 (1.00)

Your plant curves show that Doppler defect becomes less negative over core life, yet power defect become more negative. Why does this phenomenon occur?

QUESTION 5.07 (1.50)

An ECC is calculated for a startup following a reactor trip from 100%

power equilibrium Xenon (BOL). Indicate if the actual critical rod position will be HIGHER, LOWER or the SAME from the calculated position for each of the following situations. Treat each case individually.

a. Xenon reactivity curve for trip from 50% is used to calculate conditions to startup 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months after the trip.

b. The Steam Dump pressure setpoint is raised 100 psi.

c. The power defect curve for MOL is used instead of the BOL curve. ;

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

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. . 5.s THEORY OF' NUCLEAR-POWER PLANT-OPERATION.

Paga'<6

EUIDS.AND' THERMODYNAMICS LQUESTION 5.08 (3.00).

a. Discuss the behavior of reactor power and Tave during and after 2 minutes of Emergency Boration at'100..% power. ' Assume rod control is..in manual.

.b. Discuss-the behavior of reactor power and Tave after-2 minutes of Emergency?Boration at 10E-8~ amps and no. load.Tave.

QUESTION 5.09 (1.00)

' Assuming the reactor'is initially. operating.at the optimum value of

-moderator / fuel ratio, will the following changes cause the reactor core to become UNDERMODERATED, OVERMODERATED, or HAVE NO EFFECT? Consider'each change separately with all parameters constant.

a. Increase in moderator temperature,

b. Inserting group D rods from 220 steps to 180 steps.

fQUESTION 5.10 (1.00)-

L - As.the core ages, the buildup of Pu-240 causes the Fuel Temperature Coefficient (pcm/ degree F) to become~more negative. With this-change occurring, why does the Doppler Only Power Coefficient (pcm/% power) become.less negative as the core ages?

. QUESTION- 5.11 (2.00)

TWO: major factors affect differential boron worth over core life. List these TWO factors AND indicate how (MORE NEGATIVE or LESS NEGATIVE) they affect differential boron worth.

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( -

. 5.. THEORY-OF NUCLEAR POWER PLANT OPERATION. Pcgs 7 i PLUIDS.AND THERMODYNAMICS QUESTION 5.12 (1.50)

TRUE or FALSE 7

a. The faster a centrifugal pump rotates, the greater the NPSH required to prevent cavitation.

b. One of the pump laws for centrifugal pumps states that the volumetric flow rate is inversely proportional to the speed of the pump.

c. Pump runout is the term used to describe the condition of a centrifugal pump running with no volumetric flow rate.

QUESTION 5.13 (1.50)

After operating at 100% power for three months, power is suddenly lost to all of the reactor coolant pumps. EXPLAIN how each of the following enhance natural circulation.

a. Maintain pressurizer level of at least 50%.

b. Maintain adequate subcooling in RCS.

c. Maintain heat sink.

QUESTION 5.14 (0.50)

Does the Latent Heat of Vaporization INCREASE, DECREASE or REMAIN THE SAME as saturation pressure / temperature of water is increased?

QUESTION 5.15 (1.00)

The reactor is at 10% power. What is the initial effect of an increase in steam demand on steam generator level (increase, decrease, no effect)? Ex-plain your answer assuming feedwater flow remains constant.

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

J

c.1 5.. THEORY OFiNUCLEAa POWER PLANT OPERATION. -Pags=>8 e~ FLUIDS.AND THERMODYNAMICS QUESTION 5.16- l(1.50)_-

~

Match fthe heat transfer process iri Column A to an equation that

. applies-to.that process in Column B.

COLUMN A COLUMN B

a. Between' cold' leg and hot leg. - 1. Q , m c Delta-T of' reactor (normal' forced con- . .

vection flow)- 2. Q= m Delta-T

-b . Across S/G tubes (primary to .

secondary) 3. Q = U A Delta-T-

c. Across-S/G.(feedwater to steam) 4. k=mcDelta-h-

5. h=mDelta-h

- QUESTION 5.17 (1.50) a) TRUE or FALSE: During cold plant conditions, you would expect the COLD calibrated PZR level instrument to indicate HIGHER than the HOT' calibrated ~ level 1 instrument. (0.5) b) Give two different conditions involving the reference leg which could-result in a false high level indication on the.PZR level _ instrument.

(1.0).

QUESTION 5.18 (2.00);

a. Define DNBR. (0.5)

b. Since che DNBR is not a directly observable parameter, name THREE parameters the operator monitors and/or controls to ensure the DNBR limit is not violated. (1.5)

QUESTION 5.19 (1.00)

Describe the conditions under which cavitation occurs.

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

J

THEORY OF:N CLEAR P'OWER PLANT OPERATION,

~

ir ' 5. Pago - 9 i- -FLUIDS.AND THERMODYNAMICS-.

QUESTION '5.20; l(1.00)

Steam exiting:th'e HP turbine is at 785 psig, 90%' quality. Steam entering the:LP turbine is superheated to 100 F. What is the?

enthalpy' change of-the' steam?

'a. 85 BTU /lbm

b. 140 BTU /lbm

c. 154 BTU /lbm

d. 705 BTU /lbm-(***-** END OF CATEGORY 5 *****)

O .6'.m PLANT SYSTEMS DESIGN.^ CONTROL; AND~ INSTRUMENTATION ~Pcca_10 QUESTION 6 . 0'1 (1.00)

. State.the SAFETY-RELATED reason;for.having reactor vessel head vent is installed.

' QUESTION- 6.02 -(1.00)'

While. operating at 100% power a maintenance electrician is

. requested to inspect.a 13.8 KV relay cabinet. While doing so he causes an UV relay for one RCP to actuate.

Which of the following best describes the plant response.

a. The affected RCP trips, reactor trips, the other 3 RCPs trip and the plant sis on low low flow,

b. The affected RCP trips, reactor trips and the other-3 RCPs trip.

c. The affected RCP trips. The operator performs a normal controlled plant shutdown with the other 3 loops in service.

d. The affected RCP trips, the reactor trips due to low flow in the affected-loop.

QUESTION 6.03 (2.00)

a. WhatactionoccursasaresultofplacingtheSteamDumpcontr}o in the RESET position after an automatic actuation? CUn dd to~

syit

b. What indicator on the main control board will change when-resetting the Steam Dump controller?

QUESTION 6.04 (1.00)

What percentage of the decay heat removal capacity is the auxiliary feedwater system. capable of, if one motor driven pump and the' turbine driven auxiliary feedwater pump is out of service?

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

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4 W 6..' PLANT 1 SYSTEMS' DESIGN. CONTROL. AND INSTRUMENTATION - Pasa 11 cr .:

TQUESTION. - 6 .' 0 5 (1.00)

' What condition will cause the-ESF load sequencer to iake the following action?

The sequencer will'shed non-safety-related loads ONLY and sequence those 11oads required'for SI.

l

-QUESTION .6.06 (1.00)

The' function of the diesel engine after cooler is to provide cooling to:

a. intake air after the turbo charger,

b. -the jacket water cooler after engine shutdown.

~

c. standby lube oil heater-system to limit temperature.

d. the exhaust to aid in silencing the engine.

QUESTION 6.07 -(2.00)

List FIVE separate signals / methods for starting a Diesel Generator.

QUESTION 6.08 (2.00).

List the FOUR different components that-constitute the Temporary Core Loading Instrument System.

l QUESTION 6.09 (1.50)

List SIX of the EIGHT indications required to be present on either of the remote shutdown panels by Tech Specs.

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(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

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-- - . . - , . - - _ _ -,m.,-..m.-. . . - _ , _ , , _ , . . - _ - .- , . . , , - . , . . , , ,,-.. ~..~, 4._ , _ , . - . . , , , , , _ - - - - , - -

6. Pags 12 PLANT SYSTEMS _ DESIGN. CONIBOL. AND INSTRUMENTATION c

QUESTION 6.10 (2.00)

List the interlocks AND function that to allow the refueling machine to move a fuel assembly from the core to the transfer area. Start with gripper down on an assembly. ASSUME proper system operation and movements are being done - no abnormal conditions.

QUESTION 6.11 (1.00)

What are the TWO design features of the Spent Fuel Cooling System which prevents loss of the water in.the storage &enk?

Poo l QUESTION 6.12 (1.50)

State THREE signals which will either isolate or stop Steam Generator Blowdown flow?

QUESTION 6.13 (1.50) 2/3 (two out of three) pressurizer pressure channels below 1970 psig (P-11) allows the operator to manually block certain protective functions. What are the THREE protective functions blocked by this action?

QUESTION 6.14 (1.50)

Briefly explain the operation of the Intercept Valves and Intermediate Stop Valves on a generator load rejection and subsquent recovery to stable conditions at the lower load.

QUESTION 6.15 (1.00)

What interlock req rements must be satisfied to open the Letdown isolation valves? L4 59 w %

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

6. PL6NT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Paga 13 QUESTION 6.16 (2.00)

List the EIGHT reactor trips other than " General Warning" which cannot be bypassed or blocked. Include coincidence (logic). Setpoints are not required.

-QUESTION 6.17 (2.00)

a. What automatic actions does the Cold overpressure Mitigation System (COMS) take to prevent depressurization of the RCS when operating at Normal Operating Pressure?

b. How is the COMS setpoint determined during shutdown conditions?

QUESTION 6.18 (2.00)

List all PERMISSIVES which receive a DIRECT input from the power range instrument system (N41 thru N44), logic for each, if any, and briefly what each permissive accomplishes.

QUESTION 6.19 (1.50)

What are SIX functions of the Rod Control Alarm reset switch on the main control board?

QUESTION 6.20 (1.00)

According to the turbine runback procedure, 18012-1, what are the three signals that can generate a valid (as opposed to unexplained) turbine runback? (Include setpoints where applicable) l

(***** END OF CATEGORY 6 *****)

i

t4 7;. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY . Pcga.14~

- : A'ND RADIOLOGICAL CONTROL LQUESTION 7.01: ( l'. 00 )

When is the containment spray system aligned for recirculation from j the. containment sump following a LOCA7 QUESTION. ~7.02 (1.00)

Requests to exceed. administrative. radiation guides are=made by:

'a. the individual receiving the dose.

b. the immediate supervisor of the individual receiving the dose.

c. an HP Lab Foreman.

d. the Superintendent of Operations.

QUESTION 7.03 (1.00)

a. If an Orange CSF is encountered while performing an ORG, is it permissible to continue with the ORG while performing the actions of the applicable FRG7 l b. If a Red CSF is encountered while in'the FRG for the Orange CSF, is it permissible to continue the FRG for the Orange CSF while

[ performing the FRG for the Red CSF7 l

QUESTION 7.04- (1.00) r List two different conditions when an Inverse Count Rate Ratio Startup l procedure is required.

I l

QUESTION 7.05 (1.50)

List three means of minimizing the possibility of water hammer during oper-ation of the AFW system.

(***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)

Page 15

.' -7.. PROCEDURES - NORMAL. ABNQRMAL. EMERGENCY

RND RADIOLOGICAL CONTROL QUESTION 7.06 (1.50)

When initiating Procedure 1900-1, (E-0) Reactor Trip'or Safety Injection with no SI presently required, WHAT indications would require you to actuate safety injection?

QUESTION 7.07 (1.00)

A Caution Statement in ES-1.1 "SI termination" states that "without instrument air available, CCP suction should remain aligned to the RWST".

a. WHAT makes the loss of Instrument Air a problem in this situation?

b. WHAT is the basis for this caution?

QUESTION 7.08 (2.00)

Answer the following questions about the Precautions to Refueling Recovery Procedure 12000-1.

a. How many boration paths must be operable and capable during Modes 5 and 67

b. While in Mode 6, when must both RHR trains be operable and at least one operating.

c. During Mode 5, shutdown margin shall be greater than what value?

d. What is the maximum time that a (one) RHR loop may be out of service for surveillance testing in Mode 57 QUESTION 7.09 (1.00)

What is the basis for the caution note in the procedure 12001-1, Unit Heatup to Hot Shutdown, to operate PCV-131 in manual prior to starting the first RCP.

(***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)

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

7.- PROCEDURES - NORMAL. ABNORMAL. EMERGEEQY Pcca 16 AND RADIOLOGICAL CONTROL QUESTION 7.10 (1.00)

At what plant conditions (temperature and pressure) are the accumulators leak tested and placed in service?

QUESTION 7.11 (1.00)

What parameter must be checked before RCS pressure is allowed to exceed the F-11 setpoint and why? l QUESTION 7.12 (1.00)

What is the immediate operator action on a failure of a Pressurizer Pressure Instrument per AOP 18001-1C, Primary Systems Instrumentation Malfunctions?

QUESTION 7.13 (1.00)

According to a caution note of AOP 18008-1, Secondary Coolant Leakage, what are TWO concerns with the CST level dropping below 70%?

QUESTION 7.14 (1.00) j l

What condition determines whether AOP 18009-1, S/G Tube Leak Procedure, or EOP 19000-1, E-O Reactor Trip or Safety Injection will I I

be used?

l QUESTION 7.16 (1.00)

What are THREE indications used by the operator to verify that a turbine runback has been INITIATED according to AOP 18012-1, Turbine Runback?

QUESTION 7.16 (1.50)

List three items that the Reactor Startup procedure, UOP 12003-1 requires the operator to check after every 50 steps of Control Rod Bank withdrawal.

(***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)

y Pags 17

'7.~ 'PROCEDUBES - NORMAL. ABNORMAL. EMERGENCY A'ND RADIOLOGICAL CONTROL QUESTION 7.17 (1.00)

With the reactor operating @ 100% power, what immediate action is required upon loss of off-site power concurrent with the "B" diesel generator failing to tie on the bus according to AOP 18031-1, Loss of Class 1E Electrical Systems? Be specific as to how this action is carried out.

QUESTION 7.18 (1.50)

What personnel report to shutdown Panel B on a control room evacuation per AOP 18038-1, Operation from Remote Shutdown Panels?

QUESTION 7.19 (1.00)

What condition or situation dictates the number of actions completed in Procedure 18038-1, Operations from Remote Shutdown Panel, prior to actually leaving the control room?

QUESTION 7.20 (1.00)

State the criteria used to declare an imminent loss of secondary heat sink in accordance with EOP 19200-1, 1-1.1 " Heat sink critical safety function status tree". (i.e. heat sink red path summary)

QUESTION 7.21 (1.00)

During performance of E-3 " Steam Generator Tube Rupture" the procedure directs you to cooldown at the maximum rate, which you direct your RO to accomplish. Momentarily, the RO reports that his cooldown rate is

, 315 F/hr. Would it be proper to direct the RO to reduce the cooldown rate? EXPLAIN.

QUESTION 7.22 (1.00)

Why should AFW flow to intact S/G be raised prior to the maximum rate depressurization according to Procedure 19030-1, E-3 Steam Generator Tube Rupture?

(***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)

+ 7.. PRQGERHRES - NORMAL. ABEQBMAL. EMERGENCY Par;a 18

$ND RADIOLQGICAL CONTROL

-QUESTION 7.23 (0.50)

TRUE or FALSE?

If offsite power is lost after SI reset, manual SI actuation is required to restart safeguards equipment.

QUESTION 7.24 (1.00)

a. Specific RWPs may remain inactive for how many days before they are automatically terminated?

b. If an inactive RWP is to be reactivated, how is the re-activation accomplished?

i QUESTION 7.25 (1.00)

TRUE or FALSE?

a. Whole body radiation monitoring devices are to always be worn on the upper front quadrant of the body.

b. The limit to the " skin of the whole body" does not apply to the skin of the extremities.

