Text

Exam Rept 50-483/OLS-89-01 on 890123-26.Exam Results:One Reactor Operator Candidate Failed Exams.All Senior Reactor Operator Candidates Passed Exams
	ML20235U154
Person / Time
Site:	Callaway
Issue date:	02/27/1989
From:	Burdick T, Damon D; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:	;
Shared Package
ML20235U124	List: ML20235U121; ML20235U154;
References
	50-483-OLS-89, 50-483-OLS-89-0, NUDOCS 8903090074
	Download: ML20235U154 (191)
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U.Si NUCLEAR' REGULATORY COMMISSION

' REGION III n <

Report No. 50-483/0LS-89-01 Docket No. 50-483 License No. NPF-30 Licensee: Union Electric' Company. :5 Post Office-Box 149 - Mail Code 400 St. Louis, MO '63166 n

' Facility Name: Callaway Plant Examination Administered At: Callaway Nuclear Plant Examination Conducted: January 23-26, 1989-Chief Examiner: bNf7 D. O'. Damon- Date L ' Approved By: 7[

T.:M. Burdick f!uY Q L Date 7(([

1 Examination-Summary l.:

Examination' administered on' January 23-26, 1989-(Report No. 50-483/0LS-89-01) l -~ to.four Reactor Operator candidates and four Senior Reactor Operator candidates.-

Results: One P.eactor Operator candidate failed the examinations. All Senior Reactor Operator candidates passed the examinations.

1 8903090074 090228 ,/ '

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REPORT DETAILS

1. Examiners

D. J. Damon T. P. Guilfoil I. G. Kinasley

Chief Examiner

2. Exit Meeting On January 26, 1989, the examiners met with members of the facility staff to discuss the examination. The following persons attended the meeting:

J. D. Blosser, Manager Nuclear Operations, Union Electric G. Randolph, General Manager, Union Electric G. Hughes, Supr. Engineering ISEG/STA, Union Electric N. Lombardi,. QA Engineer, Union Electric M. Evans, Superintendent Training, Union Electric D. W. Neterer, STS-0PS, Union Electric B. Little, Senior Resident Inspector, NRC D. Damon, Examiner, NRC K. Shembarger, Examiner, NRC N. Maguire-Moffitt, Examiner, NRC D. Pereira, Examiner, NRC The following comments were made by the examiners during the meeting:

Strengths 1

j. a. Communications skills as displayed by the candidates during the L simulator scenarios were generally strong.

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( b. Candidates made good use of off-normal and emergency procedures l during the simulator scenarios.

c. Bistable tripping evolutions were closely controlled by the candidates during the simulator scenarios.

d. During the plant walkdowns, the reactor operator candidates were generally very knowledgeable of technical specifications and bases.

i e. All candidates were extremely knowledgeable about refueling evolutions and procedures.

Weaknesses

a. Candidates did not refer to " common-usage" procedures during the simulator scenarios. This resulted in such things as incorrect turbine loading rates and lack of proper department notifications being made during load swings.

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b. Reactor Operators and Balance-of-Plant operators did not carry out immediate action steps of emergency procedures until prompted to do' so by the Senior Reactor Operator. These items must be committed.to memory and carried out by the operators in an' expeditious manner,.

without waiting for prompting by the Senior Reactor Operator. One example of this is the requirement to trip the turbine on an ATWS condition. In'one instance, three minutes elapsed between recognition of an ATWS by the crew, and the turbine being tripped by the BOP operator.

General Comment a

The initial submittal of reference materials by the facility was- '

inadequate. In addition, materials submitted to contract examiners were'different than materials submitted to NRC examiners. Inadequacies existed in'that.not all requested materials were received by the !

examiners'in the initial shipment and that the materials were not i properly labeled. .A.second shipment of materials by the facility still '

did not result in all materials being received by the examiners. .It'is requested that the facility monitor the shipment of reference materials-more closely for future exams.

' Facility Comments Following are facility comments on the written examinations and their respective NRC resolutions.

Question 1.18 Facility Comment Locked rotor protection would cause the pump breaker to trip instantaneously. The question should be deleted - there is no reference to facility objective.

NRC Response Comment not accepted. No reference was submitted with the facility comment to support the first statement of the comment.

Knowledge of locked rotor indication is rated with a 2.9 importance l

e by NUREG-1022, Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Pressurized Water Reactors. This indicates that the commercial nuclear power industry has determined that the operator should possess knowledge of locked rotor indication. Knowledge of locked rotor indication is important for protection of safety-related equipment, especially if circuit breaker actuation fails to isolate the equipment.

Operator action would then be_ necessary to prevent further equipment !

damage. Answer key remains unchanged.

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1 Question 3.07a Facility Comment' Also acceptable high CCW return temperature from the RCP thermal barriers. )

Ref: M-22BB03 NRC Response Comment accepted. Answer key modified.

Question 3.10 Facility Comment Additional answers for Part A should include:

1. Backup heaters on.

2. PDP speed control decreasing speed.

Answer A.2 should be LCV-460 closes.

Ref: Process and Control Block Diagrams, Print 8756D37, Sheet 27 and Functional- Diagrams, Print' 7250D64, Sheet 11 and 12.

!' NRC Response L

i Comment accepted. Answer key modified.

Question 4.06 Facility Comment Beff should be .0062 to .0052. ,

Ref: Reactor Theory Review Text NRC Response Comment accepted. Answer key modified.

Question 4.10 1 Facility Comment Credit should also be given for the required change in enthalpy (Delta-H) across the steam generator being smaller at higher steam ;

generator pressures, therefore, less heat is required to convert 4

g each. pound-mass of saturated liquid to saturated vapor.

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Ref: Steam Tables Westinghouse Thero-Hydraulic Principles Pg. 7-66-through 7-69 NRC Respo_nse Comment not accepted. The facility comment discusses factors effecting primary efficiency. The question.specifically asked for a discussion of effects on secondary efficiency. Answer key remains unchanged.

Question ~4.13 Facility' Comment Also acceptable starting and stopping pumps.

Ref: 0TN-AD-00001, Precautions and Limitations, Step 2.3.1 NRC Response Comment not accepted. The answer and reference sited in the facility comment does not describe a method or means of reducing water hammer.

Answer key' remains unchanged.

Question'4.19 Facility Comment Also accept general warning causing reactor trip.

Ref: Callaway Plant FSAR Section 7.2, Figure 7.2-1, Sheet 2 NRC Response Comment partially accepted. The conditions given in the question will ,

cause a general warning to be generated in addition to causing a reactor 'l trip. The. reactor trip in this case would not be a direct result of the general warning. However, since the general warning is also generated in

.this scenario, the answer key is modified to include the general warning.

Question 4.21 Facility Comment Since there are different ways to work this problem, full credit should also be given for working with an assumed initial speed, and using formulas as long as the final power calculation was correct. Therefore, point breakdown on answer key may not be correct for different methods of calculations.

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I ii Ref: Pump lesson Plan II.B.4 NRC Response Commented noted. The candidate's answer will receive full credit if the final answer is arrived at'by any valid technique.

Question 5.05-Facility Comment Additional answers: Adverse containment as defined in emergency procedures.

Ref: E-1 l NRC Response 1~

Comment noted. Answer key already allows for other answers on a case by case basis.

Question 5.07 Facility Comment Also accept Control Room handswitch (KC-008).and local manual start.

l Ref: Fire Protection S.D., Pg. 25, Student Handout NRC Response l Comment ~not accepted. The additional reference cited in the facility comment does not show that these controls are operable when control circuits are malfunctioning. The original reference states that the emergency start lever is used to start the pump when control circuits are malfunctioning. Answer key remains unchanged.

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I Question 5.08 Facility Comment 1

The question asks to explain the basis for entering Tech Spec 3.03, therefore, the answer key breakdown should not include "This requires entry into TS 3.03 (.75)" since this statement was made in the question.

NRC Response Comment accepted. Partial credit reassigned.

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l Question'5.12 l Facility Comment Answer key point breakdown should not include "Under low RVLIS level' !

conditions (.5)" due to not being in the basis. Full credit should be given for statements'that the RCP's provide flow through the core, therefore, keep the core cool. ;

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.Ref: Westinghouse Owner's Group Emergency Response 1 Guidelines Background FR-C.2 Basis NRC Response j Comment partially accepted. Answer key modified to remove references' l to low RVLIS conditions and core uncovery. j Question 5.15 !

Facility Comment j l

Answer key point breakdown should give full credit for actions alone i (i.e., increase charging flow, decrease letdown flow). The question implies increasing pressure by stating "What actions must be taken to increase subcooling margin."

Ref: 0TO-EJ-00001 i I

NRC R,esponse Comment not accepted. Another way to increase subcooling margin is to ')

lower temperature. The parameter used to change subcooling margin is not implied in the question and the candidate must state this to receive full credit. Answer key remains unchanged.

Question 5.16 Facility Comment Answer key Part B should not include "The pump is stopped to lower the RCS pressure in' the faulted loop (.05)." This is because the pump is not stopped in Step 6.2.3 for which the basis is asked. -The RCP is stopped in Step 6.2.4.

The answer key in Part B only states that basis for the RCS depressurization to 1000 psig. Credit should also be given for the cooldown to 500 F. The basis for this is RCS subcooling.

Ref: 0T0-BB-00001, Rev.1, Pg. 2, Step 6.2.3, 6.2.4 1

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NRC Response Comment = accepted. Answer key modified to read as follows: "To increase subcooling margin (0.25) and to reduce primary to secondary leakage (0.25) to limit the increase in secondary activity (0.25)."

Question 6.05 Facility Comment-Also acceptable - charging will continue via normal charging, as well as-the RCP seals. Also, full credit should be given for stating the loss of instrument air results in letdown being isolated. No specific valves should be required. Since many air operated valves fail closed in the letdown flow path - such as letdown containment isolation valves, letdown orifice isolation valves, letdown level control valves LCV 459 and LCV 460.

Ref: M-22BG01 and M-22BG03 P&ID NRC Response l

I Comment accepted. Answer key modified.

Question 6.07 Facility Comment Answer key Part B should not have ammonia - ammonia is added for pH.

Answer key Part C full credit should be given for downstream of the feedwater isolation valve. Due to this connection being made in Area 5, while the check valve is in containment.

Ref: Plant Chemistry Fundamentals A-13 Lesson Plan, Page 20 M22AE-00002 NRC Response Comment partially accepted. Part b modified as requested. The reference submitted with the comment does not support the requested change for l Part C regarding " Area 5." The candidates were cautioned to use abbreviations and plant specific terminology only if it was common in facility literature. However, credit will be allowed if candidate states that the connection is downstream of the isolation valve and outside containment. Answer key for Part C remains unchanged.

i Question 6.11 ;

i Facility Comment 1

i Credit should be given for discussion for the basis of automatic rod control handling a 10% step change or 5% per minute ramp, without causing 1

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( . _ - _ _ _ - _ _ -

c a pressure relief actuation.

Ref: Rod Control Lesson Plan S-26, Pg. 26 NRC Response Comment accepted. Answer key expanded.

