ML20235W712

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Exam Rept 50-483/OL-87-01 on 870921-25.Exam Results:Four Senior Reactor Operators & Two Reactor Operators Took Exam. All Candidates Passed Exams.Exams & Answer Keys Encl
ML20235W712
Person / Time
Site: Callaway Ameren icon.png
Issue date: 10/09/1987
From: Burdick T, Damon D, Lennartz J, Shepard D, Sunderland P
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20235W704 List:
References
50-483-OL-87-01, 50-483-OL-87-1, NUDOCS 8710160376
Download: ML20235W712 (115)


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li U.S. NUCLEAR' REGULATORY COMMISSION:

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REGION:III p ,

. Report l No.' 50-483/0L-87-01 100ckeiNo.L50-483 License:No. NPF-30 7o

'Licenree: Union Electric Company . . . il Post Office Box 149 - Mail Code 400-  ;

zSt. Louis, M0f 63166- ,

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. Facility Name:..'Callaway '

Examination Administered-At: : Columb'ia, Missouri- (Written), Callaway Plant (0perating). ,

o Examination Conducted: September 21-25,'1987 Examiners: e ]Oh /f 7

.. [ 0/.1Damon 'Date-

. W ' f) f) .jol,)m Lennartz Date //

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'D. L. Shfpard Date Wb P. R. Sunderland ID f#l fffl Date 8 Approved By: Y /4 fM /

T. J. Burdick, Chief Dets /

Operating Licensing Section Examination ~ Summary Written and Operational examin'ations were administered on September 21-25, 1987 (Report No 50-483/0L-87-01)) to four SR0 candidates and two R0 candidates.

A written examination was administered to one additional R0 candidate.

Results: All candidates passed the examinations.

1 8710160376 871013 3 DR ADOCK 050

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DETAILS l 0

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- 1.- Examiners 1

'D..-J. Damon '

J. A. Lennartz

'D .

L. Shepard 1

  • P. ' R. Sunderland
  • Chief Examiner
2. Exit Meeting An exit-meeting was conducted at which preliminary findings were discussed. Attendees were:

. J. D. Blosser, UE, Manager, Callaway Plant M. E. Taylor, UE, Superintendent of Operations '

G. A.-Hughes, UE, Superintendent of E1gineering (ISEG/STA)

M. S. Evans, UE, Superintendent of Training E. M. Thornton, UE, QA Engineering Evaluator S. V. Henderson, UE, Operations Training-P. C. Shannon, UE, Operations Training C. H. Brown,'NRC, Resident Inspector T. M. Burdick, NRC, Chief, Operator Licensing Section J. A. Lennartz, NRC, License Examiner  !'

P. R. Sunderland, NRC, License Examiner The following areas were discussed:

a. The examination schedule was reviewed.
b. Generic Weaknesses None were noted.
c. Generic Strengths The SR0 candidates used their Emergency Implementation Procedures (EIP) well during the examination while coinbating major casualties.
d. The simulator performed well during the e).aminations. There were no instances of the simulator failing to carry out a malfunction that was entered. Also, the use of the machine to fine tune scenarios was appreciated.
e. Union Electric was concerned about the NRC using the simulator without facility representation due to the position the NRC could be put in if the simulator broke down. This will be discussed in the region and the NRC will try to work'out a way of fine tuning our scenarios with Union Electric personnel present for our next visit.

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3. Examination Review REACTOR OPERATOR EXAM COMMENTS / RESOLUTIONS Question 1.13 Comment: Should also allow. discussion on redistribution.

Ref: Westinghouse Rx Core Control For Large PWR, Chapter 3, Page 3-44, 3-45, thru 3-47.

Resolution: Partially concur. A discussion on redistribution will be allowed and given credit, provided that an explanation of how redistribution affects core flux distribution is included.

Question 2.01b Comment: No. 3 RCP Seal is a double dam seal, also allow RCDT as acceptable answer.

Ref: Callaway Lesson Plan, Chapter 9 Resolution: Partially concur. RCDT will be allowed as an acceptable answer. However, to receive full credit the containment sump and RCDT have to be included in the answer. The answer key has been modified to reflect this.

Question 2.03b Comment: Should also allow the following:

(a) system ground (b) battery monitor (c) charger failure (d) charger AC breaker open (e) switchboard bus under voltage (f) charger DC breaker open (g) charger DC over voltage

{h) charger DC under voltage si) charger AC under voltage itef: OTA-RL-RK017, ATT: F, Page 1 of 1, Item 2.1 (See also OTA-PJ-00001).

Resolution: Comment accepted. The proposed answers will be ac-cepted.

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. i Question 2.04.b 1

Comment: Answer or key states how the start failure works.but does not l state purpose. )

Purpose is: Prevent diesel from running at degraded speed j (conditions). (Lube oil pumps are engine-drive).

Ref: See Attachment 2a from Callaway FSAR, Page 8.3-10.

Resolution: Partially concur. The answer key and the reference provided say the some thing. However, the answer key has been modified to reflect the wording as in the reference.

Question 2.06 Comment: Should also allow: Administrative condition (not interlocks) from OTG-AA-00006, Step 4.2.5, Page 13 which references 0TN-EJ-00001, Section 4.3, Page 6.

Conditions (1) RCS average temperture < 350

.(2) And breakers closed for the loop isolation valves.

Ref: 0TG-ZZ-00006 and 0TN-EJ-00001 Resoluticn: Comment not accepted. The proposed answer are procedural requirements that must be met, and will not be accepted for credit.

Question 2.07 Comment: Should also allow the following which provide information about the CTMT Spray System on the main control board:

(a) RWST - CTMT spray suction valves position (b) CTMT spray pump hand indicating switches  !

(c) Spray additive tank isolation valves position Ref: Chapter 18, Page 15 & 16, Callaway Training Lesson Plan.

Resolution: Comment not accepted. The question asks for parameters that can be monitored, not indications. The pe. posed answers will not be accepted.

Question 2.09 Comment: Should also allow pressurizer level <17%, to prevent uncovering (damage) pressurizer heaters.

Ref: Callaway Training Lesson Plan Chapter 11, Page 1-3, Lines 4 thru 6.

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~ Resolution: Comment. accepted. The proposed answer will be accepted for j

. credit. However, to receive full credit the proposed answer. 1 must be included with the original answer key response. The .l answer key has been modified to reflect this.

Note: The reference supplied by the utility was incorrect, and was changed'to reflect the correct one by the examiner.

Question 2.12 Comment- a.2 i.e., permits feeding. intact generators in spite of Aux Feedwater loss due to feedline break.

Ref: Callaway Lesson Plan, CMpter 25, Page 5, Paragraph 2.

Resolution: Comment noted. The proposed response will be incorporated into answer key as an acceptable alternate response.

Question 2.13

' Comment: HB-RE-18, Known as " Plant Discharge Monitor" LE-RE-59,.Known as "0ily Waste Discharge Monitor" Resolution: Comment noted. Answer key has been modified to accept these as alternate response.

J' Question 3.03 Comment: Also the following:

(10) Hot leg wide range temperature (11) Pressurizer pressure Ref: (1) Callaway Lesson Plan, Chapter 48, TP-1 & 2 (2) Tech Specs 3.3-9,'Page 3/4 3-50 NOTE: Page 3 of Lesson, Plan inadvertently deleted these two indications.  ;

-Resolution: Comment accepted. The answer key has been modified. The utility is encouraged to revise their lesson plan to include these two indications prior to the next scheduled exam.

Question 3.07 Comment: b. Also allow: P4 required to permit SI reset.

Ref: Functional diagram - Safeguards Actuation Signals, Drawing No. 7250064, Sheet 8.

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Resolutions Partially concur. You can reset SI without P4, but if an 1

-Automatic SI signal is still present', then SI will be re-initiated. One of the functions of P4 is to prevent the i automatic re-actuation of SI after a manual reset, which is stated in the answer key. However, the-proposed response will be accepted as an alternate response for b.3. The. answer ,

key has been modified to reflect this. I Question 3.13 Comment: (1) Item 1 is no longer valid'due to Callaway Modification '

which removed the input'to CRVIS from GTRE 31 & 32.

Ref: 0TA-SP-RM-000011 Resolution: Comment noted. The answer' key has been modified to reflect this, The utility is encouraged to revise the Lesson Plan to reflect this before the next scheduled exam. i Question 4.09 Comment: (d) Should be A or C on Answer Key.

Ref: 0TG-ZZ-00001, Precautions and Limitations, Item 2.8, Page 4.

NRC Response: Comment accepted. The answer key has been modified to reflect this. 4 I

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. SENIOR REACTOR 0PERATOR' EXAM COMMENTS / RESOLUTIONS 1 - Question 5.02i M Commenti 'This'needs'to-be clarified, the, reference?for the> answer -

limplies not).a pressurizer temperatureidecreas~e of'2 degrees,.

'but a RCS' temperature decrease of 2 degrees. The: pressurizer.-

< temperature decrease with no RCS temperature decrease would-make the: Question.5.02 FALSE. -(See calculation'.below). Howeve r , .-

an:RCS temperature' decrease (outsurge),'which is'.what we' teach as part of.our course and your reference. supports a- true statement -

(as given by answer, key).

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-Assume' pressurizer at'2235 psig (2250. psia) and PSAT-TSAT;

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conditions exist:-

TSAT for 2250 psia N 652.67 F-to N1 650.67 using interpolation 3

'of steam tables,'PSXT for 650.67 F

  • 2218.88 psia. Thus ^P.for 2 F decrease
  • 40.78 psic with steam bubble.
b. For H2 ;in the pressurizer bubble, use ideal gas lawsi "Vf(648)= .02657.

Vf(652) = .02691 '02580(650.67)

Vf(656) = .02728- . J.02697(652.67)

.02680.= '9941thus V2 (650.67) ='.994V 2(652.67 ) - Water Volume .02697

. .Thus,. gas volume ~= 1.006V2 .

P2 = 2250 psia

'", "V2 = Volume of. pressurizer gas space' initial conditions 4

'T2 = 652.67 +-460 = 1112.67"R P2 = Unknown

-V2 = (1.006V3 ) = Volume of pressurizer gas space. final conditions T2 .= 650.67 + 460 = 1110.'67 R

-2250(V)~= P2 (1.006V).

Til2.67 1110.67 2250 -= 1.006P 2 1112.67 1110.67 2250(1110.67)'= 2232.5 1112.67(1.006) 1'

  • 17.52 psia under H2 bubble L__ _ _ _ _ - - _ - _ _ - - _ _ _ - _

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Resolution: LCommentinoted, since the question response implies RCS temperature'instead_oflPZR temperature, the question:is deleted.

.Questio'n 5.10..

l Comment: Answer should be y 100 steps on Bank'C, q Ref: According to Figure 5-1 Resolution: Comment. accepted. 'The. answer key was modified'to read '

96 1 10. steps on Bank C.

Question 5.18 Comment: QPTR value~ofel.02 should be acceptable due to procedure l guidance;(i.e.,~ounded r to two significant digits).

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Ref: .OSP-SE0003,-Precaution and Limitation, Item 3.5 and

' Notes following Steps.6.3.6 and 6.37.

. Resolution: Comment noted. Answer key was modified'to calculate QPTR to.two decimal points. Also noted.that-the intent of the

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question is not changed. The NRC did'not have the reference material ' referred to in the comment.

Question 6.01

Comment: Monitor HB-RE-18 is commonly known as.the " Plant Discharge

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Monitor" and LER-RE-59 is known as the " Oily Waste Discharge Monitor."

l Resolution: Comment accepted. The answer key was modified-to accept alternate answers specified by the comment.

'Qu'estion 6.03 Comment: Since the spent fuel handling tool is stored'in the spent fuel pit (which is contaminated) other administrative provisions apply such as decontamination administrative provisions apply such as decontamination or packaging. -This should be an alternative answer.

Ref: APA-AA-01000, 4.14.6 on Page 30.

Resolution: Comment not accepted. The question specifically states the use of the spent fuel handling tool out of the water and not radiological concerns.

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L i i Qi --tion 6.05 Com.,ent: c. Should also allow full credit for l (1) Trip MFP (2) Trip Turbine.

(3) Feedwater Isolation Signal (FWIS)

FWIS' includes closing the (1) Main Feed Reg Valves (2)' Bypass Feed Reg Valves (3) Main Feed Isolation Valves -

Ref: Functional diagram, Steam Generator Trip Signals, Orawing No. 7250064, Sheet 7, and Feedwater Control and Isolation, Drawing No. 7250034, Sheet 13.

i Resolution: Conaent accepted. .The answer key was modified to accept FW15 for the three answers in-the comment, Question 6.09 Comment: c. Answer incomplete, should state:

(3) Verify fast close input no longer present.

(2) Rotate switch to " Bypass" (Given in Question).

(3) Return switch to the " Operate" position (4) Go to slow open Ref: Callaway Lesson Plan, Chapter 49, Pages 16 & 17.

Resolution: Comment not accepted. The question asks for a condition to exist, not for actions to be taken. The answer fulfills what the question asks.

Question 6.12 Comment: a. Purpose for powering one rod group for Maintenance on the normal supply.

Ref: Callaway Lesson Plan, Chapter 26, Pages 1 and 2.

Resolution: Comment accepted. The answer key was modified to reflect the reference.  ;

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o Question 6.13-

Comment: a. LOCA.(1) SIS or j

.(2) CASA and (3) Either r> undervoltage "or" DG breaker closed.

Ref: Callaway Lesson Plan, Chapter 6, Pages 7-8 and TP-2.

Resolution: Comment accepted. The answer key was modified to replace "and" with "or. "' ,

Question 6.15 Comment: a. Allow also 119 to 128 psig for loading and unloading of the. lead compressor.

b. Question should read "A compressor will shut off after minutes of continuous operation UNLOAD."

Ref: Licensed Training Lesson Plan, Chapter 14, Page 4,  ;

Paragraph 2 (for Part a) and Page 5, last sentence in Paragraph 1 (for Part B).

Resolution: Comment accepted. The answer key was modified for a to accept the comment as an alternate answer. The question for b can'not be changed after the fact. From questions during the exam, the fact that unloaded.should have been included is understood.-

Question 7.03 Comment: a. NB01 which is powered.from the Safeguards Transformer never deenergized and NB02 was only temporarily deenergized until NE02 diesel starts'on undervoltage and the DG Breaker closes in the bus NB02. Therefore the operators would leave  :

-E-0 at Step No. 4.  !

Ref: Callaway Lesson Plan, Chapter 1, TP-5.

NOTE: NB01 powered from Safeguards Transformer No. 1, (

NB02 powered from the Startup Transformer via XNB02 1

'(both buses have emergency Diesels).

Resolution: Comment accepted. The answer key was modified to reflect j the comment. i Question 7.10 Comment: b. Also a fourth correct answer is Reactor Bypass Breaker A or B closed. l Ref: 0TA-RL-RK075, Attachement B, and 0TA-RL-RK076, Attachment B.

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Resolution: Comment accepted. . The comment was added as fourth answer.

This was not part of the referenced procedure; however, common sense dictates adding _it to the answer key.

Question'7.12 Comment: Also allow:"want to' isolate feed to ruptured S/G as early as I possible to prevent future overfill."

Ref: E-3' series walk. through, Pages 6-10.

i Resolution: Comment accepted. The answer key was modified to accept the comment as an additional response.

Question 8.08 Comment: a. Other acceptable answer is: " Annotated in the remarks  !

section." l Ref: ODP-ZZ-0016, Page 6, Step 4.1.2.3.

Resolution: Comment accepted. The answer key was modified to reflect the comment.

Question 8.16 Comment: Question is difficult to understand since there are no Function i "A" fire detection instruments for Zone 306. There are.only 13 function "B" instruments per Tech. Spec. 3.3.3.7, Page 3/4 3-60.

Ref: Tech. Spec. - 3/4 3-60 for Zone 306 (Room 3501 - Lower Cable Spreading Room).

Resolution: Comment noted. The Technical Specification is difficult to understand as a's ans b's in the body do not match up well with x's and y's in the tables. The intent of the question did not seem to be sacrificed, therefore, no changes were made.

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M4 STER COP U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: CALLAWAY REACTOR TYPE: PWR-WEC4 DATE ADMINISTERED: 87/0RZ22 EXAMINER: SUNDERLAND, P.

CANDIDATE:

IEEIRUCTIONS 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_ _IQTbb SCORE VALUE__ CATEGORY

_24 &Q0 _24.24 __ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS

_2E.00 _ah 25 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 25.00 _ZE Rh 7. PROCEDURES - NORMAL, ABNURMAL, EMERGENCY AND RADIOLOGICAL CONTROL

_25.00 _25.25 8. ADl4INISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS

_HRtG9.__  % Total:

Final Grade All work done on this examination is my own. I have neither given

nor received aid.

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Candidate's Signature

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NRC RULESL AND GUIDELINES FOR LICENSE EXAMINATIONS

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p During the administration of this examination the following rules apply: i E

! l '. 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 H leave. You must avoid all contacts with anyone outside the examination room to avoid even the.apoearance or possibility of cheating.

3. Use black ink or dark pencil 2Dlr 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 gn gne 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 l

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to mathematical problems whether indicated in the question or not.

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15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK. l
16. If parts of the examination are not clear as to intent, ask questions of the 2Xam1Dar only.
17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.

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18. When you complete your examination, you shall:
a. Assemble your examination as follows:

(1) Exam questions on top. l (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 i leaving, you are found in this area while the examination is still l in progress, your license may be denied or revoked.