QUESTION 7.26 (1.00)

a. PERMISSION from what TWO personnel are required to raise a fuel element with the fuel elevator?

b. How is the raising operation accomplished?

(***** END OF CATEGORY 7 *****)

8.. ADMINISTRATIVE PROCEDURES. CONDITIONS. Pago 19 AND LIMITATIONS QUESTION 8.01 (1.00)

During a plant startup with the reactor power @ 5 X 10E-9, one intermediate range channel (N0036), fails. Which of the following statements is correct? Refer to the attached Tech Specs.

a. Operation above the P-6 IRM neutron flux interlock setpoint is not allowed until the inoperable channel is repaired and declared operable.

b. If N0036 is placed in a tripped condition and the other channel is operable, you must reduce the power range neutron flux trip setpoint to =or< 85% and thermal power is restricted to =or>75%.

c. With N0036 out of service due to failure and the remaining channel is operable, power operation is limited to 10% of rated thermal power.

d. Operation above P-6 IRM neutron flux interlock may continue unrestricted up to 100% power provided the failed IR channel is placed in trip condition.

QUESTION 8.02 (1.00)

Which of the following statements is correct concerning the quadrant power tilt ratio (QPTR)?

a. No action is required within one hour regardless of the QPTR

b. If QPTR exceeds 1.02, but is less than 1.09, operation may continue indefinitely only up to the reduced thermal power (100-3% for each 1%

of indicated QPTR in excess of 1.0) allowable for the RCP combination.

c. If QPTR exceeds the maximum limit of 1.09 for 1 hr, the reactor must be immediately shutdown.

d. If misalignment of a control rod causes the QPTR to exceed 1.09, thermal power must be reduced within 30 minutes.

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(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)

Zi'8~.T ADMINISTRATIVE PROCEDURES. CQHDITIONS. Pega 20

" AND LIMITATIONS 3

. QUESTION 8.03 .(1.00) a' . What is.the minimum required level'(volume) in one condensate storage tank per plant technical ~ specifications?

b. What is the bases for the minimum level specified above?

QUESTION 8.04 (1.00)

The OSOS or SS must review all operators surveillance test performed on his shift.

a. What is.the purpose'of this review?

b '. How is the review indicated?

QUESTION 8.05 (2.00)

.What are the TWO restrictions placed on a licensed SRO assigned to supervise fuel handling?

QUESTION 8.06 (1.50)

List THREE requirements that must be met before a short-term relief of a reactor operator can take place.

QUESTION 8.07 (1.00)

According to Operation Administrative Procedure 1005-C, Operability Status Indication for Plant Safety Systems, WHAT is the operators response to an alarm on ACB04 on the QMCB due to an automatic illumination of any SSMP light?

(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)

. 8.. ADMINISTRATIVE PROCEDURES. CONDITIONS. Pcgs 21 AND LIMITATIONS t.oo QUESTION 8.08 ( B-GO)

On a preceding shift, a trip of the reactor occurred with the following conditions recorded: Plant was at 95% power, trip occurred from I&C tech. shorting reactor protection lines to trip breaker UV coil on Train "B". Train "A" breaker did not actuate. I&C tech removed the short circuit. ESF equipment responded normally. Main turbine tripped with normal indication except that one intercept valve failed to fully shut. All other secondary equipment operated properly.

a. What review condition would this trip receive?

b 'Jhcoc pciiniaalva wvuld La necdcd ta startup? % h { M QUESTION 8.09 (2.00)

WHO are FIVE people, by title, that may authorize the Shift Supervisor to have lockout relays or relay targets reset?

QUESTION 8.10 (1.00)

A maintenance person with work package and RWP comes to your office and requests issuance of a key to a high rad area. What special requirements / actions are you required to take before allowing this person to sign for the key?

QUESTION 8.11 (1.00)

What classes of keys would require that " Action for Key Compromise" be taken if it is discovered that a compromise has occurred per Plant Administration Procedure 00008-C, Plant Lock and Key Control?

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(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)

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Pcga 22

~ 3. ADMINISTRATIVE PROCEDUBES1_C9BDITIONS.

AND LIMITATIONS QUESTION 8.12 (1.00)

Unless authorized by the level or above, an individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any hour period (excluding shift turnover time).

I a. Applicable Department Superintendent, 48

b. Shift Supervisor, 24

c. Applicable Department Superintendent, 24

d. Shift Supervisor, 48 QUESTION 8.13 (1.00)

a. What action should operating personnel take if they determine that a procedure in use has erroneous directions (steps out of order, wrong equipment numbers etc.)?

b. What supervisory actions are required?

QUESTION 8.14 (1.00)

Under what TWO conditions can an OSOS or Department Supervisor release a subclearance if the subclearance holder is NOT on-site?

QUESTION 8.15 (2.00)

State the FOUR situations / conditions under which the operating shift must shut down the reactor according to Administrative Procedure 10000-C.

QUESTION 8.16 (1.00)

State the FIVE items that may constitute a " procedure intent change" per Plant Administration Procedure 00052-C.

(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)

- 8.~- ADMINISTRATIVE PROCEDURES. CONDITIONS. Pncs 23 END LIMITATIONS QUESTION 8.17 (1.00)

What are TWO exceptions allowed to the requirements of Tech. Spec.

4.0.3 relating to the failure to perform a surveillance with in the allowed time period?

QUESTION 8.18 (1.00)

Indicate whether the following statements are TRUE or FALSE.

a. The spray additivo tank contains a concentration of 20 to 25 percent by weight aodium hydroxide (NaOH) solution.

b. Spray additive tank tech spec volume and concentration ensures a pH value of between 8.5 and 10.5 in recirculated solution in containment following a LOCA.

QUESTION 8.19 (1.50)

At what emergency classification levels is assembly and accountability '

of all PROTECTED AREA personnel mandatory?

QUESTION 8.20 (1.00) i Match the times in Column B to the statements in Column A concerning E plan notifications.

Column A Column B Followup emergency message form should 1. Immediately a.

be transmitted to state and local 2. 15 minutes authorities.

b. Initial notification to offsite 3. 30 minutes authorities (state, local).

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

c. Initial notification to NRC. 5. 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

d. Followup notifications to NRC of any further degradation in plant conditions.

(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)

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Paga 24 ,

R l- 8.*' ADMINISTRATIVE PROCEDURES.' CONDITIONS.

l' AND LIMITATIONS p-u .r -

l :

. QUESTION 8.21 1(1.00)

L i What are the responsibilities of the_OSOS'after he has been relieved of-1

.the Emergency Director duties by the permanent Director according to.

Emergency Response Procedure 91001-C, Emergency Classification and

-Implementing Instructions?

~ QUESTION 8.22 (2.50)

What are FIVE responsibilities of the Emergency Director that are non-delegatable according to " Duties of the Emergency Director",

91102-07 QUESTION 8.23 - (1.00)

Who, by title, are the primary people to relieve the OSOS of the Emergency Directors position.

' QUESTION 8.24 (1.00)

You are the Emergency Director faced with a condition of having to send one'of the four~following volunteers through a radiation area to save an injured auxiliary operator. The expected dose'is calculated to be 30 rem. Choose the best candidate and justify your selection according to 91301-C, Emergency Exposure Guidelines.

MAKE NO OTHER ASSUMPTIONSI

a. Male mechanic, Age 47, with NRC Form 4 and recorded lifetime exposure of 15 rem.

b. Male auxiliary operator, age 23, with NRC FORM 4 and recorded lifetime exposure of 1.5 rem.

c. Female auxiliary operator, age 34, with NRC Form 4 and recorded lifetime exposure of 1 rem.

d. Male janitor, age 20, with previous r ecorded exposure of 20 mrem lifetime.

(***** END OF CATEGORY 8 *****)

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

__ _____________________u

t EQUATION SHEET

~.

J f = as v = s/t 2 Cycle efficiency = et ork (out) -

w = ms s=vt+o at 2 Energy (in)

E = aC -

a = (vg - y )/t 2 E

- EE = hav vg = v, + a A = AN A = A,e

, PE = agh w = g/g A = In 2/tq = 0.693/tg

"AP' . (t,.)(es) . . . .

AE = 931Am h(*")

e .. ?h+'b}

'Q = BCpaT , , y , y*,4x k=UAAT g ,g ,-yx

~ Pwr = Wg ' a"

. I=I 10"*

F = P, 10 5UR(t)

TVL = 1.3/u t

j P = P, e /T HVL

0.693/u

'SUR = 26.06/T -

T = 1.44 DT SCR = S/(1 - K,fg) fX*ggo )

SUR = 26 g, CR g = S/(1 - K,ggg)

~

T = '(1*/o ) + [(g _' g)/1,ggo ]

1( eff}1 = CR2 C1 ~ *eff)2 T,= 1*/ (p f) M " 1/(1 - K gg) = CR g/CR0 T = (I - p)/ A*ffo M = (1 - K,gg)0IN ~ Eeff)1 j

8 * ( eff-1)/K,gg = R ,gg/K*ff SDM = (1 - Keff)/Keff p= [1*/TKygg.] + [I/(1 + A,ggT )] ,

g* = 1 x 10 -5 seconds P = E6V/(3 x 1010) A,ggl e 0.1 seconds

,=. .

Idg1=Id22 WATER PARAMETERS Id =Id2 g

i 1 gal. = 8.345 lha '

R/hr = (0.5 CE)/d g,,,,,,)

2 1 gal. = 3.78 liters R/hr = 6 CE/d (feet) -

- 1 ft3 = 7.48 gal. MISCEI.I.ANEOUS CONVERSIONS ,

Density = 62.4 lbm/ft 1 Curie = 3.7 x 10 dps 10 Density = 1 gm/cm 1 kg = 2.21 1ha Heat of valorizations = 970 Etu/lbm I hp = 2.54 x 103 BTU /hr Heat of fusica = 144 Bcu/lbm 1 Hw = 3.41 x 10 Btu /hr 0

1 Atm = 14.7 Psi = 29.9 in. Ig. 1 Btu = 778 f t-lbf 1 ft. H 2O = 0.4333 lbf/in 2 g inch = 2.54 cm F = 9/5 C + 32 C = 5/9 (*r - 32)

s y.

5.' THE9RY OF NUCLEAR POWEB PLANT OPERATION. PpJa 25 EMLIDS. AND THERMODYNR1ICS 1

r 1

~

~ ANSWER 5.01 (1.0D[

d REFERENCE NUS. Nuclear Energy Training - Reactor Operation, p. 10.5 4 Westinghouse Reactor Physics, pp. I-5.77 - 79 VEGP, Training Text, Vol. 9, pp. 21-81 & 82 HER, Reactor Theory, Session 37, p. 4 ANSWER 5.02 (1.00)

. +

, x .

REFERENCE CNTO " Reactor Core Control" Section 4 -'

Westinghouse Simulator Trng book, "Rx Theory and Core Physics", Fig I-5-54 001/000; K5.38(3.5/4.,1)

VEGP Training Vol. 2,' Chapter 5, pp. 55-58 i \']

CNTO, " Reactor Core Control", pp 4-21/28 001/000; K5.38(3.5/4.1) ,

q s

.n '

ANSWER 5.03 (1.50) s

a. 1. Suberitical multiplication (M) h , defined as the ratio of the total number of fission and source resdisrphs to the total number of neu-trons which would exist due to the source only.[0.75],on 8:,g,og Mdk.4f , O R.

b. 1. CR2/CR1 = (1-Keff1)/(1-Keff2) [0.5]

500/250 = 1 .95/(1-Keff2) ( o, ge,g, ppg,,

Keff2 = 0.975 [0.25] ,, g,g .

doef suberdsal REFERENCE ,,,lg e,dw yJ,c ,do BVPS Reactor Theory Manual Chapter 5 pp 36 - #3 yI[([,

VEGP Training Vol. 2, Chapter 5, page 9 ,,,4 g j,,,,g g g4

% ,ce n. h o w w v'ie'id.3 of kdel)

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

i Pego 26

5. ' TBEQBY OF NUCLEkB POWER PLANT OPERATION.

FLUIDS. AND >THER'lODYNAMLQS_

l.,-

l}0 ANS,WNR 5.04 (1.50)

The large pellet diameter (relative to the resonance peak energy path) results in a self shielding effect for fuel in the pellet interior.

As fuel pellet temperature rises the off-resonance neutrons (for the most part that would previously pass entirely through the pellet) are more readily absorbed. [1.0)

Causes regative reactivity to be added as power increases. [0.5] ,

l REFERENCE BVPS Reactor Theory Manual Chapter 6 p 33 VEGP Training Vol. 2, Chapter 4, pp 31-37 r

i ANSFER 5.05 (2.00)

I N f

a) Decrease (+.5 ea) b) Increase c) Increase ,

d) Increase l REFERENCE SQN/WBN License Requal Training, " Core Poisons" CNTO, " Reactor Core Control", pp 6-22/28 VEGP Training Vol. 2, Chapter 4, pp 103-113 001/000; K5.09 (3.5/3.7) & K5.02 (2.9/3.4) & K5.10 (3.9/4.1) s ANSWER 5.06 (1.00) i The Mo'cerator Temperature Coefficient (MTC), (which is a component of the J Power coefficent), becomes negative over core life and thus over-

\ compensates for the effect of Doppler only.

REFERENCE CNTO, " Reactor Core Control", pp 3-37/40 VEGP Training Vol. 2, Chapter 4, pp 125-139 001/000; K5.49(3.4/3.7)

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

.f 5. THEORY OF NUCLEAR POWER PLANT OEEB611QL, Pcco 27 W76, * - EMIDS.AND THERMODYNAMICS g5 5,

ANSWER 5.07 (1.50)-

a. Higher

b. Higher

c. Higher [0.5 each]-

REFERENCE 6

TPT OP.1009.1; Plant Curves

VEGP Training Vol. 2, Chapter 5, Plant curves and Question Bank #92NB n 001/000; A2.07(3.6/4.2) r.

[ ANSWER 5.08 (3.00)

a. Power decreases initially due to the boron addition [0.5].

'The' primary to secondary mismatch causes Tcve to decrease [0.5].

The decrease in Tave inserts positive reactivity and

'- restores reactor power to a slightly lower than or the same ~

as initial power level [0.5]. Tavg will steady out.at a new-lower level [0.5].

b. Tave does not change due to the boration [0.5].

) -(Tave is determined by the amount'of pump heat and the steam-L dump setting.) After the~ initial transient, power decreases p-at a negative 1/3 DPM rate to the multiplied source level [0.5].

L REFERENCE L

MP3 Transient and A;_* dent Analysis Chapter 4.10 VEGP Training Vol. 2, Chapter 4, pp 114-122 ANSWER ' 09

. (1.00)

a. UCDERMODERATED

b. UNDERMODERATED [0.5 each]

-REFERENCE Westinghouse Nuclear Training Operations, pp. I-5.7, 9,&10 VEGP Training Vol. 2, Chapter 4, pp 74-78

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

a t 5. 5 THEORY OF NUCLEAR-POWER' PLANT OPERATIONc _ Peso - 2'8.

ELUIDS.AND THERMODYNAMICS

.e i

' ANSWER: 5.10 -(1.00) t As the core: ages, the pellet and clad creep together causing an-increase in" gap conductivityf(0.4 pts.). This causes a smaller delta T of the:

~

fuel forca given. power ~ change (0.3 pts.).

