ML20209D699

Exam Rept 50-483/OLS-86-01 on 861209-12.Exam Results:Three Senior Reactor Operators & Six Reactor Operators Passed Oral,Written & Simulator Exams

ML20209D699
Person / Time
Site:	Callaway
Issue date:	01/26/1987
From:	Burdick T, Hare S, Higgins R, Sunderland P, Weale G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20209D449	List: ML20209D444 ML20209D699
References
50-483-OLS-86, 50-483-OLS-86-0, NUDOCS 8702040531
Download: ML20209D699 (98)

Text

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U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-483/0LS-86-01(DRS)

Docket No. 50-483 License No. NPF-30 Licensee: Union Electric Company Post Office Box 149 - Mail Code 400 St. Louis, M0 63166 Facility Name: Callaway Nuclear Power Plant Examination Administered At: Jefferson City, Missouri and Callaway Plant Examination Conducted: December 9, 10, 11, and 12, 1986 Examiners: h R. L. Higgins '

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G. Weale \ Date 1-24-g~l P. R. Sunderland Date

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'5. Hare 1 Date

/-26-87 Approved By:

dThomas' . Burdic , Chief l- 26-D Operator Licensing Section Date Examination Summary Written, oral and simulator examinations were administered on December 9-12, 1986 (Report No. 50-483/0LS-86-01(DRS)) to three SR0 candidates and six R0 candidates. A written examination was administered to one additional R0 candidate.

Results: Three SR0s and six R0s passed these examinations.

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DETAILS-1 . ,

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1. Examiners

• R. L. Higgins, NRC G. Weale, Sonalysts 4 P. R. Sunderland, NRC 1 S. Hare, NRC l

• Chief Examiner

2. Examination Review Meeting l

Refer to Attachment 1 a

3. Exit Meeting i

On December 12, 1986, at the conclusion of the replacement examinations,

! the examiners met with members of the plant staff to discuss generic

findings made during the course of the examinations. The following '

j personnel attended the exit meeting:

1 G. L. Randolph, Manager, Callaway Plant

M. E. Taylor, Superintendent of Operations

! J. M. Price, Superintendent of Training j M. S. Evans, Acting Superintendent of Training i W. O. Jessop, Senior Training Supervisor T. A. Robertson, QA Engineer R. L. Higgins, NRC Region III

P. R. Sunderland, NRC Region III The following areas were discussed

a. The exaniination schedule was reviewed.

! b. Callaway personnel were very cooperative during the examination week. Bill Jessop and his staff were especially cooperative.

This was much appreciated by the NRC staff conducting the j examinations.

. c. Generic weaknesses exhibited by the candidates included:

1. A few candidates failed to keep satisfactory rough logs during i their scenarios.

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2. When asked to draw block diagrams of Instrumentation and Control Systems, many diagrams were lacking major components.

l 3. Personnel were weak on the use of radiation measuring instruments.

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d. The following observations were noted:

1. It did not appear that there was an electrical print index with the electrical drawings in the control room.

2. One candidate accidentally put non-radioactive waste in a radioactive waste bin in the counting room.

3. The plant appeared to be very clean.

e. The candidates exhibited the following strengths:

1. Their level of teamwork during simulator scenarios was superior.

2. Use of Procedures was a strong point, especially emergency procedures.

3. The candidates, especially SRO, were strong at using and interpreting technical specifications.

4. Another strength not discussed at the exit meeting was the candidates thorough preparation for their examinations.

f. Of concern to the examiners was the system description material provided the region prior to the examinations. The material seemed to lack enough depth to write a comprehensive examination. In discussion, it was learned that classes rely on handouts and notes besides the broad base that the systems material provides.

g. Cerrments from Union Electric included:

1. Union Electric is concerned about the case of an individual whose applicaticn for instructor certification was granted on November 6. Then less than a week before the examinations, he was told that he could not take an instructor certification exam but could go up for an R0 exam. This caused him some distress.

2. The R0 written examination was too long. The Licensee estimated that 354 minutes were needed to take the exam. This left a total of six minutes for the candidate to take trips to the rest room and review his answers, which in the facility's opinion is an inadequate amount of time in which to perform these tasks.

The facility did think that the R0 exam was of exceptionally high quality.

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ATTACHMENT 1 The specific facility comments concerning the R0 and SR0 examinations, followed by the NRC resolution, are listed in the following paragraphs:

R0 EXAMINATION Question 1.08 Comment: Reasons (in parenthesis) are not asked for in question.

Resolution: Information or data enclosed in parentheses in the answer key is not required to be in the candidate's answer for the answer to receive full credit. This convention applies throughout the exam.

Question 1.12 Comment:

T-H Principles, Volume II, Pages 14-25naturalcirculationflowrate-Q'am and ^ Tam' . Credit should be given for this information.

Resolution: The quoted "thumbrule" relationship can be used to derive data for separate conditions or " snapshots" of natural circulation.

The thumbrules are not useful in answering this question, which requires relating natural circulation flow to forced circulation flow.

Question 1.17 Comment: Cycle II AFD overcore life.

The question does not saecify which core cycle is being discussed. At B0L AFD 3egins slightly positive due to fuc1 depletion in the older assemblies in the prior cycle. Toward E0L AFD shifts negative due to the fuel depletion in this cycle.

Resolution: During the administration of the exam, candidates were requested to state the applicable core cycle in their answers.

The description provided in the answer key was used to grade cycle 1 answers; the description provided in the facility contention was used to grade cycle 2 answers.

Question 2.07.a Comment: "RCP Thermal Barrier CCW Outlet Temperature" increasing (MCB indicator) should also be accepted.

Resolution: Increasing RCP thermal barrier HX CCW outlet temperature was added to answer key and candidate answers were graded on "any 3, 0.25 each" basis.

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Question 2.07.b Comment: Should also accept 1) 50 gpm isolation for individual CCW return.

2) CCW surge tank make-up valves isolate if open.

Re: 1) M-22-BB03 and 2) M-22-EG01 4

Resolution: The key was changed to read:

1) The affected RCP thermal barrier HX outlet isolation valve will shut (0.25) on high flow - 50 gpm (0.25), OR the inside containment isolation valve (HV-62) will shut (0.25) on thermal barrier HX outlet flow high - 200 gpm (0.25). (These actuations are included as one answer because they are mutually exclusive; after one of these actuations occurs the other actuation probably will not occur.)

2) The CCW surge tank vent (and makeup valves) will shut 1 (0.25) on high CCW system radioactivity (0.25).

Question 2.08 Comment: This question does not specify the level of detail required.

Specific actions such as the closing of AF-BTV-12B, Bleeder Trip Valve, AF-LV8, 7A Heater Drains, AF-LV17 ' A' MSR 1st Stage Drain i Tank, AF-LV-215, 'C' MSR 1st Stage Drain Tank, AF-LV-12C, 5th Stage. Extraction, AF-LV-120, 'C' MSR Scavenging Steam, AF-LV-12E, ' A' MSR Scavenging Steam should be accepted.

Re: M22-AF-01 1

Resolution: The answer key was revised to include closure of the specific valves listed in the facility contention. Candidate answers were graded on "any,3, 0.25 each" basis.

Question 2.09.a Comment: The following additional answers should also be accepted:

Live bus transfer Main generator breakers open or main generator lockout relays tripped.

Startup transformer energized.

. Re: E-02-PA02 l

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Resolution: Answer (2) in the answer key was revised to read, " Main generatoroutputbreakersopenormaingeneratoroutgutrelays tripped." The answer "startup transformer energized was accepted, but was not required since it is included in the title (live-bus transfers require live buses, a transfer mechanism, and bus interconnections also).

Question 2.11 Comment: These are additional answers for recirculation. Question does

, not specify whether is Hot or Cold leg recirculation.

- SI pump mini-flow valves closed

- SI and RHR discharges realigned to Hot Legs for Hot leg recirculation Re: ES 1.3, Step 3b ES 1.4 Resolution: Since the question did not specify how closely the realigned flowpath had to be related to RHR or whether the shift was to cold or hot leg recirculation, the additional answers provided in the facility contention were added to the answer key and candidate answers were graded on "any 3, 0.333 each" basis.

Question 2.15 Comment: These are also correct answers:

- GT-RE 31/32 Containment Atmosphere Monitors

- GG-RE 27/28 Fuel Building monitors during normal Fuel Building ventilation operation.

Re: M-22-GG01 and M-22-GT01 Resolution: The additional answers provided in the facility contention were added to the answer key and the candidate answers were graded on "any 6, 0.167 each" basis.

Question 2.16 Comment: 21.71 psia is per setpeint document.

Resolution: 21.71 psia was accepted as an alternative to psig.

Question 3.01 Comment: Alarms should be acceptable as answers.

- Pzr. low level deviation 6

- Pzt. 17% Heaters Off/ Letdown Isolated Re: 0T0-BB-00007 Symptoms Resolution: The question asked for " component actuations", which does not normally include alarms. Candid 4te answers that included the alarms. listed in the facility contendtion received 60% of full l

credit (0.15 vice 0.25).

Question 3.03 Comment: The following are not directly caused by the actuator signals noted:

a. Turbine trip - fed from P-4 Rx trip

b. Turbine trip - fed from P-4 Rx trip

c. MDAFAS - fed from trip of both MFP's Re: AFW, page 7, MFW, Pages 13 and 15 Resolution: Since the meaning of the word "directly" in the question may have confused the candidates, the three actuation signals listed in the facility contention were deleted. Therefore, each part of the answer has 3 correct answers, worth 0.333 each.

Question 3.11 Comment: NRC exam proctor pointed out that calculations were not necessarv.

Resolution: The NRC exam proctor was correct.

4 Question 3.14 Comment: ' Backup heaters energizing at +5% deviation reason is incorrect: The outsurge is an assumption based on a subsequent pressure drop due to some unrelated transient.

Re: CPSH0 RCS Instruction, page 29 Resolution: The answer key's reason for the backup heaters energizing at

, +5% deviation is correct. However, the following"more general alternative reason was added to the answer key: to heat subcooled insurge water, thus minimizing possible pressure i transients."

Question 4.07 Comment

If the break is inside containment, the containment High-2 signal would provide protection.

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Resolution: The facility contention was added to the answer key as an acceptable alternative answer to parts e and f.

Question 4.08.b Comment: To prevent injection of N2 from the accumulators into the RCS is correct answer and should be awarded full credit.

Re: ECA-0.0, page 12 Resolution: Candidate answers which stated the reason for the caution similarly to the statement provided in the facility contention received full credit. The answer key breakdown of point values was used for partial credit assignment when candidate answers did not provide the full statement of the reason.

Question 4.12 Comment: d. 583 F is also acceptable. (Sec. 2.6 of 0TG-ZZ-00001)

e. <350 F is also acceptable. (Sec. 2.9 of 0TG-ZZ-00001)

Re: Tech Spec RCS Flow Requirements Resolution: The values provided in the facility contention were added to the answer key as acceptable alternative answers to parts d and f.

Question 4.33.a Comment: New revision to HDP-ZZ-01400 changes Part a.1 and a.3 to 6000 mrem (a.3) and should also be accepted.

Resolution: The values provided in the facility contention were added to the answer key as acceptable alternative answers to parts 1 and 3.

Question 4.13.b

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Comment: HDP-ZZ-01400 references Supt. HP and Plant Manager as two separate approval requirements. Other possible acceptable answers may be exposure history.

Resolution: " Exposure history" is not listed as a requirement for increased exposure in HDP-ZZ-01400 or Attachment 2 to HDP-ZZ-01400. The approval of the Superintendent of Health Physics was added to the answer key and candidate answers were graded on "any 2, 0.25 each" basis.

Note: A facility review of HDP-ZZ-01400, sections 2.6 and 3.4.3.1.3, for possible conflicting requirements is recommended.

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Question 4.14.b Comment: TCN-86-677 dated 5-9-86 includes Review of UR0 Log and completion of crew assigment sheets as two other answers that should be acceptable.

Resolution: The question asked for additional watch relief checklist actions besides the UR0 log review discussed in part a.

Review / completion of the Crew Assignment Sheet was added to the answer key and candidate answers were graded on "any 6, 0.167 each" basis.

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SR0 EXAMINATION Question 5.06 Comment: c. Should get credit for increasing suction pressure to increase available NPSH.

Re: T/H, Chapter 10 Resolution: Comment accepted. Answer key changed to reflect comment.

Question 6.03 Comment: 1) Rate setpoints are 14% in two seconds vice 5% in two seconds.

Re: Tech Specs, Section 2 (Table 2-2-1)

Resolution: Comment accepted. Answer key changed to reflect comment.

Question 6.08 Comment: c. No is correct. But P-7 Deenergizes which blocks low flow in one loop, RCP UF and RCPlV.

Re: CPSHO, Reactor Prote: tion Resolution: Comment accepted. Question was changed during exam and answer key after the exam to reflect comment.

Question 6.10 Comment: a. The key assumes a loss of voltage as well as a LOCA has occurred (diesel does not normally close in on LOCA - the load shed signal explained is for an undervoltage condition only).

SIS sheds all non-safety loads from safeguard buses. LOCA sequencer starts loads required to mitigate an accident.

Re: Student Handout - Safeguards Power P-6 Resolution: Comment accepted. Answer key changed to reflect comment.

Question 6.13 Comment: 0TN-EP-00001, Page 3, allows use of either "SI" pump to fill accumulators.

Resolution: Comment accepted. answer key changed to reflect comment.

Question 7.05 Comment: a. Question does not specifically address a procedure

- should also get credit for condenser off gas radiation (E-0, Step 23, kickout to E-3) 10

. O Resolution: Comment accepted. Answer key changed to reflect comment.

Question 7.06 Comment: a. Question only asks for four annunciators, but Key requires five for credit. Also, setpoints were not asked for but were required for credit in Key.

c. OT0-BB-00002, Page 4, lists six times pump must be tripped vice only this one.

Resolution: Comment accepted. Answer key changed to reflect comment.

Setpoints were eliminated from the key.

Question 7.08 Comment: b. should be >160 F.

Re: E-0, Page 3 Resolution: Comment accepted. Computer does not print this figure, exam and key must have these hand drawn in.

Question 7.14 Comment: a. Low speed mechanical stop at 47 gpm vice 32 gpm.

Re: CPSH0, CVCS, Page 1-10 Resolution: Comment accepted. Answer key changed to reflect comment.

Question 8.01 Comment: f. APA-ZZ-00010 no longer requires 0A, credit should be given for either.

Resolution: Comment accepted. Question clarified during exam so that candidates would answer with current response. Key was changed to reflect comments.

Question 8.02 Comment: c. should be <.95

d. should be 75%

f. shouldbe{200F Re: T/S, Pages 1-9 Resolution: Comment accepted. Computer will not print these characters.

Answer key manually changed to reflect comment.

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Question 8.03 Comment: a._ APA-ZZ-01000 (new procedure) requires ED0 approval for emergency entry. Should take SS or E00 for full credit.

Re: APA-ZZ-01000 Resolution: Comment accepted. Question was clarified during exam so that candidates would answer with current response. Key was changed to reflect comment.

Question 8.06 Comment: b. Answer is incorrect, in accordance with Tech Spec 3.2.1, although Action C is satisfied, the actions re must still be satisfied prior to exceeding 50% RTP, quired i.e., in "B" must have <60 penalty minutes in the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before increasing >50% RTP. Should be able to increase at 1615 on 12-9-86.