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. EQUATION SHEET -

f = ma ,

y a s/t Cycle efficiency = (Mettork out)/(Energyin) 2 w = esg s's V ,t + 1/2 at 2

E=e I

a = (V A = A,e*

  • f - V,)/t KE = 1/2 av A = AN PE i sph Vf = V, + at v = e/t A = anZ/tg = 0.693/t1/2 2 t

. W = v aP-A= a0 yfg =[Xtm (g1/2I *MI*b)3 aE = 931 am

""Y avAo 7 , gc , -h Q = aCpat 6 = UAa7 I = 1,e #

Pwr = Wfah I = I ,10'* O TYL = 1.3/u P = P 10sur(t) HVL = -0.693/u t

P = P,e /T

' SUR = 26.06/T SCR=$/(1-K,ff)

CR, = S/(1 - K,ffx)

SUR = 26e/s* + (s - p)T CRj (1 - K,ffj) = G2(1 ~ Ieff2)

T = (1*/s) + [(s - eyTo] M = 1/(1 - K,ff) = CR /CR j ,

T = s/(e - s) M = (1 - K,ffa)/(1 - K,ffj)

T = (s - e)/(Ie) SDM = (1 - K,ff)/K,ff -

e = (K ,ff-1)/K ,ff = AK /K ,ff ,ff s* = 10 seconds I = 0.1 seconds-l e = [(**/(T K,ff)] + [a,ff /(1 + IT)]-

I ldj = 1 d P = (14V)/(3 x 1010) Id j 2 =2Id 2 g2 2

E = eN R/hr = (0.5 CE)/d (seters)

R/hr = 6 CE/d2 (feet) ,

Water Parameters Miscellaneous Conversions 1 gal. = 8.345 inn. 1 curie = 3.7 x 1010dps

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1 gal. = 3.78 11 tai-s lkg=2.21lba 1 ft3 = 7.48 gal. 1 hp = 2.54 x 10 3 Stu/hr Density = 62.4 lbg/ft3 1 sw = 3.41 x 106 Stu/hr Density = 1 ge/enP lin = 2.54 cm Heat of vaporization = 970 Stu/lem *F = 9/5'C + 32 Heat of fusion = 144 Stu/1bm 'C = 5/9 ('F-32) 1 Ata = 14.7 psi = 29.9 in. Hg. 1 BTU = 778 ft-Ibf 1 ft. H 2O = 0.4335 lbf/in.

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)S; IHEQBYlQE_HU9LEAB POWER _P_LANT OPERAIION, FLUIDS, AND.

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-QUESTION,s5.01 (1.00)-

'Given a pressurizer that maintains constant level,-does:the. indicated J1evel INCREASE ~, DECREASE,'or REMAIN ~THE SAME if containment. temperature is

. raised by 50 degrees'F.

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QUESTION- 5.02' (1.00)-

'TRUE or FALSE?

s Subcooled nucleate. boiling normally occurs 'during power. operations in the reactor core.

QUESTION 5.03 -(l'.'00)

Which one of the following radioactive isotopes found in the reactor Ecoolant would NOT indicate a leak through the fuel cladding?

a. 1-131
b. Xe-133
c. Co-60
d. Kr-85

.' QUESTION' 5.04- (1.00)

TRUE or FALSE 7 For a centrifugal pump, .if'the-AVAILABLE Net Positive Suction Head (NPSB) is negative, then the pump will not cavitate.

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l' L5. THEORY'OF NUCLEAR POWER _ PLANT ~ OPERATION1 FLUIDS , iA@ .PAGE

-IBERMODYNAMICS

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QUESTION. 5.05L (1.00)

Fill in theLblank with'the most correct answer:

During natural circulation, excessive steam flow causes an interruption.of natural circulation flow. This is best indicated by .

a. Decreasing delta-T accross the core
b. Decreasing primary cold leg temperature
c. Steam pressure decreases with decreasing decay heat-generation:
d. Decreasing core outlet temperatures-QUESTION 5.06 (1.00)

Choose the correct phrase to correctly complete th'e follcwing sentence.

As the core ages from BOL to EOL, the ratio of Pu-239 to atoms U-235 atoms increases. This changing ratio causes the .

a. Reactor period to decrease
b. Void coefficient to become less negative
c. Moderator temperature coefficient to become'less negative
d. Delayed neutron fraction to increase '

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5. THEORY OF NMgLEAR POWEB_ELANT OPERATION. FLUIDS. AND PAGE' 4:

IBEBOODYEAMICS QUESTION 5.07 (1.00)

Fill in the blank with one of the choices given.

During a Xenon-free reactor startup, critical data was inadvertently taken two decades below the required Intermediate Range (IR) level. Assuming that RCS temperatures and boron concentrations were the same, the critical rod position taken at the proper IR level the critical rod position taken two decades below the proper IR level.

a. Is less than
b. Is the same as
c. Is greator than
d. Cannot be compared to QUESTION 5.08 (1.00)

Fill in the blanks concerning Reactor Vessel Integrity.

a. Neutron embrittlement of the reactor core causes to increase. This means that brittle fracture can occu'c at higher temperatures. (0.5)
b. Allowable pressures inside the reactor vessel are most limiting during (HEAT-UP, COOL-DOWN, or STEADY-STATE) operations.

(0.5) i QUESTION 5.09 (1,60)  !

l Use Figure 5-1 (Rod Worth Curves) toestimatethecriticalrodpositionifI counts double between 0 and 50 steps on bank C. (Assume the plant is at  ;

the conditions specified on the figure) l I

QUESTION 5.10 (1.50)

The Callaway Plant was at full power for 57 days when it tripped at 3:00 a.m. this morning due to operator error on maintenance of the protection system. An Estimated Critical Condition (ECC) for a 5:00 p.m. startup was, performed. The startup is delayed until 8:00 p.m. Briefly describe HOW and WHY the ECC changes.

(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)

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1 s: j 5, THE0BI OF' NUCLEAR POWERfELANT'OPEBATION. FLUIDS. AND PAGE' . !j

'IHERMODYN6MICS j

' QUESTION 5.11- (1.50)

Callaway'is operating at 67% power at BOL uhen a'ateam dump valve OPENS.

State.what happens to the following parameters if all. controls:are in manual and no' operator action is taken. (Limit your answers to INCREASES, DECREASES, or REMAINS THE SAME)

a. Power Defect
b. Steam Generator Level (answer for initial short term effect only)

.c' . Pressurizer. level QUESTION 5.12 (2.00)

An' incident at ANO-1-resulted in fuel damage when a control rod was'found to be 90 inches further into the core than the remaining rods in its group for a poriod of 12 days. The rod was withdrawn to align it with the rest of the group within one hour.while the plant continued to operate at full 1 power. Why is fuel. damage likely to occur in such a situation?

' QUESTION 5,13 (2.00)

Indicate whether-the following will cause the differential rod worth of one control rod to INCREASE, DECREASE, or have NO EFFECT. (Analyze each case separately.)

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 5.14 (1.00)

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ECCS Systems are designed to limit clad temperature to 2200 degrees F.

LIST TWO undesirable consequences resulting from the zirconium-water reaction'if clad temperature is allowed to exceed 2200 degrees F?

(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)

L______.____.__

5. THEQBY_QE_UUCLEAR POWER PLANT'OPEBATION, FLUIDS1_AHD PAGE t IHEBdQDXHAMICS QUESTION 5.15 (2.00)
a. Define DNBR. (0.5)
b. The plant is at full power steady state conditions. If plant pressure decreases, without any operator action, would DNBR INCREASE,,

DECREASE, or REMAIN THE SAME7 (0.5) EXPLAIN your answer. (1.0)  ;

l l

QUESTION 5.16 (1.50)

At the Callaway Plant, describe the change (direction and reason for) in running current in an operating centrifugal charging pump after a second centrifugal charging pump is started. (1.5)

QUESTION 5.17 (1.50)

Using the information provided below, calculate the Quadrant Power Tilt Ratio (QPTR). (Show all work)

N41 N42 N43 N44 Upper detector current 159.7 139.5 157.0 147.1 100% upper detector current 266.6 234.5 262.9 241.6 Lower detector current 166.0 145.1 160.7 150.3 100% lower detector current 278.6 236.7 270.0 252.3 QUESTION 5.18 (1.50)

Fill in the blanks to complete EACH of the following statements concerning suboritical multiplication. (limit answers to INCREASE, DECREASE, or REMAIN THE SAME.)

a. The period of time for a constant neutron level to be reached following a positive reactivity insertion at Keff = 0.9 will from an equal reactivity insertion at Keff = 0.8. (0,5)
b. If a neutron source strength decreases, the rod height for criticality will . (0.5)
c. If a neutron source strength decreases, neutron level will

. (Assume original neutron count rate to be 100 cps.)

(0.5)

(***** END OF CATEGORY 05 *****)

I.

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i-R '^ '6l PLANT SYSTEME_ DESIGN.-CONIROL. AND INSTRUMENTATION. PAGE V .

i L

. QUESTION 6.01~ (1.50)'

h N. hat are:THREE Radioactive Liquid Effluent Monitors (contained in L Technical; Specifications) that provide AUTOMATIC termination of a release?

QUESTION 6.02- (1.00)

What'happens if'a TOTAL MAKEUP FLOW DEVIATION alarm is generated for

makeup water in the DILUTE mode?

QUESTION 6.03 (1.00')

'During an outage, you plan on using the Spent Fuel Handling Tool out of the water. What special provisions must be made?

QUESTION 6.04 (1.50)

Why do accumulator tanks B and C have two vent valves while accumulator tanks A and D only have one vent valve?

QUESTION 6.05 (2.00)

Answer the following concerning P-14 (HI S/G Level):

a. What is the setpoint of P-14? (0.5)
b. What is the coincidence of P-14? (be specific)-(0,5) c.- What five functions does P-14. accomplish when actuated? (1.0) I 1

QUESTION 6.06 (1.25)

List five(5) emergency-diesel engine trip signals that will not cause the engine to trip when an SI signal is present.

(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

1 . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _____

I 6; EL6NT SYSIEMS DESIGH, COHIBQL. AND._INSTRUMENTATIQH PAGE (]

i l

l l

l QUESTION 6.07 (1.50) ]

t For the following, indicate what effect, f* any, a failure of the ]

Pressurizer Master Pressure Controller (PC-455A)-HIGH would have on that {

equipment. (Assume the plant is at full power operations except for item f when.the plant is in MODE 4 operations.) (Limit your answer to OPEN, CLOSE, ACTIVATE, DEACTIVATE, or NO EFFECT.)

a. PORV (PCV-455A)
b. PORY (PCV-456A)
c. Spray Valve (PCV-455B)
d. Spray Valve (PCV-4550)
e. Lo Pressure Block Interlock (For PORV Operation)
f. Cold Overpressure Control System (Assume system active)

I QUESTION 6.08 (1.00)

Of the conditions necessary to initiate a MOTOR DRIVEN Auxiliary 4 Feedwater Actuation Signal (AFAS), which one may be blocked on the l Main Control Board? (1.0)

QUESTION 6.09 (2.50)

Answer the following questions concerning the Main Steam Isolation Actuation System:

a. List the three Main Steam Isolation Signals. Include setpoints and  ;

coincidences. (1.5)

b. What is the normal closure time for an MSIV using the FAST mode? (0.5)
c. If the operator wishes to override the FAST CLOSE command, what I condition must exist so that rotating the valve rotary switch ~to the BYPASS position will cause the MSIV to SLOW OPEN7 (0.5)

QUESTION 6.10 (1.50)

Each reactor coolant pump has an ANTI-REVERSE ROTATION DEVICE:

a. How does this device work? (0.75)
b. What condition does this device prevent? (0.75) l

(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

6. ELANT' SYSTEMS ~ DESIGN, CONTROL, AND INSTRUMENTATION 'PAGE !

I ;'

~I-

' QUESTION,=6.11 (2.00)

Answer the following questions concerning the' Containment' Spray System:

a. List the'two conditions / actions that generate a Containment 1 Spray Actuation Signal (CSAS). Include setpoints and' coincidences where applicable. (1.25)
b. List the three automatic actions that occur in the system when a CSAS is generated.-(0.75)

' QUESTION 6.12 (2.00)

Answer the following questions concerning the Rod Control System:

a. What'is the purpose of the Rod Control System DC Hold Cabinets? (1.0)-
b. In the Rod. Control System, there are the same number of-Slave Cyclers-as there are . (Choose the most correct answer below)

(0.5).

1.- Logic' Cabinets

2. Rod Groups
3. Control Rods
4. Power Cabinets
c. The Bank Overlap Unit decides which control rods to move utilizing direct input from . (Choose the most correct answer  !

below)(0,5)

1. Master Cycler 1
2. Step Counters
3. Slave Cyclers-
4. P/A Converter ] ;

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h- ,

8 j 6; ELANT' SYSTEMS DESIGN, CONTROL.'AND INSTRUMENTATION PAGE 1C e

f QUESTION 6.13 (2.25)

Answer the following questions concerning the LOCA and SHUTDOWN sequencers:

a. List the conditions-(other than test) necessary to activate each-sequencer. (1.0)

'b. If centrifugal charging pump A were running and the Train A LOCA sequencer activated,-how would the pump respond? (0,5)

c. If the conditions to activate both the LOCA and-Shutdown sequencers:

-were to occur at the same time, what three responses occur? (0.75)'

QUESTION 6.14 -(2.50)

Recently at the Callaway. Plant, it was found that incorrect scaling of the delta-I input to the OTdelta-T trip setpoint ocurred due to'a.

misunderstanding of the adjustments required to account for incore-excore calibrations,

a. What is the Technical Specification design basis'for the 0Tdelta-T reactor trip? (0,5)
b. How is-delta-I generated? (0,5) c List two(2) plant parameters other than delta-I used to generate the OTdelta-T trip setpoint. (0.5)
d. Why is'the improper scaling incident a potentially more serious detriment to reactor safety than a single channel failure.in one of the inputs to the delta-I calculation? (1.0)

. QUESTION 6.15' (1.50)

Fill in the blanks concerning the Instrument Air System (IAS):

a. Normally, the IAS maintains pressure at _ psig.
b. A compressor will shut off after minutes of continuous operation.
c. The compressor that is first to start as IAS pressure decreases is known as the compressor.

(***** END OF CATEGORY 06 *****)

i 7 .~ PBQQEDURES - NOEMAL1_ABHQBMALi EMEBGENCY AND PAGE 10 w

B6D1QLOGICAL CONIBQL

-l

~.

1

- QUESTION > 7 .01 (1.00) f"

~Callaway.is operating l steady state at 75% power when:the plant experiences a dropped control rod. - Tave/ Tref mismatch is initially maintained by

taking control of which ONE"of the following per OTO-SF-000037.
a. -Individual control rod banks
b. Individual control rod groups s
c .- RCS boron concentrations

~

d. _ Turbine load' QUESTION. 7.02 (1.00)

Answer _the following-(f311-in) questions concerning a Plant-Startup from 5-to 20% power (OTG-ZZ-00003).

a. Following a startup, within one hour, the is. sampled and analyzed for principle gamma emitters,
b. The maximum rate of power increase shall be held to.  %/hr.

unless limited by existing Westinghouse guidelines.

QUESTION 7.03 (2.50)

~

An inadvertent reactor trip occurs during power operation. All systems function normally except.for the startup transformer, which trips on a fault when V-53 and V-55 trip.

a. When will the operators leave E-O during this event? (0.5)-
b. What five conditions will the operators check to verify natural  !

circulation? (1.5) l

c. Fifteen minutes after this event occurs, a tube ruptures in the "B" i steam generator. What should the operators do to insure that they i are in the correct procedure? (0,5) {

i

)

(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

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1 L 7. PRQgEDURES - NORMAL & ABNORMAL. EMERGENCY AND / PAGE l '1 l 1

86DIOLOGICAL CONTROL ,

1 1

u 1

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. . s I

. QUESTION ~ 7.04. . (1.50) T j Answer the following TRUE-FALSE questions.concerning sthrkins a RCP in the1i

~

J Natural Circulation-Cooldown Procedure (ES-0.2): e l fa. Two (2):RCPs may be started ~at one' time provided that(they are.not-l

' powered from the'same electrical bus. '

b. Two;(2) successive RCP' starts may be made only if the motorLis~

allowed to.coastit'o a stop between starts, Jc. If three'(3)' starts have been. attempted within antwo hour peri 5d, e pump must be. idle for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> before the fourth attempt ist.made. <

f' O'

.' QUESTION,'7.05L "

(1.50)

During an' accident recovery procedure, .the operator who has been.

'Lmonitoring Critical Safety Functions reports the.following: ,e

-a. Containment , Yellow' i iDj

-b' . Core Cooling --Orange-

.c. Heat Sink - Red d-. Integrity - Orange

e. .Suberiticality - Red
f. Inventory - Yellow Rank <the above conditions according to their importance11n order from th'e.

most~to least important.

(t**** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

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q. PBQQE'd ES - NOEMAL ABNQ M % c EMERGENCi AND PAGE 1.)

[hj J7.Q " f lS$D. 'QAL CONTRQk j ,

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r

, r'- QUEST. ION!'7.06L (1.25) frT(li. .

HV/U C"APA-ZZn00100'(Procedure Requirements) defines the terms to'be used in 3

JU procedures. Match the following terms: MAY, MUST, SHALL,. SHOULD, and WILL.

with the following-definitions: -j y ,

1 ar. This term is. considered equivalent to "shall" except when used to c -

' denote simple futurity. (0.25)

'V M p,'.

i G

' ~ IE ' "b. This' term is used to-denote' legally binding requirements to'which UE l xL f'//

r' management has committed.(e.g~., in the FSAR)'. (0.25) 9 TU' c. This term is used to denote permission to perform activities. The term, when used,'is'neither a requirement nor a recommendation. ~

. , , .(0 25)-

l

@[. d. This' term is used to denote requirements imposed by UE management on its employees, contractors, and agents which are above and in excess of the legally binding requirement of the appropriate regulatory body. (0.25)

e. This, term is used to denote recommendations imposed by UE management onLits-employees, contractors, and agents which should be met unless-there is sufficient reason not to perform the activity. (0.25)

QUESTION 7.07 (1.50)

List six1(6) indications / symptoms that;would identify,a RCS leak as being

. located INSIDE containment.

a s -

QUESTION 7.06 (2.00)

ECA-0.0 (Loss of All AC Power) states in Note-2 that "CSF Status Trees should be monitored for information only. FRGs should not be implemented." Give TWO reasons why this note is included.

QUESTION 7.09 (1.00)

Callaway is. preparing to refuel.the reactor. After removing the gate between the spent fuel pool and the fuel transfer canal, what additional ,

means for monitoring refueling pool level is available per OTG-ZZ-000077 i

(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

c

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7. PBOCEDURES - NOBdbL1 ABNORMAL. EMERGENCY AND PAGE 1el

. B6DIOLOGICAL CONIB9L I

i QUESTION 7.10 (2.50)

Answer the'following questions concerning Reactor Trip Bypass Breaker l Operation (OTS-SB-00001):

a. How long may reactor trip breakers be bypassed? ( 0 r,)
b. When bypassing a reactor trip breaker, LIST THREE indications that are verified to make sure that the bypass breaker is closed. (1.5)
c. What lines of communication must be established prior to bypassing a reactor trip breaker? (0.5)

QUESTION 7.11 (2.00)

) '

In procedure FR-S.1 (Response to Nuclear Power Generation), one of'the stopa-has the-operators initiate Immediate Boration. As a sub-step, PZR pressure is checked to see that it is less than 2335 psig.

a. What is the basis of this sub-step? (1.0)
b. If PZR. pressure is greater than 2335 psig, one of the " response not obtained" actions is to ensure that the containment purge and mini-purge dampers are closed. What is the basis for this action?