Question 6.13 Facility Comment Question C should accept RCS integrity due to question stating "The average RCS temperature is 540 F and decreasing." If the cooldown continues at this rate, RCS integrity could be challenged.

Ref: Critical Safety Function Station Tree Integrity NRC Response i Comment partially accepted. Answer key expanded to include "If the cooldown continues at this rate (.25), RCS integrity could be challenged (0.25)."

Question 6.14 Facility Comment 1

Answer B.2 should accept RCP under voltage. Answer B.3 should accept RCP under frequency.

Ref: L.P. Reactor Protection S-27 Student Handout, Figure 7 NRC Response Comment accepted. Answer key expanded.

Question 6.15 Facility Comment Also acceptable: (1) Seal between stainless steel seal ring and the )

reactor pressure vessel flange, and (2) Seal between the stainless steel l seal ring and the refueling pool wall. These two inflatable seals are i distinct sealing mechanisms. Credit should be given for each. Also acceptable for Item 3, the gate to the new fuel elevator is synonymous with cask loading pit. The cask loading pit contains the new fuel elevator. The cask loading pit gate has been removed and the pit flooded during both refuelings at Callaway for fuel reconstitution.

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L Ref: Reactor Building Stainless Steel Reactor Cavity Seal Ring C-2L2929, Print M-26040, Rev.1, Fuel Building Equipment Locations i NRC Response

. Comment accepted. . Answer key expanded.

Question 6.19 Facility Comment-Answer key Part A, classification of events will depend on assumption made by student as directed by the examiner during exam.

Answer key Part B, protective actions recommendations will depend on event classifications (i.e. , for Site Emergency or lower classifications, protective actions will be decided by the Emergency Coordinator). For a general Emergency, shelter and evacuation is required.

Ref: EIP-ZZ-00102, EIP-ZZ-00212 NRC Response Comment accepted. Grading of this question depends on assumptions as stated by the candidate on the answer sheet.

Question 6.20 Facility Comment Closing power fuses are also tagged for large 4160 volt breakers and should also be acceptable.

Ref: APA-ZZ-00310, 4.1.15 e NRC Response Comment not accepted. The step referenced in the facility comment specifically refers to the isolation of switchgear, while the question l and original reference refer to the isolation of electrical motors.

I Answer key remains unchanged.

Question 6.21a Facility Comment The question could be misinterpreted to mean required full staffing vice minimum required personnel in Control Room. An acceptable answer for Part A~should be an R0 and SR0 for full staffing or R0 for minimum staffing.

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NRC Response Comment not accepted. The question was very specific in asking for the

. minimum' staffing requirement. Answer key remains unchanged.

Question 6.21b Facility Comment The minimum local staffing requirement implies Callaway administrative staffing requirements per APA-ZZ-00010, 6.1. This answer should also be acceptable per part for Part B.

NRC Response Comment not accepted. Reference not submitted witti facility comment and is not available in original reference material submitted. Answer key remains unchanged.

Question 6.22 Facility Comment Federal limits apply to answer key only. Student answers will vary depending on using federal limits or local Callaway limits.

Ref: HDP-ZZ-01400, 10 CFR 20.101 NRC Response Comment partially accepted. Answer key modified to include local facility limits.

Question 6.24 Facility Comment Also acceptable for Answers 1 and 2, Technical Specification 6.12.1.

L Technical Specifications, Rev. 2 Ref:

NRC Response Comment partially accepted. In place of the original answer key Part 2, the concept as described in Technical Specification 6.12.1 will be accepted.

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SIMULATION FACILITY REPORT Facility Licensee: Callaway Facility Licensee Docket No. 50-483 Operating Tests Administered At: Callaway During the conduct of the simulator portion of the operating tests, the following items were observed:

ITEM DESCRIPTION 1 Pressurizer level indication BB-LT-461 was out of service for the last two simulator scenarios.

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I y Q[*g4,.h) b b sSln NUChra#F6&E8UbATO Y COMMISSION

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. SENIOR REACTOR OPERATOR LICENSE EXAMINATION REGION 3 FACILITY: Gallaway_________________

REACTOR TYPE: PWB-Wgg@_________________

I' DATE ADMINISTERED: @2fglfgg_________________

p INDIBygIlgN5_IQ_Q6NQlD81El Use separate paper for the answers. Write answers on one. side only.

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

% OF CATEGORY % OF CANDIDATE'S CATEGORY

__YBLUE_ _19186 ___QgQBE___ _y@69E__ ______________g@IEGQBy_____________

_2dt99__ _2dtgg ___________ ________ 4. REACTOR PRINCIPLES (7%)

THERMODYNAMICS (7%) AND COMPONENTS (10%) (FUNDAMENTALS EXAM)

________ 5. EMERGENCY'AND ABNORMAL PLANT

._21t2D__ __1tb5 3 .___________ EVOLUTIONS (33%)

________ 6. PLANT SYSTEMS (30%) AND

_d2tE9__ _SSz99 ___________

PLANT-WIDE GENERIC RESPONSIBILITIES (13%)

________% TOTALS

_2@tZM__ =_ _______

' FINAL GRADE AlI work done on this examination is my own. I have neither given nor received aid.

Candidate's' Signature

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

1. Cheating on the examination means an automatic denial of your application and cou'A d result in more severe penalti es.

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) .

u ,- Use only the paper provided for answers.

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

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

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

10. Gkip at least three lines betwc-en each answer.

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

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

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

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

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

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-l c ^.- 18, When you' complete.your' examination, you shall . 1

a. Assemble-your' examination as fallows:

(1): Exam questions on top.

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

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

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

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

d. . Leave the examination area, as defined by-the examiner. If.after I l eavi ng , you are found in this area while.the examination is still

, in progress, your license may be denied or revoked.

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Drawing No. Rev. SECTION 21

' UNION ELECTRIC' COMPANY ECI-001 -47 DESIGN CONTROL PAGE' 21-1.

CLASS I ELECTRICAL CIRCUIT INDEX '

NN CLASS lE' INSTRUMENT AC POWER NN01~ (E-03NN01) 01 FEEDER FROM NK0lll VIA INV NNil

'02 FEEDER FROM NG01A CR3 VIA XNN05 08 NF039A LSELS PANEL 06 NF039B LSELS PANEL 05 SENY60A NEUTRON FLUX MONITOR AMP 16 RP053AC BOP INSTRUMENT RACK 20 RP053AC BOP INSTRUMENT RACK 22 RP081A SUBCOOLING MONITORING CABINET I 03 SA036A ESFAS CABINET.

I 10 SB029A SSPS TRAIN A INPUT TO RX PROTECTION 12 SB029D SSPS TRAIN A-#2 OUTPUT CABINET 14 SB030A SSPS TRAIN A #1 TEST 09 SB032A SSPS TRAIN B INPUT TO RX PROTECTION 11 SB038 PROCESS PROTECTION SET 1

.18 SB078 RVLIS PROCESS CABINET ~

13 SE054A NIS 1 15 SE054A NIS 1

'04 SA066A STATUS IND CABINET 07 SENY60B. NEUTRON FLUX MONITOR SIGNAL PROCESSOR 17 SPARE 19 SPARE 21 SPARE ,

NN02 (E-03NN 01) 01 FEEDER FROM NK0211 VIA INV NN12 02 FEEDER FROM NG02A FF3 VIA XNN06 06 NF039A LSELS PANEL 1 04 NF039B LSELS PANEL 08 RP053DA BOP INSTRUMENT RACK 03 RP147A BOP INSTRUMENT RACK 05 SA036C ESFAS CABINET 10 SB029A SSPS TRAIN A INPUT TO RX PROTECTION 09 SB032A SSPS TRAIN B INPUT TO RX PROTECTION 12 SD042 PROCESS PROTECTION SET 2 17 SB148A PANEL FIRE ISOLATION 11 SE054B NIS 2 13 SE054B NIS 2 07 SPARE 14 SPARE 15 SPARE i

16 SPARE l 18 SPARE -

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j' - 18. When you complete your examination, you shall:

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a. Assembl e your examination as follows:

-(1) Exam questions on top.

, (2) Exam asds - floures. tables, etc.

l Ar.s wer pages includi ng figures which are part of the answer.

l (3) 1

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

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c. Turn an all scrap paper and the balance of the paper that you d)d not use for answering the questions.

d. Leave the ex aminat i on area,~as defined by the examiner. If after Jeaving, you are found in this area ohile the examination is s t i 'l l in progress, your license a.ny be denied or revoked.

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- At A = AN A=AO" CQu'iEBSIQuS:

1 gm/cm = 62.4 lbm/ft Density of water (20 C) = 62.4 lbm/ft 1 gal = 8.345 lbm 1 ft = 7.48 gal Avogadro's Number = 6.023 x 10 1 gal = 3.78 liters Heat of Vapor (H O) = 970 Btu /lbm 2

1 lbm = 454 grams Heat of Fusion (ICE) = 144 Btu /lbm e = 2.72 1 AMU = 1.66 x 10~ grams n = 3.14159 Mass of Neutror. = 1.008665 AMU 1 KW = 738 ft-lbf/sec Mass of Proton = 1.007277 AMU 1 KW = 3413 Btu /hr Mass of Electron = 0.000549 AMU 1 HP = 550 ft-lbf/sec One atmosphere = 14.7 psia = 29.92 in. Hg 1 HP = .746 KW 'F = 9/5 *C + 32 1 HP = 2545 Btu /hr 'C = 5/9 (*F - 32) 1 Btu = 778 ft-lbf 'R = 'F + 460

-16 1 MEV a' 1.54 x 10 Btu 'K = 'C + 273

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17. Chrome - Steel- 35.00

+ Aluminum- -118.00 Copper 223.00 Silver 235.00 Water (20 psia, 200 degrees F) 0.392 SteamE(1000 psia, 550 degrees F) 0.046 Uranium Dioxide 1.15 Helium 0.135 Zircaloy 10.0 51SCELLONEQUS_INEQBdGIlON:

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10 CFR 20 Appendix B Table I Table II Gamma Energy Col I Col II Col I Col II MEV per Air Water Air Water Material Half-Life Disintegration uc/ml uc/ml uc/ml uc/ml

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Ni-65 2.52 h 0.59 S 9x10 4x10 3x10 1x10 4 ~12 -14 -6 Pu-239 2.41x10 y 0.008 S 2x10 1x10~ 6x10 5x10

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1 10 ------

Any single radionuclides with T > 2 hr -9 -5 -10 -6 which does not decay by alpha or/2 1 3x10 9x10 1x10 3x10 spontaneous fission Neutron Energy (MEV) Neutrons per cm Average flux to deliver equi val en t to 1 rem 100 mrem in 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months 6 670 thermal 970x10 6

0.02 400x10 280 (neutrons) 0.5 43x10 30

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x sec 10 24x10 17 cm Linear Absorption Coef#icients p (cm~I)

Energy (MEV) Water Concrete Iron Lead 0.5 0.090 0.21 0.63 1.7 1.0 0.067 0.15 0.44 0.77 1.S O.057 0.13 0.40 0.57 2.0 0.048 0.11 0.3" O.51 2.5 0.042 0.097 0.31 0.49 3.0 0.038 0.088 0.30 0.47 l .