Resolution: Comment accepted. Answer key changed to reflect comment.

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MASTER COPY U. S. NUCLEAR REGULATORY COMMISSION

SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _QALLAWAy________________

REACTOR TYPE: _PWR-WEg4________________

DATE ADMINISTERED _@6412[99________________

EXAMINER: _gyNDERLANQ2_Pz__________

CANDIDATE: _________________________

INSIByCIlgNS_Ig_C6Np196IE1 Use separate paper for the answers. Write answers on one side only.

Staple question sheet en 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 si:: (6) hours after the examination starts.

% OF .

CATEGORY  % OF CANDIDATE'S CATEGORY

-_YB6UE_ _19166 ___SCgBE__, _y@6yE__ ______________C6IEGQBy_____________

_2Ez99-- 2Ezgg ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 2Ez99__ _2pt99 ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 25z99-- 2Ez99 ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL

________ B. ADMINISTRAT IVE PROCEDURES, 2Ez99-_ 25199 ___________

CONDITIONS, AND LIMITATIONS 100.00 Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.

Candidate's Signature

A h

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

1. Cheating on the examination means an automatic denial of your application cnd 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 gely to facilitate legible reproductions.

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

examination.

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

6. Use only the paper provided f or answers.

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

B. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a ngw page, write gnly gn gng sidg of the paper, and write "Last Page" on the last answer sheet.

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

9.

10. Skip at least thrgg 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 commonl y used in f acili ty lij,erature.

13. The point value f or each question is indihated in parentheses after the l question and can be used as a guide f or the depth of answer requ2 red.

14. Show all c al cul ati ons , methods, or assumptions used to obtain an answer to mathematicel problems whether indicated in the questi on or not.

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

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16. If parts of tne examination are not clear as to intent, ask questsons ci I the gami_ner on1y.

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 I completing the ex amination. 7his must be done af ter the examination has been completed.

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i 4 I Er. When you comp)ete your ex ami nat i ore , you shal1:

c. Assemble your examination as follows:

(1) Exam questions on top.

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

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

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

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

d. Leave the exam 2 nation area, as defined by the examiner. li after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.

• D2..TBEQBy_QE_ NUCLE 98_EQWgB_EL9NT_QEgB911gy2_E(ylpS,_gND PAGE 2 ISEBd99fN9DICS

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OUESTION 5.01 (1.00)

Why as Prescura:er Level programmed?

QUES 110N 5.02 (1.50)

Dr. an increase in steam demand f rom 50% to 75%, state whether the following parameters increase, decrease, or stay the same: (.25 each)

a. T-steam ___________________

b. T-average ___________________

c. K-eff _______ ________

d. P-steam ___________________,

o. PZR-level ___________________

f. S/G-level ___________________

QUESTION 5.03 (2.00)

Answer the following concerning Rod Insertion Limits:

a. What are the three bases for Rod Inserti on Li mi ts. (1.5)

b. What two plant parameters are used by the RIL computer to calculate R1Ls. (0.5)

QUESTION 5.04 (2.50)

e. Define Duadrant Power Tilt Rati o. (1.5)

b. What two conditions (thermal limits) does OPTR protect against?

(1.0)

QUESTION 5.05 (2.00)

During a reactor startup before going critical, a f ault in the S/G 1evel control system allows S/G A to be overfed. What effect does this have on the parameters listed below (i ncr ease , decrease, no effect). Explain each cnswer briefly.

G. T-average

b. Reactivity

c. Estimated Critical Position

d. Control rod worth

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

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-D:__ISE9BY_DE_N99LEBB_EDWEB ELONI_DEEBBI19N2_EL91pS,_eND PAGE 3 ISEBU99XNSDIQS QUESTION 5.06 (2.00) i 1

Rsfer to the diagram at the end of the exam that shows a closed cooling water system (similar to CCW) with a single centrifugal pump, a heat onchanger, a surge tenk (wi th i sol ati on val ve) , and a flow control valve.

The isolation valve for the surge tank is shut inadvertently and the pump discharge flow control valve is f ully opened. Soon afterward, the pump otarts operating noisely and vibrating excessively.

o. What is thi s phenomenon called? (0.5)

b. What are two reasons why shutting the pump discharge flow control valve half way may stop the abnormal pump operation? (1.0)

c. What is one reason why opening the surge tank isolation valve halfway may stop the abnormal pump operation? (0.5)

QUESTION 5.07 (2.00)

a. List four factors that affect axial power distribution. (1.0)

b. Define axial flux difference. (0.5)

c. If delta-1 is out of the programmed band to the left;

1. Is more power produced in the upper or lower half of the core? (0.25)

2. What must be done by the operator to restore delta-I to the target band short of changing power? (0.25)

DUESTION 5.08 (1.50)

The plant experiences a 7% load rejection at 100% power . Describe in detail what occurs in the pressuri er to maintain RCS pressure control.

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l DUESTION 5.09 (2.50)

Answer the f ollowing concerning xenon production and removal:

List the two methods that contribute to xenon production. (0.5) a.

b. List the two methods that contribute to xenon removal . (0.5) j c. Describe how and why xenon concentration varies (f or an extended period of time) after a reactor trip from 100% power. (1.0)

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d. You plan to go critical 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after the trip. If the startup is i

l delayed two hours, how and why is the estimated critical position affected. (0.5)

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

_ _ __ -2. _ _ _ - . _ _ _ _ _ _ _ _ . . _ . . _ _ - _ . - _ . _ _ , _ _ _ _ _,_ ___

PAGE 4 Lt__1SEgBl_QE_Nyg6E@B_BQWEB_EL@NI_QEEB8IlgN3 _ELylpg3_@ND IMEBDQpyN8bigS QUESTION 5.10 (2.00)

Zirconium has desirable physical and heat transfer characteristics, however, at high temperatures circonium reacts with water in the following roaction:

2r + 2 H2O = ZrO2 = 2H2

c. List two reasons that this reaction is undesirable. (1.0)

b. List the reactor design base that prevents this reaction from occurring. (1.0)

QUEST 1DN 5.11 (2.00)

R2 ferring to the curves in the back of the exam, calculate the final boron concentration (in ppm) of the reactor coolant system to keep the control rods in a constant position on orders f rom the Load Dispatcher to decrease power from 100% to 70%. The present baron concentration is 600 ppm (MOL) cnd the reactor has been at 100% power for sixty days.

QUESTION 5.12 (2.00)

With the plant operating at approximately 100% power, a reactor trip and turbine trip occur. Provide a full explanation of what initially happens to steam generator levels.

QUESTION 5.13 (1.00)

A variable-speed, motor driven centrifugal pump is operating at 1/2 speed in a " closed" cooling water system with the following parameters

a. Current = 40 amps

b. Flow = 50 gpm What are the new values for these parameters when the pump is shifted to full speed.

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

g___IMEgBy_gE_Nyg6geB_EgWgB_EleNI_gegBeligg,_E691DE2_9ND PAGE 5 IdgBb9DyNeDIGE QUESTION 5.14 (1.00)

During a reactor startup, the first reactivity addition caused the count rate to increase from 20 cps to 40 cps. The second reactivity addition ecused the count rate to increase from 40 cps to 80 cps. Which of the following statements is correct? (There is only one correct answer.)

a. The first reactivity addition was larger

b. The second reactivity addition was larger

c. The first and second reactivity additions were equal

d. There is not enough data given to determine the relationship of reactivity values.

(***** END OF CATEGDRY 05 *****)

6___EL9NI_@YSIEUS_DgglGN3_QQNIBQL3_@NQ_lNgIBUdENI@IlgN PAGE 6 QUESTION 6.01 (2.00)

List the conditions that will cause both main feedwater pumps to trip.

QUESTION 6.02 (1.00)

What are the power supplies f or the f ollowing equipment?

o. RCP C

b. Aux Feed Pump A

c. Inverter NN11 (dc supply)

d. S1 Pump B

o. PZR Backup Heaters, group A QUESTION 6.03 (2.40)

Li st all trip signals that are derived from Nuclear Instrumentation.

Indicate which range the trip signal is generated in, and the setpoint at which the trip occurs.

QUESTION 6.04 (2.00)

Answer the f ollowing concerning the Auxiliary Feedwater System:

a. Which of the following 4 conditions will cause a motor driven auxiliary f eedwater pump to start? (1.0)

1. Safety Injection

2. Reactor Trip

3. Trip of one main feedwater pump

4. S/G Low Low Level (2/4) on any S/G

b. What is the emergency source of water to the Auxiliary Feed Pumps?

(.5)

c. When is this source automatically put into service? (.5)

OUESTION 6.05 (2.00)

A "PZR HI LEVEL DEV HTRS ON" alarm is actuated in the control room.

o. What does this tell you about Pressurizer level? (1.0)

b. Why are the Back Up Heaters energized? (1.0)

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

.. PAGE 7 6___ELON1_@y@l@d@_QE@lgN3 _CQN1BQ62_BNQ_lN@lBUDEN1611QN s

QUESTION 6.06 (1.50)

Concerning the Rod Control' and Rod Posiition Indication Systems, answer the following. (.25 each)

c. The ______________ generate the requi red sequence of current orders to the CRDM coils for rod motion.

b. A l ogic cabinet __________ f ailure occurs when any one of six DC power supplies in the logic cabinet f ails.

c. Normal speed for manual operati on of the shutdown banks i s ________

SPM.

d. If a rod does not move when demanded, the step counter will/will not (choose one) reflect its true position.

o. If a failure (DRP1) eliminates the data from one of two coil sets on a control rod position detector, the system places that rod's posi ti on indication to _____________.

f. A "RPI DEV/PR TILT" alarm is generated if a deviation of +/-

___________ steps occurs between any rod and its bank demand or between any two rods in the same bank.

DUESTION 6.07 (2.00)

Concerning the P-4 permissive:

a. How is P-4 generated? (.8)

b. List any four functions performed by P-4. (1.2)

QUESTION 6.08 (2.00)

Answer the f ollowing concerning RCP's:

a. During a plant start-up, when would you expect the P-8 permissive to energire? (.5)

b. What does the P-B permissive do? (.5) .

c. During a startup, RCP's A and B trip when you are at 7*/. power. Will the reactor trip? (.5) Why or why not? (.5)

QUESTION 6.09 ( .60)

The spent f uel pool occasionally will yield a low level alarm.

a. What is the probable cause of this alarm? (.20)

b. What are the normal and emergency sources of water for filling the spent fuel pool? (.40) l I

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

= . _ . . .

b___ELONI_gy@IEdg_QEg1GN2 _QONIBQ63_6NQ_INSIBUDENI@IlgN PAGE O QUESTION 6.10 (2.00)

Answer the f ollowing questions concerning load t hedding and sequencing during a LOCA.

What are the f unctions of load shedding and sequencing? (1.2) c.

b. List B loads sequenced by the LOCA sequencer. (Don't be train specific) ( . 8)

QUESTION 6.11 (2.00)

Answer the following concerning CCW system loads:

List five loads that are included in a safety loop. (.5) a.

b. How is the service loop supplied with water? (.5)

c. What does the " VALVES MISALIGNED" alarm tell you? (.5)

d. What is the source of emergency makeup to CCW and from what location can you perform the manipulations to supply that source to the CCW system? (.5)

QUESTION 6.12 (2.50)

Answer the f ollowing concerning RCP seals.

a. What is the approximate pressure drop across the No. 1 and No. 2 seal s? (.5)

b. What happens if a RCP is operated with less than a 200 psid pressure drop across its No. 1 seal? (.5)

c. In the event of a No. 1 seal failure, what must be done and why?

(1.0)

d. How much water is supplied by CVCS and where does it go? (.5)

QUESTION 6.13 (1.50)

Briefly describe the flow path to fill an accumulator from the Safety Injection System (including valves) .

DUESTION 6.14 (1.50)

I I

l Callaway Technical Specifications state that two of three boron injection flow paths shall be operable.

[ a. What are these flow paths? (.75)

b. Why are they necessary? (.75) l l

l l

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

l

Z___EBgggpUBES_ _NQBd9b2_@pNQBDSL,_gdEBGENgy_8ND PAGE 9 ESP 196991986_G9NIBQL QUESTION 7.01 (2.00)

What are the immediate operator actions for a loss of shutdown margin?

QUESTION 7.02 (1.50)

Answer the following concerning Critical Safety Functions:

a. Arrange the f ollowing conditions in order of priority. (1.0)

1. Orange condition in containment

2. Red condition in heat sink

3. Yellow condition in sub' criticality

4. Red condition in primary integrity.

b. Fill in the blanks (.5)

Status trees should be continuously scanned if a condition higher than _________ exists. If this condition does not exist, scanning frequency may be reduced to _____ ______ minutes unless some significant change in plant status occurs. ,

DUESTION 7.03 (1.00)

During a Safety Injection, reactor coolant pumps are tripped if two conditions are met. State these conditions.

QUESTION 7.04 (1.00)

a. At the Callaway Plant, how is a high radiation area defined? (.5)

b. A valve must be repaired in the letdown system in an area where a .25 curie source is located. The repairman must stand 6 feet from the valve to accomplish the job. The estimate of work time is 45 minutes. What dose should the repairman receive? (.5)

QUESTION 7.05 (2.50)

Answer the following concerning a S/G Tube Rupture: i

c. List four ways a ruptured S/G is identified. (1.0)

b. What six things must be done to isolate flow f rom a ruptured S/G7 (1.5)

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

1

2s__BBgCgggBgg_ _NgBdBL3_B@NQBd862_gdEBQENCy_ANQ PAGE 10 BB9196991C66_QgNIBQL QUESTION 7.06 (2.00)

Concerning a loss of charging, answer the f ollowing:

o.. List 4 annunciator alarms that may be symptoms of a loss of charging.

(1.0)

b. Describe what happens in the VCT should this loss of charging continue due to tripped charging pump? (include setpoints) (.5)

c. When would RCP's have to be tripped? (.5)

DUESTION 7.07 (2.00)

For the following symptoms, indicate whether they can occur for a major LOCA and/or a Steam Line Rupture by placing an X in the appropriate space.

(Each symptom could be applicable to either, neither, or both accidents.)

STEAM RUPTURE Symptom LOCA

o. Containment Spray Initiation ____ _____________

b. D S/G depressurized completely ____ , _____________

c. Condenser Air Removal - high radiation ____ _____________

d. High Containment Sump Level ____ _____________

e. High Containment Radiation ____ _____________

DUESTION 7.08 (2.50)

Concerning emergency procedure E-0, answer the followings

o. What are SI actuation criteria? (1.0)

b. Define an adverse containment. (.5)

c. How does an adverse containment affect the SI actuation criteria?

(1.0)

OUESTION 7.09 (2.00)

Answer the f ollowing concerning a reactor startup:

c. List the initial conditions required f or the f ollowinc if a reactor startup is to be performed. (1.0)

1. Source range channels operable

( 2. Power Range channels operable l

l

3. Steam Generator Level I 4. Pressurizer Level

5. RCS Temperature

b. During shutdown bank withdrawal, you notice that the source range counts more than double. What must you direct the RO to do? (.5)

c. What are the Callaway limits f or sustained and transient startup rates during rod withdrawal? (.5)

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

l

Z___E6pCEpgBES_ _NgBd@63_9pNgBd86,_EdgBGENgy_@ND PAGE 11 69D196991Ce6_CgNIBQL QUESTION 7.10 (1.50)

What are the immediate actions f or a Continuous Rod Withdrawal Accident?