(1.0)

I QUESTION 7.12 (2.00)

Concerning a Steam Generator Tube Rupture at the Callaway Plant, LIST FOUR.

reasons why isolation of the ruptured steam generator is necessary.

(***** CATEGORY 07 CONTINUEL ON NEXT PAGE *****)

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7. EBQQEDURES _FORdakt_ARNORMAh, EMEBGENCY AND fab 2 1!i

, BADlQLOGICok_RedIBQL  ;

I l

I QUESTION 7.13 (2.00) i Answer the following concerning a Reactor Startup (OTG-ZZ-00002),

a. If criticality is not reached by the time that the control rods reach the maximum limit specified in the ECP calculation, how is the startup termiaated? (0.5)
b. As reactor power passes 1E-10 amps and the P-6 permissive light is energized, LIST FIVF actions / verifications that are performed to ensure proper transition from the source range to the intermediate range. (1.0)
c. At 1E-8 amps, data must be recorded. LIST TWO parameters (other than time) that are entered into the UR0s logs. (0.5)

QUESTION 7.14 (1.25)

It is determined the.t the control room must be evacuated due to a fire.

According to the Control Room Inaccessibility Procedure (OTO-ZZ-00001),

who is responsible for the following actions?:

a. Manually trip the reactor at RLOO3 or RLOO6
b. Go immediately to the Auxilir.ry Shutdown Panel
c. Trip all RCPs (if possible) from the Main Control Room
d. Locally verify that the turbine has tripped.

i

c. Determine if a Loss of Offsite Power has occurred at NB02 QUESTION 7.15 (2.00) l l

Per emergency procedure E-2 (Faulted S/G Isolation), LIST SIX actions taken to isolate the faulted S/G. (Assume that the MSIVs/ bypasses are already shut and checked.)

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(***** END OF CATEGORY 07 *****)

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81 APEINISTRAllYElBQQEDURES. CON'DITIONS. AND-LIMITATIONS PAGE 10 l}

1 l ' QUESTION ~ 8;01 (1.00)

~(Choose!the correct response)

With reactor ' power at . 65%,1 penalty deviation outside the delta-I t arget band?shall be~ accumulated on a time basis of .

a. One minute penalty for'each minute outside of the. target band.
b. One-half minute penalty.for each-minute.outsiderof the target band.

c.

~

lOne minute penalty.for each one-half minute outside of the target band.

d. Zero penalty for.. time.outside the target band.

I'

, QUESTION- 8.02. (1.50)

a. Per plant safety limits (Technical Specification 2.1.2), what. limit N

.is placed1on Reactor Coolant Pressure? (0,5)

L b. .What TWO (2) actions, per Technical Specifications,.are required WITHIN 1 HOUR if-this' limit is exceeded in MODE 1 conditions? (1.0)

-QUESTION 8.03 (2.00)

For each component / instrument listed below, how many must be operable per >

Callaway Technical Specifications for the operational mode stated? .!

a. 'PZR Code Safety Valve (Mode 4) f
b. PZR Pressure Channels (Mode 1)

(for Rx Trip Function)

c. PZR Water Level (Mode 1)

-(for Rx-Trip Function)

d. RCS Flow. (Mode 1)

(Single Loop (above P-8)) '

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

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?8; ADMINISTRATIVE PROCEDUBES. CONDLIl0HSi_AND_LitU.IAIIONS -PAGE _.1 QUESTION '8.04- -(1.50)

TRUE'or FALSE 7 La . ' Personnel do not have to be designated by name tofbe included in a General Radiation ~ Work Permit.

lh Entry'into'a' Radiological Controlled. Area' requires an RWP.

c. Personne1'are required to sign the RWP Sign-In Sheet and review the RWP only on the first daily entry on that particular RWP.

m

-QUESTION 8.05 -(l'50)

Answer the following.concerning.EIPs:

a. While implementing Callaway EIPs,.who, by title, may authorize radiation' exposures in excess of 10 CFR 20 limits? (0. 5 ). q

'b.. Under what FOUR conditions must the Technical Assessment Procedure' (EIP-ZZ-00213)'be initiated?'~~(1.0)~

-QUESTION 8.06 ( 1. 00 ')

Which ONE of the following items IS NOT the responsibility of the Shift Supervisor regarding Containment Access and Integrity (ODP-ZZ-00019)?

a. Granting permission to establish access to the reactor building.
b. Retaining the key and locking device for the access portal.

c Ensuring that proper security logs are concluded upon completion of.

containment access.

d. Determine the condition of the environment inside containment.

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

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_.e 28, =ADMINIEIBATIVE PROCEDURES. CONDITIONS 1 _AND LIMITATIONS PAGEL 11 t

QUESTION 8.07 (1.50)-

JTRUE or FALSE?.

Ca. The Hold ^0ff. Tag.is used where hazardousEor, abnormal conditions exist.

~

b.c A Switching Order-should be' issued on 480:V'and above breakers

-and disconnect switches.

Lc. L0nly'one (1): Restraint may be used:on a given-piece of equipment'ato one time.

QUESTION' 8.08 '

'(1.00)'

Fill in-:the. blanks for the followingustatements concerning.0DP-ZZ-00016.

(WatchstationLEquipment Logs and Practices) a.. All--abnormal / unusual readings on log sheets"should be-and . .(0. 5)'

(Each. blank contains-more'than one word).

. b. As a minimum, complete watchstation; tours should be made during each' shift. (.25)' ,

1

c. If the BOP computer is down for or more consecutive' hours, then Control Room Logs - Computer'DownLAddendum must be taken, (.25)j; 1

1 QUESTION- 8.09 (1.50) l Briefly DESCRIBE what le included in a Mode Change Checklist. (1.5)

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

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L' , 8. bDdlHISIBAIIVE PRQQEDHEES. CONDITIONS, AND LIMITATIONS PAGE. 1Ll r

QUESTION 8.10 -(2.30) l:

Technical Specification 3.7.1.3 states that the Condensate Storage Tank

'(CST) must 1xt operable. .

, i l a. What constitutesvan' operable CST 7 (0.5)

L

b. .The action if the CS's is inoperable is to restore the CST or be in

. Hot; Standby in at leaet 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. To avoid this' action, what system must you demonstrate. operability of? (0,5) l' c '. What is the basis for the CST Technical Specification? (1.3)

QUESTION 8.11' (1.50)  ;

i Technical Specification 3.9.10.2 states that at least 23 feet of water shall be maintained over the top of the irradiated fuel assemblies within the reactor pressure vessel. What is the basis for this specification?

QUESTION 8.12 (2.00)

Answer the following-questions concerning DNB Parameters Technical Specification 3/4 2.5:

a. LIST the three (3) DNB parameters that must be monitored. (1.5)
b. Each of the above parameters must be monitored every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Given;

'the following times that the parameters were monitored, when is the latest time that the parameters can be monitored and still comply 2

with' applicability requirement 4.0.2 surveillance requirement (time intervals).

Day Time 1 1 1000 1 2358  ;

2 0956 2 2041 1 3 1110 i l

l QUESTION 8.13 (2.50)

Technical Specification 3/4.5.2 states that two (2) independent ECCG l subsystems shall be operable at greater than 350 degrees F. LIST what l comprises an operable ECCS subsystem.

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

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!H. -ADMINIS'TRAIlYE PROCEDURES. CONDITIQHS, AND-LIMITATIONS PAGE 24 i

i QUESTION _ 8.14 -( 1. 50 ).

ALDesignated Management Representative (DMR) has the. authority to approve

' hot work on the Hot-Work Permit. LIST THREE conditions when the DMR must' only'be-the; Shift Supervi~sor or Operations Supervisor.

' QUESTION - - 8'.15 -(1,20)

LIST what Work Permit Priority Number.(E, 1, 2, 3) would: apply for'the following conditions. (Each condition may have more than one priority)
a.  : Requires an approved work permit.

.b. Used to.: satisfy'a Technical Specification Limiting Condition for Operation,

c. . 'Used to; maintain the plant at full power.
QUESTION 8.16 (1;50)

Control! Room' personnel receive.an alarm on Fire Detection Zone 306 (Cable.

Spreading' Room). Ituis determined that the alarm.was caused by srinding

.and welding being performed in the area. More than half of the. function-A

' fire detection instruments for Zone 306 are declared inoperable. What does Technical Specifications require you to do? (Include applicable time limits)-

?

(***** END OF CATEGORY 08 *****) i

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

c .c i 5

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5. _THEgRY OF: NUCLEAR POWER PLANT OPERATION. FLUIDS. AND .PAGE~ 2:

, , :IBEBM0 DYNAMICS. 1

--ANSWERS ---CALLAWAY -57/09/22-SUNDERLAND' P. ,

)

. ANSWER 5.01 <(1.00) J

' Increase.

REFERENCE' . . ..

-Westinghouse. Thermal-Hydraulic' Principles and Applications to the PWR-II,. l pg. 11-27 .

i 191002K108 ...(KA'S).

.t ANSWER 5.02 (1.00)

True.

REFERENCE

' Westinghouse' Thermal-Hydraulic Principles and Applications to the PWR-I, pg. 3-73 193008K101- 19S008K103 ...(KA'S)

I ANSWER. 5.03 (1.00)

c. (1.0)' t REFERENCE i Westinghouse Reactor Core Control for Large PWRs p.4-5 004000A101 ...(KA'S)-

ANSWER. 5.04 (1.00)

False.

REFERENCE Westinghouse Thermal-Hydraulic Principles and Applications to the PWR-II,

,pg. 10-47 191004K101 .191004K106 ...(KA'S) l

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5. -IHEQBY;QE_HERLEAR POWER-EL6NT OPEBATION. FLUIDS. AND PAGE 2:

L-

, IBEBdODYNAM19S-

!s. ' ANSWERS -- CALLAWAY- -87/09/22-SUNDERLAND, P.

ANSWER- 5.05- (1.00) b REFERENCE Westinghouse Thermal-Hydraulic Principles and Applications to the PWR-II, pg.14-27f 193008K121 ...(KA'S)

ANSWER 5'06 (1.00)

a. _]

REFERENCE .]

Westinghouse Reactor Core Control for Large:PWRs, pg.'2-23 192003K106 ...(KA'S) f I

LANSWER 5.07: (1.00) e b.

REFERENCE' Westinghouse FundamentalsLof Nuclear Reactor Physics, pgs. 7-20, 21 '

192008K112 ...(KA*S)

ANSWER 5.08 (1.00)

(0.5 pts each)

a. Nil Ductility Temperature (accept NDT, RT-NDT, NDTT)
b. Cool-down' REFERENCE .

l

. Westinghouse Thermal-Hydraulic Principles and Applications to the PWR-II, '

pgs.-13-56,-60, 67  !

193010K104 193010K105 ...(KA'S) ,

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u _-- - - - J

ig

5. . THEQBLQE_HILCLEAE_E0 WEB _EL6HLQEEBAll0N. ELHIDS.- AND PAGE 2"
IBERMODYNAMIGS H p ,

ANSWERS ---CALLAWAYL -87/09/22-SUNDERLAND, P.

" ANSWER 5.09L (1.50)

a. 96 steps on bank D (+/- 10 steps)L REFERENCE . . .

Westinghouse Reactor Core Control'for Large Pressurized. Water Reactors, pgs. 9-15, 9-16 001000K501 001000K502 015000K505 ...(KA'S)

ANSWER 5.10' -(1.50)

The ECC is lower-[0.75) because the three hour delay allows-Xenon concentration to decrease'[0.75].

' REFERENCE Westinghouse ~ Reactor Coro Control'for Large PWRs, pg. 4-22

.l001010K526 192008K105 ...(KA'S)

.k

.ANSVLA' 5.11 (1.50)

(0.5 sta each). .

-a.-  : INCREASES j

b. INCREASES q
c. DECREASES a l

REFERENCE  !

. Reactor Core Control for Large PWRs p. 3-12 through 3-43 l Thermal Hydraulic Principles and Applications to the PWR I p. 4-69, 4-70, and 6-30 002000K507 002000K511 192004K103 192004K107 ...(KA'S)

)

l

6 g p

! .i .* >

15. THEQBY'OF NUCLEAR _EQWEB_Ek&MT OPEBAIION. FLUIDS. AND PAGE D-

.IBEBMQDXNAMICS ANSWERS'-- CALLAWAY -87/09/22-SUNDERLAND, P.

o ANSWER 5.12 (2.00)'

' Fuel'in the; vicinity of the inserted _ rod experienced lower Xe and Iodine Leoncentrations'due to flux depression [0.5]. When.the rod was pulled back:

to position, flux in.the region increases markedly [0,5]. Xenon burns out rapidly ~in the higher flux [0.3]. This all results in cevere power-peaking in1the region [0.7]. [

(Partial credit of up to 1.0 pts may be given for an answer that talks j Laboutlthe rest of the core.being at higher power because flux is suppressed in the region with the stuck rod.)

REFERENCE Westinghouse Reactor Core Control for Large PWRs, pg. 8-32

000005K103 ' , 001000A203 001000K507 192005K116 192006K108

...(KA'S)*

ANSWER 5.13 '(2.00)

(0.5 pts each) a.- Decrease

b. Increase
c. Increase
d.  : Increase-REFERENCE Westinghouse Reactor Core Control for Large PWRs, pgs. 6-22 through 6-27 001000K509 192005K106 102005K107 ...(KA'S)

ANSWER 5.14 (1.00) i (Any 2 @ 0.5-pts each)

' 1. Hydrogen generation

2. Cladding degradation (credit should be given for the results of cladding degradation contamination, etc.) such as fission product release, high radiation,lI
3. Heat (exothermic reaction)

REFERENCE Westinghouse Thermal-Hydraulic Principles and Applications to the PWR-II, l pg .' 13-6  !

_193009K105 ...(KA'S)

o m .

  • 1 1 Si THEQBY 0F-NUCLEAB POWER' PLANT OPERATION, FLUIDS 'AND .PAGE' 2!l

.- IHEBMODYNAMICS-A'NSWERS'- CALLAWAY -87/09/22-SUNDERLAND P.

.j 1 ANSWER 5.15 (2.00) i

a. DNBR :.(Predicted) CHF-Actual HF. (0,5) a
b. Decrease. -( 0. 5 ) . . .

)

Decreasing pressure lowers the margin to DNBR.by allowi'ng the coolanti to approach saturation saturation pressure more closely. (1.0) j

' REFERENCE ,

Westinghouse Thermal-Hydraulic Principles and Applications to the PWR-II, j pgs. 13-22, 24 i 193008K105 193008K110. ...(KA'S) 1 ANSWERL l 5.16 - (1.50)

Flow rate per pump decreases to below the flowrate achieved with one-pump

[0.75), therefore, .the power'needed to operate the pump' decreases and.'

' fewer ampsfare' drawn [0.75].

REFERENCE

  • Westinghouse'e-Thermal-Hydraulic Principles and Applications to the PWR-II, pgs. 10-36,'44, 46 191004K107 191004K109 ...(KA'S) i ANSWER 5.17 (1.50)

N41 N42 .N43- N44 Upper 159.7 139.5 157.0 147.1 100% 266.6 234.5 262.9 241.6. . j Normalized .599 .595 .597 .609 Ave =.60.

QPTR 1.02 Lower 166.0 145.1 160.7 150.3 300% 278.6 236.7 270.0 '252.3 LNorma14.ze'd- .596 .613 .595 .596 Ave:.60 I QPTR 1.02 QPTR = 1.02 REFERENCE

' Westinghouse Reactor. Core Control for Large Pressurized Water Reactors,

) '

pgs 8-29 through 8-31 l

l .

l

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

7 ie -  ;-,

9 lc ;)--i

, 3 p? "

s

f. .

a L- ,5. .THEQRY DE_NHCLEAE_E0WER PLANT OPEBAIION1 _ELUIDS2_AND .. PAGE.' 28

,. IBERMODYunMICS L ' ANSWERS:-- CALLAWAY -87/09/22-SUNDERLAND,:P.

015020K504 015020K508 ...(KA'S)

II (

L ANSWER' - 5'.'18 -(1.50) l

.a. INCREASE

'b. " REMAIN THE SAME' c.

DECREASE (0.5 pts;each)~

REFERENCE

' Westinghouse FundamentalsLof Nuclear Reactor. Physics, Chapter 8, pgs. i 12-19' i 015000K506 192008K104, ...(KA'S) l,

~

')

I I

.j l

l

.lE--.-----._ - . - - - . .

{

. I

6. PLANT SYSTEMS DESIGN, CQNTROL, AND_INSTRUMENTAIIQN PAGE 2 ',

ANSWERS -- CALLAWAY -87/09/22-SUNDERLAND, P.

ANSWER 6.01 (1.50)

(any 3 @ 0.5 pts each)

1. Liquid Radwaste Discharge Monitor (HB-RE-18) (Plant Discharge Monitor)
2. S/G Blowdown Discharge Monitor (BM-RE-52)
3. Turbine Building Drain Monitor (LE-RE-59) (Oily Waste Discharge Monitor)
4. Secondary Liquid Waste System Monitor (HF-RE-45)

(Identification numbers accepted for full credit, but not required)

REFERENCE Technical Specifications, Table 3.3-12 073000K401 ...(KA'S)

ANSWER 6.02 (1.00)

Valve FCV-111B shuts (1.0)

REFERENCE Licensed Training Lesson Plans, Ch.11, p. 2-9 000022G005 ...(KA*S)

ANSWER 6.03 (1.00)

You must lubricate (close fitting moving parts) with water. (1.0)

REFERENCE OTS-KE-00006 034000K601 ...(KA'S) l

h I '

+

_ :6. PL6NT - SYSIEMS: DESIGH CONTROL'.;AND INSTRUMENTATION PAGE. 2E

~

' ANSWERS '-- CALLAWAY . -87/09/22-SUNDERLAND, P.

ANSWER- 6.04 '( 1. 50 ).

The. motor operated gate valves are' located at an elevation that is susceptible 7to flooding during cold shutdown (0.75), hence, two vent valves assure the capability to vent B and C accumulator tanks. (0.75)-

REFERENCE

' Licensed Training Lesson Plans, CH. 19, P. 3 Learning Objective 190 006020K603 ...(KA'S).