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REVIg;9y 3

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3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICA8ILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the -

succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.

3.0.2 Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.

3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ACTION shall be initiated to place the unit in a MODE in which the specification does not apply by placing it, as applicable, in:

e(s.

a.

b.

At least HOT STAN08Y within the next 6' hours, At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

c. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Where corrective measures are completed that permit operation under the ACTION requirements, the action may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for.

Operation. Exceptions to these requirements are stated in the individual specifications. 1 This specification is not applicable in MODE 5 or 6.

3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not he madi unless the conditions for the Limiting Condition for Operation This are met I without reliance on provisions contained in the ACTION requirements.

provision shall not prevent passage through or to OPERATIONAL MODES as required j to enmply with ACTION requirements. Exceptions to these requirements are stated in the individual specifications.

t

) 1 4

CAIl.AWAY - UNIi 1 3/4 0-1 .

1 L ____-__ __-_______ _ ___ - ______ _ A

]

REVIStay , , _ '

3/4.3 INSTRUMENTATION

+

3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 1

3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be'0PERA8LE with RESPONSE TIMES as shown in Table 3.3-2. 1 APPLICA81LITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLANCE REQUIREMENTS 4.3.1.1 Each Reactor Trip System instrumentation channel and interlock and the automatic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1.

'(

4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of'each Reactor trip function shall be demonstrated to be within its limit at least once per la months.

Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific Reactor trip function as shown in the

" Total No. of Channels" column of Table 3.3-1.

CAL!AWAY - UNIT 1 3/4 3-1

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TABLE 3.3-1 (Continued)

J3BLE NOTATIONS

0nly if the Reactor Trip System breakers happen to be in the closed position and the Control Rod Drive System is capable of rod withdrawal.

- The boron dilution flux doubling signals may be blocked during reactor startup in accordance with approved procedures.

The provisions of Specification 3.0.4 are not applicable.

1. Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

1. 1. Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

(1) The applicable MODES and ACTION statement for these channels noted in Table 3.3-3 are more restrictive and, therefore, applicable.

ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel .

. to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANOBY within I the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , l k b. The Minimus Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months l for surveillance testing of other channels per Specification 4.3.1.1, and

c. Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months per Specification 4.2.4.2.

ACTION 3 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level: I

a. Below the P-6 (Intermediate Range Neutron Flux interlock)

Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 ,

Setpoint; or

b. Above the P-6 (Intermediate Range Neutron Flux interlock)

Setpoint but below 10% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10% of RATED THERMAL POWER.

(

CALLAWAY - UNIT 1 3/4 3-5 Amendment No. 17

l l

TABLE 3.3-1 (Continued)

(. ACTION STATEMENTS (Continued)-

ACTION 4 - With the number of OPERA 8LE channels one less than the Minimum Channels OPERA 8LE requirement suspend all operations involving positive reactivity changes.

ACTION 5 - a. With the number of OPERABLE channels one less than the l Minimum Channels OPERABLE requirement, restore the in-operable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the Reactor trip breakers, suspend all operations involving positive reactivity changes and verify Valves BG-V178 and 8G-V601 are closed and secured in position within the next hour. ,

b. With no channels OPERA 8LE, open the Reactor Trip Breakers, suspend all operations involving positive reactivity changes and verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter, and verify valves BG-V178 and BG-V601 are closed and secured in position

, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and verified to be closed and secured in I

position every 14 days.

ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and l

{. b. The Minimum Channels OPERA 8LE requirement is met; however,

-the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months l for surveillance testing of other channels per Specification 4.3.1.1.

ACTION 7 - Deleted. l ACTION 8 - With less than the Minimum Number of Channels OPERA 8LE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by observation of the associated permissive annunciator window (s) that the intarlock is in its required state for the existing plant condition, or apply Specification 3.0.3. ,

ACTION 9 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE.

ACTION 10 - With the number of OPERABLE channels one less than the Minimum Channels OPERA 5LE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the Reactor trip breakers within the next hour.

ACTION 11 - With the number of OPERABLE channels less.than the Total Number of Channels, operation may continue provided the inoperable channels are placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . >

l

( I CALLAWAY - UNIT 1 3/4 3-6 Amendment No. 17 I _. .

. I

i

1' TABLE 3.3-1 (Continued)

I ACTION STATEMENTS (Continued)

ACTION 12 - With one of'the diverse trip features (Undervoltage or Shunt Trip Attachment) inoperable, restore it to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the affected breaker inoperable and apply ACTION 9. The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERA 8LE status.

d,.

l i

4 9

k .