QUESTION 7.11 (2.50)

You begin procedure E-O to evaluate a Loss of Coolant Accident and find that you can't trip the reactor. List 5 actions specified in FR-S.1 (Response to Nuclear Power Generation) to trip the reactor?

QUESTION 7.12 (1.50)

Answer the f ollowing concerning a loss of instrument air.

a. What automatic actions do you expect as air header-pressure decreases? (Includes setpoints) (.75)

b. What air pressure would you expect air operated control valves to star' drifting? (.25)

c. What is the fail position of the f ollowing valves? (.5)

Valve

1. Charging header Flow Control (HCV-121)

2. Steam Dumps to Condenser

3. Condensate Recirc Valves

4. PZR spray valves

5. Letdow, isolation valves l

OUESTION 7.13 (1.50) l The turbine trips due to loss of condenser vacuum:

o. List the immediate operator actions. (1.2) l

, b. What is used to maintain no-load steam pressure? (0.3) l DUESTION 7.14 (1.50) l l

You have experienced a loss of letdown while operating with the positive displacement charging pump.

l o. How will the positive displacement charging pump control system l respond? (.5)

b. What happens to VCT level? (.5)

c. Procedure has you establish excess letdown flow. Briefly describe how this i s done. (.5)

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

i

~ PAGE 12 Hz__0Dd1NISIB9Ilyg_BBQCEDUBES 3 _CgNplIlgNH3_9Np_Libil@IlgNS QUESTION 8.01 (1.50)

Fill in the blanks f or the f ollowing table concerning minimum shif t mcnning requirements for Mode 1 operation according to Procedure APA-ZZ-OOO10.

Shift Supervisor a) _______

Operating Supervisor b) _______

Unit Reactor Operator c) _______

Equipment Operator d) _______

Shift Technical Advisor e) _______

Shift Advisor f) _______

OUESTION B.02 (1.50)

Fill in the blanks for the f ollowing table concerning operational modes. T-ave Mode Reactivity Condition %_ Rated _ Power

>5% 1 350 F

1. Power Operation 1 0.99 d) 2 350 F

2. Startup 1 0.99 _____

O 350 F

3. Hot Standby b) 2

< O.99 0 e) _________

4. Hot Shutdown

< 0.99 0 f)

5. a) __________

c) O < 140 F

6. Refueling ______

QUESTION 8.03 (1.50)

a. Who may authorize an EMERGENCY CONTAINMENT ENTRY?

b. If only two people enter containment, what should at least one of the persons' area of expertise be?

c. What are the requirements for wearing Self Contained Breathing Apparatus (SCBA) during the emergency entry?

QUESTION B.04 (1.00)

Concerning classifications of emergencies (Procedure EIP-ZZ-OO101), answer the following:

a. Who is responsible for the initial implementation of the procedures?

(0.5)

b. Within how much time should initial classification take place after recognition of initiating conditions? (0.5)

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

• • • • • )

i

~

9t__9Db1NigIB@IlyE_BBgCEgyBgg,_QQNp111gNp2_@Np_LidlI@IlgNS PAGE -13 I

l i

QUESTION 8.05 (2.00)

Fill in the blanks for the f ollowing statements:

o. Reactor Coolant System leakage through a steam ger.erator to the Secondary Coolant System is known as _________ leakage.

b. Technical Specifications limits primary to secondary leakage to

_______ 9pm total through all S/Gs not i sol ated from the RCS ...

c. ... and _____ gallons per day through any one S/G.

d. Performance of a Reactor Coolant System water inventory balance is required at least once every ____ hours.

DUESTION O.06 (2.50)

a. Assume that it is 0300 on 12/9/86 and the reacter is presently at 45%

power. Considering the Delta-I target band history listed below, calculate the associated Delta-1 penalties. (1.5)

Date Time (Out) Time (In) Power (%) Penalty (Min)

1. 12/8/86 0300 0318 85 ____________

2. 12/8/86 1557 1633 65 _____ _ _ ____

3. 12/9/86 0138 0300 45 ____________

b. When may power be increased above 50%? (1.0)

OUESTION 8.07 (2.00)

The concentration of the boric acid sol uti on in the Boric Acid Storage System must be verified once a week in accordance with Technical Specification 4.1.2.6. The chemist sampled the boron concentration on the following schedule at 1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br /> each time.

Nov 10 --- Nov 18 --- Nov 26 Dec 4 Explain why surveillance time interval requirements WERE or WERE NOT exceeded on:

a. Nov 26

b. Dec 4

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

PAGE 14 B___BDDINiglB911Mg_EBQQ[QQBgg3_QQNQlligN@3_9NQ_LIDlIBIlgNE l

QUESTION B.08 (1.00)

Fill in the blanks for the f ollowing ref ueling situations.

o. The reactor shall be subcritical for at least _______ hours before movement of irradiated fuel in the reactor vessel.

b. Direct communications shall be maintained between the _________ and personnel at the refueling station during core alterations.

c. At least ________ ft of water shall be maintained over the top of the reactor vessel flange during movement of irradiated fuel assemblies in containment.

d. The refueling canal shall be borated to a concentration of _______

ppm or greater in Mode 6 operations.

QUESTION 8.09 (1.00)

o. Personnel on the Callaway plant staff should not be permitted to:

(Fill in the blanks) (0.5)

1. Work more than ______ hours straight .

2. Work more than ______ hours in any 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period

b. Who may authorize deviations to the above restrictions? (0. 6)

QUESTION 8.10 (2.00)

Answer the f ollowing questions concerning containment spray.

c. What is the Technical Specification limit for NaOH volume and concentration in the spray additive system? (0.5)

b. Why are these limits needed? (1.5) i QUESTION B.11 (2.50)

! Technical Speci f i cati on 3/ 4. 5. 2, ECCS Subsystems with T-ave > 350 degrees Fahrenheit, limiting conditions for operation requires two independent ECCS subsystems to be OPERABLE in Modes 1, 2, and 3. What does Technical Specifications require an OPERABLE ECCS SUBSYSTEM to be comprised of?

l l

l l

l

• • • • • )

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

Ez __8Dd1NISIBSIlyE _BBQQ E DUB E @2_QQNp111 gNg, _9Np_ QldlI@llgN_5 PAGE 15 QUESTION O.12 (1.50)

o. What are the steady state and transient limits f or dissolved oxygen in the Reactor Coolant System? (0.5)

b. How long do you have to restore the dissolved oxygen concentration back to within violated steady state limits before you must shutdown?

(0.25)

c. What is the normal sample frequency for dissolved oxygen? (0.26)

What is the design basis for the oxygen limit? (0.5) d.

QUESTION B.13 (1.50)

Answer the following questions:

a. Health Physics personnel have the specific authority to stop work whenever RWP radiological conditions change. (True or False) (0.5)

b. Plant administrative radiation exposure limits may be exceeded with authorization of the Shift Supervisor. (True or False) (0.5) individuals may be allowed

c. In Modecontainment inside 1 operations, ataany maximum of ____(Fill in the blank) one time.

DUESTION O.14 (1.00)

If an individual performing some activity cannot follow or believes the procedure governing that activity should not be followed as written, what chould that individual do?

DUESTION 8.15 (1.00)

Answer the f ollowing concerning the Fire Brigade.

a. Who functions as the Fire Brigade leader? (0.5)

b. Who else is on the Fire Brigade? (0.5)

QUESTION 8.16 (1.50)

If a motor operated valve must be tagged with caution tags for repairs to o component, what locations must tags be hung?

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

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

3 3 PAGE 16 Et__IHEQBY_QE_ NUCLE @B_EQWE8_E(@N1_QEEB@llQN _E(Q1QS _@NQ IMEBdQDYN@ MIGS

-86/12/09-SUNDERLAND, P.

ANSWERS -- CALLAWAY ANSWER 5.01 (1.00)

Pressuri:er level will change with reactor coolant temperature due to the volumetric change of the coolant. Therefore, mass transients are minimized and the amount of water to be reprocessed is reduced.

REFERENCE Student Handout - RCS Instrumentation and PZR Level and Pressure - P.5 ANSWER 5.02 (1.50)

o. decreases

b. increases

c. stays the same

d. decreases ,

9. increases

f. stays the same REFERENCE Westinghouse Thermal Hydraulic Principles, P.4-62 Student Handout - RCS Instrumentation and PZR Level & Pressure SNP-RC-13 Student Handout - MFW, P.19 ANSWER 5.03 (2.00)

o. (0.5 pts each)

1. Acceptable power distribution limits are maintained.

2. Minimum shutdown margin is maintained.

3. The potential effects of rod misalignment on associated accident analyses are limited.

b. (.25 pts each)

1. Delta T

2. T-average REFERENCE Student Handout - Rod Control, P.5, SNP-CR-O Technical Specifications, P. B, 3/4, 1-3 L

Dz__ISE96Y_QE_ NUCLE 66_EQWEB_E6@N1_QEEB@IlQN 3 _EbQlgSi _6BD PAGE 17 IHEBdQQXN@d1CS 8

ANSWERS -- CALLAWAY -86/12/09-SUNDERLAND, P.

ANSWER 5.04 (2.50)

c. OPTR shall be the ratio of the maximum upper excore detector calibrated out to the average of the upper excore detector calibrated outputs (0.5), or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs (0.5), whichever is greater (0.5).

b. (0.5 pts each)

1. DNB

2. Exceeding maximum linear heat generation rates.

REFERENCE Technical Specifications, P. 1-5 and P. B, 3/4, 2-4 ANSWER 5.05 (2.00) .

a. decreases (0.25) due to cooler water in the S/G cooling down the reactor coolant (0.25).

b. increases (0.25) due to moderator temperature coefficient (0.25).

c. decreases (0.25) because adding reactivity due to the temperature decrease is reactivity that does not have to be added by rods (0.25).

d. decreases (0.25) because cooler water is denser and the probability of neutrons reaching a rod decreases. (0.25)

REFERENCE Westinghouse Reactor Core Control, P. 7-27, 7-32, 6-22.

ANSWER 5.06 (2.00)

a. cavitation (0.5)

b. (0.5 pts each, any two)

1. Reduce minimum required NPSH (by reducing pump flow).

2. Increase available NPSH (by reducing system flow).

3. Increase available NPSH (by increasing suction pressure).

c. Restores source of required NPSH (0.5)

REFERENCE Westinghouse Thermal - Hydraulic Principles, Ch. 10 f

l

~

D___ISE98Y_9E_NWGLE9B_E9WE6_EL@NI_QEg69119N 2_E6919@2_@Up PAGE 18 INESd92YNed1GS

- ANSWERS -- CALLAWAY -86/12/09-SUNDERLAND, P.

ANSWER 5.07 (2.00)

o. Any four at 0.25 pts each

1. Rod Height

2. Moderator density variations

3. Power l evel

4. Fission Product Poison concentration

5. Fuel depletion variations along the length of the fuel rod.

b. Axial Flux Difference is the difference between the upper and lower detector signals. (0.5)

c. (0.25 pts each)

1. Iower half

2. borate (assuming rods are in auto)

REFERENCE W3stinghouse Reactor Core Control, P. B-25 through B-32.

ANSWER 5.08 (1.50)

A large volume of water enters the pressurizer due to an increase in T-ave. (0.3) The steam is compressed and system pressure may increase cbove the programmed band (0.3). Sprays initiate, reducing the pressure cpike and hasten condensation. (0.3) The water is cooled by the insurge cnd will cause saturation at a l ower pressure. (0.3) Heaters will heat the pressurizer back to original saturation conditions in the programmed band. (0.3)

REFERENCE W3stinghouse Thermal-Hydraulic Principles, P. 12-55.

i i

e l

l I

i

\ ' t, t m

T wJ,

,.. PAGE 19

p. THEORY OF NUCLEAR POWER PLANT OPERATIDt4 2 _FLUlp52_AND ISERDggyNAdlCS s

-86,/12/09-SUNDERLAND, P.

ANSWERS -- CALLAWAY s .

ANSWER 5.09 (2.50)

c. (0.25 pts each)

1. Fission Products

2. Iodine decay

b. (0.25 pts each)

1. Burnout

2. Decay

c. Xenon increases initially (0.2) due to no burnout (0.1) and decay of iodine is greater than decay of xenon (0.1). Xenon reaches its peak as the rates of decay of iodine and xenon equalize (0.2) Now xenon decay occurs at a greater rate than iodine decay (0.1) and xenon concentration decreases (0.2). Finally the iodine is gone and xenon concentration decreases at a rate proportional to its concentration (0.1).

d. Xenon concentration is decreasing at a high rate, adding positive 3

reactivity (0.25) so the rods are at a lower positi,on to achieve criticality (0.25).

' REFERENCE Westinghcuse Reactor Core Control , P. 4-11 through 4-22.

t i

ANSWER 5.10 (2.00)

a. (any 2 at 0.5 pts each)

1. produces hydrogen

2. produces heat (exothermic)

3. causes Zr embrittlement ~

4. release FPP to coolant

b. (0.5 pts)

Limit peak clad temps to <22OOoF.

REFERENCE Westinghouse Mitigating Core Damage, P. B-12.

o

Ez__IBE98Y_DE_NW96E96_EDWEB_EL9NI_QEg6611QN 3 _ELyJpS2 _9ND PAGE 20 INEBD99YN9 digs ANSWERS -- CALLAWAY -86/12/09-SUNDERLAND, P.

s >

ANSWER 5.11 (2.00)

Use figure to calculate:

100% total power defect -1817 pcm 70% total power defect - (-1283)pcm

~

~~ 554 pcm~'

(0.5)

From baron worth vs. boron concentration .

Diff. Boron Worth = -9.45 ( + .1) (0.5)

Dividing, we calculate as follows (give credit here for calculation and boration for numbers attained above).

600 ppm + (-534 pcm/-9.45 pcm/ ppm) = 600 + 56.5 (plus or minus 3)=656.5 (0.5 for calculation) and (0.5 f or boration)

REFERENCE Plant Curve Book - Curves attached f

ANSWER 5.12 (2.00)

1. When the steam flow out of each S/G stops, S/G temperature increases rapidly to match RCS temperature (0.5), causing S/G pressure to increase rapidly (0.5), compressing / collapsing the steam bubbles in the riser (0.5). As a result, the S/G 1evels rapidly drop or

" shrink". (0.5) or (either answer acceptable)

2. The loss of steam flow and recirculation flow through the S/G moisture separators removes the recirculation flow delta-P (0.5),

removing the need for recirculation flow driving dead (0.5), allowing the downcomer level to drop (0.5), and equalize with the lower level in the riser. (0.5)

REFERENCE Westinghouse Thermal Hydraulic Principles, P. 12-53.

a D___IHE96Y_DE_NyC(g@B_EQWEB_EL@NI_QEgB@IlgN 2 _ELylpS 2 _@NQ PAGE 21 IHEBdggyN@dlCS

. ANSWERS -- CALLAWAY -86/12/09-SUNDERLAND, P.

ANSWER 5.13 (1.00)

(.5 each)

a. Current = 40 amps x (100/50) E3 = 320 amps

b. Flow = 50 gpm x (100/50) = 100 gpm REFERENCE Wsstinghbouse Thermal-Hydraulic Principles, P. 10-36.