JANSWER 6.05' (2.00) a '. >78%' level in a S/G (0.5) b., .2/4 levels (0.25) on 1/4 S/Gs (0.25)

c. (0.2'ench)~
1. Closes FRVs Closes FRV bypass valves 2.
3. . Trips MFP

'4. Trips Turbine

5. Closes feed-isolation valves (Note: Accept FWIS for answers 1, 2, and 5) i

-REFERENCE Licensed Training Lesson Plans Ch. 27, TP4 Learning Objective 27E

.012000K610 ...(KA'S) i ANSWER 6.06 (1.25)

(Anyl5 at'0.25 pts each).

. l'. Neutral ground overcurrent relay

2. Voltage restrained overcurrent i
3. Reverse power 4'. Underfrequency l
5. Loss of field ,

6..Overcurrent- l

' REFERENCE

!. Licensed Training Lesson Plans Ch. 3, TP22 5

E . Learning Objective 3J 4

,---,g ..

' ~

s IE .,.

G:

y " 6 '. PLANT' SYSTEMS DESIGN, COBIBQL, AND INSTRUMENTATION PAGE 2f ANSWERS 1--1CALLAWAY' -87/09/22-SUNDERLAND, P.

064000K402 ...(KA'S)

E

-( ANSWER 6.07: (1.50)

(0.25 pts each).

.a. open b: no effect

'c, 'open d '. - open,

e. no effect 4
f. no effect-  !

i REFERENCE'. .

I LicensedLTraining Lesson ~ Plans Ch 30, SNP-LOG-11-1

' Learning ObjectiveL30N-016000K108 ...(KA'S).

. ANSWER 6.08 .(1.00)

Trip ofKboth main feedwater pumps.(1.0)

REFERENCE Licenaed Training LeLaon Plans Ch~25, p. 7 .

061000K107' 061000K402 ...(KA'S) 1 ANSWER '6.09 (2.50)

a. .(0.3 for signal, 0.1 each for setoint and coincidence) i
1. Containment Hi(-2) Pressure, 17 psig, 2/3

.2. Low Steamline' Pressure, 615-psig, 2/3 any steam line

3. High' Steam Pressure Rate,-100 psig in 50 sec, 2/3 any steam line  ;
b. 5 sec (.5)
c. The' fast close input' signal must be cleared. (0.5)

' REFERENCE Licensed Training Lesson Plans Ch. 4, p. 5,7,16 I Learning Objective 49D,F,I 039000A302 039000K405 ...(KA'S)

L l'

L L

l

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

[h .

, i

? -

t D'

L 16 .- ELANT-SYSIEMS DESIGN, CONTROL, AND INSTRUMENTATION' '

PAGE ' 3(

V.

l' ANSWERS'-- CALLAWAY -87/09/22-SUNDERLAND,JP;

'dNSWER. 6.10 (1.50).

a: Pawls' engage the ratchet plate preventing reverse rotation (0.75)-

b. Reverse' flow would cause reverse rotation of the pump. If you attempted to start the pump, high-starting currents would result.' this could cause overheating of the pump motor. (0.75)

(Accept variations,1however, for full credit, overheating ~of the pump motor must be' covered)

REFERENCE Licensed Training Lesson Plans Ch.9, p. 15 Learning Objective 9H 003000K405- 003000K608 ...(KA'S)

ANSWER 6.11 (2.00)

a. 1. Containment Pressure High-High (0.5) at 27 psig (0.15), 2/4 (0.05)
2. Manual Actuation (0.5), 2/2 (0.05)
b. (0.25 pts each) ,

1., Containment Spray Pumps Start

2. Loop Isolation Valves Open (HV-6,12)
3. Spray Additive Tank Isolation Valves Open.(HV-15,16)

(Valve #s accepted but not required)

REFERENCE-Licensed Training Lesson Plan Ch. 18, p. 4-7 Drawing M-22EN01 (Q)

Learning Objective 18G,H 000009A107 022000A102 026000A301 026000G015 ...(KA'S)

ANSWER 6.12 (2.00)

a. Provides holding power for any one rod group during maintenance.

(1.0)

b. 4 (0.5)
c. .1 (0.5) 1

~

~

6. ELANT SYSTEMS DESIGN. CONTROL. AND_lBSTRUMENTAIION PAGE 3:
' ANSWERS -- CALLAWAY -87/09/22-SUNDERLAND, P.

REFERENCE Licensed Training Lesson Plans Ch. 26, n. 2,16,17, TP SNP-CR-16

, Learning Objective 26D 001000K403 ...(KA'S)

ANSWER 6.13 (2.25)

a. Shutdown 1. Diesel output bkr shut (0.25) and
2. Preferred Power Supply Bkrs Open (0.25)

(Give 0.25 pts for undervoltage)

LOCA 1. SIS (0.2) or

2. CSAS (0.2) and
3. either no undervoltage (0.05) or DG breaker closed (0.05)
b. The pump would continue to run. (0.5)
c. (0.25 pts each)
1. Starts the diesel
2. Sheds all loads from the effected bus
3. Blocks the LOCA sequencer from starting directly from SIS or CSAS (also accept delays the LOCA sequencer until DG breaker closed)

REFERENCE Licensed Training Lesson Plans Ch. 6, p. 7-8, TP2 Learning Objective 6F 064000K410 064000K411 ...(KA'S) h 9

9 s -

a J

s y

6. ELadT SYSTEd.S DESIGE1_GQHTROL, AND INSTRUMENTATION PAGE 3; ANSWERS -- CALLAWAY -87/09/22-SUNDERLAND, P.

, ANSWER 6.14 (2.50) a.. Prevents DNB (0.4) for slow transients (0.1)

b. By' comparing the difference between the upper and lower power range NIs or (P-top) - (P-bottom) (either for 0.5 pts)
c. 1. Loop Average Temperature (0.25)
2. Pu. . Pressure (0.25)

. d. The improper scaling results in all four channels being affected (as opposed to one channel for the instrument channel failure) (0,5).

'OTdelta-T could be above the value where DNB would occur and the plant would be potentially unprotected (depending on the magnitude of the scaling error) (0.5).

REFERENCE Plant Technical Specifications p. B 2-5 LER 87-012-01 Westinghouse Reactor Core Control for Large PWRs p. 8-25 Licensed Training Lesson Plans Ch. 27 Learning Objective 27D 012000A101 012000A206 012000K402 012000K602 ...(KA'S) s4 i ANSWER 6.15 (1.50)

(0.5 pts each)

.l

a. 125 (between 119 to 128 psig (the span of the lead compressor) is

, acceptable)

b. 20 l c c. lead REFERENCE Licensed Training Lesson Plans Ch. 14, p. 4,5 Learning Objectives 14C(2e,4), 14D 078000A301 078000K401 ...(KA'S) i we 4F " [ . [ , '
7. EBQGEDUBES - hxEMAL&_ABHQBMAL&_EMEBGENCY AND PAGE 3' BARIOLOGIQaL CONTBQL ANSWERS -- CALLAWAY -87/09/22-SUNDERLAND, P.

ANSWER 7.01 (1.00) d.

REFERENCE OTO-SF-00003, pg. 1 000003K101 ...(KA'S)

AUSWER 7.02 (1.00)

(0.5 pts each)

a. plant vent or containment purge (either for full credit)

.b. 10 REFERENCE OTG-ZZ-00003, pgs. 1, 2 002000K111 002020K508 029000G010 039000G010 039000K104

...(KA'S) 1 ANSWER 7.03 (2,50) l

a. Step 4 or when SI is determined not to be cativated or needed. (0.5)
b. (0.3 pts each; 0.2 for the parameter and 0.1 for the condition)
1. RCS subcooling - more subcooled than instrument error steam J pressure table.
2. RCS hot leg temperatures - stable or slowly decreasing .
3. RCS cold leg temps - near saturation temperature for steam i pressure
4. Core exit thermocouple - stable or slowly decreasing ,
5. S/G pressure - stable or decreasing l
c. Use ES-0.0 (Rediagnosis) (0.25) Or if an SI has occurred, use E-0 at i step 1. (0.25)

REFERENCE Emergency Procedures E-0 p.3, ECA-0.0, ES-0.0 p.1, ECA-0.1 p.13 ,

Callaway SRO Annual Requal Exam (6/4/86) 000007K301 000038G012 000038K306 000055A202 000065K302

...(KA'S)

g:- y; r- - - -

4 t

17. PEQCEDUBES - HORMAL. ABNORMAL EMERGENCY'AND PAGE 3 L BAD 19 LOG 196L CONIEQL ANSWERS *-- CALLAWAY -87/09/22-SUNDERLAND,:P.'

1 ANSWER 7.04' (1.50)

(0.5; pts each)

a. ' False  !

l b. True- l

-c. .True

. REFERENCE ES-0~.2,LAttachment 1, pg. 1 i

'003000K614- ...(KA'S)

ANSWER 7.05: (1.50)

(.25 pts each)

1. e. v
2. c
3. b
4. d
5. a
6. f

, (Note to grader: Do not penalize the examinee for having one item-out of order. .

If one is out of order and the rest of the five-items are in order when the incorrect one is' thrown out, then deduct only 0.25 pts.)

REFERENCE .

I EOP CSF-1, p. 2 000011SG11 ...(KA'S) l ANSWER 7.06 (1.25)

a. WILL ,
b. SHALL  ;
c. MAY
d. MUST I
e. SHOULD (0.25 pts each)

REFERENCE L

APA-ZZ-00100, definitions

'194001A102 ...-(KA'S)  ;

V

7. PRQCEDUBES - NOBdala_ABBQBBAIu_ EMERGENCY AND PAGE 3!

. RADIOLOGICAL QONTBQL L ANSWERS -- CALLAWAY -87/09/22-SUNDERLAND, P.

ANSWER 7.07 (1.50)

(Any 6 @ .25 pts each)

1. Containment atmosphere radiation monitor above normal or alarming
2. Containment area monitors above normal or alarming
3. Containment temperature above normal
4. Containment humidity above normal
5. Containment air cooler drain levels increasing
6. Containment sump level high alarms
7. Increased frequency of containment sump pump operation
8. Containment pressure above normal REFERENCE OTO-BB-00003 000009A202 000009A210 000009A211 022000A405 ...(KA'S)

ANSWER 7.08 (2.00)

1. This procedure has priority over all FRGs [0.5) and is written to implicitly monitor and maintain CSFs [0,5],
2. All FRGs are written assuming at least one AC emergency bus is energized. (1.0)

REFERENCE ECA-0.0, pg. 2 ECA-0, Series Walkthrough, pg. 1 000056K302 ...(KA'S)

ANSWER 7.09 (1.00)

Spent fuel pool level alarms and indicators REFERENCE OTG-ZZ-00007, pg. 12 034000A102 ...(KA'S)

I I I h  :: 7 '. ~PRogEDUBES~- NORMAL LABHQBMAL, EMERGENCY AND' PAGE.L3!'

_. <,BbDIOLOGICAL CON.IBQL fl.

ANSWERS - CALLAWAY ' 87/09/22-SUNDERLAND, P.

p r

I:$

IANSWER .7.'10 (2'.50)

a. 2 hrs.(0.5) zb.- 1. ' Red closed flag on breaker
2. . Bypass breaker. indicates closed ~on NCB 3 General' warning.on SSPS
f. - '4. -Reactor Bypass Breaker ALor.B closed' alarm (Any 3"of 4.@ 0.5 pts'each)
c. RO'tolperson-performing procedure (0.5)-

REFERENCE OTS-SB-00001- H 000007K202 000007K203 012000A203 ...(KA'S)

' ANSWER. 7.11 '(2.00.)

a. Alert operatorito a condition where charging or SI pump flow into the'RCS is reduced [.'25] and, therefore, boration is reduced [.75].
b. The rupture disc.could have burst (from the FORV lift) [.25). The ,

closed dampers would be a barrier to radiation release [.75), i REFERENCE '

o . Red Path FRG Walkthrough Guides, FR-S.1, pg. 5-6 000024K302 000029K311 000029K312 ...(KA'S)

) l I

. 1

\

.t

7. EBQQEDUBES NORMAL ABNOBMAL. EMEBQSNCY-AND- 'PAGE . 3 '.

., .B6210 LOGIC 6L CONTBQL'

_ ANSWERS - ;CALLAWAY -87/09/22-SUNDERLAND, P.

ANSWER: 7'12

. (2.00)

(Any 4 @ .50 pts.each)

' 1. , Minimize' radiological releases

2. Maintain ruptured SG' pressure greater than non-ruptured SG. pressure-
3. RCS pressure reduction will stop the leak when. delta - P = 0
4. Non-ruptured S/Gs will_ remove heat 5< Subcooling will be maintained 6.- Prevent future overfill

. REFERENCE

.E-3 Series Walkthrough..pg. 6-10 000038A132 000038K306- ...(KA'S) 1 4

l I

i i l

I i

L , .- :_-_____-_-______-__ - - _ _ _ _ _ _ - _ - _ - _ .

. V

) L7i PEOCEDUBES' : NORMAL ABNQRMAL, EMERGENCY AND PAGE St 7

. ;BADIOLOGIgAL CONIB9L

+

ANSWERS -- CALLAWAY -87/09/22-SUNDERLAND, P.

4 f ' l ANSWER ' '7.13 (2.00)

a. Control bank D is fully inserted. (0.5)

'(.25 credit for inserting all control rods)

' b .' '(Any 5'@ 0.2 pts each)

1. Verify one decade of overlap
2. Indication on both SR + IR SUR instruments

(.125 for only.IR)

3. Transfer the inputs to recorder NR-45 to the irs.

'4. Block the source range HIGH FLUX REACTOR TRIP y

5. . Verify the source range trip A and B block status lights illuminate
6. Verify that the high voltage source has been removed from oi the source range detectors, i
c. (Any 2 at'.25 pts each) i
1. RCS T-avg
2. RCS Boron Concentration 4

-3. Control rod bank position REFERENCE OTG-ZZ-00002, pgs. 9, 10 r 001010A207 192008K112 ...(KA'S)

' ANSWER 7.14 (1.25)

(0.25 pts each)'

ca. BOP RO b, SS

c. SS  :

.d.' BOP RO

e. URO 4 REFERENCE OTO-ZZ-00001, Att. 2, pg. 1; Att. 3, pg. 1; Att. 4, pg. 1 000068G006 000068K312 000068K318 ...(KA'S) i L _ . _ _ _ _ _ _ _ _ _ _______________d

n'-.m

.- y:- . , , ;-=- - -- - - - ---_a;7,--

3 .

hi 1

- r i 1L '

, -7. PBQDEDUEES" 'NoBMAL1_ ABNORMAL 2_ EMERGENCY AND PAGE 3:!

. . BADIRLOGICAL CONTROL.

-ANSWERS - CALLAWAY -87/09/22-SUNDERLAND, P.

ANSWER 7.15 .(2.00) f' . (Any 6 @ .33 pts each).

1.. Isolate 1MFW isolation valves

2. Is'olate FRVs 3'. < Isolate FRV bypasses g 4. Isolate all AFW flow 5; Isolate steam supplies to TD AFW-pumps

'6. Verify l atmospheric steam dumps closed 7

7.. Isolate FW chemical ^ injection j 1:.

B.- Isolate'S/G" blowdown isolation. valves H REFERENCE.

Emergency Procedure-E-2, pg. 4 of 5 000040A102 000040A103 ...(KA'S) l i

)

p l-l.

U, t

l 4 , 85 ADMINISTRATIVE PROCEDURES. CONDITIONS.'AND LIMITAIION'L S PAGE^ 4(

c: .

ANSWERS -- CALLAWAY -87/09/22-SUNDERLAND, P. .

i 1

' ANSWER 8.01 (l.00)

a. i LREFERENCE Technical Specifications, pg. 3/4 2-2

'015000G011 ...(KA'S)

ANSWER 8.02 (1.50)

a. 2735 psig (0.5)'
b. 1. :Be in hot standby [0.3] with pressure below its limitE[0.2]:
2. Notify the1NRC. (0.5)

REFERENCE Technical Specifications, pgs. 2-1, 6-14 010000G005 010000G014 ...(KA'S)

ANSWER 8.03 (2.00)

(0.5 pts each)

a. 1
b. 3'
c. 2
d. 2/ loop;

' REFERENCE Technical Specifications, pgs. 3/4 4-7, 3/4 3-2, 3/4 3-3 000028G008 002000A105 002000G011- 010000G011 ...(KA'S)

ANSWER 8.04 (1.50)

(0.5 pts each)

a. True L b. True
c. False

4

8. ADdlBISIBATIVE EBQQEDMBES. CQNDlIlQNS, AND_DIMIIATIONS PAGE 4:

ANSWERS -- CALLAWAY -87/09/22-SUNDERLAND, P.

REFERENCE APA-ZZ-01002, pg. 1-6 194001K103 ...(KA'S)

ANSWER 8.05 (1,50)

a. Emergency Coordinator (0,5)(either answer for full credit)

Recovery Manager

b. Alert (0.25)

Site Emergency (0,25)

General Emergency (0.25)  !

As determined by the Recovery Manager or Emergency Coordinator (0.25)

REFERENCE EIP-ZZ-00213, pg. 1  !

EIP-ZZ-00210, pg. 1 194001A116 194001K103 ...(KA'S) l ANSWER 8.06 (1.00) 1

b. ,

I REFERENCE  !

ODP-ZZ-00019, pg. 1-4 i 194001A103 194001A112 ...(KA'S) 1 I

I ANSWER 8.07 (1.50) i

a. False (0,5) )
b. False (0,5)  !
c. True (0.5)

REFERENCE APA-ZZ-00310, pg. 1-8 194001K102 ...(KA'S) l l

1

1 3 ADMINISIBATIVE PROCEDURES, CONDITIONSz_AED LIMIIAIIONS PAGE 4:)

. \

ANSWERS -- CALLAWAY -87/09/22-SUNDERLAND, P.

i I

1 ANSWER 8.08 (1.00) i

a. (Any 2 @ 0.25 pts each)
1. Circled in red
2. Annotated in the remarks section
3. Reported to the Control Room (In Any Order)
b. Two (2) (.25)
c. Four (4) (.25)

REFERENCE ODP-ZZ-00016, pgs. 6, 7, 12 194001A106 ...(KA*S)

ANSWER 8.09 (1.50)

Listing of conditional surveillance [0.4], surveillance with a periodicity of less than one week [0.4], and administrative reviews [0.4]

which are required to be performed prior to changing modes [0.3]. l REFERENCE ODP-ZZ-00014, pg. 1 002000G001 ...(KA'S)

ANSWER 8.10 (2.30)

a. Water volume of 281,000 gallons (0.5)
b. Essential Service Water (ESW) (0.5)
c. Sufficient water available to maintain the RCS in HSBY for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />

[0.5] with steam discharge to atmosphere (0.15] and total loss of offsite power [0.15] and then a cooldown to 350 degrees F [0.5].