The smaller delta T of-the fuel'causes a smaller change in reactivity for a given power change (0.3 pts.).

'(pcm/F).X F/% power)= pcm/% power more neg less less neg

=. REFERENCE .

Westinghouse Nuclear. Training Operations,. p. I-5.22 VEGP Training.Vol. 2- Chapter.4,'pp 142-144 ANSWER 5.11 (2.00)

1. . Boron Concentration Decreases (0.5 pts.) - MORE NEGATIVE (0.5-pts.)

2. Fission Product Buildup (0.5 pts.) - LESS NEGATIVE (0.5-pts.)

-REFERENCE- ,

' Westinghouse Nuclear Training Operations, p. I-5;31-VEGP Training Vol. 2, Chapter 4, pp 122-124

. ANSWER' 5.12 (1.50)

~

a..True ,

b. False

c. False- [0.5 each]

REFERENCE

. Westinghouse Thermal Science, Chapter 10, Pp. 41-49 General Physics HT & FF, Pp.319-326.

4 4

h i

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

t w we yv..r..-- *, - en ,m -w w r m s , =,c~,,, -- , - < a- - - - - - + - - - --

,*,--y-- - w- wr-

~

^ 5.5 !

THEORY OFLNUCLEAR POWER' PLANT OPERATION. Poca'29

_ FLUIDS.AND THERMODYNAMICS-ANSWER 5.13 (1.50)

a. To ensure that no vapor (steam) pockets form (in the loops), O R hoides suGSi4. t inven4evy cotr.s3 h maWk nc- pressune c.whol Co.zs']

b. 'To prevent vapor (steam) pocket formation

c. To help assure. adequate thermal driving head

REFERENCE General Physics HT & FF, Pp.356-357.

3.4 000 015 EK 1.01~ 4.4 ANSWER 5.1'4 (0.50) decrease

. REFERENCE-CNTO, " Thermal / Hydraulic Principles and Applications, I", pp 2-58/59 Steam Tables 000/027;-EK1.03(2.6/2.9) c ANSWER 5.15 (1.00)

Increase-( 25). As steam flow increases the number and size of steam bubbles ~in1the steam generator increases (0.375). This reduces the density, thus increasing the specific. volume of the mass in the l

steam generator, which causes a swell in level (0.375).

REFERENCE i HTTFF pgs. 46-50 IOP-3 pg. 14 Westinghouse Thermal Science, Vol. 2, Chapter 12, p. 48 s

3.4 035 010 K 5.03 2.8 i

l' f -.

l- (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

'1'5Y' THEORY OF NUCLEAR-POWER PLANT OPERATION. Page 30.

._ FLUIDS.AND.THEBMODYNAMICS

' ANSWER -5.16 (1.50)

.a. 1 (or-5)-

-b. 3

c. 5 [0.5]

REFERENCE General Physics, HT & FF, pp.'180 and 181 ..

.CNTO, " Thermal / Hydraulic' Principles and Applications, I", pp 4-61/64lk 5-47 001/000-K5.45 (2.4/2.9)

~

' ANSWER ' 5 '.17 (1.50):

a) False b) HPost) accident heating of Reference Leg

-Referencel Leg leakage (vosDid6) [0.5 ea]

-REFERENCE i

NRC IE-Info Notice 84-70.(4 Sep 1984)

.TPT Lesson Plan for Requal. Cycle-II-1985 011/000; K4.03(2.6/2.9)

ANSWER 5.18 ( 2 '. 00 )

'a. DNBR = Heat flux (power) to cause DNB / actual heat flux (power) (0.5).

b '. 1. ' Pressurizer Pressure

2. RCS^Tavg

3. RCS Flow (Will.also accept Rx Power, AFD, QPTR, Rod sequencing / position /over-lap, others on a case-by-case basis.) [3-@ 0.5 ea] (1.5)-

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

1

'l

a. -

Paga 31 e35.*-THEORY OF-NUCLEAR POWER PLANT OPERATION.-

au FLUIDS.AND THERMODYNAMICS.

. REFERENCE VCS,-TS, pp. 3/4:2-15.and B-3/4 2 5 and B 2-1.

~

' General Physics HT & FF,.P.262.

.CNTO, " Thermal / Hydraulic Principles ~and Applications, II", pp 13-20/24L

. VEGP . TS - 3/4'. 2. 5~

'015/020; K5.09(3.5/3~.7)

ANSWER ~ 5.19 (1.00)

. Cavitation occurs when the pressure of the fluid is reduced to a. press-ure below that of its: saturation pressure for a given temperature which will allow boiling to occur.

' REFERENCE' General Physics HT & FF, P.319.

. ANSWER 5.20 (1.00) e REFERENCE SQNP,.HTFF t' ext,-pp. 23 - 24 Cook,zWestinghouse: Thermal Science, Chapter 7, Pp 32-46.

General Physics HT & FF,.Pp.83-96.

Steam Tables t-i.

a l

l:

(***** END OF CATEGORY 5 *****)

.._s._.. . :- - - - , .- .._;. , - - . . . . . - . . . . . _ , _ . _ _ , , .. _ .- _ - . _ - _ _ - -

t l6.'*: PLANT SYSIEMS DESIGN: CONTROL'. AND INSTRUMENTATION ~ Pega 32 ANSWER- 6.01 (1.00)

'The' presence of steam or nonconde'nsibles could-impede natural circulation

-h;d5 E sa% gm.1c leMm podh (Joe Ab gcde dJ ddkd4 - sq@ .

REFERENCE

-VEGP,.LO-LP-09202-00, PP. 13, 15

-VEGP, LO-LP-16001-00, P. 9-ANSWER- 6.02 (1.00) d REFERENCE VG,' RO-TP-217-005. 'P

' ANSWER 6.03. (2.00) i a; Reset Loss-of-Load circuitry.

b. " Steam Dump Unblocked" indicator deenergizes. oF-

. c-1 x4.udar IQU epes od ,

~

, REFERENCE

- VEGP, . Training Text, : Vol. 7, Ch. 12b, P. 5 ANSWER' 6.04 (1.00)'

100%

- REFERENCE VEGP, Training Text, Vol. 7, Ch. 13d, P. 3 ANSWER 6.05 (1.00)

Loss of power with'SI AFTER initial loads sequencing complete.

4

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

(

or g- -g - ,-,y e, - e s--e f.-+=,w- --n--w- .,w- - e. ,, -w+,-, ,-ytv--- - +vn--s

6.5 PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Paco- 33 REFERENCE VEGP, Training Text, Vol. 8, Ch. 16a, P. 30 ANSWER- 6.06 (1.00) a REFERENCE VEGP Training Text, Vol. 8, Chapter 16c., p. 9 ANSWER 6.07 (2.00)

1. SI

2. Loss of voltage on 4160 Bus

3. Manual control room

4. Manual D/G room S. Emergency manual D/G room [0.4 each]

REFERENCE VEGP, Training Text, Vol. 8, Ch. 16c, P. 20 ANSWER 6.08 (2.00)

1. (3) BF3 neutron monitors

2. (3) Countrate meters

3. (3) Scalers

4. (1) Master timer REFERENCE VEGP, Training Text, Vol. 8, Ch. 18a, P. 12

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

P;a 34 6.*

P.LANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION ANSWER 6.09 (1.50)

1. WR SG level tall loops)"---

2. SR-neutron fl'~ _ SR. ne, tron Ch.

3. In neutreen flux s rh wide q e

-4. Tc (-011 loops 4-a. wid% e,

5. WR RCS pressure

6. Par level

'7. Core exit thermocouples teclculated>_ sf. Ll-- j

8. n flux (6 @ 0.25 each)

Neutrga 44m wge,m(radicticn element h Wnd hiek.

REFERENCE VG, SR-LP-122, p 4.

Y%P T5 3,3,3,s 4, aj 74k 3 g e ,3 _g ANSWER 6.10 (2.00)

O do -

[0. each '"'^Q u.,ma M +u %F.6 w md e.

1.$qaMaged(p,,2)m-1n mS made to allow raising hoist. c or4 e W c

b(1g tosJd di% can oh +ruuse. wLu (LM fJJ ue un, LS uh4de. "'d 9mh only on hoist rcice unti3 eleavad -

2. Bottom Core Slee Zone-(LS-10) te=ded 46 gr.rrv w on Am whm fiAss. n Q y d' M %Livlly/vridge m.c'ramant /.

[do mas

3. C rip:;'e r tm'uc up (LC 3) gi e v Ein,s _

siman% +r+4trsIng W hoidag ceudb.(LR-7T ri.F dir -sed all ows bridge movement trun%

4. Fuel transfer syste. center _ine trrensfer sy-+^-

5. /M c..tk' bIoc hi Cw bNb REFERENCE VEGP, Training Text, Vol. 8, Ch. 18a, PP. 13-15 ANSWER 6.11 (1.00)

1. Anti-siphon hole in return line.

2. Suction lines enters pool near normal water level.

REFERENCE VEGP, Training Text, Vol. 8, Ch. 18b, P. 2

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

.Pago 35 6.~ PLANT'SYST2MS DESIGN. CONTROL. AND' INSTRUMENTATION 6.1

-ANSWER [.5a4 $~2rynf (1.50)Q o.G d]

1 '. AFW auto start &m883 asso

f. He94M (o.t ed.)

n , , . , ,Tu=V I n

2. High energy line break monitors sOR [ 7{ { }[g* .

3. Radiation monitor-RE-2112n52 [0. 5 c r-h]t 4

REFERENCE He<tLM d busns 5". Wu Tw from Aux stad dittA pedifeer

-VEGP, Training Text, Vol. 7, Ch'. 13c, Fig. 13c-1, PP. 4, 6-ANSWER 6.13 (1.50)

1. Pressurizer pressure low pressure safety injection.

2. Steamline low pressure safety injection.

-3. Steamline isolation.

@ 0.5 points each.

REFERENCE VG, RO-LP-213, p 12; VG, RO-LP-215.

ANSWER 6 (1.50) gg.A gy vdf,3 q 4 p o.z r d-z.,

All six valves close to limit overspeed [0.t5] with 1, 2, 3 intercept valves Then valves 4, 5, 6 reopening'immediately following load reject [0.5].

reopen when 1 2, and 3 reach full open [0.5].

REFERENCE VEGP Training Text, Vol.7, Chapter 14c., p. 6 ANSWER 6.15 (1.00)

1. Orifice isolation valves shut. [0.5]

2. Pressure level > 17% (No pressure low level). [0.5]

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

Pega 36

~6. 4 PLANT SYSTEMS DESIGN. CONTROL. AND' INSTRUMENTATION REFERENCE VEGP,.Lo Lesson Plans,. LO-LP-09101~-00, P. 11 w

' ANSWER 6.16 (2.00)

1. Manual 1/2.

2. Power range high'setpoint' neutron flux 2 4.

-3. Power ~ range neutron flux, high positiv egative rate 2/4.

l4. Overtemperature deltaT 2/4.

'4.

p4

5. Overpower deltaT 2/4. - 9* I'Weh8 N

6. High. pressurizer pressure 2/4.

7..SG lo-lo level 2/4 in 1/4 SGs.

8. Safety injection 1/2.

((p;125. each,.- name and coincidence must be- correct)

L 8 e49 qu e. u(@

REFERENCE VG, TS, Table 2.2.1 Mic. 3. s -1 ANSWER 6.17 (2.'00)-

a. (@2185) COMS shuts'PORV [0 ] andPORVBlockValve[0.k.3kM' a, q) 'd ^'#"

- coms w.ckcJ .- CW us no est WK temperature,-inputs to the pressure programmer an is

.b.fThelowKUdthen compared to RCS WR pressure [0.5) to produce an error signal twhich wi11 nnne COMS onerntinn [0.5].

Auc.bn e-..t W ecs WE-%+eadure.y93 4 =~

4 S 4 p.4=F. co.5s]

c Jp3 to Wove. pw .4 (

REFERENCE VEGP,' Training: Text, Vol. 5, Ch..lc, PP. 12-13 ANSWER 6.18 (2.00) p g

1. P-8, 2/4,Blockssinglelooplossofflowbelowh9YPWR.

. Pr4C #

2. P-9,,2/4, Blocksreactor. trip'onturbinetripbelow(50%)PWR.

3. P-10, 2/4, Backup to P-6 for Source Power, allow blocking of some reactor trips (auto instates),

e:ea ski

'^

[ Permissive,ilogic 0.3% each, function 0.33 each]

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

e_ _ , . .

Pcga-! 37.

16.

PLANT SYSTEMS DESIGN.~ CONTROL. AND INSTRUMENTATION.

y g..

LREFERENCE

VEGP, Training. Text ANSWER 6.19 (1.50)

-(6_@ 0.25 each]

1. jResets slave cycler' logic

2. . Resets' memory buffer

3. Resets' oscillator failure

4. Calls for1fullicurrent

5. ' Resets phase.

6. -Resets regulation.

~

7. Resets multiplexing.

8. Resets-logic failures.

. REFERENCE

-VEGP, LO' Lesson Plans, LO-LP-27102-00, P. 7 ANSWER. -6.20 (1.00) 1.:overtemperature deltaT within 3% of trip setpoint.

~

2. overpower'deltaT-within 3% of trip setpoint.

3. . circulating water pump trip with plant load above 50%.

(0.33 each - both signal and setpoint must be correct).

REFERENCE .

VG, 18012, p 1.

(***** END OF CATEGORY 6 *****)

m._L..

Pcga 38

'7.' PROCEDURES - NORMAL. ABNORMAL. EMEEGENCY AND RADIOLOGICAL CONTROL ANSWER 7.01 (1.00)

At RWST level of approximately 16%, OR upon receipt of RWST Empty alarm. (Either answer accepted for full credit.)

REFERENCE VEGP, Exam Bank, 531NA ANSWER 7.02 (1.00) b REFERENCE VEGP, 00920-C, p.6.

ANSWER 7.03 (1.00)

a. No.

b. No. [0.5 each]

REFERENCE VEPG, EOP Training Text, 10A-8.

ANSWER 7.04 (1.00)

1. Following a refueling outage.

-2. When the actual critical rod position is outside the ECC calculated window.

[0.5 each]

REFERENCE VEPG, 12003-1.

(***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)

m._ _ -.s- -.

'

7. PROCEDURES- - t[QBtfAL . ABNORMAL . EtiERGENCY-Paga 39-

.AND RADIOLOGICAL CONTROL

i -

ANSWER- 7.05- (1.'50)

- Any 3 @-0.5". points each.

1. SG level in NR { prevents steam in FW bypass)

2. Maintain forward flow through bypass feed nozzle.

3. Maintain continuous flow in lieu of intermittent flow.

4.-Refill piping at low flow rates (45 gpm) if steam-or hot water has entered the piping.

_ REFERENCE

.VEFG,-13610-1, p.2.

' ANSWER 7.06 (1.50)

i. h ( < j g N D'

~

l1. Subcooling monitor < 28 F [0.75]

-2. W% -

2. Pressure level cannot be maintained above 4%. o. s Ito p

[NI Rf Of'u8 %6

[34% ADVERSE cont.] [0.75] 3 L Tp ,[*3cc.

oe Y. by ~* b %n REFERENCE __ % ra.fe (c.s d,g VEGP, EOP- 19000-1, Foldout pages ANSWER 7.07- (1.00)

a. There is no makeup to VCT without instrument air. [0.5]

~b. To prevent loss of suction to CCP. [0.5]

REFERENCE VEGP, EOP 19011-1, ES-1.1, P. 7

(***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)

L

f f 7.