ANSWER 5.14 (1.00)

c. (1.0)

REFERENCE Westinghouse Reactor Core Control, P. 9-10 ,

PAGE 22 6 t__ELONI_SygIgD5_pgSIGN _CgNI696,_@ND_INSIBUDENI@IlgN 3

-86/12/09-SUNDERLAND, P.

ANSWERS -- CALLAWAY ANSWER 6.01 (2.00)

c. Hi-Hi S/G water level on any S/G (.5)

b. Safety Injection signal (.5)

c. All condensate pumps trip (.5)

d. High feed pump discharge pressure (.5)

REFERENCE Student Handouts - Main Feedwater System, P. 15.

ANSWER 6.02 (1.00)

(.2 each)

o. PA02

b. NB01

c. NKO1 (f ul l credit) or Battery Charger NK21/125 V Battery NK11 (helf credit each)

d. NB02

o. NB01 or PG 21 REFERENCE Student Handbook - Service Power (SE), Safeguards Power SA)

c. TP 2 (SE)

b. P. 4 (SA)

c. TP 3 (SA)

d. P. 4 (SA)

9. TP 1 (SA)

ANSWER 6.03 (2.40)

Power (.1) +4% in 2 sec. (.1)

(+) Hi Flux Rate (.2) (.1)

(-) Hi Flux Rate (.2) Power (.1) -4% in 2 sec.

High Flux (Hi) (.2) Power (.1) 109% (.1)

High Flux (Lo) (.2) Power (.1) 25% (.1)

High Flux (Lo) (.2) Intermediate (.1) 25% current equivalent (.1)

High Flux (.2) Source (.1) 10 E 5 cps (.1)

REFERENCE Student Handouts - NIS - Instrument Block Diagrams

- Rod Control TP-3

6___ELON1_Ey@l@d@_Q@@l@N3_ggNIBQ63_9NQ_lN@lBUdENI@l1QN PAGE 23 ANSWERS -- CALLAWAY -86/12/09-SUNDERLAND, P.

ANSWER 6.04 (2.00)

o. (.25 each)

1. yes

2. no

3. no

4. yes

b. Essential Service Water (.5)

c. AFAS Signal (.25) / Low suction pressure on pump 2/3 (.25)

REFERENCE Student Handouts - AFW, P.7 ANSWER 6.05 (2.00)

c. Level is 5% above programmed level. .

b. Heats the (subcooled) water entering the pressurizer from the RCS, minimizing possible pressure transients.

REFERENCE Student Handout - RCS Instruments and PZR Level anc Pressure

c. P. 27, SNP-RC-13

b. P. 29 ANSWER 6.06 (1.50) ,

c. slave cyclers

b. non-urgent

c. 64

d. will not

a. half accuracy

f. 12 REFERENCE Student Handouts - Rod Control a) p. 17, b) P. 17, c) P. 8, d) P. 10, e) P. 13, f) P. 24

A___EL9NI_fYSIgdp_pESIGN 3_gpNIBQL,_9Np_JNSIBQUENJBIlgN PAGE 24 ANSWERS -- CALLAWAY -86/12/09-SUNDERLAND, P.

ANSWER 6.07 (2.00)

o. Whenever a reactor trip breaker and its bypass breaker are both open

(.4 for stating " reactor trip")

b. (any f our of the f ollowing) (.3 each)

1. Trips the main turbine

2. Init2 ates f eedline isolation below Low-Tave

3. Prevents automatic re-actuation of SI after a manual reset

4. Transfer the steam dumps from the load reject controller to the plant controller

5. Prevents opening the feedwater isolation valves af ter closure due to SI signal or a high-high S/G 1evel.

REFERENCE Student Handouts - Reactor Protection, P. 17.

ANSWER 6.08 (2.00)

c. 48% power

b. Automatically unblocks low flow in one loop trip

c. No (.5), P-7 (.2) is deenergized which blocks low flow in more than one loop (.1), RCP under frequency (.1) and RCP UV trips (.1).

REFERENCE Student Handouts, Reactor Protection, TP-4.

ANSWER 6.09 ( .60)

a. (.20) Evaporation

b. (.20) N-Make-up Water System

(.20) E-Essential Service Water REFERENCE Student Handouts - Fuel Pool Cooling and Cleanup, P3, TP1

PAGE 25 6t__PLeNI_gygIgdg_gEg]GN _CgNIBgL,_@Np_JNSIBydENIeIJgN 3

-S6/12/09-SUNDERLAND, P. l ANSWERS -- CALLAWAY

)

ANSWER 6.10 (2.00) l

c. 1. (.6) Load shed signal sheds all non-safety loads from the safeguards buses.

2. (.6) The LOCA sequencer ensures that all loads required to mitigate a severe accident are sequentially started.

b. Any 8 at .1 pt each CCA S1 Pumps RHR Pumps CCW Pumps Containment Spray Pumps ESW Pumps AFW Pumps Control Room AC Units Class IE Electrical Room AC Units Containment Coolers REFERENCE Student Handouts - Safeguards Power, P. 5-7.

ANSWER 6.11 (2.00)

a. Any 5 ( .1 each)

1. SI Pump (oil coolers).

2. Cent. Charging pump (oil coolers).

3. Spent Fuel Pool Heat Exchangers.

4. RHR pump (seal c ool er s ) .

5. Residual Heat Exchangers

6. Post Accident Sample Station Sample coolers.

b. Connected to only one of the safety loops

c. Both Safety loops become cross-connected through the non-saf ety l oop

d. Essential Service Water (.25), Auto valves on MCB (.25) l REFERENCE Student Handouts - CCW P. 3, 6 l

(

6t__Eb8N1_@Y@lgd@_QE@l@N2_QQNIBQL,_@NQ_lN@lBQUENI@IlQN PAGE 2o ANSWERS -- CALLAWAY -86/12/09-SUNDERLAND, P.

A ANSWER 6.12 (2.50)

c. No. 1 - 2200 psid (+ 200) (.25)

No. 2- 35 psid (+ 15) (.25)

b. (Hydrostatic forces are insufficient to float the seal.) Contact occuru, causing damage to the seal (.5)

c. Shut HV-8141 (Seal leakoff isolation valve) (.5) Isolates the leakoff path which causes No. 2 seal to be the primary pressure drop for the seal system.

d. O gpm (.3), 5 gpm to the RCS (.1) and 3 gpm through the No. 1 seal

(.1)

REFERENCE Student Handbook - RCS, P. B-11 ANSWER 6.13 (1.50) ,

To fill the accumulators, an S1 pump (.3) is used to take a suction from the RWST (.3). Discharge flows through a common fill valve (.3) to the cccumulator fill header which splits into separate lines (.3) with an icolation valve (.3) for each accumulator.

REFERENCE Student Handout - Accumulators, P. 2.

ANSWER 6.14 (1.50)

o. (.25 each)

1. Boric Acid Storage System via BAT pump.

2. RWST via centrifugal charging pump to the RCS

3. RWST via other centrifugal charging pump to the RCS/

b. (.75)

This ensures that enough boron is available to provide a SDM of 1.3%

delta k/L if a single failure occurs.

REFERENCE Callaway Technical Specification, P. 3/4 1-0, B 3/4 1-2

Zz__EB9GE99 BEE _ _N9Bd862_BgNQBd863_EdgBggNgy_9ND PAGE 27 B99196991G96_G991896 ANSWERS -- CALLAWAY -86/12/09-SUNDERLAND, P.

ANSWER 7.01 (2.00)

1. Open the Emergency Borate to charging pump suction valve BG-HV-8104.

2. Start both BA transfer pumps.

3. Verify boric acid flow on BG-F1-183A

4. Place the RCS M/U CTRL, BG-HIS-26 in stop.

5. To maximize water turnover rate, ensure a cent. charging pump is running and establish 120 gpm letdown.

REFERENCE OTO-ZZ-OOOO3 ANSWER 7.02 (1.50)

c. (.25 each) 2 1 -3

b. Yellow (.25) 10-20 (.25)

REFERENCE Procedure CSF-1, p. 3 of 4 ANSWER 7.03 (1.00)

1. One charging pump or one SI pump running (0.5)

2. RCS pressure < 1400 psig (0,5)

REFERENCE Procedure E-0, p. 11 of 18 ANSWER 7.04 (1.00)

c. Any area accessible to personnel where an individual could receive a whole body dose in excess of 100 mrem but less than 1000 mrem in any one hour.

b. Assume pt. source Use (DR1 ) (R1 ) E2= (DR2) (R2) E2 Assume at 1 meter, dose rate (DR1) = .25 R/hr

(.25 R/hr) ( (1E2) / (2E2) )x (3/4 hrs)= 47(+/- 5) mrem

PAGE 23 Zz__EB9CEDWBgS_:_Ng6d@L2_9pNg65962_EDEBGENCy_9ND B99196991G96_QgN1896

• -B6/12/09-SUNDERLAND, P.

ANSWERS -- CALLAWAY REFERENCE Procedure APA-27-OOl60, p. 6 ANSWER 7.05 (2.50)

c. (Any 4 at .25 each)

1. Unexpected increase in S/G 1evel

2. High radiation from any S/G sample

3. High steamline radiation

4. High radiation in S/G blowdown line

5. High radiation from condenser off gas

b. (.25 each)

1. Adjust ruptured S/G atmospheric steam dump controller setpoint to 1125 psig

2. Close ruptured S/G MSIV's and bypasses

3. Check ruptured S/G atmospheric steam dump closed

4. Close steam supply from ruptured S/G to turbine AFW pump

5. Ensure blowdown isolation valves f or ruptured S/G are closed

6. Ensure chemical injection valves for ruptured S/G are closed REFERENCE Procedure E-3, p. 6, 7 of 31 ANSWER 7.06 (2.00)

(a) Any 4 (.25 each)

1. Charging Pump Trouble

2. Ltdn Regen Ht Temp Hi

3. Chg Line Flow Hi-Lo

4. VCT level Hi-Lo

5. PZR Lo Level Deviation

6. Seal Inj to RCP Flow Lo

b. As VCT level increases to 70%, the letdown divert valve (BG-LCV-112A) starts to open and directs flow to the RHT.

c. (Any 5 at .1 pts each)

1. No. 1 seal outlet temp > 235 degrees-F

2. Seal injection temp > 150 degrees-F

3. Thermal barrier cooling water inlet temp > 105 degrees-F

4. Pump lower bearing > 225 degrees-F

5. Motor hearing cooling water temp > 120 degrees-F'

6. Motor bearing temp > 195 degrees-F REFERENCE Procedure OTO-BG-OOOO2, p. 1-3

Zz__EBQGEDW6EE_ _NQ6d@(3_@ENQ656(a_EDE6@ENgy_@NQ PAGE 29 B6019690lGOL_QQNI6QL ANSWERS -- CALLAWAY -86/12/09-SUNDERLAND, P.

ANSWER 7.07 (2.00)

(.2 for each blank correct)

LOCA Stm. Rupt O. X X

b. X c.

d. X X

o. X REFERENCE Procedure E-1, p. 1, 4, 5, 8 ANSWER 7.08 (2.50)

o. Either

1. Loss of subcooling (per RCS subcooling curves) (.5)

2. PZR level cannot be maintained > 5% (.5)

b. Containment temperature ,>,160 degrees F. (.5)

c. 1. RCS subcooling curves are more restrictive (.5)

2. PZR level cannot be maintained > 28%. (.5)

REFERENCE

o. E-0, p. 1 of 1 (Foldout)

b. E-0, p. 3 or 14

c. E-0, P. 1 of 1 (Fol dout )

ANSWER 7.09 (2.00)

a. (.2 each)

1. 2

2. 3

3. 50 plus or minus 5%

4. 25 plus or minus 2%

5. 552 - 558 degrees F

b. Suspend withdrawal (.3) and evaluate core reactivity (.1) and reason for the change (.1).

c. (.25 each)

Sustained = 1 DPM Transient = 1.5 DPM REFERENCE Procedure OTG-22-OOOO2, p. 4, 6, 10

' PAGE 30 Zz__EB9CED9 BEE _ _NQBD9(2_8pNQBD962_EdgBQENQY_9Np 899196991996_G9NI696

-86/12/09-SUNDERLAND, P.

. ANSWERS -- CALLAWAY ANSWER 7.10 (1.50)

1. Place the rod bank selector switch in MAN

2. Insert the control rods as required to match Tave and Tref.

3. If rods continue to withdraw, trip the reactor.

(0.5 each)

REFERENCE OTO-SF-OOOO2, p. 2 ANSWER 7.11 (2.50)

(0.5 each)

1. Manually trip the reactor

2. Manually trip the reactor from the BOP console *

3. Trip supply breakers to LC PG19 and LC PG20

4. Manually insert RCCA's

5. Dispatch operator to locally trip the reactor REFERENCE Procedure FR-S.1 ANSWER 7.12 (1.50)

a. (.25 each)

1. Backup Air Compressors start (115 and 110#)

2. Service Air Header Isolation Valve shuts (110#)

3. All compressors fail safe at 105#

b. 70# plus or minus 10

c. 1. Open

2. Closed

3. Open

4. Closed

5. Cl osed (0.5 each)

REFERENCE Procedure OTO-MA-OOOO1, p. 2, 3, 5, 6

' PAGE 31 Z___PBQgEggBES_r_NQBd@63_6ENQRd@62_EdEBGEN[Y_9ND 89D196991996_G9NIBQ6

- ANSWERS -- CALLAWAY -86/12/09-SUNDERLAND, P.

ANSWER 7.13 (1.50)

c. (0.3 each)

1. Ensure turbine stop, control, and combined intercept valves are closed.

2. Ensure switchyard breakers V53 and V55 open about 30 sec after trip.

3. Ensure motor operated disconnects V53b and V55a open f ollowing V53 and V55 trip

4. Verif y T-ave is approaching no load value, if in manual rod control, adjust rods to achieve no load value.

b. Atmospheric steam dumps (0.3)

REFERENCE Procedure DTO-AC-OO901, p. 1, 2 ANSWER 7.14 (1.50)

a. Speed decreases to 47 gpm to maintain seal injection.

b. VCT level decreases due to no letdown flow. Assuming auto make-up, level will rise again and cycle as necessary to provide seal water flow,

c. (.25 each)

1. Opening the Reactor Coolant to Excess letdown heat exchanger valves. (HV-8153A and B, HV-8154 A and B)

2. Throttle open HCV-123 to establish excess letdown flow.

REFERENCE Procedure DTO-BG-OOOO2, p. 2, 3

PAGE 32 Ez__BDd1N1QlE911yE_ESQCEQQEEQi_QQNQlI1QNQi_@NQ_(ldll@llQNQ

-86/12/09-SUNDERLAND, P.