REFERENCE i Technical Specifications, pgs. 3/4 7-6, B 3/4 7-2 061000A104 061000G006 061000G011 ..(KA'S)

8. -ADMINISIBAIIVE PROCEDURES. CONDITIONS. AND LIMITATIONS' PAGE 4 ANSWERS - -CALLAWAY- -87/09/22-SUNDERLAND, P.

L:

l ANSWER 8.11- (1.50)

To ensure that sufficient water depth is available [0.5] to remove:99% of L the assumed 10% iodine' gap activity [0.5] released from-the rupture of an

! irradiated fuel assembly [0.5].

l REFERENCE .

. Technical Specifications, pgs. 3/4 9-13, B 3/4.9-3 .

'000036G004 ...(KA'S) -I 3

ANSWER- 8.12 (2.00)

a. '1) .Tavg- (0,5)
2) PZR Pressure . (0.5)'

3)- RCS Flow' Rate (total) (0.5)

b. 0056 on day four. (0,5)

' REFERENCE Technical Specifications, pgs. 3/4 2-13, 3/4 0-2

'002000G011 002020A301 193000K105 ...(KA'S)

ANSWER' 8.13 (2.50)

a. One' operable centrifugal charging pump (0.5)
b. One operable safety injection' pump (0.5)
c. One operable RHR heat exchanger (0.5)
d. One operable RHR pump (0.5)

.e. One operable flow path [0.3] capable of taking suction from the RWST on an SI signal [0,1] and automatically transferring suction to.the_ containment. sump during recirculation [0.1].  ;

REFERENCE Technical Specifications, pg. 3/4 5-3

~006000G011 ...(KA'S)

l Ez__6Dd1NISTBATIVE PROCEDUBES2 CONDlIlONS2 _AND_LldlIAll.ONS PAGE 4 ANSWERS -- CALLAWAY -87/09/22-SUNDERLAND, P.

ANSWER 8.14 (1.50) l (Any 3 of 4 @ 0.5 pts each)

a. No fire watch is provided.
b. Work not initiated by a Work Request.
c. Large amounts of ignitable material cannot be removed from the area.
d. Fire suppression out of service in that room.

REFERENCE APA-ZZ-00742 086000A204 194001A110 194001K108 . . . (KA'S)

ANSWER 8.15 (1,20)

a. 1, 2, 3 (0.4 pts)
b. 1 (0.4 pts)
c. 2 (0.4 pts)

REFERENCE APA-ZZ-00320, pgs. 4, 5 194001A110 ...(KA'S)

ANSWER 8.16 (1.50)

Within one hour [0.5] establish a fire watch [0.5] to inspect the zone at least once per hour [0.5].

REFERENCE Technical Specifications, pg. 3/4 3-57 LER 86-003 086000A203 086000G011 ...(KA'S) l i

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g-- --  ;- - -

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n TEST CROSS REFERENCE PAGE

.b QUESTION VALUE REFERENCE l 05.01 1.00 SUN 0000127 05.02 1.00 SUN 0000128 05.03 1.00 SUN 0000129 05.04 1.00 SUN 0000130 05.05 1.00 SUN 0000131 05.06 1.00 SUN 0000132 05.07 1.00 SUN 0000133 05.08 1.00 SUN 0000134 05.09 1.50 SUN 0000135 05.10 1.50 SUN 0000136 05.11 1.50 SUN 0000137 j l- 05.12 2.00 SUN 0000138 05.13 2.00 SUN 0000139 05.14 1.00 SUN 0000140 05.15 2.00 SUN 0000141 05.16 1.50 SUN 0000142 05.17 1.50 SUN 0000143 05.18 1.50 SUN 0000144 24.00 06.01 1.50 SUN 0000145 <

06.02 1,00 SUN 0000146 06.03 1.00 SUN 0000147 06.04 1.50 SUN 0000148 06.05 2.00 SUN 0000149 06.06 1.25 SUN 0000150 06.07 1.50 SUN 0000151 06.08 1.00 SUN 0000152 06.09 2.50 SUN 0000153 <

06.10 1.50 SUN 0000154 06.11 2.00 SUN 0000155 j 06.12 2.00 SUN 0000156 1 06.13 2.25 SUN 0000157 06.14 2.50 SUN 0000158 1 06.15 1.50 SUN 0000159 l l

25.00 l l

07.01 1.00 SUN 0000160 07.02 1.00 SUN 0000161 07.03 2.50 SUN 0000162 1 07.04 1.50 SUN 0000163 o 07.05 1.50 SUN 0000164 07.06 1,25 SUN 0000165 07.07 1.50 SUN 0000166 07.08 2.00 SUN 0000167 07.09 1.00 SUN 0000168 07.10 2.50 SUN 0000169 07.11 2.00 SUN 0000170 j


a

TEST CROSS REFERENCE PAGE o

, QUESTION VALUE REFERENCE 07.12 2.00 SUN 0000171 1 07.13 2.00 SUN 0000172 07.14 1.25 SUN 0000173 07.15 2.00 SUN 0000174 25.00 08.01 1.00 SUN 0000175 'i 08.02 1.50 SUN 0000176 08.03 2.00 SUN 0000177 08.04 1.50 SUN 0000178 08.05 1.50 SUN 0000179 08.06 1.00 SUN 0000180 08.Li 1.50 SUN 0000181

  • 08.08 1.00 SUN 0000182 Uo.09 1.50 SUN 0000183 08.10 2.30 SUN 0000184 cc.11 1.50 SUN 0000185 08.12 2.00 SUN 0000186 08.1: 2.50 SUN 0000187 08.14 1.50 SUN 0000188 n8.15 1.20 SUN 0000189 e0.16 .1.50 SUN 0000190 25.00 99.00
lT? U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR L'ICENSE EXAMINATION

/t

~

FACILITY: (bAddAld/Ik REACTOR TYPE: PWR-WEC4 ]

DATE ADMINISTERED: 87/09/22 EXAMINER: LENNARTZ, J. __

. l CANDIDATE INSTRUCTIONS TO CANDIDAIE1 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.

i

% OF CATEGORY  % OF CANDIDATE'S CATEGORY VALHE_ TOTAL SCORE VALUE CATEGOBY 25.00 25.00 ._ I . PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 25.00 25.00 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 25.00 25.00 3. INSTRUMENTS AND CONTROLS 25.00 _g5.00 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL i CONTROL i

100.0  % Totals Final Grade i All work done on this examination is my own. I have neither given i nor received aid. .

Candidate's Signature MASTER CO?Y

l l' 'NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS i

Daring 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

'f the paper, and write "Last Page" on the last answer sheet.

o

9. Number each answer as to category and number, for example, 1.4, 6.3.
10. Skipsat'least three lines between each answer.
ll. 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.

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

l. work is your own and you have not received or been given assistance in I

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

been completed.

oy >

.1,.

': 4

's18. WhenjyouLcomplete'your examination, you shall:

5 . .

a. Assemble =your' examination as follows:

(1)- Exam. questions on top.

~

(2) Exam _alds - figures, tables, etc, f.3) 1 Answer.pages including figures which are part of.the answer.

Jb . Turn'in your copy of the examination and all pages used t'o 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 i

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f' L

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a

=,

  1. EQUATION SHEET -

f = ma y = s/t Cycle efficiency = (Net work out)/(Energyin).

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

E = ac 2

L KE = 1/2 av a = (Vf - V,)/t A = AN A = Ag e"**

PE = mph Vf =,V, + at w = e/t A = an2/t.1/2 = 0.693/t1/2

~

W = v aP.

A= ID, 2 t1/Z'I#

  • EI*M'Ib}3

[(t1/2)*I*b)3 aE = 931 am *

""Y avA '- -Ex t=y Q = aCpat

'6 = UA4T I = I,e~"*

Pwr = Wfah I = I,10~*/U ' i TVL = 1.3/u 8

HYL = -0.693/u P = P*10 "'III '

p ,p o,t/T SUR ='26.06/T SCR = S/(1 - K,ff)  ;

CR, = S/(1 - K,ffx)

SUR = 26a/t* + (s - p)T CRj (1 - K,ffj) = CR 2 I ~ Inff2)

T = (t*/s) + [(s - s yIo] M = 1/(1 - K,ff) = CR j/ G ,

T = s/(p - s) .M = (1 - K ,ff,)/(1 - K ,ff))

T = (s - e)/(Is) SDM = ( -K,ff)/K,ff a = (K,ff-l)/K,ff = AK,ff/K,ff 8*

  • 10 seconds I = 0.1 seconds-I e = [(t*/(T K,ff)] + [i,ff /(1 + IT)]-

Idlj=1d i Id 2 ,2 gd 2

P = (s4V)/(3 x 1010) jj 22 2 E = eN R/hr = (0.5 CE)/d (meters)

R/hr = 6 CE/d2 (feet) ,

Water Parameters Miscellaneous Conversions 1

1 gal. = 8.345 lem. 1 curia = 3.7 x 1010dps I ga; . = 3.78 liters i kg = 2.21 lba 1 ft< = 7.48 gal. I hp = 2.54 x 103 stu/nr Density = 62.4 Itg/f t3 1 nw = 3.41 x 106 8tu/hr Density = 1 gm/cm3 lin = 2.54 cm Heat of vaporization = 970 Stu/lom *F = 9/5'C + 32 Heat of fusion = 144 Stu/lbm 'C = 5/9 (*F-32) 1 At:n = 14.7 psi = 29.9 in. Hg. 1 BTU = 778 ft-lbf 1 ft. H 2O =.0.4335 lbf/in.

s

'T . ,; .

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  • l

. Attachment No. Rev. 3 Proced. No. ,

NATU*AL CHCULATI(Mi PofWmm p{ .,, E5-0.2  :

f

. RCS SOC 00 LUG METER ERROR C0thdON II

1. Determine subcooling margin, taking into account the ef fects .

of instrument error.

~

a.

Enter the subcooling meter reading in the appropriate column.

b.
  • Enter the subcooling meter error from the table in the same column.
c. Add algebraically.

$UBC00 LED (SCJ, SUPERNEAT(SH1

- t --

INIDICATION

+ .

+

ERROR-O - t1

  • CORRECTED READING
  • Negative results indii:ste degrees subcooled. * ., ,

. : Positive results indicata degrees superheat.

4 M

I N

f i .

CSM ERRORS (DEG. F)

SUBC00 LING 84 000 87 84 84 175 95 m ER 39 (DEG. F) SE 200 63 49 44 40 160

< $0 23 q[ 0.

137 55 39 33 27 sC

, 200 3000 l 300 500 1000 O 100 200 ,

i I

l RCs PRESSURE (PsIG) (

r Page 1 of 1 ,

Page 4 QBlHCIPLES OF NUCLEAB_EQW5,1MHI .9PJRATION.

IHEJRs) DYNAMICS. H2AT TR&m JE] LAND ;uUID FFQR o 1 L  !

j

QUESTION- 1.01- (1.00)

TRUE.or FALSE 7 ,

- a. One of the pump laws for centrifugal pumps states that-the volume q flow: rate is proportional to the speed of the pump.- (0,5)

. b. As VCT temperature decreases, volume flow rate from the centrifugal '

charging pump increases. (0.5)

QUESTION 1.02~ (1.00)

In-regards to DNBR,. choose the ONE correct statement: (1.0)

a. DNBR increases with decreasing flowrate

-'b . DNBR decreases with decreasing reactor power

c. DNBR decreases with increasing pressurizer pressure
d. DNBR increases with decreasing Tavg QUESTION 1.03 (1.50)

Answer EACH of the following either TRUE or FALSE:

. a. When the system characteristic curve is considered with the two 1 pumps-in-parallel curve, the operating point represents a higher system volumetric flow rate and a greater system head than there would be with one pump operating. (0.5)

b. In order to increase the volumetric flow rate in a system, centrifugal pumps are often used in series. (0.5) q
c. If a centrifugal pump's flow rate is reduced by throttling the discharge valve, the pump's minimum REQUIRED NPSH DECREASES and the AVAILABLE NPSH INCREASES. (0.5)

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

1 y

j f. -PRINCI?LES QF NUCLEAR POWER PLANT OPERATION. :Page 5. .,

THERMODYNAMICS. HEAT _ TRANSFER ARD FLUID RQE- J

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' QUESTION. 1.04 (2.00) a .- If.'the' reactor-is. operating in the power range,'how'long will '

it take to raise power from 20% to 40% with a +0.5'DPM startup )

rate?; (Choose the ONE correct answer). (1.0) . 1

1. 12' seconds
2. 21 seconds
3. 36 seconds' 3
4. 54 seconds

'l

'b. How<long will ititake to raise power from 40% to 60% with the same-

+0.5 DPM startup' rate? (Choose the ONE correct' answer) (1.0)'

- 1. 12 seconds

2. 21: seconds
3. 36 seconds
4. I
4. 54 seconds i

o QUESTION 1.05 (1.00)

Which one of the following best describes the effect that DECREASING RCS

.Tave from 585 degrees'F to 560 degrees F will have on MTC. (1.0)

a. 'It-becomes less negative because boron and water molecules are

~

~

swept into the core as a result of the cutsurge from the pressurizer,  !

therefore, neutrons spend more-time'in the resonance region.

b. 'It becomes less negative because the rate of change in the density of water per' degree temperature change is less at lower temperature

.which causes a' lesser change in rate in resonance escape probability.

]

c.

- It becomes more negative because thermal utilization increases and resonance escape probability decreases.

d. -It becomes more negative because as temperature is lowered, the moderator becomes more dense, this increases'the amount of water molecules in the core therefore neutrons have a greater probability ,

of colliding with a water molecule and this is an increased negative l reactivity effect.

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

1 PRINGIPLES OF NUQLEAR POWER PLANT OPERATION, Pego 6 THERMODYNAMICS. HEAT TRANSFER AND FLUID EkGd -

i t

QUESTION 1.06 (1.00)

Delayed neutrons play a major role in the operation of the core because they ... (Choose the ONE correct answer) (1.0)

a. are born at (thermal) slow energy levels (less than'1 ev) and therefore are more apt to cause a fission as compared to being absorbed by a poison.
b. are considered as thermal neutrons and therefore they will not travel far enough to leak out of the core.
c. are born so much later than the prompt neutrons and provide controllability during steady state operations and power transients,
d. provide 70% of the fission neutron inventory and have higher importance factors associated with them as compared to prompt ,

neutrons.

QUESTION 1.07 (1.50)

Compare the calculated Estimated Critical Position (ECP) for a startup to be performed in'accordance with OTG-ZZ-2, "Rx Startup," 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> after a trip to the Actual Critical Rod Position (ACP) if the following events / conditions occurred. Assume reactor power had been at 100% for 12 days prior to the trip. Consider each independently. Limit your answer to:

a. ACP higher than ECP
b. ACP lower than ECP l
c. ACP would not be significantly different than ECP.  !

}

l

1. One Reactor Coolant Pump is stopped one minute prior to criticality. {

(0.5) l

2. The steam dump pressure setpoint is increased by 100 psig. (0.5)
3. The startup is commenced 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> earlier than anticipated. (0.5)  !

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(***** CATE3ORY 1 CONTINUED ON NEXT PAGE *****) l

( 7 7-_ ,

L,........_PRIERIELES OF NUGLEAR POWEB PLANT OEERATION. Pcgo 7 IHERMODYHAMICS. HEAT TRANSEER_AND FLUID FhgW

(

1 QUESTION 1.08 (1.00) j In order to maintain a 200 degrees F subcooling margin in the RCS when  ;

reducing RCS pressure to 1600 psig, steam' generator pressure must ]s be reduced to approximately; (Choose the ONE correct answer) (1.0)

a. 245 psig
b. 445 psig
c. 645 psig
d. 45 psig

. QUESTION 1.09 (1.00)

During a reactor startup from an initial Keff of .90, the first reactivity l '

addition caused count rate to increase from 10 cps to 16 cps. The second reactivity addition caused count rate to increase from 16 cps to 32. cps. e Which ONE of the following statements BEST describes the relationship between the first and second reactivity additions. (1.0) )

l

a. The first reactivity addition was the larger of the two.
b. The second reactivity addition was the larger of the two.
c. The first and second reactivity additions were equal.
d. There is not enough data given to determine the relationship.

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

I

_ _ _ _ _ _ _ _ _ _ __ I

IO Page .8;

, 1 1.

PRIlLQIELES OF NUCLEAR-POWER' PLANT OPERATION.

i:l .IBEB11QDYNAMICS. HEAT TRANSFER AED FLUID FLOW 4

l L ,

r.

QUESTION 1.10 (1.00) l i

r ,

Indicate whether.EACH of the'following will cause shutdown margin to INCREASE, DECREASE, or REMAIN THE SAME. (Consider each independently) u I

L

1. Control rods withdrawn 10 steps with no change in turbine. power orLboron. (.25) l

- 2. Turbine power increased :10 percent with no change in boron or rod position. (.25)

3. Normal power increase from 50 to 60 percent with no boron change, l rods in automatic. (.25)
4. Boron concentration is increased 10 ppm with rods in automatic, tubine power ~is constant. (.25)

QUESTION. 1.11- (2.50)

Brittle fracture of your reactor's pressure vessel can occur ~at  !

a.

stresses well below its yield stress. TWO conditions must be present besides high stress. What are these TWO conditions? l

. ( 1. 0 )

b. How do heatup/cooldown rate limits on the reactor coolant system i reduce the probability of brittle fracture? (0.5) )

l;

c. Why does the concern about-brittle fracture of the reactor pressure vessel increase as the plant ages? Include in your answer the )

specific material PROPERTY that is affected. (1.0) {

1 QUESTION 1.12 (2.00)

a. What happens to the power required by a motor driven centrifugal pump as the discharge valve is throttled in the SHUT direction? (1.0)
b. What are TWO effects of operating a motor driven centrifugal pump at or beyond a runout condition? (1.0) .

l

)

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

Page 9 L PBitLQlELEE_QF NUQLEAR POWEB__PLANIEERATION<

i THERMODYNAMICS. HEAT TRANSEER AND FLUID FLOW QUESTION 1.13 (1.50)

-The reactor is operating at 100% power with all rods out, near BOL with equilibrium Xenon conditions when power is to be reduced to 50%. The operator observed that AFD is -7, and within its band, and decides to lower power by borating, leaving rods in the ARO position. Actual Tavg follows programmed Tavg, DESCRIBE the change that will occur in AFD and i WHY it occurs, prior to changes in Xenon having a noticeable effect.