PROCEDURES - N_QRMAL. - ABNORMAL, EMERGENCY - Paga.40 611D_86DIOLOGICAL~ CONTROL i-ANSWER 7.08' -(2.00)

[0.5 each] _

a. One

b. When' water level < 23 ft. above reactor vessel flange.

c. 1.5% Delta-K/K

-d. 2 Hours or (o ./,k. 6. ;.-l ., dans/4 / Af4 gond 4. Cffen,y/,-

REFERENCE-VEGP, Procedure 12000-1 Precautions, PP. 1 and 2 VEGP, Tech. Specs. 3.4.1.4.1, 3.4.1.4.2, 3.9.8.2 3

i' ANSWER 7.09 ~(1.00)

With S/G temperature hi,gher [ lower] than,RCp temperature, a pressure' rise

~

[ drop] is expected.---

REFERENCE VEGP, Unit Operating Procedure, 12001-1, P. 15:

ANSWER- 7.10 (1.00) ,

435 +/ -15 [0.5] and 925-+/- 50 psig [0.5] acceptable range

also accept - prior to exceeding 450 F and 1000#-

REFERENCE VEGP, Unit Operating Procedure, 12002-1, P. 10

' ANSWER 7.11 (1.00)

Steam line pressure > 585 To prevent safety injection [0.5 each]

(***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)

9 i

, . - - ,--.v-,----, -m , ~ - - - p . -. e.

Pcge 41

' T* 7 '. U PROCEDURES - NORMAL.~ ABNORMAL. EMERGENCY L~_~*' &ND RADIOLOGICAL CONTROL-e -

, REFERENCE

.VEGP, Unit;OP-Procedures, 12002-1, P. 13

-ANSWER- 7.12; ( l'. 00 )

Verify RCS pressure stable or. rising.

REFERENCE ~

VEGP,_' Abnormal Procedures 18001-1, P. 5

' ANSWER- 7.13 (1.00)

1. Loss of vacuum drag' capability. [ 0 .' 5 ]

2. Loss of cond. vacuum. [0.5]

. REFERENCE-

-VEGP, AOP 18008-1, P. 3

' ANSWER 7.14 (1.00)

Imminent loss of pressurizer level.

. REFERENCE

-VEGP,-AOP 18009-1, P. 2

' ANSWER. .7.15 (1.00)

1. Turbine control valves closing.

2. Generator Load lowering.

3. Loss of TURB~ LOAD INTLK C7 STATUS LIGHT - Energized. [0.33 each]

(***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)

j

Pegs 42...

, *'7.'lPROC$DUBES - NORMAL. ABNORMAL'.' EMERGENCY-

1. * &ND RADIOLOGICAL CONTROL 1

REFERENCE-

-VEGP,jAOP 18012-l', P. 2.

ANSWER.' 7.16' - (1.50)

'Any[3.@ 0.'5 points each:

~1. Proper group alignment.

2.'DRPI and group demand in agreement.

3.- Proper bank overlap. '

4. Count. rate stabilization.

REFERENCE

.VEPG,~12003-1,'p.9.

4.

ANSWER. 7.17 (1.00)-

1. Depress both Em. stop pushbuttons to trip the B diesel.

REFERENCE VEGP, AOP 18031-1, P. 2 ANSWER- 7.18 (1.50)

[3-@ 0.5 each]

1. Shift' Supervisor

2. Reactor Operator

-. 3 . Extra Shift Personnel REFERENCE AOP 18038-1, P. 4 ANSWER 7.19 (1.00)

If-' staying will hinder personnel safety or not.

(***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)

I. -

A ,,- .._..,,.-_.-..,.,m,, -

p-

7. ' PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pags 43 AND RADIOLOGICAL CONTROL REFERENCE VEGP, AOP 18038-1, P. 2 ANSWER 7.20 (1.00) z, All SG levels less than NQs narrow range -(27 percent fut udversc a_

contairmer_t)q[0.5] AND AFW flow is less than 550 pgm [0.5].

REFERENCE VEGP, Question Bank, Section 7, Question #42NB ANSWER 7.21 (1.00)

No. [0.25] Because the total cooldown required is < 100 F [0.75]

(thus if cooldown is terminated and temperature held cooldown will not exceed 100 F in one hour) . or cooldom should not be //m%f /y> 7;M..G,e. (f.g]

REFERENCE VEGP, NRC Exam Bank Question #217NA Procedure 19030-1, P. 11 Steam Tables ANSWER 7.22 (1.00)

To prevent initiation of AFW flow to the ruptured steam generator.

REFERENCE VEGP, EOP, 19030-1 (E-3), P. 11 ANSWER 7.23 (0.50)

FALSE REFERENCE VEGP, EOP 19011-1 ES-1.1, P. 2

(***** CATEGORY 7 CONTINUED ON NEXT ? AGE ** ***)

Page 44

7.

PROCEDURES - NOEMAL _6BNQEUAL1_ EMERGENCY AND RADlQLQQICAL CONTHQL ANSWER 7.24 (1.00)

a. 6 days

b. Upon notification to HP by work group supervisor (or work planning group).

REFERENCE VEGP, HP Procedure 43007-C, P. 2 ANSWER 7.25 (1.00)

a. FALSE

b. TRUE REFERENCE VEGP, HP Procedures Form 43001-1, Part 13 ANSWER 7.26 (1.00)

a. Shift Supervisor AND Fuel Handling Engineer

b. By use of key bypass switch on the raising hoist.

REFERENCE VEGP Refueling Procedure, 93210-C, p. 1

(***** END OF CATEGORY 7 *****)

gg... ,

k.k ..k $bl Pcgs:45.

d78. *~ ADMINISTRATIVE PROCEDURES. CONDITIONS".

..J ;':" AND LIMITATIONS CANSWER ' 8. 01, .(1.00)'

c

REFERENCE

~

.VG, TS, p 3/4'3-6.

' ANSWER 8.02 (1.00) d.

. REFERENCE VG,.TS, p 3/4-2-11.

ANSWER- 8.03 (1.00)

a. :340,000 gallons (or accept 68%)

b. EnoughLwater to maintain hot' standby conditions for four-hours +k- - - + -

followed by.a cooldown to where RHR can.be initiated.'-

><..u.,,, .

REFERENCE VEGP,~ Exam Bank Question # 46NB ANSWER' -8.04 (1.00)-

.a. Ensure completeness and accuracy.

b. -Signing and dating. (signature required, initials will not be accepted.) [0.5 each]

REFERENCE.-

.VEGP,-OPS Admin. Procedure 1000-C, P. 21

(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)

Pega 46 8'.

ADMINISTRATIVE PROCEDUEES. CONDITIONS.

'AMD_ LIMITATIONS ANSWER 8.05- (2.00)

1. Shall remain in Containment [0.5] on fuel handling floor [0.5].

2. Have no other concurrent responsibilities. [1.0]

REFERENCE VEGP, OPS Admin. Procedure 10003-E, P. 1 ANSWER 8.06 (1.50)

1. Relieving operator knowledgable of plant condition.

2. Both operator perform joint walkdown of boards.

3. Permission granted by SS.

REFERENCE VEGP, OPS Admin. Procedure 10003-C, P. 3 ANSWER 8.07 (1.00)

Manually illumine all SSMP lights for systems that are potentially rendered inoperable.

l REFERENCE VEGP, OPS Admin. Procedure, 10005-C, P. 1

/.00 ANSWER 8.08 (Er00)

a. Condition Er

=b. A Shift Opc. Supervisor. (de/ed REFERENCE VEGP, OPS Admin. Procedure 10006-C, PP. 1 and 3

(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)

Pega 477 i' 8.E iADMINISIRATIVE PROCEDURES. CONDITIONS.'-

-": &BD LIMITATIONS L

- ; ANSWER' 8.09 (2.00)

[0.4 each]

1. -Operations Supervisor

2. Superintendent'of Plant Engineering and Services

- 3 '. Superintendent of Maintenance

- 4. Operations Superintendent

5. Maintenance Supervisor REFERENCE

- VEGP, Operations Admin.- Procedure. 10007-C g ANSWER 8.10 . ( 1. 00 ) :

Consult with HP department [0. 5] 1x) determine if escort'is required

' [i0.5].

REFERENCE b VEGP, Plant Administration Procedure 00008-C, P. 8 L

' ANSWER'- 8.11 (1.00)

Controlled key [0.5],- Priority key [0.5]

-REFERENCE VEGP, Plant Administration Procedure 00008-C, P. 11 ANSWER 8.12 (1.00)

-C

(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)

'r[ - -

8? Pcga 48 ADMINISTBATIVE PROCEQUEES. CONDITIONS.

" 'A N D L I M I I A T I O N S REFERENCE VEGP, Plant Administration Procedure, 00005-C, P. 1 ANSWER 8.13 (1.00)

a. 1. Cease activities associated with the performance of the procedure

2. Contact his supervisor.

b. 1. Ensure associated plant systems and equipment are restored to conditions, such that plant and personnel safety are not jeopardized.

2. Revise the procedure (IAW 00051-C or 00052-C) prior to resuming associated activities.

REFERENCE VEGP, 00054-C, P. 2 ANSWER 8.14 (1.00)

Either he is contacted by phone [0.5] or if not contacted, he is notified upon his return to the site [0.5].

REFERENCE VEGP, Plant Administration Procedure 00304-C, P. 4 ANSWER 8.15 (2.00)

[0.5 each]

1. The safety of the reactor is in jeopardy.

2. Operating parameters exceed any of the reactor protection circuit setpoints and automatic shutdown does not occur.

3. Shutdown is required to protect personnel, equipment and the public.

4. Warranted by unusual circumstances.

(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)

' ' 8 '-- ADMINISTRATIVE ' PROCEDURES. ~ CONDITIONS.

Pags'49 S'" .hND LIMITATIONS REFERENCE. .

~VG, AP 10000-C, p 13'.

-ANSWER ~ -8.16 .(1.00) ho{ b rgvWd

1. a-change in-the purpose % scope-of the affected procedure

2. Cha.nge. nn <

3K.~ a change to plant administrative procedures e crerscncy oper=+1ng

~

e---.

'l b. , a reduction in the established level of safety (as adNised mTec4. Spec.uJ r~s4

{ y. a change in acceptance criteria that is less conservative ' than previously established

. gaug% m c.onbis e2 446skal in plad adme'n peedwus

~ kg. a cher.ge to Twh Cpece er the FSAR.q REFERENCE VG, AP 00052-C, pp 1,7.

ANSWER 8.17 (1.00)

1. stated in the individual spec., and

2. surveillances do not have to be performed on inoperable equipment.

REFERENCE VG, TS, p 3/4 0-2.

' ANSWER 8.18 (1.00)

[0.5 each]

a. FALSE i b. TRUE

(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)

y- .

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-I87~" ADMINISTRATIVE PRQCEDURES. I CONDITIONS. .- Pega~50

4: 'AND LIMITATIONE J

I LREFERENCE'.

.VEGP,! Exam { Bank Question' # 530NA-

,y ANSWER'- -

8.19

(1~.50)~

[ 0'. 5 e a'c h ]

1. Alert

2. : Site area

-3. . General emergency REFERENCE-VEGP, Exam Bank Question-# 52NA

~ ANSWER- 8.20- (1.00)

. a .- 3 --

'b.

2

.. c. 4-

d. 1 REFERENCE

-VEGP, Emergency. Response Procedure 91002-C, P. 3 ANSWER '8.21 ~- ( 1. 00 )~

1 ~. Recognize changes'in: plant conditions----[0.5]

AND-

2. Advising the Emergency Director concerning classification'of events. [0.5]

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.(***** CATEGORY 8 CONTINUED ON NEXT PAGE **-***)

,r, ,, e,r.-, ,- ,, -r , , , , _ ,

Pc:ga 51 Ef." ' AQULNISTBATIVE PROCEDURES. CONDITIONS.

o &ND LIBIT6TIONS REFERENCE VEGP, Emergency Response Procedure 91001-C, P. 1 ANSWER 8.22 (2.50)

[5 @ 0.5 each]

1. Classifying and declaring the emergency, including upgrading, downgrading or termination.

2. Recommending protective actions to offsite authorities and content of messages.

3. Authorizing personnel radiation exposures in excess of 10CFR20 limits, if necessary.

4. Deciding to evacuate non-essential personnel from the site at the Alert classification level.

5. Deciding to request sssistance from federal support groups.

6. Deciding to notify offsite authorities responsible for emergency measures.

REFERENCE VEGP, Emergency Response Procedures, 91102-C, P. 2 ANSWER 8.23 (1.00)

1. VP Nuclear Ops. [0.5]

2. General Manager, Vogtle Nuclear Ops. [0.5]

REFERENCE VEGP, Emergency Response Procedure 91102-C, P. 1 ANSWER 8.24 (1.00)

a. [0.5)

"A" because he is over 45 with the most exposure in bank due to 5(N-18) rule. [0.5]

\

(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)

2. . . '

78 " ADMINISTRATIVE' PROCEDURES. CONDITIONS, -:Paga:52-

4 ~' '. 9tND' LIMITATIONS >

4

' REFERENCE.

. VEGP Emergency Response Procedure, 91102-C,-p. 7-rs

.(***** END OF CATEGORY 8 *****)

(********** END OF EXAMINATION *4********)

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U. S. NUCLEAR REGULATORY COtt!ISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: VOGTLE 1 REACTOR TYPE: PWR-WEC4 DATE ADMINSTERED: 86/10/21-EXAMINER: JENSEN. N.

CANDIDATE U iii.o MDV IYIMJILII WI I 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 704 in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after-the examination starts.

% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTA _L_ SCORE VALUE CATEGORY-27.50 22.73 1. PRINCIPLES OF NUCLEAR POWER

! PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 32.00 26.45 2. PLANT DESIGN INCLUDING SAFETY

!- AND EMERGENCY SYSTEMS l

l 31.00 25.62 3. INSTRUMENTS AND CONTROLS l

30.50 25.21 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL-CONTROL 121.0 Totals All work done on this examination is my own. I have neither given nor received aid, j

i

! Candidate's Signature l

J

o NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS 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.

2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.

3. Use black ink or dark pencil only to facilitate legible reproductions.

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

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

6. Use only the paper provided for answers.

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

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

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

10. Skip at least three lines between each answer.

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

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

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

l

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

[

E 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE I

QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

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

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

work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has i been completed.

A

+# .

A

18. When you complete your examination, you shall:

a. Assemble your examination as follows:

(1) Exam questions on top.

(2) Exam aids - figures, tables, etc.

(3) Answer pages including figures which are part of the answer.

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

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

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

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'1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Page 4 M THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.01 (1.50)

True or False?

a. The faster a centrifugal pump rotates, the greater the NPSB required to prevent cavitation.

b. One of the pump laws for centrifugal pumps states that the volumetric flow rate is inversely proportional to the speed of the pump.

c. Pump runout is the term used to describe the condition of a centrifugal pump running with no volumetric flow rate.

QUESTION 1.02 (2.50)

a. After operating at 100% power for three months, power is suddenly lost to all of the reactor coolant pumps. Briefly explain how each of the following acts to ENHANCE natural circulation.

1. Maintain pressurizer level of at least 50%.

2. Maintain adequate subcooling in RCS.

3. Maintain heat sink.

b. State how the following parameters will be trending if natural circulation is lost (INCREASE, DECREASE, or REMAIN THE SAME):

1. RCS differential temperature (Delta-T)

2. Steam generator steam pressure QUESTION 1.03 (0.50)

Does the Latent Heat of Vaporization INCREASE, DECREASE or REMAIN THE SAME as saturation pressure / temperature of water is increased?