ANSWERS -- CALLAWAY ANSWER B.01 (1.50) 01 b.1 c.2 d.4 0.1

'f .1 or O (Position no longer required due to > one year of operation)

REFERENCE Procedur e APA-2 2-OOO10, p. 17 ANSWER 8.02 (1.50)

(0.25 each) -

c. Cold Shutdown

b. < 0.99

c. f 0.95

d. $ 5%

o. 350 F > T-ave > 200 F

f. < 200 F REFERENCE Technical Specifications p. 1-9 ANSWER 8.03 (1.50)

a. Either the Shift Supervisor or the EDO for full credit (0.5)

b. Health Fhysics (technician) (0.5)

c. SCBA must be worn until the atmosphere is tested and evaluated to be safe. (0.5)

REFERENCE Procedure CDP-22-OOO19 p. 5-6 l

ANSWER 8.04 (1.00)

a. Shift Supervisor

b. 15 minutes REFERENCE Procedure EIP-22-OO101, p. 1, 2, 1 of Attachment 1 l

Ez__6DDINIEIBBIlYE_EB99EDWBES,_QQNp111pNg3_9Np_(1611911gNE PAGE 33 ANSWERS -- CALLAWAY -85/12/09-SUNDERLAND, P.

ANSWER B.05 (2.00)

c. Identified ,

b. 1

c. 500

d. 72 REFERENCE Tcchnical Specifications p. 1-3, 3/4 4-19, 3/4 4-20 ANSWER 8.06 (2.50)

c. 1. 18 minutes.

2. 36 minutes.

3. 41 minutes. (0.5 each)

b. Must have < 60 min of penalty time in previous 24 hpurs before increasing to > 50% power. Should be able to increase at 1615 on 12/09/86.

REFERENCE Tcchnical Specifications p. 3/4 2-2.

ANSWER B.07 (2.00)

a. Not exceeded (0.5). Eight days does not e::ceed 1.25 times the specified interval. (0.5)

b. Exceeded (0.5). The three consecutive intervals exceed 3.25 times the specified interval. (0.5)

REFERENCE Tcchnical Specifications, p. 3/4 1-12 ANSWER B.08 (1.00)

a. 100

b. control room

c. 23

d. 2000 (0.25 each)

REFERENCE Technical Specifications p. 3/4 9-1, 9-3, 9-5, 9-12 4

- 9 _~m_- _ , _ - - _ _ - . _ _ _ , - . e -. , . _ _ ~r , , - - . . . ,

l U___9901Nig16811Mg_66QQgQu$gg3_QQNQlligNg2_@NQ,gid11gljgNE PAGE 30 ANSWERS -- CALLAWAY -86/12/09-SUNDERLAND, P.

s ANSWER B.09 (1.00)

o. (0.25 each)

1. 16

2. 24

b. Callaway Plant Manager (0.25) or Emergency Duty Officer (0.25)

REFERENCE Procedure APA-2Z-OO13, p. 2 ANSWER 8.10 (2.00)

c. 4340-4540 gallons (0.25) .

28-31 wt % (0.25)

b. Ensure a pH value between 8.5 and 11.0 for the sol'ution recirculated within the containment following a LOCA (0.5). This pH band minimizes the evolution of iodine (0.5) and minimizes chloride and caustic stress corrosion on mechanical systems and components. (0.5)

REFERENCE Technical Specifications p. 3/4 6-14 and B 3/4 6-3 ANSWER B.11 (2.50)

s. One operable Centrifugal Charging Pump (0.5)

b. One operable Safety Injection Pump (0.5)

c. One operable RHR Heat E>: changer (0.5)

d. One operable RHR Pump (0.5)

o. An operable flowpath (0.3) capable of taking suction from the RWST (0.1) and automatically transferring suction to the containment sump during recirculation (0.1).

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

B___99DINJ@lB@llyE_EBQQEQUBE@2_ggNQllipN@2_@NQ_LidlI@IlQNS PAGE 35 1 ,

t.

ANSWERS -- CALLAWAY -86/12/09-SUNDERLAND, P.

o ANSWER 8.12 (1.50)

o. < 0.10 ppm steady state state (0.25), < 1.00 ppm transient (0.25)

b. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (0.25)

c. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (0.25)

d. Reduce corrosion (0.25) and reduces potential of RCS leakage or failure due to stress corrosion (0.25)

REFERENCE Technical Specification p. 3/4 4-22, 23 and B 3/4 4-5 ANSWER 8.13 (1.50)

s. True

b. False

c. 20 REFERENCE Procedure APA-ZZ-OOl60, p. 10, 13, 17 ANSWER 8.14 (1.00)

1. Place the system / components in a safe and stable condition. (0.5)

2. Notify the responsible supervisor. (0.5)

REFERENCE APA-ZZ-OO100, p. 6 ANSWER B.15 (1.00)

a. Operating Supervisor (0.5)

b. Two equipment operators (0.25)

Two Red Chem helpers (0.25)

REFERENCE Procedure APA-ZZ-OO743, p. 1, 10

PAGE 36 Ez__eDUlN1EIE@llME_E6QQEDQBEEx_QQNQlligNgg_@NQ_(idil@llQNE

-86/12/09-SUNDERLAND, P.

-ANSWERS -- CALLAWAY ANSWER B.16 (1.50)

(0.5 each)

1. Handwheel

2. Supply breaker

3. Control switch (es)

REFERENCE Procedure APA-ZZ-OO310, p. 11

MASTER U. S. NUCLEAR REGULATORY COMMISSION COPY o

REACTOR OPERATOR LICENSE EXAMINATION FACILITY: CALLAWAY REACTOR TYPE: PWR-WEC4 DATE ADMINISTERED: 86/12/09 EXAMINER: WEALE. G.

CANDIDATE:

INSTRUCTIONS TO CANDIDATE:

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

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

% OF CATEGORY  % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 25.00 25.00 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 25.00 25.00 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 25.00 25.00 3. INSTRUMENTS AND CONTROLS 25.00 25.00 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 100.00 Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.

Candidate's Signature

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

l 1.s NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS

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

1. Cheating on the examination means an automatic denial of your application end 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 cypropriate, start each category on a Rag page, write onlY 2R 2RR side i 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.

4

11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table, i 12. Use abbreviations only if they are commonly used in facility literature.

l

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 i

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.

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 i

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

been completed.

l

18. When you complete your examination, you shall:

c. Assemble your examination as follows:

(1) Exam questions on top.

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

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

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

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

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

I I

l t

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 2

. THRMODYNAMICS. MAT TRANSFER AND FLUID FLOW 4

QUESTION 1.01 ( .50) )

Which of the following combinations describes the changes that take place in the steam between the inlet and outlet of the main turbine? l

1) Enthalpy decreases, entropy decreases, quality decreases.

2) Enthalpy constant, entropy constant, specific volume increases. l

3) Enthalpy constant, entropy decreases, quality decreases.

4) Enthalpy decreases, entropy increases, specific volume increases.

QUESTION 1.02 (2.00)

Compare the Estimated Critical Position (ECP) calculated for a startup, expected to be performed 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after a trip from steady 1004 power, to the ACTUAL critical rod position (ACP) if the following events / conditions occur prior to reaching criticality. .

Indicate whether the ECP is HIGHER than, LOWER than, or the SAME as the ACP rod position AND briefly EXPLAIN the reason for your answer.

Consider each occurrence separately.

o. The fourth RCP is started two minutes prior to criticality. (0.50)

b. The startup is delayed until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the trip. (0.50)

c. The steam dump pressure setpoint is increased to a value just below the Steam Generator PORV setpoint about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> before criticality. (0.50)

d. All Steam Generator levels are rapidly being raised by 5% just prior to criticality. (0.50)

\

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

__ __~ -

PAGE 3

1. PRINCIPTRR OF NUCr.rAR POWER PLANT OPERATION.

. THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW s

QUESTION 1.03 ( .50)

Prior to reaching criticality during a reactor startup, an initial rsactivity addition causes count rate to increase from 20 cps to 40 cps.

A second reactivity addition causes count rate to increase from 40 cps to 80 cps. Which of the following statements is CORRECT 7

c. The first reactivty addition was larger.

b. The second reactivity addition was larger.

c. The first and second reactivity additions were equal.

d. There is not enough data given to determine the relationship of the reactivity additions.

QUESTION 1.04 (1.00)

TRUE or FALSE 7 t

c. As Keff approaches unity, the fractional changes in neutron level resulting from identical changes in Keff will be smaller. (0.5)

b. With Keff greater than unity, a constant positive startup rate i

will occur only if net REACTIVITY is NOT changing. (0.5)

QUESTION 1.05 (2.00)

o. List the THREE (3) largest contributors to total power coefficient in order of INCREASING magnitude. (1.5)

b. How does total power coefficient vary as the core ages? (0.5) i f

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

1. PRINCIPTFR OF NUCtFAR POWER PLANT OPERATION. PAGE 4

• THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW e

QUESTION 1.06 (2.00)

The plant is operating at 70 percent power with all systems in automatic.

Indicate how SHUTDOWN MARGIN is changed (INCREASED, DECREASED, NO CHANGE) by the following conditions / situations. Assume that the reactor does NOT trip and that no operator actions are taken unless stated.

a. The Reactor Coolant System is borated by 10 ppm. (0.50)

b. A control rod in a shutdown bank drops. (0.50)

c. Power is increased to 90 percent WITHOUT dilution. (0.50)

d. Pressurizer level increases by 10 percent. (0.50)

QUESTION 1.07 (1.00) .

How does the positive reactivity insertion rate during a' reactor startup offect the source range count level at which criticality is achieved?

EXPLAIN your answer.

QUESTION 1.08 (2.00)

o. The plant has been operating at 100% power / load for several weeks.

Power / load are now being reduced to 50% during a 2-hour period with all systems in automatic. List the rod motion directional trends that will occur over the next 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> (beginning at start of the load change) to maintain Tave on its program assuming no change in boron concentration is made. For each directional trend indicate the time period that the rods will be trending in that direction. (1.5)

b. The equilibrium (at power) value of samarium reactivity is (DEPENDENT ON or INDEPENDENT OF) power level. (0.5) i l

l l

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

1. PRINCIPLFR OF NUCLFAR POWER PLANT OPERATION. PAGE 5 l

• THRMODYNAMICS. WAT TRANSFER AND FLUID FLOW l

l QUESTION 1.09 (2.00)

Collaway has just restarted following a refueling outage; Wolf Creek, with cn identical reactor and fuel loading scheme, is near EOL and has just started up following a 3-week shutdown period.

c. Critical data has just been taken at 10 E-8 amps at both plants and the operators have added small, equal amounts of reactivity to continue the power ascension. Which plant will have the HIGHER steady startup rate from this equal reactivity insertion? WHY7 (1,0)

b. Shortly after 50 percent power is reached during these startups, rod control is placed in manual at both plants. Shortly afterward, a shutdown bank control rod worth -150 pcm drops into the core at both plants. Assuming that no operator action is taken for these casualties and that neither reactor trips, which plant will end up with the HIGHER steady-state Tavg? WHY7 (1.0)

QUESTION 1.10 ( 1. ",G )

A centrifugal chargina pump is operating at a low flow condition. The downstream flow centrol valve opens in response to a low-level signal from the pressurizer level controller. How will each of the following parameters be affected? (INCREASE, DECREASE, or NO CHANGE)

Assume VCT level remains constant.

! o. Pump Discharge Pressure (0.5)

b. Required NPSH (0.5)

c. Motor Amps (0.5) i 5

QUESTION 1.11 (1.00)

With the plant steady-state at 100% power, reactor coolant flow rate through D steam generator is about 10 times larger than the feed / steam flow

rates in the steam generator while the feed temperature change is only cbout 2 times larger than the reactor coolant temperature change. At steady otate the heat lost / gained by the two fluid systems should be equal.

EXPLAIN what happens to the apparent additional heat being lost from the reactor coolant.

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

t

1. PRINCIPLER OF NUCLEAR POWER PLANT OPERATION. PAGE 6

., THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.12 (1.50)

The reactor is producing 100% rated thermal power at a core delta-T of 60 degrees and a RCS mass flow rate of 100% when a station blackout occurs. Natural circulation is established and core delta-T goes to 28 F.

If decay heat is equivalent to 2% rated power, what is the natural circulation mass flow rate (in %)?

QUESTION 1.13 (2.00)

Will the Departure from Nuclear Boiling Ratio (DNBR) INCREASE, DECREASE, or REMAIN THE SAME if the following plant parameters DECREASE during power operation? Consider each parameter separately.

m. Reactor Coolant System (RCS) Pressure (0.5)

b. RCS Temperature (0.5)

c. RCS Flow . (0.5)

(0,5)

d. Reactor Power QUESTION 1.14 (1.50)

With the Callaway plant at 100% power and all systems in automatic, RCS Tavg is approximately 588 F and steam generator (SG) pressure is cpproximately 980 psig. The programmed Tava at 0% power is 557 F.

Uaing these given conditions, determine the following:

a. SG temperature at 100% (0.25)

b. RCS Tavg at 75% (0.50)

c. SG temperature at 75% (0.50)

d. SG pressure at 75% (0.25)

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

1. PRINCIPLRR OF NUCLEAR POWER PLANT OPERATION. PAGE 7

. THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.15 (1.50)

The plant is in Hot Standby with the RCS pressure being maintained at 985 psig. A pressurizer PORV is leaking to the pressurizer relief tank which is at 5 psig.

c. The leakage entering the PRT is a: (SELECT ONE) (0.50)

1) Superheated vapor

2) Wet vapor

3) Saturated vapor

4) Subcooled liquid

b. What is the enthalpy of the leakage entering the PRT7 (0.50)

c. What is the temperature on the tail pipe downstream of the PORV?(0.50)

QUESTION 1.16 (1.50)

Indicate how each of the following parameters changes (INCREASE, DECREASE, or NO CHANGE) if C SG main steam isolation valve closes with the plant at 50% load. Assume all control systems are in automatic and that no trip occurs.

1) C SG level (INITIAL change only)

2) C SG pressure

3) C RCS loop cold leg temperature

4) D SG level (INITIAL change only)

5) A SG pressure

6) B RCS loop cold leg temperature (0.25 each)

QUESTION 1.17 (1.50)

If the reactor operates continuously at nearly 100% for an entire cycle, describe how the axial flux distribution will trend over core life. INCLUDE reasons for each trend direction. g g g g g g g g g A me% &(cp i n cys z)

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

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 8 QUESTION 2.01 (1.00)

Ench reactor coolant pump motor has an anti-reverse rotation device.

What does this device do AND why is it needed?

QUESTION 2.02 (2.00)

e. Describe the effect of a Containment Phase A Isolation signal (CISA) on:

(1) Normal RCP seal injection flowpath. (0.25)

(2) Normal RCP #1 seal leakoff flowpath. (0.25)

b. Describe how RCP seal injection / leakoff flow is maintained under CISA conditions. (0.5)

c. What additional RCP-related actuations occur upon a Containment Phase B Isolation signal (CISB) ? (0.5)

d. Describe how an RCP could be operated under spurious / faulty CISB conditions if necessary? (0.5)

QUESTION 2.03 (2.00)

c. With the plant operating at 95% power, list FOUR (4) conditions that will generate a Safety Injection Signal. INCLUDE setpoints and coincidences where applicable. (1.0)

b. Indicate the pressure at which each Emergency Core Cooling System will inject during a continued RCS depressurization caused by a LOCA during normal "at power" operations. (Assume each ECCS system was in standby until the Safety Injection Signal occurred.) (1.0)

QUESTION 2.04 (2.00)

o. List FOUR (4) adverse effects - WITH locations - that could occur if the continuous pressurizer spray valve bypass flow were isolated for several days during "at power" operations. (1.00)

b. List TWO (2) annunciators available to alert the operator that minimum spray flow is not being maintained. (0.50)

c. List the TWO (2) driving forces for pressurizer spray flow. (0.50)

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

PLANT DESIGN INCLUDING SAFUY AND EMERGENCY SYSTEMS PAGE 9 I L l l

QUESTION 2.05 (2.00)

State how the following components respond (FAIL OPEN, FAIL CLOSED, REMAIN FUNCTIONAL (i.e., remote control still available), DIVERTS TO .... ETC.)

when instrument / control air pressure is lost.

a. Letdown pressure control valve (PCV-131) (0.25)

b. Charging header flow control valve (HCV-182) (0.25)

c. TDAFP discharge flow control valve (AL-HV-6) (0.25)

d. TDAFP steam supply valve (AB-HV-6) (0.25)

e. Pressurizer spray valve (PCV-455C) (0.25)

f. Boric acid flow control valve (FCV-110A) (0.25)

g. Excess letdown flow control valve (HCV-123) (0.25)

h. Steam generator PORV(AB-LV-7) (0.25)

QUESTION 2.06 (1.00)

Match the following symptoms or causes in column "B" to the specific Rod Control System failure or error in column "A".