CALLAWAY - UNIT 1 3/4 3-6a Amendment No. 19

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4___BE6GIRB_ESINCIELEg_JZ31_IUgBdggyN8blGS Page 4 IZ31_6NQ_CQUEQNENIg_11931_1EUNp8dENIS6$_EX8dl QUESTION 4.01 (1.00)
Will the shutdown marg.n INCREASE, DECREASE, or REMAIN THE SAME for the following conditions? (Consider each case seperately)
~~a. Load rejection f rom 100% to 90% power, control rod position lowers and T-avg equals programmed T-avg.~~
~~b. The reactor is steady state at 50% power when a small dilution occurs with control rods in automatic.~~
~~c. The reactor is steady state ,at 50% power when a small dilution occurs with control rods are in manual.~~
~~d. Reactor power holding at 50%, xenon is increasing.~~
QUESTION 4.02 (1.00)
For two equivalent positive reactivity additions to a critical reactor, will the SUR be the same, larger, or smaller at EOL as compared to BOL? Explain
~ your choice. Use of appropriate formulas is acceptable.
QUESTION 4.03 (1.00)
While operating at 100% power, the plant experiences a circulating water pump trip resulting in a turbine runback to 75% power. The reactor operator places roo control in automatic per OTO-MA-OOOO1. The rods stop stepping in with Bank C control rods at 208 steps and Bank D control rods at 93 steps.
~~a. What immediate action must be taken by the reactor operator? (.5) !~~
I
~~b. What supplemental action must be taken in order to dampen xenon oscillations? (.5) ,~~
OUEST10N 4.04 (1.50)
The curve of critical boron concentration vs. burnup has an initial steep drop f ollowed by a relatively constant concentration until approximately 2000 MWD /MTU burnup, after which concentration drops linearly with burnup.
Explain the reason for the initial steep drop AND for the level concentration out to 2000 MWD /MTU.
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) .
4 __BEGGIOB_EBINg1ELES_JZ31_IHEBdQQyN9 DIGS Page 5 IZ31_8N9_G9dEgNENI@_ fig 31_1EUNp9dENI66@_E18dl J
DUESTION 4.05 (1.50)
~~1. As the reactor core ages, Differential Boron Worth (DBW) changes. {~~
~~a. Explain TWO effects which cause this change, and~~
~~b. HOW each changes (i e. more negative, less negative,...)~~
~~2. Explain the overall effect on DBW as the core ages (ie. more negative, less negative,...) AND why.~~
, OUESTION 4.06 (1.00)
The reactor is critical in the intermediate range, BOL in Cycle III, when the reactor operator positions control rods to achieve a 1 DPM start-up I rate. Assuming no f urther operator action, at what level will reactor power stabilice? State all assumptions and show all work.
1 QUESTION 4.07 (1.00)
(
.Which DNE of the following prevents DNB from occurring during normal plant operations?
~~a. Minimum temperature for criticality (550 degrees F) i~~
~~b. Heat Flux Hot Channel Factor~~
~~c. RCS pressure safety limit (2735 psig) 1~~
~~d. Enthalpy Rise Hot Channel Factor f OUESTION 4.08 (1.00)~~
~~Which ONE of the following is NOT characteristic of fast neutron irradiation on reactor vessel metals?~~
~~a. Increased strength~~
~~b. Decreased ductility~~
~~c. Nil-ductili ty Temperature (NDT) decreases.~~
l 1
1 1 d. A change in the lattice structure of the metal.
1 I
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) ,
Sz__BE8CI98_EBINCIELES_f231_IdESdggyN9 DIGS Page 6 ]
IZ31_8Np,ggdEQNENIS_figZl_IEUNQ8dENI66@_EK8dl 1
l QUESTION 4.09 (1.00)
For the following changes in plant status, indicate whether the DNBR-in the core will INCREASE, DECREASE, or REMAIN THE SAME. Consider each change separately, and assume all other parameters remain unchanged. The reactor is initially at 50% power.
~~a. Increase reactor power~~
~~b. Increase CVCS charging and letdown.~~
~~c. Increase pressurizer pressure~~
~~d. Increase core inlet temperature OUESTION 4.10 (1.00)~~
Assume that during the last ref ueling outage, impulse pressure to the rod control system was calibrated, so that Tref reads five degreec higher than it normally would read. What effect will this have on secondary plant
' efficiency? Why? Assume rod control and EHC are in automatic.
QUESTION 4.11 (1.00)
Why shoul d verification of natural circulation conditions be performed on a TRENDING basis versus an individual set of readings?
s DUESTION 4.12 (1.00)
The reactor is in a no-load condition with average temperature at 557 degrees-F and PZR pressure at 2250 psig prior to start-up. Calculate the subcooling margin using the steam tables (provided).
1 OUESTION 4.13 (1.00)
Describe THREE means used to minimize water hammer in steam or water pipes.
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) .
Page 7 4t__BE8GI9B_EBINGlELE@_12Zl_ISEBdQDYN8 DIGS IZZl_8ND_GgdEgNENI@_11Q31_1EUND8dENI66@_Ef8dl OUESTION 4.14 (2.50)
Answer each of the f ollowing regarding temperature detection circuits:
a) If an RTD fails OPEN, what happens to indicated temperature? (.5) b) If an RTD SHORT CIRCUITS, what happens to indicated temperature? (.5) c) If a thermocouple fails open, what happens to indicated temperature?
(.5) d) If a thermocouple fails due to a SHORT CIRCUIT, what happens to indicated temperature? (.5) e) Which of the two temperature detection elements, RTD or thermocouple, develops its own electrical output? (.5)
QUESTION 4.15 (1.00)
Safety valves are used in the main steam system to protect the the system ;
' piping from damage due to overpressurization. Which of the following i statements concerning the operation of safety valves is correct? (CHOOSE ONE) l
~~a. When the activating pressure for a safety valve returns to the lifting !~~
~~set point, a combination of steam and air pressure above the valve disk closes the valve.~~
~~b. As steam pressure increases to the safety set point, the pressure overcomes spring tension on the valve operator, causing the valve to open.~~
~~c. As the disk on a safety valve lifts, less pressure is exerted on the disk, reducing the upward force on the disk, preventing " valve sl ammi ng " .~~
~~d. A safety valve is lifted cff its seat, then is forced fully open by an air-operated piston.~~
(4**** CATEGORY 4 CONTINUED ON NEXT PAGE *****) .
4 t__BE@CIgB_EBINCIE6gS_JZZl_IdgBdggyN@ DIG @ Pace 8 ]
lZ31_6ND_GQUEQNENIg_flQ31_JEUND@dgNI66S_gf@dl {
DUESTION 4.16 (1.00)
Which ONE of the f ollowing statements is TRUE f or the MIXED-BED demineralizers?
~~a. The mixed-bed demineralizers are designed to remove only ionic isotopes-and not particulate.~~
~~b. Maximum flow rate is limited to 75 gpm to prevent resin bed channelling.~~
~~c. An unsaturated mixed-bed demineralized can decrease RCS boron concentration by approximately 100 ppm.~~
~~d. A mixed-bed demineraliter can operate for only two months with 1%~~
failed fuel.
QUESTION 4.17 (1.50)
A pressurizer level instrument develops a leak in its reference leg.
' Explain HOW and WHY pressurizer level indication changes for this instrument.
QUESTION 4.18 (0.50)
Why is a starting duty limit (i e. restarting motor, consecutive starts) placed on the Reactor Coolant Pumps?
QUESTION 4.19 (1.00)
Reactor Bypass breaker A is racked in and closed for testing with Reactor Trip breaker A open. Reactor Bypass breaker B is also racked in. Describe the sequence of actions that occur when bypass breaker B is closed.
QUESTION 4.20 (1.50)
What is the purpose of TCV-129 (CVCS demineralized inlet divert valve)?
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) .
4z__850GIRB_EBIURIELES_1ZZl_ISEBdQQYN6d1G@ Page 9 1231_0ND_G9dEQUENI@_11931_1EU$p@dEUI@6@_@f8dl OUESTION 4.21 (1,00)
A centrifugal pump with a variable speed motor uses 2 KW to develop a discharge pressure of 50 psi. It is desired to increase the discharge pressure of the pump to 100 psi by changing the pump speed. How much power will the motor require to develop the increased discharge pressure?
i I
)
(***** END OF CATEGORY 4 *****)
l
I Page 10 gz__gdEB@gNCY_@ND_G@NgBdB6_P6@NI_EyO6UIIgNS ll2*41 l
l OUESTION 5.01 (3.00) h For each of the following cases, state whether or not IMMEDIATE boration
(.25 ea) is required:
~~a. Two control rods f ail to insert on a reactor trip~~
~~b. Reactor is in mode 6 with boron concentration of 1900~~
~~c. Reactor power is 100% with control bank D at 78 steps~~
~~d. Delta I is at -17. Target delta I is -3~~
~~e. RCS temperature is at 551 degrees F and decreasing with a steam generator faulted not requiring a safety injection~~
~~f. Reactor is in mode 3 with Keff of .98~~
~~g. During a startup, the reactor goes critical with the " Rod Bank Lo Limit" alarm lit~~
~~h. During a reactor shutdown, a steam dump fails open and will not shut, resulting in a cooldown QUESTION 5.02 (2.25)~~
~~Concerning'the Emergency Procedures, explain why each of the following statements are FALSE: (0.75 each)~~
~~a. In general, a required task as stated in a procedural step MUST be completed prior to proceeding to the next step.~~
~~b. The procedures are applicable only in modes 1 through 3.~~
~~c. The onl y entry points into the EP's are through E-0, ECA O.O and FR-S.1.~~
QUESTION 5.03 (3.00)
FR-S.1, " Response to Nuclear Power Generation /ATWS," immediate action Step 4 has you initiate immediate baration. Describe the FOUR methods used in order to initiate immediate boration.
(***** CATEGORY 5 CONTINUED CN NEXT PAGE *****) .
~ - -- -- _ - _
5 __EdEBGEugy_8dD_8BUQBd86_B68dI_Eyg6UIlgNS Page 11 I?IZ1 QUESTION 5.04 (1.50)
Reactor vessel upper internals are installed in the reactor and the head is off. A man is locally monitoring water l evel in the cavity and level remains above the reactor vessel flange throughout a pump-down operation using the RHR pumps.
Explain why cavitation may occurr in this situation, despite the fact that cavity water level remains above the reactor vessel flange.
QUESTION 5.05 (2.00)
Indicated pressurizer level is not always an indication of actual RCS inventory. Several events have occurred in the industry where actual RCS inventory has been less that that indicated by pressuri=er l evel . Assuming that pressurizer level instrumentation is functioning correctly, describe TWO conditions that can cause this phenomenon to occur.