(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

1

t 1. ' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Paga 5

% THERMODYNAMICS. HEAT TRAMSFER AND FLUID FLOW QUESTION 1.04 (1.50) ho an Match the heat transfer process in Column A en the equation that applies to that process in Column B.

COLUMN A COLUMN B

a. Between cold leg and hot leg 1. k = m o Delta-T of reactor (normal forced con- . .

vection flow) 2. Q= m Delta-T

b. Across S/G tubes (primary to ,

secondary) 3. Q = U A Delta-T

c. Across S/G (feedwater to steam) 4. k=mcDelta-h

5. k=mDelta-h QUESTION 1.05 (2.00)

a. Define DNBR. (0.5)

b. Since the DNBR is not a directly observable parameter, name THREE parameters the operator monitors and/or controls to ensure the DNBR limit is not violated. (1.5)

QUESTION 1.06 (1.00)

Describe the conditions under which cavitation occurs.

(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

_ - _ _ _ _ _ _ J

i 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Paca 6 k THERH0 DYNAMICS. liEAT TRANSFER AND FLUID FMNF QUESTION 1.07 (2.00)-

The reactor is at 30E of-full power. Briefly explain how AND why each of=

the following parameters will behave during the first several minutes, fif one reactor coolant pump (RCP) is shut off with rods in manual and the turbine is in P-impulse mode.

a. Reactor coolant flow in the unaffected loops

b. Indicated reactor coolant flow in affected loop c.- Steam flow in the unaffected loops

d. T-avs of'an' operating loop QUESTION 1.08 (1.00)

Multiple' Choice OP-12003-1, " Reactor Startup", requires that the critical rod position be taken at 1.0E-8 amps on the intermediate range. If, during a Xenon-free reactor startup at MOL, the operator " overshot" 1.0E-8 amps and instead L leveled off at 1.0E-7 amps, which of the following statements is correct?

a. At 1.0E-7 amps, there are little or no ef.fects from nuclear heat but i

since-the reactor is a decade higher in power, the critical rod

! position will be higher.

l

b. At 1.0E-7 amps, there are little or no effects from nuclear heat; therefore, the critical rod position will be the same as at 1.0E amps.

l

c. At 1.0E-7 amps, there are substantial effects from nuclear heat; therefore, the critical rod positions will be higher than at 1.0E-8 amps.

j d. At 1.0E-7 amps, nuclear heat, xenon and the decade higher in power

! level will result in a higher critical rod position.

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A'

1 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Pasa 7 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.09 .(2.00)

Multiple Choice Indicate whether the following will cause the differential rod worth to- .,

INCREASE,4 DECREASE or have NO EFFECT. I

a. An adjacent rod is inserted to the same height

b. Moderator temperature is INCREASED

c. Boron concentration-is DECREASED u

.d. An-adjacent burnable poison rod depletes

~ QUESTION 1.10 (1.50)

a. Define suberitical multiplication (M).

b. During a reactor startup, count rate is 250 CPS with a corresponding-.

K-eff of 0.95. The count rate increases to 500 CPS. What is the re-sultant K-eff? Show all calculations.

QUESTION 1.11 (0.50)

Multiple Choice Concerning equilibrium Samarium-149 (Sm) reactivity, which of the following statements is correct?

a. 50% equilibrium Su reactivity is one quarter of 100% equilibrium Sa reactivity.

b. 50% equilibrium Sa reactivity is one-half of 100% equilibrium Su reactivity.

c. 50% equilibrium Sm reactivity is three-quarters of 100% equilibrium Sm reactivity.

d. 50% equilibrium Sm reactivity is equal to 100% equilibrium '

Sm reactivity.

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

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81. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Pasa 8 W- THEHH0 DYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.12 (2.00)

TWO major factors affect differential boron worth over core life. List-these TWO factors'AND indicate how (HORE NEGATIVE or LESS NEGATIVE) they affect differential boron worth.

-QUESTION 1.13 (1.00)

Multiple Choice If reactor Power increases from 1000 cps to 5000 cps in 30 seconds, what is the SUR7

a. 0.8 DPH

b. 1.0 DPM

c. 1.2 DPM

d. 1.4 DPM QUESTION 1.14 (3.00)

a. Discuss the behavior of reactor power and Tavg during and after 2 minutes of Emergency Boration at 100 % power. Assume rod control is in manual.

b. Discuss the behavior of reactor power and Tavg after 2 minutes of Emergency Boration at 10E-8 amps and no-load Tavg.

QUESTION 1.15 (1.50) ,. g ggj What is Self Shielding"d AND how does it affect reactor operations as

' power changes?

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(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) '

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1. ' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Page 9 R THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.18 (0.50)

True or' False?

One reason for overlapping rod groups is to minimize the effects of rod shadowing on TOTAL ROD WORTH.

QUESTION 1.17 (1.50)

An ECC is calculated for a startup following a reactor trip from 100%

power equilibrium Xenon (BOL). Indicate if the actual critical rod positio will be HIGHER, LOWER or the SAME from the calculated position-for each of the following situations. Consider each case separately,

a. Xenon reactivity curve for trip from 50% is used to calculata conditions to startup 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months after the trip.

b. The Steam Dump pressure setpoint is increased 100 psi.

c. The power defect curve for MOL is used instead of the BOL curve.

4 QUESTION 1.18 (1.00)

Assuming the reactor is initially operating at the optimum value of moderator / fuel ratio, will the following changes cause the reactor core to become-UNDERMODERATED, OVERMODERATED, or HAVE NO EFFECT7 Consider each change separately, with all other parameters constant.

a. Increase in moderator temperature.

, b. Inserting group D rods from 220 steps to 180 steps.

(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

e l'. 'PRTNCIPLES OF NUCLEAR POWER PLANT OPERATION. Paga-10

-% THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.19 (1.00)

Multiple Choice When performing a reactor S/U to full power that commenced five hours after a trip from full power equilibrium conditions, a 0.54/ min ramp was used. .How would the resulting xenon transient vary if instead a 2%/ min ramp was used?

a. The Xenon dip-for the 24/ min ramp would occur sooner and the magnitude of the dip would be smaller.

b. The Xenon dip for the 24/ min ramp would occur later and the magnitude of the dip would be smaller,

c. The Xenon dip for the 2%/ min ramp would occur sooner and the magnitude of the dip would be larger.

d. The Xenon dip for the 2%/ min ramp would occur later and the magnitude of the dip would be larger.

(***** END OF CATEGORY 1 *****)

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'2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pasa 11

'e BYSTEREl QUESTION 2.01 (1.50)

Valve HV-15214 in the CVCS letdown line is sometimes known as a Letdown Leak Protection Valve. List THREE of the four conditions indicative of-a leak in the letdown system, which will cause automatic closure of valve HV-15214. Setpoints not required.

QUESTION 2.02 (0.75)

How does the positive displacement charging pump (PD pump) respond to a Safety Injection (SI) signal if it was operating prior to the SI?

QUESTION 2.03 (1.00)

State the SAFETY-RELATED reason for-having incorporated the reactor vessel head vent, which removes steam and/or non-condensibles from the head area, into the Vogtle plant design.

QUESTION 2.04 (1.00)

What is the purpose of the interlock bypass switches, which are essociated with the Steam Dump System?

QUESTION 2.05 (2.00)

List FIVE separate signals / methods for starting a Diesel Generator.

i QUESTION 2,06 (1.50)

List THREE of the four permissives which must be met in order for the l diesel to accept an emergency start signal from the Control Room.

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(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)

'2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Paco 12 6 SYSTEMS QUESTION 2.07 (1.50)

State THREE of the four automatic protective trips which are NOT disabled following an emergency start of the emergency diesel.

Setpoints not required.

QUESTION 2.08 (1.00)

The function of the diesel engine after cooler is to provide cooling to:

a. intake air after_the turbo charger.

b. the jacket water cooler after engine shutdown.

c. standby lube oil heater system to limit temperature.

d. the exhaust to aid in silencing the engine.

QUESTION 2.09 (1.50)

With regard to.the component cooling water surge tank:

a. State the " Normal" and the " Emergency" makeup water sources to the surge tank. (1.0)

b. Which makeup water source can be automatically controlled by surge tank level? (0.5)

QUESTION 2.10 (0.75)

State the THREE loads which are cooled by the Component Cooling Water (CCW) system.

QUESTION 2.11 (1.00) l l State the purpose of the mini-flow recirculation lines installed between i the downstream sides of the RHR heat exchangers and the suctions of the RHR Pumps, t

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(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)

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i' 2. PLANT DESIGN INCLUDING SAFETY AND EMERQENCY P o '13 i .. SYBTEMS 4 QUESTION _ 2.12 (3.00)

Concerning the Containment Spray System:

a. State the TWO purposes of the Bodium Hydroxide (NaOH) which is added to the spray solution'.

b. What flow path provision is made for running a pump performance test?

c. State TWO suction sources for each spray pump. (Do not include the eductors.)

QUESTION 2.13 (2.50)

Where in the reactor coolant system are the following penetrations located?

NOTE: Specify Loop 1, 2, 3, or 4, AND hot leg, cold leg, or intermediate leg.

a. Normal letdown line.

b. Pressurizer surge line.

c. Normal charging lines (2 required).

d. Excess letdown line.

e. Pressurizer spray lines (2 required).

QUESTION 2.14 (1.00)

a. Where on the Reactor Coolant System loops are the flow transmitters-

. located?

b. How many flow sensors are there per loop?

4 QUESTION 2.15 (0.50)

What percentage of the decay heat removal capacity is the auxiliary feedwater system capable of, if one motor driven pump and the turbine driven auxiliary feedwater pump is out of service?

(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)

9 - < - -

y ,e-.+. -r w---- y---- '

em m w$- -

2.. , PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Paca 14 4 SYSTEMS i

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' QUESTION 2.16 (1.00)- l State TWO purposes for providing a small amount of continuous bypass flow around each pressurizer spray valve.

QUEETION. 2.17 (2.00) a.- What automatic actions does the Cold overpressure Mitigation System (COMS) take to prevent depressurization of the RCS when operating at Normal Operating Pressure?

b. How is the COMS setpoint determined during shutdown conditions?

l QUESTION 2.18 (1.00)

During a routine manipulation on QMCB-A, the reactor operator inadver-tently turns.off the-running turbine plant closed cooling water (TPCCW) pump. Briefly explain the response of the standby pump, inc.1.uding what-the pump will do AND when. ;

QUESTION 2.19 (1.00)

List four of the five liquid process monitors with automatic actions.

Giving the automatic action is NOT required.

QUESTION 2.20 (1.00)

Explain how more than one SG is prevented from blowing down if the A SG is

. faulted and the A SG main steam isolation valve fails to shut.

QUESTION 2.21 (0.50)

True or False?

During a double-ended cold leg shear, the INITIAL cooling mechanism-is provided by the injection of ECCS Accumulator water into the core.

(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)

' ' it . PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pago 15 SYSTEMS QUESTION 2.22 (2.00)

What is the purpose of the following limitations-on ECCS accumulator parameters?

a. Low water level

~b. High water level

c. Low nitrogen pressure

d. High nitrogen pressure

QUESTION 2.23 (1.00)

The plant is operating at 100% power when an electrician inspecting a downstream RCP breaker (the class 1E breaker) accidently causes the underfrequency relay for the breaker to actuate its protective function.

.Which of the following best describes the plant response.

a. The affected RCP trips. The operator performs a normal controlled plant shutdown with the other 3 loops in service.

b. The affected RCP trips, the reactor trips due to low flow in the

! .affected loop.

I c. The affected RCP trips, reactor trips and the other 3 RCPs trip.

d. The affected RCP trips, reactor trips, the other 3 RCPs trip and the l Plant sis on low low flow.

QUESTION 2.24 (2.00)

List the FIVE signals or conditions which will cause automatic closure of the feedwater regulating valves. Setpoints and coincidences required where applicable.

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(***** END OF CATEGORY 2 *****)

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

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

'3. IMBTRUMENTS AND CONTROLS _

Pasa 16-J

-QUESTION 3.01 '(1.50) 2/3 (two out-of three) pressuriser pre.ssure channels below 1970 psig (P-11) allows the operator to manually block certain protective, functions. What are the THREE protective functions blocked by this action?

QUESTION 3.02 (1.00)

What interlock requirements must be satisfied to open the letdown isolation valveag (f.V 'is9 lff,o)

. QUESTION 3.03 (2.00)

List all PERMISSIVES which receive a DIRECT input from the power range-instrument system (N41 thru N44), logic for each, if any, and briefly what

each permissive accomplishes.

QUESTION 3.04 (2.00)

a. What is the purpose of the turbine impulse pressure signal used in the S/G Level Control System? (0.5)

b. Considering only the Feedwater Regulating Valve Control portion of the S/G Water Level Control System, indicate whether feedwater flow would initially INCREASE, DECREASE, or NOT CHANGE, if the controlling S/G pressure transmitter failed high during 50% power operation. Briefly explain your answer.

l (Assume turbine feed pump speed control is in MANUAL.) (1.5) 1 QUESTION 3.05 (1.50)

State the THREE variable ple.nt parameters which provide an input to the feedwater pump speed control circuitry.

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(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

3.' INSTRUMENTS AND CONTROLS Pasa 17 QUESTION 3.06 (1.50)

Indicate whether the following statements concerning operation of the Reactor Trip (RT) and Bypass (BY) breakers are TRUE or FALSE.

u. If one train is placed in test while the other train's bypass breaker is closed, then both reactor trip breakers and both bypass breakers will trip.

b. If it is attempted to close both bypass breakers at the same time, then both bypass breakers will trip but the reactor trip breakers will remain closed.

c. A " train A" reactor trip signal will trip RTA and BYA breakers.

QUESTION 3.07 (1.50) ,

Assume the reactor is operating at 45% power with all systems in auto-matic control. For each condition listed below, give the initial direction of rod motion, AND state the initial reason for this rod motion.

a. Loop 3 Tc fails high.

b. A main steam atmospheric relief valve (ARV) fails open.

c. The turbine is ramped to 100% power at 5% per minute.

QUESTION 3.08 (1.00)

Why is a variable-gain circuit included in the rod control system power mismatch circuitry?

QUESTION 3.09 (2.00)

List FOUR of the five conditions which will cause a rod drive M-G set trouble alarm.

(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

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3.'.-INSTSUMENTS AND CONTROLS Paga 18 QUESTION 3.10. (3.00)

List the SIX rod stop control interlocks. With each, include its setpoint, i- _ coincidence (if applicable), and whether it affects manual, auto, or both modes of rod motion.

1 QUESTION 3.11 (0.50)

True or False?

The low steam line pressure safety injection signal is lead-lag compensated

'and will initiate safety injection at greater than 585 psig if the rate of pressure decrease is sufficiently high.

4 QUESTION 3.12 (2.00)

State the nominal pressure setpoints for the following automatic actions.

(Coincidence NOT required.)

1. ' Pressurizer proportional heaters full on.

! 2. PORV's open.

3. Low pressure reactor trip.

4. High pressure reactor trip.

5. Spray valves full open.

8. Safety injection.

QUESTION 3.13 (1.50)

The pressurizer PORY and block valve interlock closes the PORV's and block valves on low pressure.

! a. State the SETPOINT and COINCIDENCE for this interlock.