.. A B "

a. Logic Cabinet Urgent Failure 1. Caused by simultaneous zeroe fh current order to stationaryg and 4meetbdt gripper coils.

b. Power Cabinet Regulation Failure 2. Caused by unselected rod having current to gripper coils.

c. Power Cabinet Phase Failure 3. Caused by failure of redundant power supply module,

d. Power Cabinet Logic Error 4. Caused by pulser failure.

5. Caused by coil current exceeding demand current. .

(There is only 1 correct numerical 6. Caused by SCR bridge failure.

Enswer for each lettered error or failure at 0.25 each)

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

2. PLhMT DESIGN INCLUDING SAFETh iWD EMERGENCY SYSTEMS PAGE 10 4

1 QUESTION 2.07 (1.75)

o. List THREE (3) Component Cooling Water (CCW) system indications or symptoms which may be observed before an alarm setpoint is reached if a 10 gpm tube leak occurs in a RCP thermal barrier heat exchanger.

(0.75)

b. If the leak rate continuously increases, list TWO (2) CCW system AUTOMATIC actuations that should occur; INCLUDE the actuating condition or parameter trip point for each automati actuation. (1.0)

J QUESTION 2.08 ( .75)

List THREE(3) automatic actions that occur coincident with the HI-HI level alarm on 6A High Pressure Feedwater Heater.

l.

QUESTION 2.09 (2.00) ,

n. List the conditions required for a live-bus fast transfer of a 13.8 KV bus to the startup transformer after a turbine trip. (1.0)

b. List the conditions required for a dead-bus fast transfer of a 13.8 KV bus to the startup transformer after a turbine trip. (1.0)

QUESTION 2.10 (2.00)

e. Excluding the SI signal, list THREE(3) other automatic signals that will cause a Feedwater Isolation Signal (FWIS). INCLUDE coincidences and setpoints. (1.2)

b. Excluding the automatic FWIS, list THREE (3) other events / actions that will cause fast closure of the feedwater isolation valves. (0.8)

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

-r-.e- - --- - - - - -

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 11 l i

QUESTION 2.11 (2.50)

c. During normal "at-power" operations, what is the status / position of the following RHR-related valves / components?

1) RWST suction valves

2) RHR flow control valves out of heat exchangers

3) CCW flow to RHR heat exchangers

4) RHR discharge train crosstie isolation valves

5) RHR cold leg discharge valves (1.5)

b. State THREE (3) RHR-related flowpaths that must be realigned by the operators when shifting to recirculation flow for emergency core cooling. (1.0)

QUESTION 2.12 (1.50)

Describe the boric acid flowpaths (valves, components) from the boric acid transfer pumps to the charging pumps for the following operations:

1) Borate

2) Immediate Borate

3) Alternate Immediate Borate (0.5 each)

QUESTION 2.13 (1.50)

The AFW lines to the C steam generator have 3 orifices: one in the supply from the TDAFP, one in the supply from the A MDAFP, and one in the combined AFW supply. List all design purposes of the TDAFP supply orifice.

QUESTION 2.14 (1.00)

If the B diesel generator output breaker has closed automatically after power was lost to NB02 bus, list FIVE (5) conditions that will cause it to trip open automatically.

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

i

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 12 QUESTION 2.15 (1.00)

List SIX (6) unique / separate radiation monitors that will automatically ,

isolate their associated radioactive effluent flowpaths when their l clarm setpoint is reached.

QUESTION 2.16 (1.00)

State the signals that will cause the Essential Service Water (ESW) pumps to autostart while leaving the ESW return valves to Service Water open.

Include any applicable setpoints and coincidences.

l l

(***** END OF CATEGORY 02 *****)

i

3. INSTRUMENTS AND CONTROLS PAGE 13 ,

l l

l QUESTION 3.01 (1.50)

With the plant operating at 100% power and all control systems in auto, the detector for the controlling channel for the pressurizer level control system fails low.

e. List FOUR (4) immediate component actuations that will be initiated by the detector failure. (1.00)

b. If no operator action is taken for this casualty, what Reactor Protective System trip will eventually trip the plant? (0.50)

QUESTION 3.02 (2.50)

a. Explain how gross overcompensation of both intermediate range channels can cause a premature reactor trip during a normal reactor shutdown.

(0.75)

b. What device / component in the power range NI system is. adjusted to match nuclear power to secondary power following a calorimetric? (0.25)

c. Identify each of the following as related to the SOURCE RANGE, INTERMEDIATE RANGE, or POWER RANGE instrumentation. (1.50)

1) Summing and level amplifier

2) Pulse shaper

3) Idling current

4) Auctioneering circuitry

5) Scaler-timer

6) Preamplifier l

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

1

3. INSTRUMENTS AND CONTROLS PAGE 14 QUESTION 3.03 (3.00)

With the reactor operating at 40% power, list the actuation / trip signals that are immediately/directly caused by the following occurrences.

Consider each occurrence separately. (Alarm signals not required)

a. Low-low level on 3 channels for C SG (1.00)

b. Low-low level on 3 channels for A SG and on 2 channels for C SG (1.00)

c. Hi-hi level on 3 channels for D SG (1.00)

QUESTION 3.04 (2.00)

The plant is operating at 90% power with all control systems in automatic.

B2nk D rods are at 200 steps. Given the following conditions / situations, how will final rod height be changed (INCREASE, DECREASE, NO CHANGE)?

Assume no operator action is taken and that the reactor does NOT trip.

Consider each case separately. ,

n. A PORV on B steam generator fails open. (0.50)

b. Loop A narrow-range Thot instrument fails high. (0.50)

c. Loop C narrow-range Teold instrument fails low. (0.50)

d. Turbine load is reduced to 50%. (0.50) i i

QUESTION 3.05 (1.75)

The reactor is critical at the point of adding heat during a plant startup.

List SEVEN (7) reactor trips which are still disabled or should have been blocked by this point.

l QUESTION 3.06 (2.00)

List FOUR (4) unique turbine runback signals initiated from outside the l

turbine control system and describe the resultant runback / load decrease characteristics for each signal.

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

3. INSTRUMENTS AND CONTROLS PAGE 15

, I QUESTION 3.07 (1.00) l List FOUR (4) unique / separate control room indications of the axial flux difference value exiting the target band.

QUESTION 3.08 (1.00)

List the conditions required for an automatic start of the C Component Cooling Water (CCW) Pump with its switch in AUTO.

QUESTION 3.09 (1.00)

Ragarding the steam dump control system, describe the difference between the Loss of Load (or Load Rejection) controller and the Turbine Trip (or Plant Trip) controller relative to the following:

a. Compared temperature signals ,

(0.5)

b. Deadbands (0.5)

QUESTION 3.10 (2.25)

During a plant power increase from 40% to 100% the steam pressure detector for the selected steam flow channel on SG 1A sticks at the 40% position.

EXPLAIN the effect of the stuck steam pressure detector on the following:

a. 1A SG indicated steam flow versus actual steam flow (0.75)

b. 1A SG level versus normal level at 100% power (0.75)

c. 1B main feed pump speed versus normal speed at 100% power (0.75) l

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

3. INSTRUMENTS AND CONTROLS PAGE 16 I

1 QUESTION 3.11 (2.25)

For the following actions / occurrences, state the:

1) Steam dump control system response,

2) The plant response, and

3) The resulting method of RCS temperature control Assume all systems operate normally except as stated and that no operator action is taken. CONSIDER EACH CASE SEPARATELY.

a. With the plant in Hot Standby at normal operating temperature and pressure, steam dumps in steam pressure mode of control, waiting for a reactor startup, the steam pressure setpoint is reduced by 200 psi.

(0.75)

b. While in the steam pressure mode of control at 5% power, the train A Steam Dump Interlock Bypass Selector Switch is placed in the OFF position. (0.75)

c. The train B reactor trip breaker fails to open on a reactor trip from 78% power. NOTE: The train A breaker opens properly. (0.75)

QUESTION 3.12 (1.00)

The Rod Insertion Limit (RIL) computer uses the following equation to calculate the RIL for Control Bank D:

RIL = K'*(Tava - 557) + K*(% delta-T) + K'

where K'= 0.0, K: 2.28, and K'= -64.0

a. With the reactor at 50% power and normal operating temperature / pressure, i what is the calculated RIL for Bank D? Show calculations. (0.5) 1

b. If the calculated RIL for Bank C at 25% power is 107 steps, what is the setpoint for the following alarms on Bank C rod position? (0.5) l

1) Rod Bank Lo

2) Rod Bank Lo-Lo

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

3. INSTRUMENTS AND CONTROLS PAGE 17 4

QUESTION 3.13 (1.00) )

What is the function of the variable-gain circuit in the rod control system AND why is it needed?  ;

QUESTION 3.14 (1.00)

List AND explain TWO (2) pressurizer level conditions (including setpoints) under which the pressurizer level control system controls the pressurizer hanters.

QUESTION 3.15 (1.75)

o. List all control, protective, and permissive functions which use individual loop Tavg signals and NOT auctioneered Tavg. (1.00)

b. Excluding the control board indicators, list all other systems that use RCS loop wide-range temperature signals.

(0.75) 1 l

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

m

4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 18

. RADIOLOGICAL CONTROL

(

QUESTION 4.01 (1.50)

According to step 2 of FR-S.1, Response to Nuclear Power Generation, the operator must ensure turbine trip. List the THREE(3) control board actions, in the proper sequence, the BOP operator should take if the turbine has NOT tripped, and does NOT respond to his initial actions.

QUESTION 4.02 (2.00)

a. The Foldout for Procedure E-0, Reactor Trip er Safety Injection, lists TWO (2) conditions that TOGETHER require the operator to trip all RCPs.

What are these 2 conditions? (1.00)

b. The Foldout for E-0 also lists TWO (2) RCP conditions, either of which

require tripping the affected RCP. List these 2 conditions. (1.00)

QUESTION 4.03 (2.00)

e. List FOUR(4) separate / unique switches / controls / components that the Off-Normal Op Proc for a dropped control rod, OTO-SF-00003, requires the operators to manipulate when realigning a dropped control rod with the other rods of its group. (1.0)

b. According to OTO-SF-00004, the Off-Normal Op Proc for misalignment of control rods, the procedure for realigning a rod that is misaligned LOW with respect to its group is significantly different from the procedure for realigning a rod that is misaligned HIGH. State the basic difference between the 2 realignment procedures. (1.0) l l QUESTION 4.04 (1.50)

The off-normal operating procedure OTO-ZZ-00003, Loss of Shutdown Margin, provides 5 symptoms of this abnormal condition and requires emergency l boration as an immediate operator action. List the FIVE (5) symptoms.

i l

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

i

4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 19

. RADIOLOGICAL CONTROL QUESTION 4.05 (1.50)

With the plant operating at 100% load, you are at the turbine-driven aux fcedwater pump (TDAFP) to observe a surveillance procedure. Although the TDAFP has not been operated for at least 6 days, you note the smell of hot pipe lagging or hot pipe paint in the area of the TDAFP.

State the type of leakage that is indicated and describe the possible resulting events that could disable the TDAFP (make it unavailable for cmergency use during a station blackout) if the hot pipe lagging or paint 10 found at either of the following locations. (Hint - These events have occurred in AFPs at other PWR plants.)

Consider each location separately.

1. #- C' (0.75;

2. St::: ;;;;17 i..lm o Lv IDAFE

b. TDAFP discharge outlet to AFW feed piping (0.75,/.SCQ s

QUESTION 4.06 (2.00)

List the FOUR (4) RCS leakage detection systems having operability rcquirements stated in the Tech Specs.

l l

l l

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

i

4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 20 l

. RADIOLOGICAL CONTROL l l

l l

QUESTION 4.07 (3.00)

c. During a controlled plant cooldown per OTG-ZZ-00006, what action does the operator take to block a Steam Line Isolation (SLI) due to low steam line pressure? (0.5)

b. What indication does the operator use to determine that he can take this SLI block action? (0.5)

c. What is the limit given in OTG-ZZ-00006 before which this SLI block action must be taken? (0.5)

d. After the operator has taken this SLI block action, how does he know it worked? (0.5)

o. Explain how an SLI can still occur for a large steamline break after the operator has taken this SLI block action. (0.5)

f. Explain how an SLI can occur for a large steamline break if RCS pressure is inadvertently increased by 100 psig shortly after the operator has taken this SLI block action. .

(0.5)

QUESTION 4.08 (2.50)

The procedure for loss of all AC power, ECA-0.0, requires the operators to depressurize intact steam generators to 250 psig with the SG PORVs if AC power cannot be restored,

s. ECA-0.0 contains a note stating that the SGs should be depressurized at maximum rate. Explain the reason for this note. (1.0) l

b. ECA 0.0 also contains a caution stating that SG pressures should not be decreased below 150 psig. Explain this caution. (1.0)

c. If pressurizer level is lost during the SG depressurization, the depressurization should be (SELECT ONE).

i stopped -

continued -

accelerated -

slowed (0.5) l t

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

I

4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 21 RADIOLOGICAL CONTROL QUESTION 4.09 (1.50)

a. When should the RHR system be aligned for hot leg recirculation following a LOCA7 (0.50)

b. List TWO (2) reasons for requiring hot leg recirculation following a LOCA. (1.00)

QUESTION 4.10 ( .75)

Classify the following as either IDENTIFIED, UNIDENTIFIED, CONTROLLED, or PRESSURE BOUNDARY leakage. (0.25 each)

a. A RCP #1 seal leakoff

b. Pressurizer safety valve seat leakage

c. Steam generator tube leakage -

1 QUESTION 4.11 (1.50)

ES-1.3, Transfer to Cold Leg Recirculation, contains a caution stating that "SI pumps should be stopped if RCS pressure is greater than 1710 psig" Explain the reason the cold leg recirculation procedure contains this caution.

l l

l l

l

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

l I

, n-- - .- .