buESTION 5.06 (1.50)
During a refueling outage, while fuel is being transferred from the core to the spent fuel pool, an SRO in the control room is informed that the specific activity of the primary system is 110/E bar microcuries per gram.
The Tech Spec limit is 100/E bar microcuries per gram. What is the bases for the Tech Spec limit? (1.5)
QUESTION 5.07 (1.50)
Fill in the blanks:
A fire loop leak results in low water inventory in the fire protection system. The jockey pump fails to automatically start at a level of 2.5 f inches bel ow the center of the ______ ______. Investigation indicates that a malf unction has occurred in the control circuits. As a result, the
______ ______ ______ may be used to start the jockey pump.
I i
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) ,
l
Qz__EDEBGEUGy_6ND_8pNQBd86_B68SI_EVg6UllON@ Page 12 13331 QUESTION 5.08 (3.00)
A recent LER describes an event where an aparator inadvertently closed valve EJ-HV-8716B (RHR train B hot leg recire isolation valve). The
. operating crew entered Technical Specification 3.03 for this event.
Explain what the basis is for entering Tech Spec 3.03 under these conditions.
DUESTION 5.09 (2.00)
Following a reactor trip and uncontrolled depressurization of all steam generators, attempts are made to re-establish a secondary pressure boundary and control feedwater flow. List or describe FOUR reasons why it is important for the operator to manually control auxiliary feedwater flow in this situation.
QUESTION 5.10 (1.00)
During 100% power operations, a control rod drops into the core. In preparation to recover the rod, the step counter for the affected group is manually reset to nero and the auto-manual switch at the Pulne-to-Analog ,
Converter cabinet is held in the manual positon. Why i s the auto-manual switch held in this position?
DUESTION 5.11 (1.00)
A large break LOCA occurs inside containment, which results in the pressure in containment exceeding the design pressure limit. Besides the initial ;
high containment pressure due to the energy release from the LOCA, what I
other challenge to containment integrity potentially exists during this accident?
l l
OUESTION 5.12 (2.00)
FR-C.2, " Response to Degraded Core Cooling", requires that RCP's be kept running even if normal conditions for running RCP's are not established or l maintained. What is the basis f or keeping RCP's running under these conditions?
i
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Page 13 5t__EUgBGEUCy_6UD_6BUOSU66_E68UI_Eyg6UllgNS S;;%L QUESTION 5.13 (1.50)
Results of eddy current inspection of RCCA rodlets at Callaway has shown two effects which were not taken intb account during safety analysis.
These are:
~~1. swelling of the rodlets caused by haf nium hydriding, and~~
~~2. indications of cladding wall loss caused by flow induced vibration fretting of the rodlets against reactor upper internals.~~
List or describe TWO potential safety issues concerning the mechanical functionality of the RCCA's which were not addressed in the initial safety analysis. based on these inspection results.
QUESTION 5.14 (1.50)
Assume that the reactor has tripped, and one control rod has failed to insert into the core. How is the actual shutdown margin affected, as compared to the calculated shutdown margin?
QUESTION 5.15 (1.00)
During a loss of RHR flow, the subcooling margin falls to 40 degrees F while the RCS is filled.
~~a. If the RCS is solid, what actions must be taken to increase the subcooling margin to a minimum of 50 degrees?~~
~~b. If the pressurizer has a bubble, what actions must be taken to increase the subcooling margin to a minimum of 50 degrees?~~
DUESTION 5.16 (1.50)
A steam gener ator blowdeen high radiation monitor elarm (channel 32) is received as a result of a steam generator tube leak.
(.75)
~~a. What automatic actions will occur as a result nf the al arm~~
b. Step 6.2.3 in OTO-BB-OOOO1, " Steam Generator Tube Leak", instructs the operator to beg 2n depressurizing the RCS to approximately 1000 psig(.75) and 500 degrees F. What is the basis for this step?
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Y Page 14 5t__EMEBGEdGY_8ND_6BU9Bd86_ELONI_E296U11985 I??Zl-DUESTION 5.17 (1.00)
An event is occurring in the plant which results in the f ollowing indications:
LP turbine exhaust pressure INCREASING LP turbine exhaust hood temperature INCREASING Gener ator output DECREASING
~~a. Is condenser vacuum INCREASINGs DECREASING, or REMAINING THE SAME?~~
~~(.25)~~
~~b. List or describe THREE probable causes for this event. (.75)~~
~~QUESTION 5.18 (1.00)~~
~~a. Per Technical Specification 3.4.6.2, "RCS- Operational Leakage", what is the daily limit f or leakage through any one steam generator? (.25)~~
~~b. What i s the basis for the daily limit on steam generator leakage? (.75)~~
(***tt END OF CATEGORY 5 *****) 1
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LQUESTION. 6.01- (3.00)-
Permissive:P-4 is actuated when a reactor trip breaker.and-its associated
' bypass breaker-are both opened. Which FOUR of the following are functions
~~that.P-4 performs?~~
~~'1. Actuates'a main 1 turbine ~ trip.~~
~~2. Main 'f eedwater isolation coincident with an SI or-hi-hi S/G 1evel.~~
~~3. Auxiliary Feedwater actuation on 2/4 lo-lo S/G 1evels -in' any S/G.~~
~~4. Locks in circuit.to prevents re-opening the MFW valves which were closed by either an SI'actuatt.on or a.hi-hi'S/G 1evel.~~
~~5. Provides a signal to the Safety-injection block and reset logic circuit to prevent re-actuation of. auto SI after'a manual.SI reset.~~
6 .. Actuate turbine trip steam durnp controller in Pressure mode.
. '7 a Actuate turbine trip steam dump controller in Tavg-mode.
'O.- Causes'. turbine intercept valves to shut and MSR drain valves to open.
~~9. Actuates a main generator trip. ,~~
QUESTION 6.02 ( 1.' 00)
. List ~or describe 2 controls or indications that.use an auctioneered LOW temperature signal as.an input.
r DUESTION 6.03 (2.00)
Rod Control is in manual and.the disconnect switch for the movable grippers in the power cabinet is open'in preparation for fuse replacement when the RD^ attempts to move rods. A reactor trip results.
State what the cause of the reactor trip i s, AND explain why the trip occurs.
A
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L E BESEONSIP161IIES_11231 j DUESTION 6.04 (1.75)
' During a reactor startup, the f ollowing indications were observ'ed on the nuclear instruments:
N31 -3x 10E+4 and increasing, positive SUR N32 - oscillating around 2 x 10E+3, oscillating SUR N35 - 2 x 10E-11 and increasing, SUR oscillating positive N36 - 1.5 x 10E-11 and increasing, SUR asci11ating positive N41 - O and steady N42 - O and steady N43 - O and steady N44 - O and steady The RO and OS decided to suspend the reactor startup.
Five minutes after the startup is suspended, source range nuclear ;
instrument channel N31 fails low.
Evaluate per Technical Specifications, what actior.s'must be taken. Include time limits. Assume that repairs will take 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to complete.
OllESTION 6.05 (2.50)
Instrument air valve V584 is inadvertently shut by an equipment operator.
This valve supplies air to the Auxiliary Building and Containment.- A reactor trip occurs as a direct result of the loss of air.
Explain why a reactor trip will occur in this instance. , Assume that reactor power is 95%, all systems are in automatic, and no operator action is taken. Include setpoints where applicable.
1 i
DUESTION 6.06 (1.50)
When the reactor is in mode 5, a loss of instrument bus NNO1 will result in i a reactor trip signal being generated. Explain why a reactor trip signal !
is generated in this case.
i 1
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.BEEE9N5121LII1ES_1153) ;
b.
Is QUESTION 6.07. t(2.00)
'a.. Why is it important to control the oxygen content of the steam generator?. (.75). i!
~~b. Assuming, after anl air leak into the. secondary system, that the source. j~~
~
~
of oxygen has been eliminated. How'can the oxygen content:of the steam generator.be reduced? (.75)
c. At what point in the feedwater piping is injection "of wet 1^ayup chemicals to steam generators made? '( . 5 )
f QUESTION 6.08 (2.50) t-While in. mode 4 operation and cooling down the Reactor Coolant System f or a refueling. outage, an inadvertent dilution occurs which results in a reduction of the shutdown margin below that required by. Tech Specs.
a';Per.
) Tech Specs, what is the minimum required shutdown margin.for this!
avent? (Mode 4); (1.0) b)'Why is it important to:ma,intain an adequate shutdown margin? Address
.all modes in your answer. (.75) .l
~
i c) Per Tech Specs, what actions should be taken by the operating crew?(.75)- !
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_ - . _ . . - - - _ _ _ _ l
l 1 6m__eLegI_gygIgdg_1gOZ1_ ego _E6edI:WIDg_GgggBig Page 18 BESEONSIB161I155_113Z1 DUESTION 6.09 (1.50) l The letdown portion of the CVCS system is aligned as follows:
VALVE VALVE NAME -VALVE POSITION LCV-460 Letdown Isolation Valve open LCV-459 Letdown Isolation Valve open HV-8149A Letdown Orifice Isolation Valve closed HV-8149B Letdown Orifice Isol ati on Valve open HV-8149C Letdown Orifice Isolation Valve closed HV-8152 Letdnwn Containment Isolation Valve open HV-8160 Letdown Containment Isolation Valve open The reactor operator attempts to close valve LCV-460, but the valve will not close.
a) What is preventing the valve from closing? (0.5) b) Why is the system designed to prevent closure of LCV-460? (1.0)
OUESTION 6.10 (2.50)
A large break LOCA occurs, which results in automatic actuation of the Safety Injection System. After approximately one day, an SRO instructs the j RO to perf orm the f ollowing. ,
i
~~1) Stop SI Pump A~~
~~2) Close HV-8821A (col d leg isolation)~~