(1.5)

QUESTION 1.14 (0.50)

(Fill in the Blank)

During a Xenon-free reactor startup, critical data was inadvertently taken two decades below the required Intermediate Range (IR) level. Assuming RCS temperatures and boron concentrations were the same, the critical rod position-taken at the proper IR level the critical rod position j taken two decades below the proper IR level. (0.5) 1

a. Is Less Than
b. Is the Same As
c. Is Greater Than
d. Cannot be Compared To QUESTION 1.15 (2.50)

The Unit 1 reactor is operating at 50% power, BOL, when a steam dump fails open. Assume rods are in manual, no operator action is taken, and no ,

reactor trip occurs. Explain HOW and WHY reactor power and Tave will I' change, and WHERE they will stabilize in relation to the initial values.

(2.5)

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

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_ $1". 'PRidCIPLES OF NUCLEAR POWER PLANT' OPERATION.

-Page 10-1 THERMODYNAMICS. HEAT TRANEEER-AND FLUID FLOW-

'l QUESTION 1.165 (2.50)

~

-a. Which one of the following descriptions best supports'the' reason why Xenon reactivity increases. sharply after a trip from 1000 hrs. at 100% power 7- ( 0. S )' ,

~

1) Xenon decays less rapidly due to'a reduction ~'in.the neutron.

flux. ,

-2)- ~ Iodine half-life is much shorter than' Xenon half-life.

3) Iodine production is greatly reduced and Xenon production is greatly-increased.due to the reduction in neutron flux.

s

<4 ) Due to reduced neutron absorption,' Iodine concentration t . increases, and. Xenon decays directly from Iodine, thus' Xenon increases.

'b. Give two reasons why Sm-149 is not as much of a concern to an operator after a: reactor. trip as is Xe. (1.0)

c. A Xenon ~ oscillation in a reactor core might be produced by certain types =ofurod motion. How would the Xe oscillation resulting from the following two-cases be different?' Explain. (1.0)
1) A turbine runback occurs with rods in auto. Rods drive 60 steps.

'2 ) Rods are driven 60 steps starting from~the same position as in Case 1,-but slowly over 4 days time.

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

PR1HQ1ELEE_QF.NUQLEAR POWEB_ELaHI_QEEBAIl0th. Pego 11 1

IRERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW

. QUESTION 1.17 (1.50)

Fill in the blanks to complete EACH of the following statements concerning s subcritical multiplication. (Limit answers to INCREASE, DECREASE, or

' REMAIN THE SAME.)

a. The period of time for a constant neutron level to be reached following a positive reactivity insertion at Keff = 0.9 will from an equal reactivity insertion at Keff = 0.8. (0.5)
b. If a neutron source strength decreases, the rod height for criticality will . (0.5)
c. If a neutron source strength decreases, neutron level will

. (Assume original neutron count rate to be 100 cps.)

(0.5) l i

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

' PLANT' DESIGN INCLUDING _BAFETY AND EMERGENQ1 Page'12 l

1

, SYSTEMS 1 99 k

1 I

LQUEST' ION' 2.01~ ( 1.~ 50 )

' Answer 1the!following. questions concerning the Reactor' Coolant Pump Seals:.

a; What forces the 3 gph leakoff from the #2' seal, AND.where does the leakoff from the #2 seal go? (0.5)  ;.

b. Where is the #3 seal. injection supplied fromLAND where does the #3-seal. leakoff.go? (0.5) ;j

.c. What are the TWO reasons for the injected flow to the #3 seal? .(0.5) a 1i

-QUESTION 2.02 (0.50) 'l LTRUE~or-FALSE 7 -

j L Any fault condition'that trips open-the Station Service Transformer-Feeder  ;

Breaker ~(PA0104) must be manually reset at the switchgear before that breaker can'be closed from the control room.

L QUESTION 2.03-  : ( 2.' 00 )

The Emergency Seal Oil Pump automatically started due to a fault in the

=

Main Seal Oil Pump breaker, and is drawing a' constant 60. amps. 1

a. If battery charger PJ21 is out of service for maintenance, and l battery charger.PJ31.cannot be placed in service, how long will the Emergency Seal Oil-Pump be able to run before it totally discharges the 250.VDC battery? (Assume that the Emergency Seal Oil Pump is the 3 only load running off the~ battery.) (0.5)

'li. What are THREE conditions which would cause the "250 VDC PJ01 PWR TROUBLE" annunciator to energize? (1.5)

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

Pegs 13

_2;. ' PLANT DESIGN INCLUDING SAEETY AND EMEEGENCX SYSTEMS

-~ ,

QUESTION 2.04 (2.00)

Answer the following questions concerning the Emergency hiesel Generator (EDG):

a. What are FOUR signals that would energize the Engine Shutdown Relay i following a NORMAL start of the EDG7 (1.0)
b. What is the purpose of the Start Failure Relay? (1.0) l l

l QUESTION 2.05 (1.50) I List FIVE of the SIX AUTOMATIC actions that occur in the Essential Service Water System upon receipt of a Loss of Offsite Power eignal. (1.5)

QUESTION' 2.06 (2.00)

What FOUR conditions / interlocks must be met in order to OPEN the RCS Hot Leg Loop 1 to RHR Pump A Suction Valves PV-8702A and HV-8701A? (2.0) l QUESTION 2.07 (1.50)

What are FOUR PARAMETERS that can be monitored / read on the Main Control Board that are associated with the Containment Spray System? (Redundant indications count as a single response.)

QUESTION 2.08 (2.00)

Answer the following questions concerning the Safety Injection System:

a. What are TWO reasons that hot leg recirculation is performed? (1.5) ;

TRUE or FALSE? (0.5) l b. The SI pumps mini-flow isolation valves (EM-HV-8814A and B) fail as I is on a loss of power, while the mini-flow common isolation valve ]

(BN-HV-8813) fails open on a loss of power. ]

i 1

b 1

l (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) j l 1

_ _ - - - - - - l

m; , . - - - - - .

16 ]

ddLANT' DESIGN INCLUDING ~ SAEETY AHp__EMERGEH91 L Page?14 SYSTEMS

/

1 QUESTION: 2.09l '( 2. 00 ) ,

i m '

Answer the following questions concerning the Chemical ~and Volume' Control l System: ,

a. 'Stateinwhat:positioneachofthe:following' valves [willfailif control air.is lost. (1.0).
1. Charging Header Flow Control Valve (HV-182)
o'

'2n ' Letdown Heat Exchanger Outlet: Pressure Control Valve (PCV-131)' i

3. Auxiliary Pressurizer Spray Valve (HV-8145)-

4: CVC Demineralized' Inlet Divert 1 Valve (TCV-129) y

.b. .What.is-the reason-for the interlocks associated with the operation

,of.the letdown isolation valves (LCV-459 and 460)7 '(1.0) 1 fQUESTION.. '2.10 .(1.50) aBriefly explain why.you would AGREE or DISAGREE with the following statement:

Should a containment isolation occur, there will not be any RCP #1 seal leakoff. (1.5)

QUESTION 2.11 (1.50)

Answer the following questions concerning the Main Feed System:

a '. What are FOUR signals which will cause ONE Turbine. Feed Pump to trip? (Setpoints not required) (1.0) I

b. What is the reason for the Lo-Lo S/G level Main Feedwater Isolation  !

Signal-(FWIS)? (0.5) i 1

a l

l

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

1

.6 LPLANTJESIGN INQLUDING ' SAFETY ' ANDl EMERGEEQ1; P:ge 15 L SYSTEMS

QUESTION! 2.12 f(2'.00) inswer'the'fo11owing questions concerning the Auxiliary Feedwater System (AFW):
a. .What are TWO of the THREE' safety bases'for the. flow restricting orifice downstream of each flow control valve? (1.5)-

TRUE or FALSE 7

b. The automatic shift from the CST.to.ESW will occur when.an SI signal is present and low suction. pressure to the AFW pump exists on two of three sensors. (0.5)

QUESTION 2.13 (1.50)

What are THREE Radioactive Liquid Effluent Monitors that are contained in

. Technical Specifications and provide AUTOMATIC termination of a release?

' QUESTION 2.14 (0.50)

How does the positive displacement charging pump (PD pump) rerpond'to a Safety Injection (SI) signal if it was operating prior to the SI? (0.5)

QUESTION 2.15 (1.50)

Answer the following questions concerning the con.ponent cooling water <

system (CCW):

a. What is the NORMAL makeup water source to the surge tank? (0.5)
b. WHAT is the EMERGENCY makeup water source to the CCW system, AND WHERE does it tap into the system? (1.0) l h

h l

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

2,.....PLABI_ DESIGN INCLUDING SAFEIJ_AHD_EMEEGEEDX Page 16 L

SYSTEMS QUESTION 2.16 (1.00) ,

Answer the following questions concerning the Reactor Makeup Water System:

a. Why is the Boric Acid Flow Control Valve, (FCV-110A), designed to fail OPEN on loss of air or control voltage? (0.5),

TRUE or FALSE b The Total Flow Counter (FY-111B) maintains a continuous record of the number of gallons of combined solution added to the primary system.

(0.5)

QUESTION 2.17 (0.50)

TRUE or FALSE 7 The RCP seal injection filter is equipped with an external bypass line which allows the filter to be changed without interrupting the injecticn flow. (0.5)

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

i 1

m3. INSIBilM LIE._6HP CONTROld Pago 17 1

(

~

{

QUESTION 3.01 (2.25)

Assume the Reactcr is at. 8% power after a reactor startup, with one main feed pump in manual, the main turbine at 1800 rpm unloaded, NOT transferred to the main feed regulator valves, and no operator action.

Describe the responde (if any) of the feedwater system to the following malfunctions. (Include in your answer the effect to the feedwater components and the final state of the plant.)  !

1

a. N41 (Power Range Channel) NI fails high. (1.0)
b. Controlling steam generator level channel for S/G B fails low.

(1.25)

QUESTION 3.02 (2.00) l ,

Answer TRUE OR FALSE for EACH of the following statements with respect to the Vibration Loose Parts Monitoring (V& LPM) System:

a. If any of the signals exceed a preset level, an alarm signal is activated and resetting of the alarms requires operator action.

(0.5)

b. The V& LPM System monitors a]l four steam generators, and the upper and lower regions of the reactor vessel. (0.5)
c. In general, loo.se parts will give off a lower frequency signal than a normal system vibration would. (04 5)
d. The operator must manually disable the upper reactor vessel sensors when moving control rods. (0.5)

QUESTION 3.03 (2.80)

List SEVEN different indications and/or controls available at the Auxiliary Shutdown Panels (RP118A and RP11BB) for monitoring and controlling the Reactor Coolant System. (DO NOT include any Steam l Generator or AFW controls or indications.) (Components which have both '

control and indications available, and redundant instrumentation count as a single response.) (2.8)

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

3. INSTRUMENTS AND CONTROLS Pego 18 QUESTION' 3.04 (2.40)

List the automatic signals which will give a Main Steam Isolation Signal.

(Include setpoints, coincidence, blocks, and when in effect where applicable.) (2.4) l i

QUESTION 3.05 (3.00)

State the control interlocks (C's) associated with the Rod Control System.

Include in your answer: the instrument that gives the interlock, coincidence, setpoint, direction of block (IN, OUT, BOTH) and whether auto, manual, or both are effected. (3.0)

QUESTION 3.06 (2.00)

For each of the following malfunctions, explain the response of the rod control system. (Assume normal lineups for each case and no operator action. Consider each case independently.)

a. During a reactor startup with power at 8%, N35 (intermediate range nuclear instrumentation) channel fails low at a rate of 2%/ minute.

(Include in your answer, the effect this failure has on the control rods, and list all inputs to the Rod Control System from N35. If this failure has no effect on the control rods, state so in your answer.)

(1,0)

b. At 100% power, feeder breaker to PG-20 trips open. (Include in your answer, the effect this failure has on the control rods. If this failure has no effect on the control rods, state so in your answer.

Briefly explain.) (1.0) i QUESTION 3.07 (2.35) ,

l Answer the following questions with respect to permissive P-4.

a. What are the minimum signals required to generate P-4? (0.75)
b. List FOUR functions of P-4. (1.6)

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

p. , o L.

3; INSTRUMENTS ~AND-CONTROLS. Page 19 9

= QUESTION 3.08 (2.00) t f State the effect.on Channel II OT delta T setpoint for the following:

(Assume initially at-100% power, all. control systems in ',utomatic a except for rods in manual'. Limit answer to INCREASE, DECREASE,t.or NO CHANGE in

'setpoint, Consider each part individually,.no explanation' required.).

a. Auctioneeredlhigh Tavg unit fails high. (0.5)
b. N42' power range lower detector fails low. (0.5)-
c. PT-456' pressurizer pressure fails high. (0,5)
d. Reduce power to 50% with normal pressure and temperature. (0.5)

QUESTION- 3.09 (1.00)

Source range instrument N31.had read 40 cps.for the last 60 minutes while in mode 5 at 1250-ppm boron. A welding spike caused N31 to reach 3000 cps for 20~ seconds. The reactor operator then noticed that N31 counts were slowly decreasing .

below 40 cps. Explain the reason for counts decreasing. (1.0)

QUESTION 3.10 (1.20)

List'THREE FUNCTIONS / SYSTEMS that wide range temperatures feed. (Do not list any indications that wide range temperatures feed.) (1.2)

QUESTION 3.11 (1.50)

In March 1986, while in. Mode 6, the following monitors lost power:

GK-RE-04 (Control Room Radiation Monitor), GT-RE-31 and 33 (Containment- j Purge Radiation Monitors), and GG-RE-28 (Fuel Building Radiation Monitor).

What automatic actions, if any, would occur if the monitors spiked high  ;

upon'reenergization? .(1.5) l

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

_ ~ - _ _ _ _ .

~3. INSTRUMENIji AND CONIBQId .Paga-20: L l

j

> 4

~

j

. QUESTION 3.12' ( 1'. 00 )

I While'in Mode 1 at 31%~ power with normal operating temperature.and )

pressure','a radwaste technician unintentionally. opened DC fused disconnect  :

switch PK-0112 which'resulted in removing power from PK-51 (AuxiliaryL '

Building distribution switchboard). , 1 1

' List TWO ESF actuations that occurred. (1.0)

QUESTION 3.13 (1.50) q List.SIX conditions / signals which will actuate a CRVIS (Control Room

~

i Ventilation Isolation Signal). (Setpoints and coincidence not required)

~

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(***** END OF CATEGORY 3 *****)

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[

Paga 21

4. PROCEDURES' - NORMAL. 'ABILQBtiAL. ...EtiERGENCY
AND RADIOLOGICAL-CQHIRQk I

QUESTTON 4.01 '( 1. 7 5 )

~Per FR-SJ1 (Response to Nuclear Power Generation), list the required

. actions if the action / expected response is NOT OBTAINED.for the following'

~

steps: (Include in your answer ALL the contingency actions that apply to Jeach step.)

a. Ensure Reactor Trip-(1.0)
b. Ensure Turbine Trip (0.75)

. QUESTION 4.02' (0.50)

Using Attachment No. 3 from ES-0.2 (Natural Circulation Cooldown),

determine the subcooling margin given:

2 Subcooling meter reading 21 degrees F subcooled RCS pressure 950 psig (0.5)

QUESTION 4.03 (1.00) i Per ECA-0.0 (Loss of All AC Power), list the immediate action steps (DO NOT include substeps or response not obtained). (1.0)

QUESTION 4.04 (2.00)

Per E-0 (Reactor Trip or Safety Injection), list the substeps required to verify the following immediate action steps:

a. Verify Reactor Trip (1.0) i i
b. Ensure Feedwater Isolation (1.0) )

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(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

i L-

4. PROCEDURES -: EQRMAL. ABNCRMAL". EMEBQENCY Pago 22

.AND RADIOLOGICAL CONTBQL QUESTION 4.05. . ( 1. 50 ) '

HDP-ZZ-01400 (External Dosimetry Program) provides guidance on the. wearing-

~

and use of TLD's andlPIC's, ano dose limits for individuals.

State where a single whole body TLD and PIC are required to be worn.

~

a.

(Normal' position only, DO NOT state exceptions.-) (0.5)

'b. State the actions for an individual who has entered a Radiologically Controlled Ares and whose PIC is discovered off-scale. (0.5)

c. List the whole body and extremity dose limits for life-saving activities. (0.5)

QUESTION 4.06 (2.00)

For each of the.following situations, state whether an RCP (Reactor Coolant Pump) may be started.

(Consider each situation independently. Limit answer to START or NOT START, no explanation is required. Assume normal operations, not an emergency situation.) (2.0)

a. Leakoff flow 0.6 gpm
b. #1 Seal Backpressure of 30 psig
c. Seal injection flow to the RCP 24 gpm
d. Immediately after the first start where the motor failed to reach full speed before it was stopped.

QUESTION 4.07 (1.50) ,

l Per ODP-ZZ-00003 (Shift Relief and Turnover), seven items must be '

completed / reviewed by the Oncoming URO on the URO Watch Relief Checklist. i List SIX of the seven items requiring completion or review. (1.5)

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(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

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N Pego 23 A: EROCEDURES - NORMAL. ABNORMAL. EMEEGENCY AND RADIOLOGICAL CONTROL QUESTION 4.08 (1.00)

APA-ZZ-01000 (Callaway Plant Health Physics Program) defines the actions (

to be taken by any plant personnel that discover a small spill of radioactive material. State the actions required for a small spill of radioactive material. (1,0) -

QUESTION 4.09 (2.50)

Fill.in the blanks for the following " Precautions and Limitations" from '

OTG-ZZ-00001 (Plant Heatup Cold Shutdown to Hot Standby).

a. The Residual Heat Removal (RHR) System should be isolated from the Reactor Coolant System (RCS) before exceeding degrees F or psig. (0.5)
b. Reactor Coolant Pumps (RCP) should not be operated when the number 1 seal differential pressure is less than psid or VCT pressure is less than psig. (0.5)
c. Three CRDM cooling fans should be in service when Tavg is greater than degrees F or the CRDM's are energized. (0.25)
d. At a minimum, either or containment cooler fan and ,

one fan must be in service whenever Tavg is greater I than 135 degrees F. (0.5) {

l

e. At least one (1) RCP should be in operation with Tavg greater I than degrees F. (0.25) i
f. Prior to exceeding a power level oi 1 MW thermal, RCS hydrogen concentration should be greater than ec/kg. (0.25)  ;

I

g. The positive displacement pump shall not be in service or available i for service when the RCS temperature is less than degrees F, l and the PORV's are used for cold overpressure protection. (0.25)

)

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(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) l l

1

Pega' 24 1 'PROCEDUBES - NORMAL. ABNOEMAL.1_ EMERGENCY AND. RADIOLOGICAL-CONTROL QUESTION._ 4.10 (2.25)

. Fill ~in'the blanks.