. b. Why is this interlock necessary?

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} (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

. 3. INSTRUMENTS AND CONTROLS .Page 19 '

QUESTION 3.14 (1.00) ,

Why are the backup heaters automatically turned on when pressuriser level is 5E above program?

- QUESTION .3.15 (1.50)

State the automatic actions which occur on Steam Generator Low-Low Water Level. ~ Include setpoint and coincidences. Do not include alarm functions.

-QUESTION 3.16 (0.50)

TRUE or FALSE 7 A break in the reference leg of the pressurizer level transmitter will cause indicated level to be lower than the actual level.

QUESTION 3.17 (1.00)

What would a low temperature alarm on TE-450 (surge line) indicate? Assume steady state plant conditions and no instrument failure.

QUESTION- 3.18 (2.00)

. State all automatic actions associated with a high radiation j signal from the following process and effluent radiation monitors. (If r none, so state.)

1. RE-2565A Containment Vent Effluont Particulate monitor.

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2. ARE-13133A Volume Reduction Rooms Exhaust Cubicle Particulate monitor.-

3. RE-0017A Component Cooling Water monitor

4. RE-12116 Control Room Intake Ventilation monitor l

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

.. 3Jr INSTHDMENTS AND COMTBOLS P c 20 QUESTION .3.19 (0.50)

True or False?

Concerning the Safeguards Sequencer:

' If a Safety In.iection occurs during load sequencing following an undervoltage (UV) condition, the sequencer will reset to the SI mode, and any loads which are already started will be shed and restarted on the SI sequence.

QUESTION 3.20 (1.00)

What is the design purpose for tripping the 230-KV circuit breakers upon sensing a reverse power condition on the main generator?

State the detrimental effect AND at least one cause of this effect.

QUESTION 3.21 (1.00)

The Cold Overpressure Mitigation System (COMS)-provides overpressure protection only if the system has been armed,

a. What signal (or condition) is necessary for the COMS system to be armed, assuming a normal plant shutdown?

b. What component (s) other than the PORY itself is affected by the arming signal?

j QUESTION 3.22 (1.50) l List the THREE Containment Isolation Phase-A (CIA) actuation signals.

I Include coincidence. No setpoints required.

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END OF CATEGORY 3 i

n .,,,.,,,.,,,,,,,,,,,..,_.,,--,...,,_.n,,_,,nn-~n., ,.,,--.,_-,----v,.,,---_,,a

l- , 4. PROCEDURES - NORMAL. ABNORMAL. EtERGENCY Pag) 21

- AND RADIOLOGICAL CONTROL I

QUESTION 4.01 (1.00)

According to System Operating Procedure 13003, " Reactor Coolant Pump Operation":

What provision must be met in order to attempt TWO successive starts of a reactor coolant pump? (1.0)

QUESTION 4.02 (2.50)

Indicate whether or not the following systems / components are Tech.

Spec. related. State either YES or NO for each item.

a. Turbine overspeed protection.

b. Condenser steam dump system.

c. Reactor makeup water pumps.

d. Steam generator atmospheric relief valves.

e. RE-12442 (A, B, 0) plant vent monitors.

QUESTION 4.03 (1.50)

Technical Specification 3.3.3.6 deals with accident monitoring instrumentation, and is applicable in Modes 1, 2, and 3.

List THREE of the four temperature measurements required to be operable by this Tech Spec.

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. . . .. = ,

4.'. PEDCEDUBES - MOEMAL. ABMOBHAL. EMERGENCY Page 22

-. AND RADIOLOGICAL CONTROL QUESTION. 4.04- (2.50)

During performance of routine surveillance, it is determined that the

-reactor coolant system total leakage rate is 2.2 gallons per minute, distributed as follows:

0.43 gym through S/G #1 tubes.

0.18 gym through S/G #3 tubes.

0.31 spa through S/G #4 tubes.

0.15 gym through Loop #1 RTD manifold hot leg-isolation valve packing The remainder of the leakage is unidentified.

Describe any technical specification limiting condition (s) for operation which is/are exceeded by these leak rates.

QUESTION 4.05 (1.00)

What are the THREE 10CFR20 quarterly limits for ionizing radiation for individuals in restricted areas? (Assume previous dose history not

!- - available.)

QUESTION 4.06 (0.50) i TRUE or FALSE?

A teletector 6112B CAN be used to detect the presence of beta radiation.

QUESTION 4.07 (1.00)

! The following limitation is taken from Unit Operating Procedure 12003,

" Reactor Startup":

"The boron concentration in the pressurizer should not be different from the RCS by more than 50 ppe. Pressurizer backup heaters may be energized as necessary to equalize the boron concentration."

How does energizing pressurizer backup heaters equalize boron concentration?

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, 4.

PEOCEDURES - NORMAL. ABMORMAL. EMERGEliCY Paga 23 AND RADIOLOGICAL CONTROL QUESTION 4.08 (2.00)

According to UOP 12004, what TWO requirements must be met prior to placing rod control in automatic? Assume power ascension in progress from low power.

QUESTION 4.09 (0.50)

Multiple Choice The maximum time that one operator may relieve another from his duties using "Short Term Relief", per Procedure 10004-C, Shift Relief, is:

a. 5 minutes

b. 15 minutes

! c. 30 minutes

d. 60 minutes

(

QUESTION 4.10 (1.00)

According to Limitation 2.2.9 of Unit Operating Procedure 12001, Unit Heatup to Hot Shutdown, the maximum allowable PRIMARY-to-SECONDARY pressure differential is a paid and the maximum allowable SECONDARY-to-PRIMARY pressure differential is b paid, during plant operation or leak tests.

QUESTION 4.11 (1.00)

Concerning the reactor coolant pumps:

a. If the No. 1 seal leakoff flow suddenly increases to >5 gpe, the No. 1 seal leakoff must be isolated promptly (<5 min.).

For what reason is this done? (0.5)

b. The pump must be stopped within minutes after the No. 1 seal leakoff is isolated. (0.5)

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. 4.* PEOCEDURES - NORMAL. ABMORMAL. EMERGENCY Pasa 24-AND BADIOLOGICAL CONTROL QUESTION .4.12 (2.25)

Step 2 of the turbine runback procedure (AOP 18012) requires the operator to " Verify that a runback has initiated". According to the rest of this step, how does the operator perform this verification?

(i.e. What THREE items must he check and what should be their conditions?)

QUESTION 4.13 (2.00)

State the immediate operator actions required for a turbine trip below P-9, according to Abnormal Operating Procedure 18011. Include both the " Action / Expected Response" requirements AND the " Response Not Obtained" requirements. Assume control rods in automatic.

QUESTION 4.14 (1.50)

Immediate action Step 2 of EOP 19211 "FR-S.1 Response to Nuclear-Power Generation /ATWT", states to verify the turbine is tripped, and if it is not, to manually trip the turbine. State the required '

operator actions if the turbine will NOT trip.

QUESTION 4.15 (2.00)

State the RCP TRIP CRITERIA, as delineated on the foldout page of EOP 19000, "E-0 Reactor Trip or Safety Injection".

l QUESTION 4.18 (0.75)

Step 8 of EOP 19001, "ES-0.1 Reactor Trip Response" directs the operator to transfer the condenser steam dump to the STEAM PRESSURE mode. According to the " Response Not Obtained" column of this step how is the operator directed to control S/G pressure if the condenser is NOT available?

4

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i

4.'* PROCEDUBES - $10BMAL. N . EMERGENCY Pasa 25 AND BADIO14GICAL CONTBOL QUESTION 4.17 (2.00)

.The following question refers to EOP 19001, "ES-0 1 Reactor Trip .

Response":

-The " Response Not Obtained" column of Step 9 states that "IF a RCP can NOT be. started, THEN verify natural circulation using Attachment D".

State FOUR of the five conditions listed in Attachment D which support-or indicate natural circulation flow.

QUESTION- 4.18 (1.00)

Place the following Critical Safety Functions in their correct order-i of priority, listing the highest priority FIRST and the towest priority LAST.

a. Heat Sink

b. Containment

c. Core Cooling

d. Suboriticality

! e. Integrity QUESTION 4.19 (0.50) i j True or False?

When operating from the Remote Shutdown Panels, if an SI actuation -

occurs, any components previously transferred to the Shutdown Panels

must be MANUALLY actuated by the operator.

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, 4.. PRDCEDURES - NORMAL. AM OMGAL. EREIGENCY Pcca 26 AND RADIOLOGICAL CONTROL QUESTION 4.20 (2.00)

While in the initial RCS cooldown during performance of EOP 19030, "E-3 Steam Generator Tube Rupture", you observe that the Technical Specification limitation on RCS cooldown rate is being exceeded. The Shift Supervisor directs you to reduce cooldown rate to within Technical Specification limits.

1 +-Is the Shift Supervisor CORRECT or INCORRECT in this

--circumstance? Briefly explain; QUESTION 4.21 (1.00)

Multiple Choice i

The control room operators are performing EOP 19222, " Response to Degraded Core Cooling", in response to an orange path condition shown on the core cooling critical safety function status tree. Which one

. of the following statements is correct in regards to transitions out of this procedure?

a. The operators taust leave this procedure before completion and go to EOP 19212-1, " Response to Loss of Core Shutdown", if the suboriticality status tree indicates a yellow path condition.

I b. The operator may leave this procedure at any step'as soon as the core cooling CSF is satisfied (Green).

I

c. The operators must leave this procedure before completion and go to EOP 19231, " Response to Loss of Secondary Heat Sink", if i the heat sink CSFST indicates a red path condition.

(

i

d. The operator must leave this procedure before completion and go to EOP 19251, " Response to High containment Pressure", if :

( the containment CSFST indicates an orange path condition.

P QUESTION 4.22 (1.00)

When is the containment spray system aligned for recirculation from the containment sump following a LOCA?

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

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)

, 1.' -PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PcCa 27 TEERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWER 1.01 (1.50)

a. True

b. False

c. False [0.5 each]

REFERENCE Westinghouse Thermal Science, Chapter 10, Pp. 41-49 General Physics HT & FF, Pp.319-326.

ANSWER. 1.02 (2.50)

a. 1. To ensure that no vapor (steam) pockets form (in the loops), OR

2. To prevent vapor (steam) pocket formation /gre

( vijn M((menf

3. To help assure adequate thermal driving head invenfory) 00 m*in fairt

b. 1. Will increase (exceed 100% full power value)

2. Will decrease [reswruer [ressure condre /.

REFERENCE General Physics HT & FF, Pp.356-357.

3.4 000 015 EK 1.01 4.4 i

! ANSWER 1.03 (0.50) i decrease REFERENCE i

CNTO, " Thermal / Hydraulic Principles and Applications, I", pp 2-58/59 Steam Tables 000/027; EK1.03(2.6/2.9) l l

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. L._* EBINCIELES_DE_HUCLEAB_EONEB_ELANT_0EEB& TION 2 Pega 2D IBgBdQDYNed1CS,_HgeI_IBeNSEEB_8NQ_ELylp_E6gy ANSWER 1.04 (1.50)

a. 1 (or 5)

b. 3

c. 5 CO.5 ea3 REFERENCE General Physics, HT & FF, pp. 180 and 181 CNTO, " Thermal / Hydraulic Principles and Applications, I", pp 4-61/64 & 5-47 OO1/OOO-K5.45 (2.4/2.9)

ANSWER 1.05 (2.00)

a. DNBR = Heat flux (power) to cause DNB / actual heat flux (power) (0.5)

b. 1. Pressurizer Pressure

2. RCS Tavg

3. RCS Flow

-(Will also accept Rx Power, AFD, QPTR, Rod sequencing / position /over-lap, others on a case-by-case basis.) E3 e 0.5 ea3 REFERENCE VCS, TS, pp. 3/4 2-15 and B 3/4 2 5 and B 2-1.

General Physics HT & FF, P.262.

CNTO, " Thermal / Hydraulic Principles and Applications, II", pp 13-20/24 VEGP TS 3/4.2.5 015/020; K5.09(3.5/3.7) l l

ANSWER 1.06 (1.00) l Cavitation occurs when the pressure of the fluid is reduced to a press-l ure below that of its saturation pressure for a given temperature which I will allow boiling to occur.

I REFERENCE l General Physics HT & FF, P.319.

1

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.LII_EBING1ELES_DENWRLE86_EQWEB_EL8HIQEEB8I1964 Pcga 29 IBEBdQDYN8 MIGS 4_UE8T TB8NSEEB_8ND_ELWID_ELQW ANSWER 1.07 (2.00)

.a. Flow will increase [0.253 due to head loss caused by lost RCP (backflow in the idle loop) [0.253 OR - Dve le d'e crewse /H core della - P E a 2s.1 -

b. Indicated flow will initially decrease to zero as the RCP coasts down

[0.253 but will increase due to backflow [O.253

c. Steam flow will increase [0.253 to account for affected steam generator [0.253

d. Tavg will increase initially then decrease [0.253 because load remains constant [0.253 REFERENCE General Physics HT & FF, Pp. 322-329, 264-266.

3.4 000 015 EK 1.04 2.9 ANSWER 1.08 (1.00) b REFERENCE CR, OP-210, Rev 16, p B; NUS, NETRO, Unit 6; W, RP, sect 3&53 VG, APNPP, Vol 2, p 5-17.

ANSWER 1.09 (2.00)

a. Decrease

b. Increase

c. Increase

d. Increase [0.5 ea3 REFERENCE SON /WBN License Requal Training, " Core Poisons" CNTO, " Reactor Core Control", pp 6-22/28 VEGP Training Vol. 2, Chapter 4, pp 103-113 001/0003 K5.09 (3.5/3.7) & K5.02 (2.9/3.4) & K5.10 (3.9/4.1)

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,L1 _EBING1ELEE_DE_NWGLE88_EQUEB_ELONI_QEEBBIIQN2 Pup 30'

-INEBNQDYNONIGE4_HEGI_IB0NSEEB_8ND_ELWID_ELQN ANSWER ~ 1.10 (1.50)

a. Suberitical' multiplication (M) is defined as the ratio'of the total number 1of-fission and source neutrons to the total numbe trons which.would exist due to the source only; OR #= q{, of og neu-

b. CR2/CR1 = (1-Kef f 1) / (1-Kef f 2) [O.53 I 500/250 = 1 .95/(1-Keff2) /-Mg y (OR any Keff2 = 0.975 [0.253 anfWer MblCb REFERENCE BVPS Reactor Theory Manual Chapter 5 pp 36 - 43 kescrifllive Corrteb kScribtS SMNf5 I'"

//g/jg,f,, gg jgg/gg VEGP Training Vol. 2, Chapter 5, page 9 consbe/g l or inerte$I Mtahek levels, Kart las %n unityr ANSWER 1.11 (0.50) 4 f"YC tr d lke ref eitterot CoWrce ned roms En bbt VICI!!if REFERENCE oh llte hef.

NUS, Nuclear Energy Training - Reactor Operation, p. 10.5 4 Westinghouse Reactor Physics, pp. I-5.77 - 79 VEGP, Training Text, Vol. 9, pp. 21-81 & 82 HBR, Reactor Theory, Session 37, p. 4 ANSWER 1.12 (2.00)

1. Boron Concentration Decreases CO.53 - MORE NEGATIVE [0.53

2. Fission Product Buildup EO.53 - LESS NEGATIVE E0.53 REFERENCE Westinghouse Nuclear Training Operations, p. I-5.31 VEGP Training Vol. 2, Chapter 4, pp 122-124 ANSWER 1.13 (1.00)

d. (Caclulated value is 1.3986 DPM... rounds to 1.4 DPM)

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v

.*-. is_LEBINGIELEH_DE_NWGLEGB_EQWEB_ELONI_QEEBeI1QN2 Pcco 31 INEBMQQYN001Gai_NE8I_IB8MBEEB_8ND_ELWID_ELQN REFERENCE VEGP_ Question Bank, Question #47BNA ANSWER 1.14 (3.00)

a. Power decreases initially due to the boron addition. CO.53 The primary to secondary mismatch causes Tavg to decrease.