~~3) Open HV-8802A (hot leg isolation)~~
4) Start SI Pump A a) After performing the above actions, what mode of operation is the A train of the Safety Injection in? (.5) b) What is the bases for perf orming these actions? (2.0) l .
l QUESTION 6.11 (2.00)
During 100% power operation, an 8% step load reduction occurs. The reactor operator is concerned about an increase in the RCS pressure to the point of lifting a PORV. Explain why, if all systems operate as designed, that the reactor operator should not be concerned aboct pressure increasing to the PORV lift setpoint.
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BEEEQN@l@lLillg@_11}Z1
'DUESTION '6.12 (1.50)
~~a) What reactor' trip signal is continuously-calculated usinglthe. measured pressurizer' pressure'as an input, and is designed to protect the reactor~~
. core'from'DNB conditions? (0.5) .s b). List;four measured parameters that are used to determine if DNB will
~~,. occur.' -( 1. 0 ) -~~
QUESTION 6.13 (1.50)
At'0805. a reactor trip occurs. At 0808, the average RCS temperature is 540' degrees F and decreasing. The reactor' coolant pumps are in operation
~~and steam generators are being fed by main feedwater flow.~~
v a) What.RPS signal has failed to perform its function? (0.50) b).What actions should this signal have performed? (0.50) c) Why is the excessive cooldown a concern?'(0.50)
QUESTION 6.14 (1.75)
Low flow trips f unction during a Loss of Coolant Flow accident to ensure adequate core protection.
~~a. Why are OP and OT-delta temperature trips not considered adequate~~
~~. protection f rom DNB during low flow conditions? (.75)~~
~~b. What'are the three meihods of sensing a-loss of flow accident? (1.5)~~
DUESTION 6.15 (1.50)
During a refueling outage, while fuel is being transferred from the
. reactor vessel to the spent fuel pool, a non-licensed operator discovers that the level in the spent fuel pool is decreasing. What THREE seal mechanisms may be the source of the l ea!c?
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l QUESTION 6.16 (0.50) 1 When RCS level is at mid loop and shifting of RHR trains is required, it is I recommended that the running RHR pump be stopped prior to starting the standby RHR pump. Why?
OUESTION 6.17 (0.50)
There is a precaution in OTN-GS-OOOO1, " Containment Hydrogen Control System", that states not to place the electric thermal hydrogen recombiner in service with containment volumetric hydrogen concentration at or approaching 6%. What is the basis for this precaution? l l
DUESTION 6.18 (1.00)
Describe the sequence of actions in the pressurizer pressure control system that will occur if a pressurizer PORV fails open. Continue the description until pressure remains stable. Assume that the reactor is at 50% power, all systems are in automatic, all plant parameters are at their nominal values prior to the failure, and no operator action is taken after the failure. Include cetpeints where applicable.
QUESTION 6.19 (0.75)
[ While operating at 100% power, a loss of offsite power automatic reactor
! trip / safety injection occurs. The following indications are observed:
~~1. pressurizer l evel at zero percent 1 2. RCS pressure 900 psig and decreasing rapidly~~
~~! 3. containment humidity increasing i~~
~~4. incore thermocouple increasing 4~~
~~5. containment normal and retirc sump levels increasing~~
~~6. RVLIS less than 40%~~
~~7. containment radiation levels increasing~~
~~8. containment pressure 2.0 psig Cl assi f y the event. (.25) a.~~
~~b. What protective action recommendations would you make to state and~~
(.5) local authorities?
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QUESTION 6.20 (1.50) ]
l Whil e in operati on, the B Component Cooling Water Pump Motor shorted out, I requiring the motor to be sent to an electric shop for repair. Prior to removal from the system, what actions must be taken to ensure the motor is electrically isolated and remains isolated from it's power supply?
QUESTION 6.21 (1.00)
During a refueling outage, while fuel is being transf erred f rom the vessel I to the spent fuel pool.
a) Per Technical Specifications, what is the minimum control room licensed personnel staffing requirement? (0.5) b) What is the minimum local refueling licensed personnel staffing requirement? (0.5) l l
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QUESTION 6.22 (2.50)
A condition arises which regt es entry into a high radiation area. The ,
operator entering the area will receive a whole body dose of 40 mrem. The i personnel li sted bel ow, with their related personal information, are available to do the work. Each candidate is technically competent and physically capable of performing the task. Emergency limits do not apply and time constraints do not permit obtaining authorization for an exposure limit increase. Which candidate (s) have acceptable exposure margins to perform the tauk? Indicate the reason (s) for rejecting a candidate for the job, if applicable. (2.5)
NOTE: Each exposure below (qtr, yr, life, etc.) includes the exposure above it. Assume the current quarter is the fourth calendar quarter. All expcsures are in mrem.
2 3 4 Candidate 1 Sex male male female female Age 27 38 24 20 Today's e>.posure 50 10 0 20 Wkly exposure 90 150 10 250 Otr exposure 1220 600 470 1230 Yr exposure 2200 2995 1110 2810 Life exposure -
54730 5200 9770 Remarks history form 4 3 months form 4 unavailable on file pregnant on file form 4 not on file pregnant DUESTION 6.23 (1.70)
Per 10 CFR 20, define each of the f ollowing terms:
a) Radiation A r e a ., (1.0) b) High Radiation Area. (0.7)
DUESTION 6.24 (1,80) 10 CFR 20.203 describes three methods used to control entrance or access to a high radiation area. Describe these three methods.
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~DUESTION. 6'.25 (1.25)
Par.APA-ZZ-OO802, " Confined Space Entry Permit Program", what is a
" confined space"?
QUESTION 6.26 . (0. 50)-
During normal plant operation, both -trains of control room HVAC'are inoperable and the temperature in the' control room is 86 degrees F'and--
increasing. What is the primary concern during high control room-temperature situations?
DUESTION 6.27 (2.00)
Per APA-ZZ-OO703l " Fire. Protection Operability Criteria and Surveillance Requirements":
~~a. What actions must'be t'aken if a required fire barrier i s inoperabl e - and~~
~~'there in an operable fire detector on one side of the barrier? Include (1.0) time limits.~~
~~b. 'What actions must be taken if a required fire barrier is inoperable'and there IS NOT an operable fire detector on either side of the~ barrier?~~
(1.0)
Include' time limits.
t t,
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(*****.END OF CATEGORY 6 *****)
(********** END OF EXAMINATION **********)
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J- 12Zl_6NQ_QQdEQUEUI@_119Zl_1EUNp8dgNI6Lg_EX8dl-ANSWER 4.01 (1.00)
(0.25 ea)
~~a. remain the same-b., decrease c -. decrease~~
~~d. increase~~
~ REFERENCE
. Tech Spec 1.29 192OO2K114 ..(KA's)
-ANSWER 4.02 (1.00)
Larger SUR at EOL (0.5) because Beta Bar effective decreases with increased Pu-239'(and decreased U-235) (0.5) credit given for use of SUR and PERIOD formulas provided explanation includes the effect of Beta Bar effective.
REFERENCE Neutron Kinetics Lesson Plan, II.C.5.e-g 192OO3K106 ..(KA's)
I ANSWER 4.03 (1.00)
~~a. Immediate borate (.25) due to being below rod insertion limits (.25)~~
~~b. Return control bank half way back to original position (.25) by boration .(.25)~~
REFERENCE OTO-ZZ-OOOO3 OTO-ZZ-OOOO4 192OOSK115 ..(KA's) x 4 CONTINUED ON NEXT PAGE *****) i
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-- - - - _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ ______J
1 fi__BE6GIQB_E31Ng1ELgg_J231_Ibg8dQDYN801GS Pago 25 IZ31_6MD_QgbEgNgNIg_flgZl_fEMNp8dgMI6Lg_gf8dl ANSWER 4.04 .11.50)
~~
J initie) steep drop is due to buildup of fission product poisons in a clean I core (.75) lovel concentration is due to depletion of burnable poisons (.75) l REFERENCE Reactivity & Fuel Temperature. Effects Lesson Plan, II.D.4 ,
192OO7K104 ..(KA's)
ANSWER 4.05 (1.50)
~~1. baron concentration decres.ses with core life (.25) which raakes DBW more negative (.25). Fission product poison buildup with core life~~
~~(.25) makes DBW loss negative (.25).~~
~~2. Boron concentration decrease is the dominant factor (.25) which makes DBW more negative with core life (.25).~~
REFERENCE Chemical Shim Control, II.F.5.a-d 192OO4K111 ..(KA's)
ANSWER 4.06 (1.00)
NOTE: answer will depend on assumptions made by the candidate regarding the values of beta-effective, lambda and power. defect. Acceptable ranges:
Power defect a 14 - 16 pcm/% power j Lambda = 0.1
! beta eff = .0052 - .0062 I
( tau =26.06/SUR NOTE: .25 credit for sol uti on ' i
. rho =(beta eff)/(1+1ambda* tau) NOTE: .25 credit for solution based on !
candidate assumptions l power l evel = rho / (pcm/% power ) NOTE: .5 credit for solution based on candidate assumptions
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j THERMODYNAMICS Page 26'
'IZZl_8Np_QQMEQNENIQ_J193 L1EyNQ@dgNI@(@_gX@dl a:n REFERENCE
' Curve Book, figure 6-1A B 192OOOK110' 192OOOK113- ..(KA's) 4-p ANSWER' 4.07 (1.00) 1 .
<d L
REFERENCE. ,
l
[ Westinghouse Thermo-Hydraulic-Principles, pg 13-36
.193OO9K105 ..(KA's)
. ANSWER- 4.08. (1.00)
C REFERENCE Westinghouse Thermal. Sciences, Mod.D-5, 2. 3' 193010K105- ..(KA's) ,
ANSWER- ;4.09 (1.00)
-(0.25 ea)
~~a. :. d ecr ea s e ,~~
~~b. remain the same ,~~
ci- increase-Jd.' . decrease REFERENCE Westinghouse Thermo-Hydraulic Principles, pp 13-22 to 13-24
'193Oo8K105 ..(KA's) k
' ANSWER- 4,10 (1.00) .
Secondary efficiency would increase ( . "5 ) , because steam temperature (and pressure) would~ increase allowing more energy to be obtained from each pound : mass of steam (.5)
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~~,,--x - _ - - - - - - - - - _ - - - - - - - - - - _ - - _ _ - - - - - - _ _ _~~