)

4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 22 j RADIOLOGICAL CONTROL QUESTION 4.12 (2.00)

State the limits provided in the GENERAL OPERATING PROCEDURES for the following parameters. (0.20 for each)

a. Maximum RCS cooldown rate

b. Maximum pressurizer cooldown rate

c. Maximum boron concentration differential between RCS and pressurizer

d. Maximum differential temperature between pressurizer and spray fluid

e. Minimum VCT pressure with RCPs running

f. Maximum RCS loop temperature with no RCPs running 6 Maximum transient startup rate

h. Maximum pressurizer heatup rate

i. Minimum Tava during critical operations J. Maximum steam generator primary-to-secondary differential pressure QUESTION 4.13 (1.50)

e. List the following exposure limits (non-emergency) for a 20-year-old UE male employee with a satisfactorily documented exposure history.

(0.25 each)

1. Maximum administrative quarterly limit for skin t

2. Maximum 10 CFR 20 quarterly limit for extremities

3. Maximum administrative yearly limit for whole body l

4. Maximum 10 CFR 20 yearly limit for whole body i b. List the TWO (2) items required before a UE employee can exceed the administrative quarterly limit for skin exposure. (0.5) 1

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

4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 23 RADIOLOGICAL CONTROL QUESTION 4.14 (1.75)

c. If you are relieving the URO after having been on vacation for SIX (6) days, how far back are you required to review the URO logs? (0.25)

b. What six additional actions must be performed to satisfy the URO watch relief checklist? (1.50)

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

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

1. PRINCIPMR OF NUCMAR POWER PLANT OPERATION. PAGE 24

. THERMODYNAMICS. REAT TRANSFER AND FLUID FLOW ANSWERS -- CALLAWAY -86/12/09-WEALE, G.

ANSWER 1.01 ( .50)

--4--

REFERENCE T-H PRIN PG 7-88 ANSWER 1.02 (2.00)

n. SAME [0.25] Steam dumps will compensate for any additional heat added by the fourth RCP. RCS temperature / reactivity unchanged. [0.25]

b. ECP LOWER than ACP [0.25] Xenon will increase to near peak at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after trip. Rods must be higher to compensate. [0.25]

c. ECP LOWER than ACP [0.25] The corresponding temperature increase must be compensated by a higher critical rod position. [0.25]

J

d. ECP HIGHER than ACP [0.25] The reduction in temperature must be compensated by a lower rod position. [0.25]

REFERENCE FNRP PG 7-24 3.1/010/001/A2.07/3.6 ANSWER 1.03 ( .50) i

--a----------------------------------

REFERENCE FNRP PG 8-39 THRU 47, RCC PG 9-10 3.9/000/015/K5.06/3.4 l

l l

ANSWER 1.04 (1.00)

a. False (0,5)

b. True (0.5)

REFERENCE FNRP PG 8-54,60 j 3.1(3.9)/000/001(015)/KS.47(K5.06)/2.9(3.4) i I

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 25

. THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- CALLAWAY -86/12/09-WEALE, G.

ANSWER 1.05 (2.00)

a. 1) Void coefficient (0.4 each, 0.3 for correct order)

2) Moderator temperature coefficient

3) Doppler power (or fuel temperature) coefficient

b. Total power coefficient becomes more negative from BOL to EOL. (0.5)

REFERENCE RCC PG 3-42 3.1/000/001/KS.49/3.4 ANSWER 1.06 (2.00)

o. INCREASED

b. NO CHANGE

c. NO CHANGE .

d. NO CHANGE (0.50 EACH)

REFERENCE RCC PG 7-13 TS DEFS 3.1010(000)/K5.35(K5.19)/3.3(3.5) i l

ANSWER 1.07 (1.00)

The faster the insertion rate, the lower the source range counts at crit-icality (0.5) due to the reduced time for suberitical multiplication.(0.5) l l

l REFERENCE l FNRP PG 8-51 3.1/000/001/K5.18/4.2 i

l l

l i

1. PRINCIPLES OF NUCWAR POWER PLANT OPERATION. PAGE 26 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- CALLAWAY -86/12/09-WEALE, G.

ANSWER 1.08 (2.00)

c. Rods insert (0.25) for about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> jpower defect compensation) (0.25)

1. L %' ithdrawRods w $ wen. redsnmen (0.25) for the next 5 (4-6) hours (Xe comp.) (0.25)

Rods insert (0.25) for remaining 33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> Xe comp.) (0.25)

M MU Aguilt'h 444444c/(d A M b daua

b. INDEPE EENT OF (0.50)

REFERENCE RCC PG 4-26 3.1/000(010)/001/K5.13(K5.26)/3.7(3.5)

ANSWER 1.09 (2.00)

c. Wolf Creek (0.50) due to a smaller Beff value (0.50)

b. Wolf Creek (0.50) due to a more negative MTC value (0.50)

REFERENCE FNRP PG 7-33, RCC PG 3-23 3.1/000/001/K5.47(K5.49)/2.9(3.4) i ANSWER 1.10 (1.50) i a. Decrease (0.5)

b. Increase (0,5)

c. Increase (0.5)

REFERENCE T-H PRIN PG 10-33,36,48 COMPONENT-VALVE /2.0

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 27 THFRMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- CALLAWAY -86/12/09-WEALE, G.

ANSWER 1.11 (1.00)

Supplies the latent heat of vaporization (0.5) for the feedwater-to-steam phase transformation / change (0.5).

REFERENCE T-H PRIN PG 8-41,6-42,4-5,60 3.2/000/002/K5.01/3.1 COMPONENT-HX and COND/2.0 ANSWER 1.12 (1.50)

To determine flow in NC:

Q = m cp delta-T => 100 = 100

• cp

• 60 => cp = 100

. 100

• 60 cp = .0167 (0.75)

THEREFORE: 2% = m * .0167

• 28 => m= 2%

---

= 4.28% (0.75)

.0167

• 28 REFERENCE T-H PRIN PG 4-60 3.2/020/002/K5.01/3.1 ANSWER 1.13 (2.00)

a. DECREASE

b. INCREASE

c. DECREASE

d. INCREASE (0.5 each)

PEFERENCE T-H PRIN PG 13-23 3.9/020/015/K5.09/3.5

1. PRINCIPLES OF NUCr.RAR POWER PLANT OPERATION. PAGE 28 THRRMODYNAMICS. HFAT TRANSFER AND FLUID FLOW ANSWERS -- CALLAWAY -86/12/09-WEALE, G.

ANSWER 1.14 (1.50)

(Tolerance is +/- 1 F on temperatures, +/- 10 psig on pressures)

c. From steam tables, 980 psig --> 544 F

b. At 100%, Tavs : Tavg(no-load) + Y , where Y = Tavg program delta-T or, 588 F = 557 F + Y, --> Y = 31 F (0.25) at 75%, Tavg = 557 + 0.75*Y = 557 + 0.75*31 = 557 + 23 = 580 F (0.25)

c. At 100%, Tstm = Tavg minus Z , where Z = SG tubewall delta-T or, 544 F = 588 F -

Z, --> Z = 44 F (0.25) at 75% Tstm = Tavg - 0.75*Z = 580 - 0.75*44 = 580 - 33 = 547 F (0.25)

d. From steam tables, 547 F --> 1005 psig . (0.25)

REFERENCE T-H PRIN PG 5-43, 12-9/10/11 STEAM TABLES 3.1/000/001/K5.45/2.4 ANSWER 1.15 (1.50)

o. --l-- (0.50)

b. 1192 BTU /LBM (1190-1195 BTU /LBM acceptable) (0.50)

c. 300 F (290-310 F acceptable) (0.50)

REFERENCE STEAM TABLES COMPONENT-VALVE /2.0

~

1. PRINCIPrRR OF NUCrJAR POWER PLANT OPERATION. PAGE 29 IHERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- CALLAWAY -86/12/09-WEALE, G. .

ANSWER 1.16 (1.50)

I') Decrease

2) Increase

3) Increase

4) Increase

5) Decrease

6) Decrease [0.25 each]

REFERENCE T-H PRIN PG 48-51 M iA WNhz,e {et.

^ '

j:,fg j 01M& den.c.N h hd5ah wbik yc/t C,or,,tg FInx ou.Lge peaks below the centerline initially [.25] due to MTC and lower core inlet temperature [.25]. Flux bulge moves upward over core life [.25]

due to lower core fuel depletion in upper regions of the core [.25].

Bulge flattens out later in life [.25] because MTC value has increased and coves some flux back down. [.25]

REFERENCE RCC PG 8-19 THRU 22

- 15>1- Cyc.de E 67f) h O*~ ffEf* -yc6 / Aeoalk/4r /5$dix D'""'  % 4 af f g% Q ps 44e & M Jia p a&n n a a psapsma ad-act)

A' @*- /d saf J ce,6sfcou b qu, ao 4jn

+-p4 M~ EOL & A}+ $ Y j a gagg 4 g/gya/3 l

,4g cmu p 4g a .dg&'sd2/fudu. Jebe,-k,(m 49%< Av fadu. goata d tot) l

1

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 30 i l

ANSWERS -- CALLAWAY -86/12/09-WEALE, G.

l ANSWER 2.01 (1.00)

Holds the motor shaft idle (or prevents shaft from reverse rotating) (0.5) to prevent motor overheating on starting (or prevent excessive starting current) (0,5)

REFERENCE CPSHO RCS PG 15 ANSWER 2.62 (2.00)

o. 1) No effect (or normal seal injection still supplied) (0.25)

2) Seal leakoff containment isolation valves close (0.25)

b. Seal leakoff goes through a relief valve (V-8121) upstream of the containment isolation valves to the PRT (injection still normal)

(0.5)

c. Component cooling water to the RCP heat exchangers is isolated (0.5)

d. Bypass valves around CCW isolation valves could be opened (0.5)

REFERENCE CPSHO CVCS PG I-13, CCW PG 7,9 ANSWER 2.03 (2.00) l c. Low pressurizer pressure (0.10) 1849 psig(0.10) 2/4(0.10) l High contaiment pressure (0.10) 3.5 psig(0.10) 2/3(0.10) j Low steamline pressure (0.10) 615 psig(0.10) 2/3(0.10)

Manual (0.10)

b. High head injection (CCPs) 1849 psig(immediate) (0.25)

Intermediate head injection (SI) 1530 + or - 50 psig (0.25)

Accumulators 625 + or - 50 psig (0.25)

Low head injection (RHR) 195 + or - 20 psig (0.25)

REFERENCE CPSHO RHR A1,PG20; SI PG1 l

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 31 ANSWERS -- CALLAWAY -86/12/09-WEALE, G.

ANSWER 2.04 (2.00) l

o. 1) Thermal shock / stress in the spray line/ nozzle (0.25)

2) Thermal shock / stress in the surge line/ nozzle (0.25)

3) Non-uniform chemistry (stratification) in the pressurizer (0.25)

4) Non-uniform temperatures (stratification) in the pressurizer (0.25)

b. 1) Pressurizer Spray Line Lo Temperature alarm (0.25)

2) Pressurizer Surge Line Lo Temperature alarm (0.25)

c. 1) Differential pressure across the reactor vessel (will also accept RCP delta-P or delta-P from spray connection to surge connection) i (0.25) ,

2) Velocity head of the RCS flow (into spray scoops) (0.25) 1 REFERENCE CPSHO RCS PG 26 ANSWER 2.05 (2.00)

o. Fail open

b. Fail open

c. Remain functional (backup N2 supply)

d. Fail open

o. Fail closed

f. Fall open

c. Fail closed

h. Remain functional (backup N2 supply) (0.25 each)

REFERENCE CPSHO CVCS I-5,11,14;2-10; RCS 26; AFW 5; MS 7

2. PLANT DESlQtL_IHCLUD1HG SAFETY AND EMERGENCY SYSTEtni PAGE 32 ANSWERS -- CALLAWAY -86/12/09-WEALE, G.

~

ANSWER 2.06 (1.00)

a. 4

b. 5

c. 6 (0.25 each)

d. 1 REFERENCE CPSHO RD CNT PG 18,19 ANSWER 2.07 (1.75)

o. 1. Rising surge tank level

2. Increasing CCW system radioactivity

3. Increasing thermal barrier heat exchanger outlet flow (0.25 each) 4, cOmeMaJvg k.P tlw al (re^^UA Cc!J oufAri Vsay's b 1. The inside containment isolation valve (HV-62)' will shut (0.25) on high thermal barrier heat (0.25 W ' CL k i/ 4 2. emdshdualcaJukepon The CCW surge tank venNwill ghut egee$xchanger on high CCW system outlet flow

/ M M ,p LW * *kdap vh( 0. )25 ) '

• • 6 % radioactivity ik4 o-EL+g otfuns ud ca.w

( 0. 25 ) .2)

,go.} REFERENCE i CPSHO CCW PG 7 ANSWER 2.08 ( .75)

1) Extraction steam valves close

2) Drain valves to heater close av Sp

3) Scavenging steam valve closes (0.25 each)

REFERENCE ' '

y CPSHO FW BTR TP-10 gf _gfy gzg gg,,dag fjj y&

\ >

AF - LV2 M Ha tes bras s A P - t. V 1 7

\ 'A ' MSR k/-s*fDsalmDe.

A F - LY2IS 'C ' lASR 1d Sh 945 7~5-Ay -L vizc 54 Sky M arhk A F - L\E D 'C' MSI2 W h 8 #b**

AF - LVizE 'A' MSR &jk &

2. PLANT DESIGN INCLUDING SAEETY AND EMERGENCY SYSTEMS PAGE 33 ANSWERS -- CALLAWAY -86/12/09-WEALE, G.

ANSWER 2.09 (2.00)

a. 1) Control switch in normal

2) Main generator output breakers open #2. 4Maud , *

3. Synchro check relay satisfied (0.333 each) d /oclodN 77 'P

b. 1) UAT output breaker open A0 6WYoh SMah tawfotaux enuoyd -

0

2) Control switch in normal

3) Undervoltage condi, tion cn 13.8 KV bus

4) Main Gen breakers open or lockout relays tripped (0.25 each)

REFERENCE CPSHO SERV PWR PG 4 ANSWER 2.10 (2.00) s

o. 1) HI-HI SG LEVEL 78% 2/4 channels 1/4 SGs *

(0.4)

2) LO-LO SG LEVEL 23.5% 2/4 channels 1/4 SGs (0.4)

3) LOW TAVE 564 F with P-4 2/4 loops (0.4)

b. 1) Manual (0.3)

2) Loss of power to active hydraulic circuit (0.25)

3) Loss of power to inactive hydraulic circuit (0.25)

REFERENCE CPSHO MFW PG 13 4

ANSWER 2.11 (2.50)

a. 1) open

2) open (for full flow)

3) isolated

4) open

5) open (0.3 each)

b. 1) cut in CCW to RHR HXs

2) open RHR to CCP suctions Tf

3) en RHR to SI pump suctions s(0.333each)

REFERENCE CPSHO RHR PG 16 6] STj' ppg gg g jg fgl ,,cyc i

r

2. PLANT DESIGN INCLUDING SAEETY AND EMERGENCY SYSTEMS PAGE 34 ANSWERS -- CALLAWAY -86/12/09-WEALE, G.

ANSWER 2.12 (1.50)

1) Thru BAFCV (FCV-110A) to blender thru VCT outlet MUFCV (FCV-110B)

2) Thru Immed Borate valve (HV-8104) (0.5 each)

3) Thru BAFCV (FCV-110A) thru manual borate valve (V-177)

REFERENCE CPSHO CVCS PG 2-17,18,19 ANSWER 2.13 (1.50)

1) Limit containment pressure rise due to ruptured C SG/ steam header (limit AFW flow to faulted C SG) (0.5)

2) Permit feeding intact SGs (prevent excessive pressure loss to C SG feedline break) (0.5)

3) Limit potentially damaging AFW pump runout ,

(0.5)

REFERENCE CPSHO AFW PG 5 ANSWER 2.14 (1.00)

1) DG lo lube oil pressure

2) DG hi .iacket water temp

3) DG hi crankcase pressure

, 4) DG overspeed

5) Differential overcurrent (.20 each)

REFERENCE CPSHO STBY GEN TP 20,22 ANSWER 2.15 ( 1. 00.)