CO.53 The decrease in Tavg inserts positive reactivity and restores reactor power to a slightly lower than or the same as initial power level. CO.53 Tavg will steady out at a new lower level..CO.53

b. Tavg does not change due to the boration. CO.53 (Tavg is determined by the amount of pump heat and the steam dump setting.) After the initial transient, power decreases at a negative 1/3 DPM rate to the multiplied source level. [0.53 REFERENCE MP3 Transient and Accident Analysis Chapter 4.10 VEGP Training Vol. 2, Chapter 4, pp 114-122 ANSWER 1.15 (1.50)

The large pellet diameter (relative to the resonance peak energy path) results in a self shielding effect for fuel in the pellet interior.

As fuel pellet temperature rises the off-resonance neutrons (f or the most part that would previously pass entirely through the pellet) are more readily absorbed. C1.03 Causes negative reactivity to be added as power increases. CO.53 REFERENCE BVPS Reactor Theory Manual Chapter 6 p 33 VEGP Training Vol. 2, Chapter 4, pp 31-37 ANSWER 1.16 (0.50)

False

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.+

- 123_EBINGIELEE_9E_NWGLE88_EQWEB_ELONI_QEEBBIIQN2 Pcg3 32 IHEBU9DYN051GS2 UEGI_IB9NSEEB_8ND_ELWID_EL9W 4

REFERENCE Westinghouse Nuclear Training Operations, p. I-5.36 - 43 VEGP Training Vol. 2, Chapter 4, pp. 103-113 ANSWER 1.17 (1.50)

a. Higher

b. Higher c.' Higher CO.5 ea3 REFERENCE TPT OP 1009.1; Plant Curves VEGP Plant Curves.

VEGP Training Vol. 2, Chapter 5 VEGP Exam Bank, 9BNB.

001/000; A2.07(3.6/4.2)

ANSWER 1.18 (1.00)

a. UNDERMODERATED

b. UNDERMODERATED REFERENCE l Westinghouse Nuclear Training Operations, pp. 1-5.7, 9, & 10 VEGP Training Vol. 2, Chapter 4, pp 74-78 ANSWER 1.19 (1.00) a j (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

i

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, 1[_EBING1ELES_QE_NWGLE68_EQWEB_ELONI_DEEBOI1QN2 Pcge 33

-' IHE850DYNed1GS4_UEGI_IBONSEE8_eND_ELUID ELQW REFERENCE CNTO " Reactor Core Control" Section 4 Westinghouse Simulator Trng book, "Rx Theory and Core Physics", Fig I-5-54 001/0003 K5.38(3.5/4.1)

VEGP Training Vol. 2, Chapter 5, pp. 55-50 CNTO, " Reactor Core Control", pp 4-21/28 001/0003 K5.38(3.5/4.1) 4 4

1

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, 2E__ELONI_DESIONINGLWDIN9_E9EEIY_0ND_EUEB9ENGX Pcro 34 EYSIENE ANSWER 2.01 (1.50)

1. Low pressure in letdown line.

2. High temperature in letdown HX roo . (R'AO l

3. High temperature in CVCS valve gallery [ ([-dO b.

4. High temperature in aux. building piping penetration roo (k-dO h.

[any 3 0 0.5 each3 REFERENCE VEGP Training Text, Ch. 5a, P. 35 VEGP PlID 1X @B -Ily ANSWER 2.02 (0.75)

Will continue to operate (no change). [0.753 REFERENCE VEGP, LO-LP-09202-OO, P. 7 4

ANSWER 2.03 (1.00)

The presence of steam or noncondensibles could impede natural circulation flaw y OR PenWes a safeG yude kW p aH (nr N , L {e / , G w (e REFERE E h~ '

VEGP, LO-LP-09202-OO, PP. 13, 15 VEGP, LO-LP-16001-OO, P. 9 i

ANSWER 2.04 (1.00)

Allow controlled cooldown below the Lo-Lo Tavg interlock (of 550 F, by allowing this interlock to be bypassed for the three steam dump cooldown valves PV507-A, -B, and -C.)

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y-PO;o 35

- 2s_*_EL8HI_

. DES 19N_INGLUDINQ_E9EEIY_8ND_EMEBQENGY SYSIEMS REFERENCE VESP, Training Text, Ch. 12b, P. 6 ANSWER 2.05 (2.00)

1. SI

2. Loss of voltage on 4160 Bus

3. Manual control room

4. Manual D/G room

5. Emergency manual D/G room [0.4 each3 REFERENCE VEGP, Training Text, Vol. 8, Ch. 16c, P. 20 AN9WER 2.06 (1.50)

1. Local / remote transfer switch in remote.

2. Mode selector switch in auto (i . e. not in the maintenance mode).

3. Starting air pressure greater than 150 psig (either header).

4. No protective device tripped. [3ce o. f ea. ]

REFERENCE VG, RO-LP-1093 VG, SR-LP-058; VG, 13145; VG, SR-LP-044 ANSWER 2.07 (1.50)

1. Low lube oil pressure.

2. High Jacket water temperature.

3. Engine overspeed.

4. Generator Differential 187A, B, C. [Any 3, e 0.5 each3 1

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i .

v2=_'_EL8HI_ DESIGN _INCLUQ1N9_EGEEIY_0ND_EMEBGENGY Pcgo 36

- SYSIEUS REFERENCE VEGP, LO-LP-011.201, P. 11.

VEGP, System Operating Procedure 13145, Precaution 2.1.4.

ANSWER 2.08 (1.00) a REFERENCE VEGP Training Text, Vol. 8, Chapter 16c., p. 9 ANSWER 2.09 (1.50)

a. Normal -

Domineralized water.

Emergency - Reactor makeup water. C1.03

b. Demineralized water. [0.53 REFERENCE VEGP, Training Text, Ch. 10b, P. 7 ANSWER 2.10 (0.75)

1. Spent fuel pit HX.

2. RHR HX.

3. RHR pump seal coolers.

REFERENCE VEGP Training Text, Ch. 10b, P.3.

ANSWER 2.11 (1.00)

Assure a minimum flow for removing RHR pump heat.

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l

2m_tEL8HI_DEH10N_INGLWDINQ_E8EEIY_8ND_EMEBGENCY POCO 37 SYEIEda !

' REFERENCE VEGP, LO-LP-12101-01-C, P. 8 VEOP, Training Text, Ch. 10a, P. 2 ANSWER 2.12 (3.00)

a. 1. Remove iodine from containment atmosphere. [0.53

2. Minimize post-LOCA corrosion of components (by maintaining spray fluid pH between 8.5 and 10.5). [0.53

b. Recirculation lines to the RWST are provided. C1.03

c. 1. RWST

2. Containment recirc sump. [0.5 each3 REFERENCE VEGP, LO-LP-15101-OO, PP. 6-10 VEGP, Training Text, Ch. 9b ANSWER 2.13 (2.50)

a. Loop 3 cold leg

b. Loop 4 hot leg

c. Loop 1 cold leg CO.253 AND Loop 4 cold leg [0.253

d. Loop 4 intermediate leg

e. Loop 1 cold leg CO.253 AND Loop 4 cold leg CO.253

[0.5 each]

REFERENCE VEGP, LO-LP-16001-OO, PP. 16-17 ANSWER 2.14 (1.00)

a. Between SG and RCP. CO.53

h. 3. [0.53

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2t_: EL8HI_DEH10N_INGLWDINQ_E9Ef?IY_8NQ_EMEBQENGY Pc;o 38 SYSIEda REFERENCE VEGP, RO-LP-217.

ANSWER 2.15 (O.50) 100%

REFERENCE VEGP, Training Text, Vol. 7, Ch. 13d, P. 3 ANSWER 2.16 (1.00)

1. Reduce thermal stresses (or thermal shock) (to spray lines, spray nozzle, and surge line).

2. Help maintain uniform water chemistry.

3. Help maintain uniform water temperature.

[2 required e 0.5 each3 REFERENCE VEGP, LO-LP-16301-01-C, P. 11 ANGWER 2.17 (2.00)

CoMs is blocked OR Cms bas no effec} (g? cred/ [or correc{ reftrence f, J/g5 3

a. 449105) COMO .t.eti. = V l0.53 e,,4 POTV ")1eck V .1 v e E O. 5 3. N 8 V /4 d r/ec ) 17 Anlioneered y

h. ThegowRCSWR$wmperatureginputstothepressureprogrammere,J ie it.e. , c ea.p., d iv gC-;y g, ...ur m, CO. p to producef; 2.,cr _i gr,;!

-i . . u . i .... m.m.. __.._ .,, _ , . u u , , . . . . .

j ebe Sj$ le m se in }, [j,O]

REFERENCE VEGP, Training Text, Vol. 5, Ch. ic, PP. 12-13

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

' . 24._ ELONI_DEH19N_INGLUDING_E9EEIY_0ND_Et!EBQENGY Pcco 39 HYSIEt!H ANSWER 2.18 (1.00)

The standby pump will auto-start CO.53 when the low header pressure setpoint is reached CO.53.

REFERENCE VG, RO-LP-149, p 5.

ANSWER 2.19 (1.00)

(RE-0016) Baron recycle liquid process monitor.

(RE-0010) Liquid waste processing offluent.

, (RE-OO21) Steam generator blowdown liquid process monitor.

(RE-0848) Turbine building drain liquid effluent monitor.

(RE-17646) Control building sump liquid effluent monitor.

(any 4 9 0.25 ea)

REFERENCE 1

VEGP Training Text, Ch. 11a, Pp. 20-22.

ANSWER 2.20 (1.00)

^

The MSLIO will provide a signal to shut all SG MSIVs. (The MCIV i iivr-to mim . . Juu-= i. ; , o --e r o li vi . L u i--, , SG". f r wm dvebiw Lrrim i. v . i riv i w b -, i = . ) OR Ib MSLf5 winfm/f a spi g/ /,o dy/ d)q gec,,,] MSfy REFERENCE in llre a((ce/e/ sheemlime' VEOP, RO-LP-215.

, VEGP, Ro -LP~2/ lop -oo, Pp /Y-/t;.

ANSWER 2.21 (0.50) f True I

l

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. 3 _* t1MI_DE81M_1ELUDIN9_98EEIY 8HD_EDEBGENGY Pcg3 40 SYSIEUS i 4

-REFERENCE VESP Training Text, Ch. 9a, P.7.

VESP TS, B3/4.5.1. -

ANSWER 2.22 (2.00) f

a. Provide sufficient core reflood volume and prevent injection of 5 noncondensible nitrogen into RCS.

i' i Provide sufficient nitrogen volume for complete discharge.

b.

c. Provide sufficient core reflood volume and prevent late discharge.

d. Prevent early discharge and injection of noncondensible nitrogen

) into RCS. ,

REFERENCE VG, TB, p B 3/4 5-13 VG, RO-LP .!O3, p 3; VG , Bit-LP-032, p 5.

t 8 w.

j ANSWER 2.23 (1.00) j s,

b REFERENCE s

VG, RO-TP-217-OO5.

l

{ ANSWER 2.24 (2.00) 4 1. Loss of instrument air. CO.33 4

2. Electrical failure of the permissive solenoid valves. CO.33

3. BI' signal. CO.33 78 % y

4. Hi-hi 8/G water level CO.33 APT, 2 out of f. CO.253 ,

5. Low Tavg following Rx trip CO.33 564 F, 2 out of 4. CO.253 !

i

(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)

4 V

i

.- Ea. :.E1MI_ DEB 19tLitMAWDitlS 58EEIY_8ND_Et$EBGENGY Pago 41 SYSIEtfB t r

i

. REFERENCE VEGP Training Text, Ch. 13a, P.5.

VE& 7"S, Tlr4/es 3. 3~3 anol 3. 3 - +'.

f" F

F h

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i l I r

i J

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! (***** END OF CATEG0ftY 2 *****) i i

4 a

c

.Iz_11NSIBudENIH_0ND_GQNIB96@ Pcco 42 ANSWER 3.01 (1.50)

1. Prossurizer pressure low pressure safety injection.

2. Steamline low pressure safety injection.

3. Steamline isolation. [0.5 points each3 REFERENCE VG, RO-LP-213, p 123 VG, RO-LP-215.

ANSWER 3.02 (1.00)

1. Orifico isolation valves shut. [0.53

2. Pressurizer level > 17% (No presourizor low level). CO.53 REFERENCE VEGP, Lo Lesson Plans, LO-LP-09101-OO, P. 11 ANSWER 3.03 (2.00) gl vg/af /

1. P-0, 2/4, Diocks single loop loss of flow below(49N Wy. p g/,, ,p

2. P-9, 2/4, Dlocks reactor trip on turbine trip below (50'2[PWR.

3. P-10, 2/4, Dackup to P-6 for Source Power, allow blocking of some roactor trips (auto instates).

[Permissivo, logic O.33 nach, function 0.33 mach 3 REFERENCE VEGP, Training Text ANSWER 3.04 (2.00)

a. To provide a programmad level sutpoint. [0.53

b. Increase. CO.753 The failed high steam pressure transmitter causes the steam flow input to 60WLC to increano. CO.753

(***** CATEGORY 3 CONTINUED ON NEXT PAGE ** * * *)

, . h_1.ItfEIButlEtlIE_8tfD_GQtfIBQLE Pcco 43 REFERENCE KNP OTM IV-10.4, 10.5 VEGP Training Text, Ch.~13b, ppi-2.

VEGP LO-LP-18301-00, pp10-11.

ANSWER 3.05 (1.50)

1. Total steam flow (summation of FI-512, -522, -532, and -542)

2. Steam header pressure (PT-507)

3. Feedwater pump discharge pressure (PT-508)

REFERENCE VEGP, Training Text, Ch. 13b, Fig. 13b-3 VEGP, Training Text, Ch. 20, P. 5 ANSWER 3.06 (1.50)

a. True

b. Falso

c. False [0.5 ea.]

REFERENCE KNP OTM IV-11.2, 11.3, 11.21, W Logic Diagram VEGP LO-LP-28101-00-C, pp15-17.

ANSWER 3.07 (1.50)

a. Rods move in, because Tavg is higher than Tref. (0.5)

b. Rods move out, because Tavg becomum less than Tref. (0.5)

c. Rods move out, because the power mismatch circuit sees turbine (0.5) power, (as sensed by impulse prennure) increasing above reactor power.

(*a*** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

,.3s_'INSIBWBENIS_960_GDNIBQLS Pago 40 i REFERENCE Cook RO-C-N804, Pg. 2.

VEGP Training Text, Ch. 6a, Pp. 7-8.

ANSWER 3.08 (1.00)

Because at higher power levels, need a. smaller change in reactivity to get the same change in % power (than that which is needed at lower power levels).

REFERENCE Cook RO-C-NSO4, Pg. 6.

VEGP Training Text, Ch. 6a, Pg. 8.

i ANSWER 3.09 (2.00)

1. The input breaker to the motor trips.

i 2. The generator has an overvoltage condition.