'a 4___BE8GI9B_EBINg1 ELE @_fZ%1_IBgBdgpyN8d1GS Page 27 IZZl_8ND_GQdEQNENIS_J1931_1EUUp8dgNI66@_EX8dl
- REFERENCE Westinghouse Thermo-Hydraulic Principles, pg 7-67
_193OO5K103 ..(KA's) l ANSWER 4.11 (1.00)
Eliminates effects of instantaneous variations in readings (Will accept: some indications of natural circulation depend on a trend in readings, not on the reading itself)
REFERENCE ERG Executive Volume, Generic Issues, Natural Circulation 193OO8K122 ..(KA's) l ANSWER 4.12 (1.00)
(654-557)=97 degrees-F (+/- 2)
REFERENCE l steam tables 193OO3K125 ..(KA's)
ANSWER 4.13 (1.00)
Any 3 9 .33 ea:
~~1. use of drain valves in steam pipes~~
~~2. steam traps in steam pipes~~
~~4. procedures to drain condensate from steam pipes~~
~~4. insure proper suction pressure to running pumps in water pipes )~~
~~5. system piping pressurized prior to operation in water pipes l~~
6. good operating procedures to avoid operation near saturation i n water 1 P i Pe5 !
~~f. REFERENCE Basic Fluid Flow. II.F.3 & 4 193OO6K104 ..(KA's)~~
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f c A N S W E P..; 4.14, (2. 50) '
'c): indicated T fails high7 (0. 5)
-.b)-- i ndi cated L T i f ail s- l ow (0.5) ic)' indicated-LT f ails low f(0.5) -
<d)l indicated T' fails low (0.5)
- e) ! T/C ' (0. 5)-
~~REFERENCE "RCS-Instrumentation, II.B.1:~~
Monitoring Critical Parameters,.II.E 2, II.E.3 191002K114 141002K113 ..(KA's)
ANSWER 4.15 (1.00)
.b REFERENCE'.
' Main' Steam Lesson' Plan, II.B.4.a.
191001K101 ..(KA's)
' ANSWER ! '4.16 (1.00)-
'c REFERENCE .
CVCS Lesson Plan, II.A.2.n.B
.191007K108 ..(KA's)
ANS'WER 4.17 (1.50)
Indicated level increases (higher than actual) (0.75) because of less mass-in the reference. leg (causing DP in the bellows to decrease) ( 0. 75 ) - j REFERE'NCE
~.RCS Instrumentation LP, II.B.11 191002K109 ..(KA's) 4 CONTINUED ON NEXT PAGE *****)
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] l ANSWER ~ 4.18 (0.50) prevent' damage (.25) to the motor windings .25)
REFERENCE OTN-BB-OOOO3, 2.8 ,
4 191005K106 ..(KA's)
ANSWER 4.19 (1.00) ,
Generai warning alarm will annunciate (.5) and both bypass breakers will trip, tripping the reactor (.5)
REFERENCE OTA-RL-RKO76 (c) 19100GK104 ..(KA's) 1 1
ANSWER -4.20 (1.50) diverts flow around the demineralized 4 75) to protect the resin from overheating (.75)
REFERENCE CVCS LP, II.A.2.m.1 191007K109 ..(KA's)
ANSWER 4.21 (1.00) head proportional to speed squared so:
if head doubles, (N squared) doubles, or speed increases by a factor of 1.414 (.5) power proportional to speed cubed so:
1.414 cubed = 2.83 or power increases by a factor of 2.83 (.25) l final p ower =i r.i t i al power
~~increase = 2~~
~~2.83 = 5.66 KW (.25) l l~~
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L 12Zl_8ND_29dEONENIS 11931_1EUND8dENI865_EX8d1 L.
. REFERENCE Pumps LP, I1.B.4 191005K104 ..(KA's) i 1
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4 Dz__EdEBEENGY_8Np_9BNQBd@6_E!.6NI_gyg6UIlgNS Psgo 31 r
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. ANSWER 5.01 (3.00)
~~c. Yes~~
~~b. Yes~~
~~c. Yes~~
~~d. No~~
~~a. Yes f.- 'No~~
~~g. No~~
~~h. Yes~~
.(.25 pts each)
REFERENCE DTO-ZZ-OOOO3 Tech Specs: 3.1.3.6, 3.9.1, 3.2.1, 3.1.1.1 FR-S.1 OOOO24A205 OOOO24K301 OOOO24K302 ..(!A's)
K ANSWER 5.02 (2.25)
~~a. Simply starting the step is sufficient.~~
~~b. Some procedures are applicable in modes 4 and 5.~~
~~c. Entr y is not allowed directly. into FR-S.1. (.75 ea)~~
I REFERENCE l
ERG Executive Volume, Users Guide, pp 5, 20-25 ;
OOOOO7G011 -000007G012 ..(KA's)
ANSWER- 5.03 (3.00) j 1._immediate baration through BG-HV-8104
~~2. alternate immediate boration through BG-V177~~
~~3. boration with the RWST~~
~~4. baration via SI initiation~~
(.75 ea)
REFERENCE FR-S.1, attachment i 000029G010 ..(KA's) t
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) .
l L
77 .
L:.
Ez__EME89ENGY_BFD_8BNQBd@(_PL@UI_gyg6pIlgNB. Pags 32
) 11}*bL h
ANSWER- 5.04- (1.50)
With the upper internals , assembly installed in the reactor. vessel, flow.is.
restricted f rom. the cavity to the reactor ' coolant hot legs. (.75)- This
~
~~f. low restriction may1cause a: vortex to form, drawing' air into the suction~~
'of the RHR pump, causing-cavitation to occur. (.75)
NOTE: credit given for concept
~~REFERENCE:~~
Reactor-Vessel & Core construction, II.A.2.e.3.b, II.A.3.b.2 Byron-LER 88-007 OOOO25K101 ..(KA's):
ANSWER 5.05 (2.00)
(any 2 of the following 9 1.0) .
J1. During cold shutdown, a combination of a vacuum formed in the PZR (.5) and gasses coming out of solution at RCS high points.(.5) may cause PZR level to indicate-erroneously high.
'2. During natural circulation cooldown (.5) a void may form in the vessel head (.5) . causing PZR level to indicate erroneously high 1
~~3. With a spray valve or.PORV stuck open (.5)'a void may form in the RCS~~
(.5) causing PZR level to' indicate. erroneously high )
Other answers accepted on a case basis' REFERENCE ERG Generic Issues, RCS Voiding OOOO28K303- ..(KA's) i c
ANSWER 5.06 (1.50) 1 J
The specific activity limit ensures that the resultingE2 hour doses (0.25) 1 at the site boundary (0.25) will not exceed an appropriately small fraction of the Part 100 limit (0.50) following a steam generator tube rupture accident (0.25) in conjunction with an assumed steady stete l primary-to-secondary steam generator leakage rate of 1.0 gpm (0.25). ]
REFERENCE Tech Spec 3.4.8 g 000076G004 ..(KA's) ,
1
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) .
--a_,__u. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _
'5___EME8GENQY_GUQ_BgNQBU86_E6@NI_gyg6911gNg Page 33 I?IZ1 ANSWER 5.07 .(1.50) accumulator tank emergency start lever (.75 each blank)
REFERENCE Fire Protection SD, pg 9 OOOO67 GOO 6 ..(KA's)
ANSWER 5.08 (3.00) closing the valve rendered both trains inoperable (.75) based on the fact that neither train coulv inject into all 4 coolant loops (.75). This violated the Tech Spec LL' action statement (of TS 3.5.2) (.75) which only allows 1 train inoperable s 75). (This required entry into TS 3.03) i REFERENCE Technical Specification 3.5.2
'Callaway LER 88-002-00 OOOO25 GOO 4 ..(KA's) l' ANSWER 5.09 (2.00)
( 5 ea)
~~1. to minimize RCS cooldown rate~~
~~2. prevent overfilling' steam generators~~
~~3. control RCS temperature when RCS cooldown is stopped~~
~~4. prevent steam generator tube dryout i~~
REFERENCE f
E-2 series-SG Tube Integrity LP, pg 6 OOOO40K304 . . (KA's) i j
ANSWER 3.10 (1.00) to prevent the converter from counting the additional steps (.75) as the rod is being individually withdrawn (.25)
I 1
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) -
l s- - _ _. . _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _
i5: _EMEBggggy_8NQ_@BUQBUS6_EL8NI_EyQLUI1QN@ Pegn 34 13331 REFERENCE OTO-SF-OOOO3 OOOOO3A102- . . (KA's)
LANSWER 5.11 (1.00) a pressure' spike (.25) could occur as a result of hydrogen i gnition. (.5)
'The total pressure could then potentially exceed the strength of containment. (.25)
REFERENCE Control Board Certification, D-9, Containment FR-Z OOOO69K301 ..(KA's)
ANSWER 5.12 '(2.00)
RCP's will provide some forced single or two phase cooling to the core (.75)'
Stopping the RCP's under these conditions (.5) may result in inadequate
' core cooling (.75)
REFERENCE FR-C.2 background document, Step 4 Note OOOO55K302 ..(KA's)
ANSWER 5.13 (1.50)
~~1. RCCA insertion capability (based on swelling) or RCCA SCRAM time (based on swelling) (.75)~~
~~2. RCCA structural integrity (based on wall loss) (.75)~~
REFERENCE Callaway_to NRC JCO dated 12/06/88 Tech Spec Table 3.1-1 000005G002 ..(KA's)
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____m.____.._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
,c5t__gMERGENCY AND'ABNgRMA6_ PLANT _Eyg6UTIQNE- :Page 35 b 3333L I
I 1
{
'ANS'WER 5.14 (1.50)
No' changes (.5), since the calculated shutdown margin. assumes the most-reactive rod stuck out of the core. (1.0) (Will:-accept: actual shutdown-margin is creater than calculated if candidate assumes that a rod other than the most reactive rod is stuck out.)
REFERENCE
~
. Tech Spec 1.29 000005G003 ..(KA's)
' ANSWER 5.15 (1.00)
~~a. raise pressure (.25) by increasing charging flow ( . 25)'~~
(accept:- decreasing letdown flow (.25))
b.; raise pressure (.25) by cycling the backup heaters'(.25)
. REFERENCE
'OTO-EJ-OOOO1 OOOO25K101 ..(KA's)
ANSWER 5.16 '(1.50)
~~a. 1. Steam generator blowdown isolation valves will close ( . 25) -~~
~~(accepts blowdown & sample process isolation ('. 25) )~~
~~2. Steam generator blowdown discharge pumps will stop (.25)~~
L 3. Steam generator blowdown dishcarge pump discharge valves will '(.25) close j
~~b. to: increase subcooling margin (.25) and to reduce primary to secondary leakage (.25) in order to limit the increase in secondary side activity (.25)~~
REFERENCE OTO-BB-OOOO1 OOOO37K305 OOOO37K308 ..(KA's) l
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) .
f Di__EDESQgNGy_@ND_@ENQEd@L_EL@NI_gyg6QIlgNE Pcgo 36 13551 ANSWER' 5.17 (1.00)
~~a. decreasing (.25)~~
~~b. (any 3 9 .25)~~
~~1. . circ water s;mtem malfunction~~
~~2. air removal = system malfunction~~
~~3. air leaks~~
~~4. improper number of circ water pumps running for the load~~
~~5. ;urbine steam seal system malfunction (other explanations evaluated on a case basis)~~
REFERENCE OTO-AD-OOOO1 OOOO51A201 ..(KA's)
' ANSWER 5.18 (1.00)
~~c. 500 gpd (.25)~~
.b. ensures that steam generator tube integrity is maintained (.25) in the
- Gvent of a main' steam line rupture (.25) or under LOCA conditions. (.25)
REFERENCE Tcch Spec 3.4.6.2 and basis 000037G003 OOOO37 GOO 4 ..(KA's)
(***** END OF CATEGORY 5 *****) .
._l-...___ _
Of 16___E66NI_gy@IgMS_'J}931_8dD_EL9BI WIDg_@gNgBIG_ Pega 37 gi RggPQNElE1LillgS_J1331
~~3. '~~
~~ANSWER 6.01 '(3.00)~~
'1,4,'5,7.(.75 ea)
. REFERENCE o
R; actor Protection SD, 3.2.4
.013OOOK115 013OOOK401 013OOOK412 ..(KA's)
ANSWER 6.02 (1.00) 1 - ,
Cold. Overpressure Protection System temperature error meter for rod-control system.
(0.5 ea)
REFERENCE RCS Instruinentation LP, II.C.2.g, II.C.1.c s OO1000A101 OO1000A102 ..(KA's)
ANSWER 6.03 (2.00).
The cause of ithe trip is negative rate due to_ multiple dropped rods. (0.5)
The rods drop because the movable grippers remains'de-energized (.75) and the stationary grippers de-energize when rod movement is demanded. (.75)
REFERENCE Ro'd.Contro1LLesson Plan, II.B.1.c.4.f Braidwood LER 88-016-00 OO1000K201 OO1000K103 ..(KA's) f ;
ANSWER 6.04 (1.75)
Within[1 hour (.25) commence a shutdown to be in hot standby (.25) within the next 6' hours (.25), hot shutdown (.25) within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 
(.25)'and cold shutdown (.25) within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . (.25)
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) ,
b___EbeUI_EYEIEMS_flQZl_6ND_E66NIWlpg_@ENEBig Peg 2 38 BEEEONSIB161IIEE_113Z1 REFERENCE Technical Specifications 3.3.1, 3.0.3 000032G003 ..(KA's)
ANSWER 6.05 (2.50)
Loss of instrument _ air results in letdown isolation valves (.5) failing shut
(.5) Charging will continue (.5) via the RCP seals and normal charging (.5)
With no letdown and continued charging, pressurizer level will rise.(.25) At 92%, (.25) high pressurizer level will cause a reactor trip.
(other explanations evaluated on a case basis)
REFERENCE drawing M-22KA03 078000K302 ..(KA's)
ANSWER 6.06 (1.50)
' Loss of bus NNO1 results in loss of control power to source range N31 and intermediate range N35. (.75) This results in the high flux bistables being de-energized, generating the reactor _ trip signal. (.75)
REFERENCE i
Safeguards Power SD, TP-4 Braidwood LER 88-003-00 015000K201 ..(KA's)
ANSWER 6.07 (2.00)
~~a. minimizes corrosion (.75)~~
~~b. add hydrazine (.5) to the feedwater (.25)~~
~~c. downstream of the FWIV (.25) and upstream of the check valve (.25)~~
REFERENCE l Lab Guide for Main Feedwater System, pg 27 059000K106 ..(KA's)
I l
1 l
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) .
b __ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _
62__ELONI_Syg1Ed@_J}OZ1_6Np_EL6dI;W1pg_GENgBIG Paga 39 BEEEONg1D1Lillgg_f1}Z1 ANSWER 6.08 (2.50) a) 1.3% delta k per k (1.0) b) A sufficient shutdown margin ensures that:
~~1. the reactor can be made subtritical (0.15) from all operating conditions (0.10)~~
~~2. the reactivity transients associated with postulated accident conditions (0.15) are controllable within acceptable limits (0.10), and~~
3. the reactor will be maintained suf ficiently subcritical (0.10) to preclude inadvertent criticality (0.10) in-the shutdown c ondi ti on (0.05) c) initiate and continue immediate boration (.5) until the required shutdown margin is restored (.25)
REFERENCE Tech Spec 3.1.1.1 OO4000K519 004000B005 OO4000 GOO 6 ..(KA's)
ANSWER 6.09 (1.50) a) An interlock which requires all letdown orifice isolation valves to be closed (0.5)
.b) The interlock insures that the water in the regenerative heat exchanger does not flash to steam (0.5) and damage the tubes (0.5)