1) liquid radwaste discharge (RE-18)

2) SG blowdown (RE-52)

3) turbine building drains (RE-59)

4) secondary liquid waste (RE-45) eg

5) containment purge (RE-22,33) d*J 4 e ,gu M axwetJf

6) radwaste building vent (RE-10)

(b0.167 each) d$40 W y}$

.cen+o n ca-wp_w&r f! f Alon) ahm esflu.2.

(kt-27{Zl)

+" m

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SXSIEMS PAGE 35 ANSWERS -- CALLAWAY -86/12/09-WEALE, G.

REFERENCE TS 3/4.3-63,72,73 ANSWNR 2.16 (1.00) n 21.71fa Low suction pressure (LSP) to Aux Feed Pumps (0.25) 7psig(0.25), 2/3(0.25) with Aux Feed Actuation Signal (AFAS) (0.25) ^

REFERENCE CPSHO ESW PG 9

3. INSTRUMENTS AND CONTROLS PAGE 36 ANSWERS -- CALLAWAY -86/12/09-WEALE, G.

ANSWER 3,01 (1.50)

c. Letdown isolation valves close (0.25)

All letdown orifice isolation valves close (0.25)

All pressurizer heater groups are turned off (0.25)

Increased ch,arging flow (0.25)

, du b.

Also W).alla ocetMadb. 8un fou .$ vel d; n1Ne,v og fys /7% }{Q ya My High pressurizer level (at 9f%). (0.50) (ato,2 u g.

REFERENCE CPSHO PP&LC LD [ "fhbdf ANSWER 3.02 (2.50)

a. Gross overcompensation could cause an erroneously low indicated flux on the IR channels (0.25) which could cause premature reenergizing of the source range detectors (0.25) thus causing a reactor trip from SR high flux trip (0.25). -

(0.75)

b. The gain potentiometer (or gain adjustment on the front of the PR B drawers) (0.25)

c. 1) PR

2) SR

3) IR

4) PR

5) SR

6) SR (0.25 each for total of 1.50)

REFERENCE CPSHO NIS TPs ANSWER 3.03 (3.00)

.33

c. Reactor trip,-turbine trip, FWIS, and MDAFAS (0.25- each)

.35

b. Reactor trip, turbi m trip, FWIS, and TD/MDAFAS (0.iHr each)

.33

c. Turbine trip, FWIS, TDMFP trip, and MBAFAG- (0.e5 each)

REFERENCE CPSHO AFW PG7, MFW PG 13,15

3. INSTRUMENTS AND CONTROLS PAGE 37 ANSWERS -- CALLAWAY -86/12/09-WEALE, G.

ANSWER 3.04 (2.00)

c. INCREASE

b. ' DECREASE

c. NO CHANGE

d. DECREASE (0.50 each)

REFERENCE CPSHO ROD CTRL PG 9 ANSWER 3.05 (1.75)

1. Source range high flux

2. RCP low voltage

3. RCP underfrequency

4. Pressurizer low pressure

5. Pressurizer high level

6. Single-loop low flow aw4y 1,

7. Turbine trip 0 (3 25 each) 9.

AferAWf2.s.Acef!asJfeta REFERENCE CPSHO RPS TP-4 ANSWER 3.06 (2.00)

1) C-3(OT delta-T) (0.25) Load decreases at a. bout 10% per minute until below setpoint (0.25)

2) C-4(OP delta-T) (0.25) Load decreases at about 10% per minute until below setpoint (0.25)

/

3) Loss of cire water pump at > 75% (0.25) Late decreases to 75% (0.25)

, 4) Loss of stator water cooling (low flow, b?>h + mp, or loss of DC control power) (0.25) Load decreases to 8(% d,6,,

REFERENCE CPSHO HN TURB CTRL PG 21

(

o

3. INSTRUMEHLS AND CQH.TROLS PAGE 38 ANSWERS -- CALLAWAY -86/12/09-WEALE, G.

l ANSWER 3.07 (1.00) l

1) Alarm printer output (0.25)

2) Delta-I meters (0.25)

3) CRT display (color change or violation time indication) (0.25)

4) Audible alarm window (0.25)

REFERENCE LER 483/85-037 ANSWER 3.08 (1.00)

1) Low header pressure with A CCWP running (0.5)

2) A CCWP fails to start on LOCA/ shutdown sequencer start signal (0.5)

$ GCCtfbt ! - A &.U,g A ccP x A kak ecap),wy Q PH C W PG 4,5

• d " C M

ANSWER 3.09 (1.00) o.-L of L compares Tavg to Tref (0.25); TT compares Tavg to Tno-load (0.25)

b. L of L has 2F deadband(0.25); TT has no deadband(0.25)

REFERENCE

• CPSHO MS PG 24,25

3. INSTRUMENTS AND CONTROLS PAGE 39 ANSWERS -- CALLAWAY -86/12/09-WEALE, G.

ANSWER 3.10 (2.25)

m. Indicated steam flow will be higher than actual flow (0.25) because steam pressure compensation is using too high pressure multiplier.(0.5)

b. Erroneous high steam flow will try to bring feed flow up(0.25), but level-dominant SGWLC system will reduce feed flow as necessary to main-tain constant programmed level or slightly increased level.(0.5)

c. Erroneous high steam flow will cause programmed feed pump delta-P to go high(0.25) causing feed pump to speed up to supply more delta-P than normal at 100% power (0.5)

(Give full credit on b and e if effect traced correctly even though initial effect on steam flow reasoned incorrectly.)

REFERENCE CPSHO MFW TP 7,8 ANSWER 3.11 (2.25)

n. 1) The steam dumps open(.25)

2) Cooling Tavg to about 550F(0.25)

3) The P-12 interlock will control RCS temperature by cycling the steam dump valves around 550 F (0.25)

b. 1) All steam dump valves will shut (0.25)

2) Steam pressure will rise to the setpoint of the main steam atmospheric relief valves (0.25)

3) MS atmospheric relief valves will cycle to maintain steam pressure and RCS temperature.(0.25)

c. 1) Normal shift to the Turbine Trip controller will not occur (0.25)

2) Plant will cooldown because Load Rejection controller will open steam dumps (0.25)

3) The Load Rejection controller will cool and maintain RCS temperature near "No Load" Tref (557F +2F deviation /deadband) (0.25)

REFERENCE CPSHO MS TP 13 l

i l _

3. INSTRUMENTS AND CQRIROLS PAGE 40 ANSWERS -- CALLAWAY -86/12/09-WEALE, G.

4 ANSWER 3.12 (1.00)

c. RIL = 0.0*( ) + 2.28*50 + -64 = 50 steps (0.5)

b. 1) 117 steps

2) 107 steps (0.25 each)

REFERENCE CPSHO ROD CTRL PG 15 ANSWER 3.13 (1.00)

It reduces the signal multiplier at higher powers (0.5) because the change in % power for a given change in reactivity is greater at higher powers (or to prevent overshoot at higher powers) (0.5)

REFERENCE .

CPSHO ROD CTRL PG 9,10 ANSWER 3.14 (1.00)

1) Turns all heaters off at 17% level (0.25) to prevent heater burnout (.25)

2. Turns on backup heaters if actual level varies above program by 5%(.25) to minimize the pressure transient during an outsurge after an insurge of relative y cold water (0.25) l REFE CE j CPSHO RCS INST PG 29 l

i

3. INSTRUMENTS AND CONTROLS PAGE 41 ANSWERS -- CALLAWAY -86/12/09-WEALE, G.

ANSWER 3.15 (1.75)

c. 1) OT Delta-T calculator (0.25)

2) OP Delta-T calculator (0.25)

3) P-12 circuitry (Hi Stm. Flow SI permissive, Steam dump block, Lo-lo Tavg signal) (0.25)

4) Feedwater isolation circuitry (Lo Tavg signal) (0.25)

b. 1) RVLIS

2) COPPS

3) Core subcooling monitor (0.25 each)

REFERENCE CPSHO RCS INST PG 11

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

4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 42

. RADIOLOGICAL CONTROL ANSWERS -- CALLAWAY -86/12/09-WEALE, G.

ANSWER 4.01 (1.50)

1) Manually trip turbine

2) Manually run back turbine

3) Fast close MSIVs and bypasses (0.4 each, 0.3 for order)

REFERENCE FR-S.1 ANSWER 4.02 (2.00)

n. At least one CVCS Pump or SI Pump running (0.5), and RCS pressure less than 1400 psig. (0.5)

b. 1) Either motor bearing temperature > 195 F

2) CCW to the motor is lost > 2 min -

(0.50 for each)

REFERENCE E-0 FOLDOUT ANSWER 4.03 (2.00)

c. 1) Bank selector switch

2) Lift coil disconnect switches for other rods in bank

3) Step counter for affected group

4) Auto-manual switch at P-A converter

5) In-Hold-Out switch (any 4, 0.25 each)

6) Alarm reset for urgent failure

b. A rod that is misaligned high is driven in to match its group (0.5);

for a rod that is misaligned low, the group is driven in to match the misaligned rod. (0.5)

REFERENCE OTO-SF-00003,00004 i

i

4. PROCEDURES - NORMAL. ABHQRMAL. EMERGENCY AND PAGE 43 RADIOLQGICAL_.QQHIBQL ANSWERS -- CALLAWAY -86/12/09-WEALE, G. .

ANSWER 4.04 (1.50) l' ) Failure of one or more RCCAs to fully insert following a reactor trip or shutdown.

2) Anytime an RCCA is at or < Lo-Lo insertion limit.

3) Unexplained or uncontrolled reactivity increase.

4) Uncontrolled cooldown not requiring safety injection.

5) Any time minimum SDM is in question. (0.30 each)

REFERENCE OTO-ZZ-00003 ANSWER 4.05 (1.50)

. ndicates ste t leaka 11iif., l i n e

'25) which dsusing d causa-significant peed <

/"[K-co n in steam

.t.s ur an e p gency rt. ,fM2 ) ,

b. Indicatesdischargecheckvalve,backleakage(0.f5) which could cause steambindingin.theTDAFP(0.g5)preventingemergencyAFWflowdueto loss of NPSH (0./5)

REFERENCE IE INFO NOTICE 86-14, INPO SOER 84-?

ANSWER 4.06 (2.00)

1) Containment Atmosphere (0.25) Particulate Radioactivity Monitoring System (0.25)

2) Containment Normal Sump (0.25) Level Measurement System (0.25)

3) Containment Air Cooler (0.25) Condensate Flow Rate Monitoring l System (0.25) l

4) Containment Atmosphere (0.25) Gaseous Radioactivity Monitoring System (0.25)

REFERENCE TS 3/4.4.6 l

l l

l I -

4. PROCEDURER - NORMAL. ABNORMAL. EMERGENCY AND PAGE 44

., RADIOLOGICAL CONTROL ANSWERS -- CALLAWAY -86/12/09-WEALE, G. .

ANSWER 4.07 (3.00)

e. Place both A an4 B train switches for steam line isolation / safety injection (SLI/SI) to block (0.5)

b. P-11 status light illuminates (0.5)

c. RCS Tava 540 F (0.5)

d. Train A and B SLI/SI BLOCKED. status lights illuminated (0.5) te is n (0.5)

e. SyI (atdue St-[to whigh steam line b.&%fxd1Cnf 3 negative M y pressure wuMol% rp&g 4*Z &ow enablep t ska)  % 424)

f. Increasing RCS pressure > 1970 psig removes P-1T and au1fomatically unblocks / reinstates the low steam line pressure SLI (0.5)

REFERENCE (m SLT w he a Of5b A fy conham/ pun.,e siaela. C M.z sef&l- du.s fa OTG-ZZ-00006 PG 11,12 shgfj ANSWER 4.08 (2.50)

c. To minimize RCS inventory loss (0.5) via the RCP seals. (0.5) l b. To prevent injection of complete accumulator (0.5) and subsequently

nitrogen into RCS (0.5)

c. Continued (0.5) -

REFERENCE ECA-0.0 PG 12 ANSWER 4.09 (1.50)

a. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the LOCA. (0.50)

b. 1) To prevent or flush out any buildup of boron crystals in the upper regions of the core (0.50)

2) To quench any steam bubble that may have built up in the top of the core (0.50)

REFERENCE ES-1.4, CPSHO RHR PG 19 4

I

. _ . . , . , - . _ _ _ _ . - _ . _ , _ . - . , _ __ _ _ _ . _ _ _ , . _ _ . _ _ _ . , . _ _ . _ . , _ . _ , . _ . . _ _ _ . _ . - . _ , . _ . , _ _ . _ . , . . _ . _ _ , , . _ _ . _ _ , , . . _ _ _ _ . , _ __ _ m

4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 45

. RADIOLOGICAL CONTROL ANSWERS -- CALLAWAY -86/12/09-WEALE, G.

ANSWER 4.10 ( .75)

c. Controlled

b. Identified

c. Identified (0.25 each)

REFERENCE TS - DEFINITIONS ANSWER 4.11, (1.50)

The miniflow isolation valves for the SI pumps are shut during this procedure (0.5). If the RCS pressure is greater than the shutoff head of the SI pumps (approximately 1536 psig), there will be no flow thru the Pumps (0.5) to remove pump heat and resulting overheating *could damage the pumps (0.5)

REFERENCE ES-1.3 PG 4, CPSHO SI PG 1,4 ANSWER 4.12 (2.00)

o. 50 F/hr

b. 100 F/hr

c. 50 ppm

d. 320 F at SF3*F

o. 15 psig

f. 160 F a2 <.3 s O

• F

g. 1.5 dpm

h. 100 F/hr

1. 551 F

j. 1600 psid (0.2 for each)

REFERENCE OTG-ZZ-00001,2,6

4.

PROCEDURES - NORMAL. ABNORMAL. EMER(FNCY AND PAGE 46 3 RADIOLOGICAL CONTROL ANSWERS -- CALLAWAY -86/12/09-WEALE, G. ~

b

\

ANSWER 4.13 (1.50)  !

a . ' 1. 6500 mrem m.4000 m

2. 18750 mrem l

3. 4800 mrem et 4000AvJu 4.

12000 mrem ot so h (not to exceed 5*(N-18)) (0.25 each)

b. 1) Plant Manager's approval

2) Stay-time calculations 0.25 each) 3)

REFERENCE HPlaybdnLifa a?hnevd ii (MI, HDP-ZZ-01 ANSWER 4.14 (1.75)

a. For last three days (0.25) .

b. 1) Review Standing and Night Orders.

2) Test control room annunciators.

3) Review annunciator defeat log.

4) Discuss significant operations or maintenance in progress.

5) Perform control board walkdown. 4,

6) Review incident reports. 25 each)

7) Ruba/wfik. c^w wifmf REFERENCE ODP-ZZ-00003 u