I

3. The generator has a reverse current condition.

4. An output phase develops a ground.

5. The generator output breaker trips on overcurrent.

Cany 4 0 0.5 each3

, REFERENCE l

.! VEGP, Training Text, Ch. 6, P. 35 t i

j i

4 4

i i

i

(***** CATE00RY 3 CONTINUED ON NEXT PAGE *****)

i I

, s L._.'.13EIBWUENIB_8HD_GQNIBAE Pcco <5 ANSWER 3.10 (3.00)

1. C-1: 1/2 [0.153 IR's greater than or equal to 20% equivalent [0.23, both CO.23.

I/f

2. C-2: As1f [0.153 PR's greater than or equal to 103% CO.23, both [0.23

3. C-3: 2/4 [0.153 OT Dmita-T w/in 3% of setpoint CO.23, both CO.23

4. C-4: 2/4CO.14}OPDelta-Tw/in3%ofsetpoint Co.23, both [0.23

5. C-5: Turbine power (PT-505) < 1 5 '/. C O . 2 3 , auto CO.23 333

6. Control bank D e 2Ger steps CO.23, auto [0.23 REFERENCE VEGP, Training Text, Ch. 25, P. 11 VEGP, Training Text, Ch. 6, P. 29 VEGP, L o - L p . ) 7 a ,3 . n o , fy, jo - y, ANSWER 3.11 (0.50)

TRUE REFERENCE VEGP, Training Text, Ch. 25, P. 9 ANSWER 3.12 (2.00)

1. 2220 psig

2. 2335 psig 1940

3. OE4B psig

4. 2305 psig

5. 2310 psig l 850

6. & Hee psig [6 e 0.333 each3

(***** CATEGORY 3 CONTINUED DN NEXT PAGE ** ** *)

.3s_21NEIBWBENIE_0NQ_GQNIBQLE Pcg3 Co REFERENCE ,

VEGP, Training Text, Ch. 20, Fig. 20-3 VEGP, Training Text, Ch. 1c, Fig. 1c-6 vuP, TS Tables 2. 2- 1 a ,( 3. 3 - Y.

ANSWER 3.13 (1.50) '

a. 2185 psig CO.253, 2/4 CO.253 l l

b. To prevent PORV from inadvertantly depressurizing the RCS. C1.03 ,

REFERENCE VEGP, LO-LP-16303-00, P. 13 ANSWER 3.14 (1.00)

To anticipate a pressure reduction due to an insurge of colder water.

REFERENCE VEOP, LO-LP-16302-00, P. 17 ANSWER 3.15 (1.50) ,

1. 2/4 indicators CO.103 on 1/4 S/G's CO.10] less than or equal to 17%

CO.103 will cause Rx trip CO.303 and auto start of MD AFW pumps CO.303

2. 2/4 indicators CO.103 on 2/4 O/O's Co.103 less than or equal to 17%

CO.103 will cause auto start of TD AFW pump CO.303.

REFERENCE VEGP, LO-LP-18301-00, P. 7 l r

ANSWER 3.16 (0.50) 1 L

FALGE ,

4

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i

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, .It_'INSIBudENIa_0ND_GQNIB9LQ Pc;D 47 REFERENCE VEGP, LO-LP-16302-OO, P. 10 ANSWER 3.17 (1.00)

Insufficient bypass spray flow.

REFERENCE VEGP, LO-LP-16301-01-C, P. 17 ANSWER 3.10 (2.00)

1. Closes containment purge supply and exhaust ducts.

2. Trips processing system.

3. Nonu

4. Switches control room ventilation to safety grade filtration train. [4 0 0.5 wach3 REFERENCE VEGP, Training Text, Ch. 11a, PP. 17-21 ANSWER 3.19 (0.50)

FALSE (Only NON safety-related loads will be shed.)

REFERENCE VEOP, Training Text, Ch. 16c, P. 31 ANGWER 3.20 (1.00)

Provents turbinn ovorbunting (and damagu) to.53 due to windmilling [0.253 in a vaporous steam atmosphuro CO.253.

(***** CATEGORY 3 CONTINUED ON NEXT PAGC *****)

t

I

.o j

, *h _.*1NBIBUDENIE_8Hk_GONIBOLS Pag 3 48

)

REFERENCE VEGP, Training Text, Ch. 16a, P. 28 i l

l ANSWER 3.21 (1.00)

a. Control board handswitch must be in ARM position.

b. PORV block valve REFERENCE I

VEOP, LO-LP-16501-00, P. 6 i

AN8WER 3.22 (1.50) !

1. Manual 1/2

2. SI 1/2

3. Containment area radiation 1/2 REFERENCE VEGP, LO-LP-28201-00, P. B 1

i l

(***** END OF CATEGORY 3 s****)

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. h._280GEDUBES_:_NQB0064_8BNQBU6La_Et!EBGENGY Pcge C9

'. 8HD_B8010LQQ1G86_GQNIBQL ANSWER 4.01 (1.00)

The motor must be permitted to coast to a stop between starts.

REFERENCE VEGP Procedure 13003, P.2 ANSWER 4.02 (2.50)

a. YES

b. NO

c. NO

d. NO

e. YES REFERENCE VEOP, Tech. Specs. 3/4.7 VEOP, Exam Bank, 553NA ANSWER 4.03 (1.50) l

1. Wide range That

2. Wide range Tcold

3. Core exit temperatures (thermocouples)

4. RCS Subcooling tany 3 0 0.5 each)

REFERENCE VEOP, Tech. Specs. 3.3.3.6

(***** CATE00HY 4 CONTINUED ON NEXT PAGE *****) ,

I t

. L _RB9GEDUBER_:_NQBUL_8BNQBUL_Et!EBGENCY PCC3 50' 8HD_B80106901 gel _GQNIBQL ANSWER 4.04 (2.50)

~1. Max. S/G tube leakage of 500 gpd, (exceeded by leakage through #1 S/G - 619.g gpd).

2. Max. unidentified leakage of 1.0 gpm, (exceeded by remaining unidentitied leakagw uf 1.13 gpm).

REFERENCE VEGP, TG 3.4.6.2 ANSWER 4.05 (1.00) 1.25 rem - whole body 7.50 rem - skin (of whole body) 18.75 rem - hands and forearms, feet and ankles (extremities)

REFERENCE 10CFR20.101(a)

ANSWER 4.06 (0.50)

TRUE REFERENCE VEGP, Exam Dank, 347NA i

ANSWER 4.07 (1.00)

Dy raining pressurizer prenmuro, causing thu spray valves to open.

REFERENCE VEGP, UDP 12003, P. 2 i VEOP, LO-LP-16301-01-C, P. 10 i

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,i h lEBQGEDQBEB_:_NQBU662_0BNQBdeL4_EMEB9ENGY Page 51 BUD _80DIDLQQ1GOL_GQNIBQL ANSWER 4.08 (2.00)

1. C-5 interlock cleared.

2. Tavg within 1.0 F of Tref.

REFERENCE VEGP, UDP 12004, P. 8 VEGP, Exam Bank, 665NA ANSWER 4.09 (0.50) c REFERENCE VEGP, Exam Bank, 497N ANSWER 4.10 (1.00)

a. 1600

b. 670 REFERENCE VEGP, UDP 12001, Limitation 2.2.9.

ANSWER 4.11 (1.00)

a. To preserve the sealing capability of the No. 2 seal.

b. Thirty (30)

REFERENCE VEGP, Procedure 13003, P. 2

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

, *S:r- TBQGEDUBEE : NQBt!8L 8BNQBt!864 Et!EBSENGY Pcga.52 8HD 8801%QQ1G86 GQNIBA.

ANSWER 4.12 (2.25)

1. .Turoine Control Valves - Closing CO.753

2. Generator Load - Lowering [0.753

3. LOSS OF TURBINE LOAD INTLK C7 Status Light - Energized. [0.753 REFERENCE VEGP, AOP 18012, P. 2 ANSWER 4.13 (2.00)

Action / Expected Response Response Not Obtained

1. Verify turbine trip. (Turbine 1. Trip the turbine EO.53 Stop Valves -- Shut. ) CO.53

2. Verify auto Rod Control CO.253 2. Adjust rods [0.253 and adjusting Tavg to No-Load establish Tavg at No-Load value of 557 F CO.253. value of 557 F [0.253.

REFERENCE VEGP, AOP 18011 ANSWER 4.14 (1.50)

1. Manually run back the turbine. CO.753

2. IF the turbine can NOT be run back, THEN shut main steam lino isolation valves and bypass valves. CO.753 REFERENCE VEGP, EDP 19211

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

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. Asu _tBQGEDUBER_ _NQBM9L4_8DNQBM862_EMEBDENGY Paco 53 i GND_8001%Q91GOL_GQNIBA -

ANSWER 4.15 (2.00)

1. CCP's or SI Pumps - At least-ONE running, (AND)

2. RCS pressure-< 1375 psig.

REFERENCE VEGP, EDP 19000, P. 26 ANSWER 4.16 (0.75)

By using the S/G Atmospheric Relief Valves-(ARV's)

REFERENCE VEGP, EOP 19001, P. 7 ANSWER 4.17 (2.00)

.1. RCS subcooling monitor indication: >28 F.

2. S/G pressures: stable or lowering.

3. RCS hot leg temp.: stable or lowering.

4. Core exit TC's: stable or lowering.

5. RCS cold leg temp.: at saturation temp for S/G pressure.

Cany 4 G 0.5 each]

REFERENCE l

VEGh, EDP 19001, P. 8, 16

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

i-

fJ ff._dBQGEDQBER_:_NQBU8L,._8BNQBU6L2_Et!EBGENQY Pag 3 'iJ 8HD_B8D1QLQQ1G86_GQNIBQL ANSWER -4.18 (1.00)

d. (Subcriticality) ~

c. (Core cooling)

a. (Heat sink)

e. (Integrity)

b. (Containment) [0.2 for each placed in correct order] 3 REFERENCE VEGP, EOP 19000, P._26 ANSWER 4.19 (0.50)

~TRUE REFERENCE VEGP, AOP 18038, P. 5 ANSWER 4.20 (2.00)

Incorrect [0.53 Because the procedure directs the operator to dump steam (cocidown) at MAXIMUM rate (to equalize RCS and ruptured S/G pressures ASAP) [1.53. OR Because Rehnical S,nctficaltan li,.,;}s may b e oceede/ l{ phreeled by Foy aefions El.5].

REFERENCE l VEGP, EOP 19030, P. 12 l

VEOP, Exam Bank, 217NA VEGP, E0f hinj,, Te~g 1-]f, }5 ANSWER 4.21 (1.00)

C I

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i L_

a 90

. dw_lEBQGEDUBER_:_NDBMOL2_eENQBdBL2_EME8@ENQY Pcg3 55

'OND_B9DIDLQ91G86_GQNIBQL REFERENCE VEGP, Exam Bank, 695N ANSWER 4.22 (1.00) lb N or lesS)

At RWST level of .pp.exi,.i-1 7 16%, OR upon receipt of RWST Empty alar (Either answer accepted for full credit.)

REFERENCE (Yb.

VEGP, Exam Bank, 531NA VEGP, Eo? 190/3-I, Pp f, 9.

vecP Tednic.d Dda Book, T;L 49 1

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'(***** END OF CATEGOF,Y 4 *****)

,- ~

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

i3

TEST CROSS REFERENCE Pcga 1 o

gyESIlgd _yOLyg BEggBENgg_

1.01 1.50 ZZZOOOOO83 1.02 2.50 ZZZOOOOO84 1.03 0.50 ZZZOOOOOB5 1.04 1.50 ZZZOOOOOB6 1.05 2.00 ZZZOOOOO87 1.06 1.00 2ZZ0000088 1.07 2.00 ZZZOOOOOB9 1.08 1.00 ZZZOOOOO98 1.09 2.00 ZZZOOOOO94 1.10 1.50 Z2Z0000092 1.11 0.50 ZZZOOOOO90 1.12 2.00 2ZZ0000099 1.13 1.00 ZZZOOOO101 1.14 3.00 ZZZOOOOO96 1.15 1.50 ZZZOOOOO93 1.16 0.50 2ZZ0000100 1.17 1.50 ZZZOOOOO95 1.18 1.00 ZZZOOOOO97 1.19 1.00 ZZZOOOOO91 27.50 2.01 1.50 ZZZOOOO102 2.02 0.75 ZZZOOOO103 2.03 1.00 ZZZOOOO104 2.04 1.00 ZZZOOOO105 2.05 2.00 2ZZ0000003 2.06 1.50 ZZZOOOO120 2.07 1.50 ZZZOOOO106 2.08 1.00 ZZZOOOOOO2 2.09 1.50 2ZZ0000107 2.10 0.75 ZZZOOOO108 2.11 1.00 ZZZOOOO109 2.12 3.00 ZZZOOOO110 2.13 2.50 ZZZOOOO111 2.14 1.00 ZZZOOOO112 2.15 0.50 22Z0000001 2.16 1.00 ZZZOOOO114 2.17 2.00 ZZZOOOOOO4 .

2.18 1.00 ZZZOOOO116 2.19 1.00 22Z0000118 2.20 1.00 ZZZOOOO119 2.21 0.50 ZZZOOOO121 2.22 2.00 ZZZOOOO122 2.23 1.00 ZZZOOOO123 2.24 2.00 ZZZOOOO124 32.00 3.01 1. tiO ZZZOOOOOO5 3.02 1.00 ZZZOOOOOO6 3.03 2.00 2Z20000007 3.04 2.00 ZZZOOOO125 3.05 1.50 ZZZOOOO126 3.06 1.50 ZZZOOOO127

e TEST CROSS REFERENCE Pcg5 2 gygsIlgN _YGLUE BEEEBENcg_

3.07 1.50 ZZZOOOO128 3.08 1.00 ZZZOOOO129 3.09 2.00 ZZZOOOO130 3.10 3.00 ZZZOOOO131 3.11 0.50 ZZZOOOO132 3.12 2.00 ZZZOOOO133 3.13 1.50 ZZZOOOO134 3.14 1.00 ZZZOOOO135 3.15 1.50 ZZZOOOO136 3.16 0.50 2ZZ0000137 3.17 1.00 ZZZOOOO138 3.18 2.00 ZZZOOOO139 3.19 0.50 ZZZOOOO140 3.20 1.00 22Z0000141 3.21 1.00 ZZZOOOO142 3.22 1.50 ZZZOOOO143 31.00 4.01 1.00 ZZZOOOO163 4.02 2.50 ZZZOOOO160 4.03 1.50 ZZZOOOO144 4.04 2.50 ZZZOOOO115 4.05 1.00 ZZZOOOOl62 4.06 0.50 ZZZOOOOl61 4.07 1.00 ZZZOOOO158 4.08 2.00 22Z0000159 4.09 0.50 ZZZOOOOl64 4.10 1.00 ZZZOOOO146 4.11 1.00 ZZZOOOO145 4.12 2.25 ZZZOOOO151 4.13 2.00 ZZZOOOO149 4.14 1.50 ZZZOOOO147 4.15 2.00 ZZZOOOO155 4.16 0.75 ZZZOOOO157 4.17 2.00 ZZZOOOO153 4.18 1.00 ZZZOOOO156

! 4.19 0.50 ZZZOOOO150 4.20 2.00 ZZZOOOO154 4.21 1.00 ZZZOOOO152 4.22 1.00 ZZZOOOO148 30.50 l ______

121.0 l

l l

.

ML20207T403

Text

Reference:

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