Per Technical. Specifications: ,

i The RWST shall be. OPERABLE with:

a. A minimum contained borated water volume of gallons. (0.25)

' b .- A minimum solution temperature of degrees F. (0.25)

c. A maximum solution temperature of degrees F. (0.25)

Each RCS accumulator shall be OPERABLE with:

d. 'The~ isolation valve and power. . (0.5)
e. A contained borated water volume of between and gallons.

(0.5)

f. A. nitrogen cover-pressure of between and psig. -(0.5)

QUESTION 4.11 (2.00)

IF the plant is operating at 100% power with rod control in automatic, control bank C at 225 steps, Tavg at 589 degrees F, pressurizer pressure at 2255:psig,. pressurizer level at 62% and Tref (for rod control) at 588 degrees F, WHAT immediate action steps, if any, are required to be performed? (2.0)

I i

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(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

i1

-4. PROCEDURES"- NORMAL. ABHORMAL, EMERGENCY Page.25- J AND-RADIOLOGICAL CONTROL l

.IJ i

QUESTION 4.12 -(2.00)  ;

'The plant is performing a cooldown while in Mode 4 at 300 degrees F and '

f 1400'psig, when all of the following conditions occur-

a. Pressurizer: level rapidly decreasing
b. Containment' air monitors indication increasing-
c. RHR loop flow low annunciator
d. Running RHR pump discharge pressure decreasing

, What FOUR immediate actions are required? Consider ALL of the symptoms-together. (2.0)

QUESTION 4.13 (2.00)

PT-505 (Selected Turbine Impulse. Pressure Transmitter) fails low with the plant at:90% power and all of the systems in normal configurations.

~

List the FOUR immediate actions required for failure of PT-505. (2.0)

QUESTION 4,14 (1.00)

While at 55% power with all systems in the normal alignment, maintenance ,

requests that valve BN-HV-8813 (SI combined recirculation valve) be tagged j SHUT to provide double valve protection to work on EM-V-121 (local sample isolation). They state the work will take 8-12 hours to complete.

a. What problem, if any, will tagging this valve (BN-HV-8813) cause with respect to Technical Specifications? (0,5)

I

b. How long may this valve (BN-HV-8813) be tagged and maintain-55%

power? (0.5)  !

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(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

1

Page 26

.A. PROCEDURES ' NORMAL.LABEQBMAL, EMERGENCY AND RADIOLOGICAL CONTROL l

L .

I

' QUESTION 4.15 '(1.00)

AnswerLEACH of'the following TRUE or FALSE concerning APA-ZZ-00310, "Workmans Protection Assurance and Caution Tagging."

a. Caution tags shall not be used in lieu of a temporary procedure except during non-safety system testing. (0.5)
b. Any tags that are destroyed while the Workers Protection Assurance is in effect shall be replaced, and given the-next sequential serial number'and tag number. (0,5)

QUESTION 4.16 (1.00)

Answer each of the following questions concerning APA-ZZ-00802, " Confined Space Entry Permit Program":

a. What precaution must a person take to immediately enter an area with a Class A atmosphere, in an emergency situation, without a Confined Space Entry Permit? (0.5)

TRUE or FALSE 7

b. A short term entry into an area can be made to manipulate a valve without a Confined Space Entry Permit. (Assume that Health Physics confirmed the area to have a Class B atmosphere just prior to entry. )

(0.5)

(***** END OF CATEGORY 4 *****)

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

p

1. PRINCIPLES OF-NUCLEAR POWER PLANT OEERATION. Pcge'27
IEEBdQDYNAMICS. HEAT. TRANSFER AND FLUID FLOW ANSWER l'.01 (1.00)'
a. True
b. -False (0.5 pts each) -

REFERENCE Westinghouse Thermal-Hydraulic Principles and Applications to the PWR II, Chapter.10, pgs. 31-38 191004K105- 191004K115 191004K112 ..(KA's)

ANSWER 1.02 (1.00)

d. (1.0)

REFERENCE Westinghouse Thermal-Hydraulic Principles and Applications to the PWR II, Chapter 13, pgs. 17-23 193008K105 ..(KA's)

ANSWER 1.03 (1.50)

a. True
b. False

. c. True (0.5 pts each)

REFERENCE Westinghouse Thermal-Hydraulic Principles, Chapter 10, pgs. 41-49 191004K119 191004K114 ..(KA's)

ANFWER 1.04 (2.00)

a. 3
b. 2 (1.0 pts each)

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

r:

T 1.- PRINCIPLES OF NUCLEAR POWER PLhNT OEEBATION, .Pcga 28

! THERMODYE6MICS. HEAT TRAHEEER AND._ELUID FLOW 4

I REFERENCE Westinghouse Fundamentals of Nuclear Reactor Physics, Chapter 7, pages 19-23 l< '192003K109 ..(KA's)

ANSWER 1.05 (1.00) i b.

' REFERENCE 4 Westinghouse Reactor Core Control for Large PWR, Chapter 3, pgs. 20-21 001000K526 192004K106 ..(KA's) 1 ANSWER 1.06 (1.00)

C.

REFERENCE Westinghouse Fundamentals of Nuclear Reactor Physics, Chapter 7, pages 29-37 192003K107 ..(KA's)

ANSWER 1.07 (1.50)

1. c
2. a
3. a (0.5 pts each)

REFERENCE Westinghouse Reactor Core Control, Chapter 7, pgs. 24-33 001000K529 001010A207 001010K526 ..(KA*s)

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

Paga 29 1!. PRINCIPLEE_QF'NUCLEAB_ POWER PLANT QEEB611Q_NN2

, IHERM0 DYNAMICS. HEAT TRANSEER AND FLUID FLOW ,

, ),

1 l ANSWER 1.08 '(1.00) i

\

a. (1.0). l iREFERENCE- .

i l Westinghouse Thermal Hydraulic Principles and Applications to the PWR II, Chapter 2, pgs. 63-70; Steam. Tables  ;

~

-193003K125 ..(KA's)

. ANSWER 1.09 (1.00)

a. (1,0)

. REFERENCE Westinghouse Reactor Core Control, Chapter 8, pgs. 39-48, Chapter 5, pgs. 21-22 192002K112 192002K108 192008K105 ..(KA's)

ANSWER 1.10 (1.00)

1. Remain the same
2. Remain the same
3. Remain the same
4. Increase (0.25 pts each)

REFERENCE Westinghouse Theory and FF 192002K113 ..(KA's) i

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

{

1. PRI'CIPLES N OF NUCLEAR POWER PLANT OEEBATION. Page 30 THERMODYNAMICS. HEAT TRANSFER Ahp_EkQlD FLOW

.s ANSWER L 1.11'- (2.50)

.a. 1. Presence of al flaw (or crack of sufficient-size). .(0,5)

2. . Low temperature (or': Temp-below NDT). (0.5)

! )

.b. ' Reduces the' thermal stress. (Reduced DT across the RV wall-(0.5)-

reduces total / thermal / tensile stress;)

c. Neutron' exposure (integrated).(0.5) makes the material more brittle I

-(raises NDT)-(Reduces-ductility) (0.5). j LREFERENCE WEC-Thermal Hydraulic Principles and Applic. to the PWR II, pg. 13; 58-62 193610K105' 193010K104 193010K101 ..(KA's) 1

' ANSWER 1.12 (2.00)

a. Power' decreases. (1,0)
b. 1. Overload (overheat) of motor (trip of the motor breaker)
2. ' Cavitation

-3. Pump efficiency decreases

4. . Overheating of pump bearings  ;

(Any 2 9 0.5 pts'each)

REFERENCE Westinghouse' Thermal Hydraulic Principles, Chapter 10, pgs. 31-38 l 191004K112 191004K107 ..(KA's)

ANSWER 1.13 (1.50)

Due'to the' greater decrease in the. temperature of the coolant exiting the core relative to the decrease of the inlet coolant [0.5), more positive reactivity will-be added to the upper core regions [0.5], resulting in a-J 1 more positive (less negative) AFD [0.5).

l

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

1n -PRIHCIPLES
OF NUCLEAR POWEB PLANT'OPERATIQEi 'Paga 31

-THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW L

?, * + '

4 REFERENCE l Westinghouse'Nucl' ear Training Operations, Chapter 8 'i' l; 193009K102- . . ( KA ' s ) -

~ ..

ANSWER '- 1.14 (0.50) /

- b. (0.5)

REFERENCE-Westinghouse Fundamental of Nuclear Reactor Physics, Chapters:5 and 7 192008K110 ..(KA's)

' ANSWER. 1.15 (2.50)

Tave decreases since more energy is being removed. (0.7)

Rx power' increases due to the positive reactivity added through MTC [0.5].

Doppler would add negative reactivity [0.3].

Power stabilizesLat a higher value. (0.5)

~

Tave stabilizes at a 1ower value. (0,5)

' REFERENCE Westinghouse Thermal-Hydraulic Principles and Applications to the PWR, ) '

-pgs. 1-9 to-1 . Westinghouse Reactor Core Control for Large PWRs, pgs. 3-29 to 3-41' i 192008K121 .192008K117 ..(KA's)

' ANSWER. 1.16 (2.50)

a. 2. (0.5)
b. Sm-149 has a smaller absorption cross section and therefere less reactivity worth than Xe-[0,5], and does not decay.away like Xe

-[0.5]. (Will also accept explanation that half-life is much greater than xenon and reactivity effect is much slower than xenon.)

c. Case 1.'would be a more noticeable XL transient [0.2] because the local power changes occurred rapidly with respect to Xenon's ability to maintain equilibrium with local power [0.8].

j

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

y l

i 4;. PRINCIPLES OF NUCLEAR POWER PLANT _QEEBATION4. Page'32

.TilERMODYEAMICS HEAT TRANSFER AND FLUID FLOW REFERENCE Westinghouse Reactor Core Control for Large PWRs, pgs. 4-21 to.4-23, 4-28  !

to'4-34 ]

'192006K122 '192006K112 192006K106 ..(KA's) 4 1

.i

. \

I ANSWER 1.17- (1.50)
a. INCREASE
b. REMAIN THE SAME
c. DECREASE l

-(0.5 pts each)

~ REFERENCE Westinghouse ~ Fundamentals of Nuclear Reactor Physics, Chapter 8, 'l

' pgr. 12-19 1015000K506 192008K104 ..(KA's) i l

i l

1 l

) '

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

l L_________:---_________

i h ' 2.' -PLANT DESIGN INCLUDING SAEETY AND EMEBGEEDX- Page 33 SYSIEMS ANSWER- .2.01- (1.50) )

a. The. backpressure from #3 seal-[0.25); the. leakoff goes to.the RCDT:[0.25].
b. Supplied.from standpipe (Reactor Makeup Water via standpipe) [0.25];

4 Part of the leakoff goes to the containment sump,- [0.125), and part goes to the RCDT [0.125].

c. To lubricate #3 seal [0'.25]; and to' prevent dissolved' radioactive gases from entering the containment atmosphere [0.25].

' REFERENCE Callaway Training-Lesson' Plan, Chapter 9, pgs. 8-14 Callaway Learning Objective, Chapter 9, Objective'G-003000K103 ..(KA's)

- ANSWER 2.02 .(0.50)

.True.

REFERENCE Callaway Training. Lesson Plan, Chapter 2, pg. 9 000055A206 ..(KA's)~

ANSWER 2.03 (2.00)-

a.

120060' amp amphour ' =.20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> (0.5) b,- 1. Fault on battery charger PJ21 f

2. Fault on battery charger PJ31 1
3. Fault on the bus (PJ01) J 4- .

Fault on the battery (PJ11)

(Any 3 @ 0.5 pts each)

REFERENCE Callaway Training Lesson Plan, Chapter 2, pg. 15 l

063000G008 000055K101 ..(KA's)

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

I,

PLANTDESIGNINCLUDIH(EAEETYANDEMERGENCX .Page 34'-

l[

-SYSTEME-1 o

-l ANSWER 2.04. (2.00)

a. 1. Low Lube Oil Pressure: l
2. HighLJacket Water Temperature
3. High Crankcase. Pressure i
4. Overspeed Trip l S. Generator. Differential Overcurrent Relay (Any.4l@ 0.25 pts each)
b. The purpose of the Start Failure Relays is to interrupt the starting of the D/G'[0.5] if a. predetermined speed is not reached [.125] or.

if a lube oil pressure is not established [.125] within a }

predetermined time following the start signal [.25].

' REFERENCE-Callaway Training Lesson' Plan, Chapter 3, pgs. 23-24  !

- Callaway. Learning Objectives, Chapter 3, Objective J Callaway FSAR, pg. 8.3-10 "

064000K405. 064000K402 .(KA's)

ANSWER' 2.05 (1.50)

1. Both ESW pumps start
2. SW supply to ESW valves close
3. SW return from ESW valves close
4. ESW return valves to Ultimate Heat Sink open 5, CCW heat exchanger outlet valves shut
6. .CCW heat exchanger inlet valves receive an open signal (Any 5 @ 0.3 pts each) l

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

l

f. I a _ _ _ = - - _ _

<L ,

j

2. ' PLANT DESIGN INCLUDING SAEETY AND EtiERGEliG'I Page 35;- )'
SYSTEMS 1 REFERENCE' J l

'Callaway Training Lesson Plan, Chapter 5, pgs. 3-9 .

rCallaway Learning Objective, Chapter 5, Objective D.3 4 076000K119 000062K302 ..(KA's)

J

.. 1

. 1 JANSWER - 2..0 6 - . (:2. 00 )- d

d 1 1 ' .. Containment sump to RHR pump "A" suction (HV-8811A) must be closed . )

1

2. RWST to RHR pump "A" suction (HV-8812A) must be closed .

'3. 'RHR-pump "A" discharge to the CCP's suction and'SI pump "A" suction

-(HW8804A).must be closed.

- 4 . ---

RCF e issure less than 360 psig.

(0.5 pts each) 4 REFERENCE

-Callaway. Training Lesson Plan, Chapter 7, pg. 11 005000K407. ..(KA's)

ANSWER 2.07' (1.50)

1. Containment Spray Pump dischargo pressure (PI-4, PI-10)
2. Containment Spray System Flor rate (FT-5, FT-11).
3. Sodium Hydroxide flow rate (FI-13, FI-14)
4. Spray Additive Tank Level (LI-17, LI-19)
5. Recirculation Sump Level (Any 4 @ 0.375 pts each)

REFERENCE Callaway Training Lesson Plan, Chapter 18

-Callaway Training Learning Objective, Chapter 18, Objective J 026000A401 ..(KA's) y (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)

l^

1

, :2. PLANT DESIGN INCLUDING SAEETY AND EMERGEEQX Page 36-L SYSTEMS i .

j j

1 t-l ' ANSWER. 2.08_ (2.00)

a. 1. Flush boron from' upper: regions of core >
)

'i

2. Collapse steam bubble _in top of the. head'(coolfupper head)  !

i

.(0.75 pts'each)

b. False. (0.5)

-i d

REFERENCE Callaway Training Lesson Plan, Chapter 17, pgs. 2, 6 . 1 Callaway Learning Objective, Chapter 17, Objective'H j 006000A208- 000011K313

~

..(KA's)

~ ANSWER' '2. 0 9 ' (2.00) .{

a. 1. .Open
2. Open  !
3. Closed ,

'4. ~ Divert to VCT (0.25 pts each)

b. -To insure that the regenerative heat' exchanger is always at'RCS

. pressure (0.25] which will prevent steam flashing [0.25] and possible damage to its tubes [0.25), and to prevent uncovering (damage)

< pressurizer heaters. (PZR level < 17%) [0.25).

' REFERENCE Callaway Training Lesson Plan, Chapter 11, pgs. 2-12 l 000004G007 004010K403 078000K302 004010A204 ..(KA's) i l

i- ,

j l

' ANSWER 2.10 (1.50) ]

Disagree [0.5]. - RCP #1 seal leakoff is assured by seal water leakoff l header relief valve (V-8121) [0.5] which will relieve to the PRT [0.5]. I 1

.1

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

i

n a .

?

22 PLANT DESIGN I'NCLUDING SAEETY AND EMERGENCY- Pega 37 rSYSTEME REFERENCE L _ .

Callaway Training Lesson Plan, Chapter 11, pgs. 1-13 ,

003000A201. 004010A205 ..(KA's)  !

I

> ]

ANSWER 2.11 (1.50) -

'i

a. 1. Low lube oil pressure to feedwater pump bearing.
2. Low lube' oil pressure-to feedwater turbine bearing.
3. Turbine overspeed
4. Thrust bearing wear
5. Exhaust vacuum low

.(Any 4 @ 0;25 pts each)

b. Ensures that if auxiliary feedwater. system starts, that flow will not reverse, but will go into the S/G. (0.5)

REFERENCE Callaway Training Lesson Plan, Chapter 23, pgs. 12-15 Callaway Training Learning Objective, Chapter 23, Objective D 059000K102 059000K416 ..(KA's) 1 ANSWER 2.12 (2.00)

a. 1. To limit containment pressure caused by a ruptured steam line inside containment by limiting AFW flow to the faulted.

S/G.

2. To limit the pressure drop on the auxiliary feedwater headers in the case of a feed line break. (Permits feeding the intact S/G's in spite of auxiliary feedwater loss due to the feedline break.)
3. To limit potentially damaging AFW pump runout.

(Any 2 @ 0.75 pts each)

b. False. (0.5)

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

Pago 38

.71_ PLANT DEElGN INGLUDING SAFETY AND EM$RGENCY.

SY.ETEME >

l -

REFERENCE

.Callaway Training Lesson Plan, Chapter 25, pgs. 5 and'8 Callaway Training Learning Objective, Chapter 25, Objective G 000061G007 061000K404 ..(KA's)

ANSWER 2.13 (1.50)

.1 . Liquid Radwaste Discharge Monitor (Plant Discharge Monitor)

(HB-RE-18)

2. S/G Blowdown Discharge Monitor (BM-RE-52)
3. Turbine Building Drain Monitor (Oily Waste Discharge Monitor)

(LE-RE-59)

4. Secondary Liquid Waste System Monitor (HF-RE-45)

(Any 3 @ 0.5 pts each)

REFERENCE Callaway Technical Specification, Table 3.3-12 073000K401 ..(KA's)

ANSWER 2.14 (0.50)

Will continue to operate (no change). (0.5)

REFERENCE Callaway Training Lesson Plans, Chapter 11, pg. 9; Chapter 17 013000K111 004000K115 ..(KA's)

ANSWER 2.15 (1.50)

a. Demineralized water (0.5) ,i
b. Essential Service Water [0.5] supplied directly to the CCW pump suction header [0.5].

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

i

'2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pege'39 L

l- (SYSTEdE l

, REFERENCE' i

, Callaway Training Lesson Plan, Chapter 10, pg. 3 '

008010K101 ..(KA's) ,

P ANSWER 2.16 .(1.00)  ;

.The' alternate emergency boration line taps off downstream of it.

a.

(0,5) l 1

b. FALSE. - ( 0,5 ). .!

REFERENCE Callaway training L6sson Plan, Chapter 11, pg. 2 004000G007 .004010KE09 ..(KA's)

. ANSWER 2.17 (0.50)

FALSE. (0.5)

REFERENCE Callaway Training Lesson Plan, Chapter 9, pg. 13 004000K104 ..(KA's) w

, :la i

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

INSTRUMENTS AND CONIROLS Page 40 r3 l

ANSWER. 3.01 (2.25)

a. N41 failing high sends a signal to all FRV bypass valves to open.

(0.25) -

Causes all-S/G levels to increase. (0.25)

When level is greater than 50%, level error sends a close signal to the FRV. bypass valves (0.25)

FINAL CONDITION:

1. Level increases > P14 (High-High S/G Level), causing'FWIS and trips the MFP.and the main turbine, S/G fed by AFW OR
2. Level maintained greater than 50% in all of the S/G's.

(Either @ 0.25 pts)

b. S/G 1evel failed low sends a signal to the B S/G FRV bypass valve to open (0.25)

Causes actual level in B S/G to increase (0.25)' )

' Level in S/G B will' increase to P14 (High-High S/G 1evel) (0.25) i Causes MFP trip, turbine trip, FWIS, and all S/G's to be fed by AFW (0,5) ,

REFERENCE-Callaway Lesson Plans, Chapter 23, "Feedwater," Objective G 035010K401 035010A203 000054A205 059000G007 059000A211 059000K419 059000K104 ..(KA's)

ANSWER 3.02 (2.00)

a. TRUE
b. TRUE
c. FALSE
d. FALSE (0.5 pts each) i

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

1

7

3. INSTRUMENTS'AND CONTROLE Pcge 41
,y i

? i i- REFERENCE ]

'Callaway Lesson Plans, Chapter 42, " Loose Parts Monitoring System,"

Objectives B, C, E, and F 016000K112 016000K101 016000G007 016000G004 ..(KA's)

L:

l ANSWER 3.03 (2.80) >

1. RCS wide range pressure indication l 1
2. Pressurizer. wide range level indication ]
3. Cold Leg wide range temperature (for each loop)
4. Source range NIS indication
5. Intermediate range NIS indication f
6. Pressurizer backup heater control and indication l CVCS Letdown control and indication for:
7. Letdown isolation valves (LCV 459, 460)
8. Orifice isolation valves (HV8149 A, B, C)
9. Containment isolation valves (HV8152 and 8160)
10. Hot leg wide range temperature (for each loop)
11. Pressurizer pressure (Any 7 @ 0.4 pts each)

REFERENCE Callaway Lesson Plans, Chapter 48, " Control Room Isolation / Auxiliary Shutdown Panel," Objective B, pg. 3 000068A112 000068K201 ..(KA's) i

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

4 3' . INSIBUMENTS AND CONTROLS Pcgo 42 ANSWER 3.04 (2.40)

Containment High Pressure.(HI-2), 17 psig, 2/3, not blockable (0.2 pts.each, 0.8 total)

RLow Steamline Pressure, 615 psig, 2/3 any steam line, manual block < P11

-(1970 psig) (0.2. pts each, 0.8 total)

High Steam Pressure Rate, 100 psig in 50 sec, 2/3 any steam line, in effect only if low steamline pressure MSLI is blocked (0.2 pts each, 0.8 total)

REFERENCE Callaway Lesson Plans, Chapter 49, " Main Steam and Feedwater Isolation -

-Actuation System," Objective D, pg. 5 000040A204 000040G011 000040A101 ..(KA's)

ANSWER 3.05 (3.00)

C-1 High Flux IR, 1/7. , 20% (amp equivalent), auto and manual, blocks outward only (0.1 each, 0.5 total)

C-2 High Flux PR, 1/4, 103%, auto and manual, blocks outward only (0.1 each, 0.5 total) l C-3 OT delta T, 2/4, 3% Below Reactor Trip Setpoint, auto and manual, blocks outward only (0.1 each, 0.5 total) 0-4 OP delta T, 2/4, 3% Below Reactor Trip Setpoint, auto and manual, blocks outward only-(0.1 each, 0.5 total)

C-5 low turbine power Pimp, 1/1 selectable, 15% , auto only, blocks outward only (0.1 each, 0.5 total)

C-11 Bank D withdrawal stop, P/A converter, 223 steps, auto only, blocks outward only (0.1 each, 0.5 total)

REFERENCE Callaway Lesson Plans, Chapter 26, " Rod Control," Objective J, pg. TP-5 001000K407 015000K402 ..(KA's)

-I L

i l

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

i

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

l' i-U ' 3. LINS'TRUMENTS AND CONTROLS .

Page 43 l-l-

i ANSWER- 3,06- . (:2. 00 )

a .' ;No effect on control rods [0.25] N35 inputs a rod stop at'20% (amp Jequivalent).and reactor trip at.25% (amp equivalent) [0.75].

Lb, -No-effect-on control. rods-[0.25). _PG-20 supplies #2 rod drive M-G' set;.however, rod drive M-G set.#1 would still be powered-[0.75].

REFERENCE Callaway_ Lesson Plans, Chapter 26, " Rod Control," Objectives.H, J, L.

001000K201' 001000K105 . (KA's)

- ANSWER 3.07 (2.35) i a. One reactor trip breaker and its bypaas breaker both open. (0.75)

b. 1. Trips the main turbine
2. Initiates a feedline isolation when Tavg decreases below low Tavg setpoint (of 564 degrees F).
3. Prevents automatic re-actuation of a safety injection after
a. manual reset. (P-4 required to permitl31 reset.)
4. Transfers steam dumps from the load reject controller to the ,

plant trip controller.

5. Prevents opening of feedwater isolation valves after they were closed by a SI signal or high-high steam generator level.

(Any 4 9 0.4 pts each)

REFERENCE Callaway Lesson Plans, Chapter 27, " Reactor Protection,"

Objective E, pg. 17 012000G004 012000K603 012000K610 ..(KA's) i i

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

___ _ ____j

3.. 1NSTRUMENTS AND CONTROLS Paga 44

' ANSWER 3.08 (2.00)

a. NO CHANGE
b. DECREASE
c. INCREASE
d. INCREASE (0.5 pts each)

REFERENCE Callaway Lessson Plans, Chapter 27, " Reactor Protection," Objective D, pg. 13, T/S 2.2.1, Table 2.2-1 012000K503 012000A205 ..(KA's)

ANSWER 3.09 (1.00)

N31 feeds Flux-Doubling / Boron Dilution Protection System. (IF counts ,

double when averaged over one minute and compared to nine previous one minute samples), CVCS system switches charging suction from the VCT to the RWST causing a boration. (1,0)

REFERENCE Callaway Lesson Plans, Chapter 28, "Excore Nuclear Instrumentation,"

Objective F, pg. 5 LER 86-004 015000G004 015000K604 015000A107 ..(KA's)

ANSWER 3.10 (1.20)  ;

a. Core subcooling monitor
b. Reactor Vessel Level Indicating System (RVLIS)
c. Cold Overpressure Protection System (COPS)

(0.4 pts each)

REFERENCE Callaway Lesson Plans, Chapter 30, " Reactor Instrumentation," Objective D, pg. 12 016000G004 018000K301 ..(KA's)

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

y n. y .,

i- Y Pago 45 p INSTR MENTS AND CONTROLS-q q.e <

w

, . .t -

Bn.

? ANSWER 3.11L '(1. 50 ) .

l 4 lbontrolRoomLVentilation. Isolation ~(CRVI)

. Fuel' Building-Ventilation Isolation (FBVI)

-Containment Purge I-solation (CPI)

+

i;.s ;. ..

s T (0.5 pts'each)- -

REFERENCE' Callaway.LER 86-005 072000K403- .

072000K402 072000K401 ..(KA's)

. ANSWER 3.12 (1.00)

Auxiliary Feedwater. Actuation (AFAS) (0,5)

Steam' Generator Blowdown Isolation (SGBIS) (0.5)

REFERENCE.

Callaway LER 86-015 061000K203 061000K202 061000A203 ..(KA's)

ANSWER 3.13 (1.50)

1. Control Room Air Supply High Radiation (GK-RE-4 and 5)

'2. Containment Purge-Exhaust High Radiation (GT-RE-22 and 33)

.r

3. SI Signal
4. . Containment Isolation Signal Phase A (CISA)
5. Fuel Building Isolation Signal (FBIS) l i
6. Manual (Main. Control Board)

, (6-@ 0.25 pts each)

REFEREWCE Callaway Lesson Plans, Chapter 39 " Ventilation Systemi," Objective C, TP-5

~073000K101 072000K403 072000K204 ..(KA's) i

(***** END OF CATEGORY 3 *****)

l I

h_ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ i

~ -

y_ _ 7, _ _

'4. PROCEDURES -' NORMAL.'ABNQRMAL. EMERGEHQY Pega 46 JAND RADIOLOGICAL CONTROL ,

& t ANSWER 4.01 -(1,75)

a. Manually trip reactor (RLO3), . ,

(If not) trip reactor from BOP console (RLO6)'

(If'not) trip power supply breakers to LC PG-19 and LC PG-20 (If not) manually insert RCCA's (0.25 pts each, 1.0 total) c ,T

b. Manually trip. turbine

~

(If not) manually run back turbinem; (If not) fast cloue main,sterm isolation and bypass valves 1 '

(0~.25 pts each, .75: total) i-

' REFERENCE Emergency Operating Procedures', FR-S.1, Response to Nuclear Power j Generation 1' 000029G010 000029A209- 000029K312 000029K306 000029K301

..(KA's) '

i;; .J_; e ANSWER 4.02 (0.50)

Indication - 21 degrees F 1 Error + 27 degrees F

+ 6 degrees F superheated (0.5 pts)

, i REFERENCE .

~ Emergency Operating Prdcodure', ES 0.2, Naturdl Circulation Cooldown 000056G012 000056A255' 4

..(KA's)

ANSWER 4.03 -(1.00)

Ensure reactor' trip  !'

Ensure turbine trip Check if RCS is isolated Ensure AFW flow - greater than 26'0,000 lbm/hr

.(0.25 pts each, 1.0 total) l ,

I i

L ,

(***** CATEGORY 4 CONT,13UED ON NEXT PAGE *****) k i

l 5

p .. .-

Pago 47

4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY

'AND RADIOLOGICAL CONTROL REFERENCE-Emergency Operating Procedure, ECA-0.0, Loss of All AC Power 000055K302 000055G010 ..(KA's)

ANSWER 4.04 '(2.00)-

a. 'All rod ~ bottom lights - LIT Reactor trip nud bypass breakers - OPEN NR-45 recorder - DECREASING FLUX NIS indications - DECREASING FLUX (0.125 pts each for indication and expected condition, 1.0 total)
b. MFW isolation valves - CLOSED MFW regulating. valves - CLOSED MFW regulating' bypass valves - CLOSED FW chemical injection valves - CLOSED (0.~125 pts each for indication and expected condition, 1.0 total)

' REFERENCE Emergency Operating Procedure, E-0, Reactor Trip or Safety Injection 000007G010 000007A106 000007K301 ..(KA's)

ANSWER 4.05 (1.50)

a. Single TLD Placement The whole body TLD shall be worn on the front part of the body at or between the shirt pocket to beltline level [0.25). The PIC a shall be worn next to the TLD so that both devices receive similar I exposures [0.25). )
b. The individual shall leave the area immediately [0.25] and report to Health Physics Access Control {0.25).
c. 100 rem whole body, 300 rem extremity (0.25 pts each)

REFERENCE i I

HDP-ZZ-01400, External Dosimetry Program l 194001K103 194001K104 ..(KA's)

(4**** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

NORMAL. ABNORMAlu__EMEBQXHQ1 Pcga 48 3.' PROCEDUBES'

'AND RADIOLOGICAL CQNTROL-

+ _

+ ,

ANSWER. 4.06 (2.00)

a. START-

.b. ' START c.. .NOT START- .

-d.. NOT START .

(0.5 pts each)-

REFERENCE OTN-BB-00003, Precautions and Limitations O '003000K103 1003000G010 ..(KA's) 0

' ANSWER 4.07 '(1.50)

. Standing / Night Orders

' Control' Room Annunciator Test Annunciator Defeat Log URO Log Walk:Down Control Boards-Incident Reports Operations /Significant Maintenance in progress (Any 6 @ 0.25 pts each)

' REFERENCE ODP-ZZ-00003, URO Checklist 194001A113 194001A106 ..(KA's)-

i I

i l

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) '

L-______-____________

24. PROCERURES - NO.BMAL. ABNORMAL. EMERGENCY 'Paga 49~

,&ED RADIOLOGICAL CONTROL ANSWER 4.08 (1.00)

1. The first action to be taken should be to stop the : spill and' control the spread of radioactive material as promptly as possible. (0.2)-
2. Warn other personnel-who could'either become contaminated by the spill or ctuld provide assistance in controlling it [0.2), and i notify the Control Room [0.2).

.3.- Keep unnecessary personnel away from the area affected by the spill.

to minimize the spread of contamination. (0.2)

4. 'If the spill is such a magnitude that radiation or airborne radioactivity levels present an unacceptable hazard to personnel, evacuate the area and restrict access until the conditions can be properly assessed. (0.2)

' REFERENCE  ;

APA-ZZ-01000, paragraph 4.11.1.1, pg. 20

-194001K103 194001K104 000059K304 ..(KA's)

ANSWER 4.09 (2.50)

a. 350 degrees F, 450 psig (0.25 each)
b. 200 psid, 15 psig (0.25 each)
c. 165 degrees F. (0.25)
d. A or C, cavity cooling (0.25 each)
e. 160 degrees F (0.25) i (0.25) i
f. 15 cc/kg
g. 350 degrees F (0.25)

. REFERENCE OTG-ZZ-00001 022000G010 004000G010 005000G010 003000G010 . (KA's)

L I

I f

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

l L_____.__:-___--- --

,5; .. t s L. PEQQED.QnES iNORMAL; ABNQBMAL; EMERGENgl .

Pcgo 50 y . MWUQLOGICAL CONTRQL ti .

1 v.

A M ER 4.10 '(2.25) a;. 394,000 (gallons) (0.25) ec <

H

'b . - 37 (degrees F) (0.25) ,

c. 100-(degrees F) (0.25) d.. open,fremoved (0.25 each) er. 6122, 6594-(gallons) (0.25 each)
f. 602, 648 (psig) (0.25 each)

REFERENCE-T/S 3.5.5, 3.5.1 006000G011 006020A107 006020A109 ..(KA's)

ANSWER -4.11 (2.00~)

(Rods are below rod insertion limit)

1. Open, emergency borate to charging pump suction (BG-HV-8104) (0.4)
2. Start both BA Transfer Pumps (0.4)
3. Verify' boric acid flow (on BG-FI-183A) (0.4)
4. Place RCS M/U CTRL (BG-HIS-26) in stop (0.4)
5. (To maximize RCS water turnover rate), ensure centrifugal charging pump is running [0.2) and establish letdown' flow of 120 gpm [0.2].

(0.4)

' REFERENCE T/S 3.1.3.6 .I OTO-ZZ-00003, TP 5.0 000024G010 000034G011 000024K302 000024K301 ..(KA's)

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

7 , ,_ _. - _

4' PBQQEDUBES - NOBS L. ABNORMAL. EMERGENCY Page~51 AM: RADIOLOGICAL _QQNTROL

)'

' ANSWER '4.12' .(2.00)

1. StopLthe respective RHR pump. (0.5)
2. Close the RCS suction. valve for the secured pump, (BB-HV-8702B or EJ-HV-8701A.).- (0,5) -
3. If RHR trains are cross-connected', shut either EJ-HV-8716A or D.

(0.5)

4. Trip any running RCP [0.25] if RCS < 325 psig :[0.125] and #1 seal delta P < 200 psig [0.125].

. REFERENCE OTO-BB-00003, TP'5.2 000009A223 000009A215 000009A202 000009G011 000009G010

...(KA's).

ANSWER- 4.13 (2.00)

1. Identify the failed channel (by comparing meter indications and

< identifying alarmed annunciators)' .

2. Select alternate impulse pressure channel (by utilizing the Impulse Pressure Selector. Switch).

3.- Maintain-Tave at corresponding Tref value for existing plant power.

4. Stop steam dump actuation (by turning Steam Dump Interlock Selector switches to OFF/ RESET position).

(0.5 pts each)'

REFERENCE OTO-AB-00005, TP 5.0 '

001000G014 001050K501 041000G014 ..(KA's)

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

4 J ..... . PROCEDURES - NORMAL . ABNORMAL. EMERGENCY Paga 52.

y AND RADIOLOGICAL' CONTROL'.

l 'ANSWERL -4.14 (1.00)

.a. Tagging this valva places both SI. pumps inoperable ;l(T/S 3.5.2, p 3.0.3). (0.5)

l. - -b. T/S 3.0.3 requires within 1-hour action initiated.Li(0.5) 3 (T/S interpretation 20 requires load reduction started within-1 hour)

REFERENCE T/S.3.0.3, 3.5.2 Callaway Event No. 09543 of 8/5/87 006000G011 006000K603 ..(KA's)

ANSWER' . 4 '.15 - . ( 1. 00 )'

a. FALSE b.- FALSE-(0.5 pts each) ,

, REFERENCE L__ '

APA-ZZ-00310, pgs. 8, 10 194001K102 ..(KA's).

ANSWER 4.16. (1.00)

a. Must wear a self-contained breathing apparatus. (0.5)
b. FALSE. (0.5) i i

REFERENCE APA-ZZ-00802 )

194001K114 ..(KA's) o i

(***** END OF CATEGORY 4 *****)  ;

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

.l 1

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