REFERENCE-Chemical & Volume Control LP OO4010K403 . . (KA's )
ANSWER 6.10 (2.50) a) Hot Leg Recirculation (1.0) b) To flush boron (0.5) from the upper regions of the core (0.5) and quench any steam bubble (0.5) in the top of the head (0,5)
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) .
6___E6eNI_gygIEdg_Jgg31_6ND_E68NI;NIDE_@ENEBIC Pegn 40 BEgEQNg1316111Eg_J1]Z1
' REFERENCE SI System LP, II.F.5 OO6000K103 ..(KA's)
ANSWER- 6.11- (2.00) because the RCS system is designed with 2 spray valves (.75) that will provide (900 gpm) spray flow (0.5) to prevent pressure from reaching the PORV setpoint (.75) (f ollowing a maximum step load reduction of 10%)
(accept: auto rod control can handle 10% step or 5%/ minute ramp without g pressure relief actuation) V REFERENCE Lesson Plan / Lab Guide f or ' Reactor Coolant OO2OOOK410 ..(KA's)
ANSWER 6.12 (1.50) a) overtemperature delta T (0.5) b) 1. RCS temperature (0.25)
~~2. RCS pressure (0.25)~~
~~3. Core Power (0.25)~~
~~4. Coclant Flow (0.25)~~
REFERENCE Lesson Plan / Lab Guide for Reactor Instrumentation 012OOOK501 . . '( K A ' s )
ANSWER 6.13 (1.50) a) FWIS (0.5) b) Close feedwater isolations, (.125) main (.125) and bypass (.125) reg valves and chemical injection (.125) valves c) Cooldown is adding positive reactivity (0,5)
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) .
1
6m__ELegI_sygIgds_1;ggl_8ND_EleNIWIDg_QgNEBig. Pcge 41 BESEQN$lglLII1Eg_flgZ1 REFERENCE l RCS Instrumentation LP, II.C.2.d 012OOOK108 ..(KA's)
ANSWER 6.14 (1.75) a) delta temperature trips do not respond fast enough (.75)
(also accept: low flow through bypass manifolds) b) 1. Measured flow in the reactor coolant piping (0.5)
~~2. Sensing an undervoltage condition (0.3) on the reactor coolant pump buses (0.2) (accept RCP undervoltage)~~
~~3. Sensing an underfrequency condition (0.3) on the reactor coolant pump buses (0.2) (accept RCP underfrequency)~~
REFERENCE J
Reactor Protection System Description, 3.2.3 i 012OOOGOO4 ..(KA's)
ANSWER 6.15 (1.50) any 3 9 .5 ea
~~1. Cavity seal at reactor pressure vessel flange~~
~~2. Steam Generator Nozzle Dam Seals~~
~~3. SFP Transf er Gates to cask loading pit (accept gate to new f uel elevator)~~
~~4. cavity seal at refueling pool wall REFERENCE Fuel Handling LP, II.c.9.)~~
034000G007 ..(KA's) l I
ANSWER 6.16 (0.50)
Prevent possible loss of suction to both RHR pumps l
1
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i i
l
b 41__ELONI_My@IEd@_JgOZl_8ND_EL8NI:WIDg_GENEBIC Pegn 42 BEEEONgiglLillEg_flgZ1 REFERENCE OTN-EJ-OOOO1, 2.8 OO5000G010 ..(KA's)
ANSWER 6.17 (0.50) to prevent combustion of hydrogen in containment REFERENCE OTN-GS-OOOO1, 2.2 028000G106 ..(KA's)
ANSWER 6.18 (1.00) at 2220 psig (.125) variable heaters will be full on (.125) at 2210 psig (.125) backup heaters will energize (.125) at 2185 psig (.125) (2/4) PORV block valves will shut (.125) pressure will stabilize at 2185 psig (+/-) (.125) riding on the block valve interlock (.125)
REFERENCE RCS Instrumentation LP, II.C.5 010000 GOO 7 ..(KA's)
ANSWER 6.19 (0.75)
~~a. si te emergency (.25)~~
~~b. shelter 2 mile radius (.25) and 5 miles downwind (.25)~~
REFERENCE EIP-ZZ-OO101 EIP-ZZ-OO102 194001A116 ..(KA's)
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) ..
Page 43 6 __ELBUI_SYSIEUS_J}OZl_@ND_E6@NI: WIDE _GENESIG BEgEONgig161IlgS_11}Z1 ANSWER 6.20 (1.50)
~~1) Put handswitch on main control board panel (RLO!9) to pull to lock (0.25) and tag (0.25)~~
~~2) Put handswitch at (NB02) switchgear to stop (0.25) and tag (0.25)~~
~~3) Open or rack out supply breaker (0.25) and tag (0.25)~~
REFERENCE APA-ZZ-OO310, 4.1.16 CCW SD 194001K102 ..(KA's)
ANSWER 6.21 (1.00) a) One licensed operator (0.5) b) A licensed SRO or an SRO limited to fuel handling (0.5)
(Will accept: an RO and an SRO)
REFERENCE Tech Spec 6.2.2 194001A103 ..(KA's)
ANSWER 6.22 (2.50)
Candidate #1: Rejected (0.25). Will exceed 1250 mrem /qtr (0.25).
Candidate #2: Rej ect ed (0.25). will exceed yearly alert limit (.25) and has exceeded quarterl y alert limit (.25)
Candidate #3: Rej ect ed (0.25). Will exceed 500 mrem / pregnancy (0.25),
and has exceeded quarterly alert and admin limits (.25)
Candidate #4: Rejected (0.25). exceeded quarterly alert limit ( 25)
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) .
1 1
~~fi__PLANTigygTEMS (30Z)' AND PLANT iWIDg_ GENERIC- 'Pege 44 .~~
'RESPONSIBI61IlgS_J1331
+
t
. REFERENCE-10 CFR 20.101 HDP-ZZ-01400 194001K103 . . (KA's):
ANSWER; 6.23 ( 1. 70) -
a) Any. area, accessible to personnel, . (.1) in which there exists radiation,-originating in.whole or in part within license material, (.15)-
at such levels that.a major portion of.the body (.15) could receive in any' one hour (.15).a dose in excess.of 5 mrem, (.15) or in any 5 consecutive days (.15).a dose.in excess of 100 mrem. (.15) b) Any area, accessible to personnel, (.1) in which there exist radiation, originating in whole or in part within license material,. (.15) at such levels .that a major portion of the body (.15) could receive in.any one hour
(.15).a dose in excess of 100 mrem. (.15)
REFERENCE 10 CFR 20.202 194001K103 . . (KA's)
ANSWER '6.24 ~(1.80)
-(.6 ea)
~~1. a control device which shall cause the level of radiation to'be reduced below that at which individuals might receive a done of 100 mrem in i hour.~~
upon' entry into the area 2.1 a control device which shall energize a conspicuous visible or audible-alarm signal in such a manner that the individual entering the high radiation area and the licensee or a supervisor of the activity are made' aware of the entry (accept: barricaded and posted as a hi rad area and entrance controlled by RWP)
~~3. maintained locked except during periods when access to the area is required, with positive control over each individual entry.~~
(concept, not exact wording, required for full credit)
L REFERENCE 1'
i 10 CFR 20.203 l' 194001K103 . . (KA's) l l
l-(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) .
____.___._mm_...
Pagn 45 s___E66NI_SYSIEMS_IsOZl_BND_E60NI: WIDE _ GENESIC BESE0NSIB161IIES_11231 ANSWER 6.25 (1.25)
A space which has a good potential (.25) for.having an oxygen deficiency j
(.25) or toxic substance concentration (.25) in a workers breathing zone
(.25) or flammable' concentration (.25) in the space. I REFERENCE APA-ZZ-OOB02, 2.1 194001K114 ..(KA's)
ANSWER 6.26 (0.50) inoperability of electrical equipment (.5)
REFERENCE Tech Spec 3.7.12 194001K108 ..(KA's)
ANSWER 6.27 (2.00)
~~a. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (.5) establish an hourly fire watch patrol (.5)~~
~~b. within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (.5) establish a continuous fire watch (.5)~~
REFERENCE
'APA-ZZ-OO703, Attachment 6 194001K116 . . (KA's)
(***** END OF CATEGORY 6 *****)
(********** END OF EXAMINATION **********)
A -_m. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ , _
f, _ j, (m,. q ' " r% p% . - - ~ . , 9, r&X ((N/Ni.D 1 [NL (N.II )f U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION REGION 3 FACILITY: .Callaway REACTOR TYPE: PWR-WEC4 DATE ADMINISTERED: 89/01/23 INSTRUCTIONS TO CANDIDATE: i Use separate paper for the answers. Write answers on one side only.
Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 23.00 23.00 1. REACTOR PRINCIPLES (7%)
THERMODYNAMICS (7%) AND
. COMPONENTS (11%) (FUNDAMENTALS EXAM) 29.50 29.50 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS (27%)
47.50 47.50 3. PLANT SYSTEMS.(38%) AND PLANT-WIDE GENERIC RESPONSIBILITIES (10%)
100.0 % TOTALS FINAL GRADE All work done on this examination is my own. I have neither given nor received aid.
Candidate's Signature u
$ p...-.... .n. ,r. . .
d (< /g b ., .. E , . . (, ' / 9x-;_j g h [ . . h bp [a -ht.? O %
~~rn -, (+ c A, ,~~
Nwi Q# l 1
a4 8 ' qq .
NRC RULES'AND GUIDELINES FOR LICENSE EXAMINATIONS t JDuring the administration of this examination the following rules apply:
~~1. . Cheating on the examination means an automatic denial of your application and could result in more severe penalties.~~
2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyono 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 i appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer sheet.
~~9. Number each answer as to category and number, for example, 1.4, 6.3.~~
~~10. Skip at least three lines between each answer.~~
~~11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.~~
12. Use abbreviations only if they are commonly used in facility literature. 1
~~13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.~~
~~14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.~~
~~15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.~~
~~16. If parts of the examination are not clear'as to intent, ask questions of the examiner only.~~
17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.
r--._- - _ _ _ . - - - . _ -.
18 When you complete your examination, you shall:
~~a. Assemble your examination as follows:~~
(. (1) Exam questions on top.
i (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. ;
< , ' o
-EQUATION SHEET
- f = ma v = s/t . Cycle efficiency = (Net work out)/(Energy in) 2 w = mg s = V,t + 1/2 at 2
E = mc KE.= 1/2 mv a=(Vf - V,)/t 'A'= AN A=Ae"*
g PE = mgn Vf = V, + at w = e/t 1 = an2/t1/2 =.0.693/t1/2 W = v aP.
A= nD 2 t 1/2*" * @ M Y 4 [(t1/2)
~~II b)3-AE = 931 am m = V,yAo ,~~
-Ex Q = mCpa Q = UA'a T I = I,e~"*
Pwr = Wfah I = I,10~*/ M TVL = 1.3/u P = P 10 sur(t) HVL = -0.693/u P = P ,e*/I SUR = 26.06/T SCR = S/(1 - K,ff)
CR x = S/(1 - K,ffx)
SUR = 26p/t* + (s - p)T- CR)(1 - K ,ff)) = CR2 (I ~ Eeff2)
T = ( t*/o ) + ((a - o V io ] M = 1/(1 - K,ff) = CR)/CR, T = t/(o - a)- M = (1 - K ,ff,)/(1 - K,ff))
l T = (8 - o)/(Io) SDM = ( - K ,ff)/K,ff a = (K ,ff-1)/K ,ff = AK ,ff/K,ff 1* = 10 seconds I = 0.1 seconds-I o = ((t*/(T X,ff)] + [i,ff /(1 + IT)]
Idjj=Id P = (I6V)/(3 x 1010) I)d) 2 ,21d 2 22 2 I = oN R/hr = (0.5 CE)/d (meters)
R/hr = 6 CE/d2 (f,,g)
Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lem. 1 curie = 3.7 x 1010 eps 1 ga]. = 3.78 liters 1 kg = 2.21 lbm .
1 ft4 = 7.48 gal. Inp=2.54x103Stu/nr Oensity = 62.4 lbm/f:3 1 m . 3,41 x loo Stu/hr Density = 1 gm/cm 3 lin = 2.54 cm Heat Of vacornation = 970 5tu/lem "F = 9/5'C + 32 Heat of fusion = ila 3:u/ltm 'C = 5/9 (*F-32) ,
1 Atm = 14.7 asi = 29.9 in. Hg. 1 BTU = 778 ft-lbf 1 ft. H 2O = 0.4335 lbf/in.
1
\
i.# REACTOR PRINCIPLES (7%) THERMODYNAMICS Page f4' '
(7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM) i QUESTION 1.01- '(1. 00) .
MULTIPLE CHOICE (Select the best answer)
Within minutes following a reactor trip from 100%' power, Startup Rate (SUR) attains an equilibrium value of approximately -1/3 dpm. Which ONE_of the following is responsible for this SUR value?
< a. The shortest-lived group of delayed neutron precursors.
~
~~b. The longest-lived group of delayed neutron precursors.~~
~~c. Spontaneous fission of U-235.~~
~~d. Spontaneous fission of PU-239.~~
u I
1
~~j. -~~
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l-
i RdCTOR" PRINCIPLES (7%) THERMODYNAMICS Page' 5-(7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)-
I! QUESTION 1.02 .(1.00)
MULTIPLE CHOICE _(Select the best answer)
Which ONE of the following. conditions will cause.the differential
!. worth'of a-given1 control rodsto DECREASE?
.a._ _An adjacent burnable poison rod depletes.
~~b. Moderator temperature is increased.~~
~~c. Boron concentration is decreased.~~
~~d. An adjacent control rod is' inserted to'the same height.~~
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El. REACTOR PRINCIPLES (7%) THERMODYNAMICS- Page -6 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
QUESTION 1.03 (1.00)
' MULTIPLE CHOICE (Select the best answer)
During a reactor startup, a reactivity addition by rod withdrawal causes stable count rate to increase from-200 cps to 400 cps. A second' reactivity addition causes stable count rate to increase from 400 cps to 800 cps.
Which one of,the following statements is CORRECT?
~~a. The first' reactivity addition was larger,~~
~~b. The second reactivity addition was larger.~~
~~c. The first and second reactivity additions were equal amounts.~~
~~d. The stabilization times are needed to determine the ~~~
relationship of the reactivity additions.
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'1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 7 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
, QUESTION 1.04- (1.00) .,
Assume Plant X is.near-the beginning'of its fuel cycle, and Plant Y, with an identical fuel load, is near the end of its fuel cycle. Both Plants are being started up after 2-week shutdown periods.
Both Plants are stabilized at 50 percent power during their startups with rod control in Manual. At this point, one shutdown bank control rod with identical reactivity worth drops (fully inserts) at both Plants. Assume that NO operator action is taken and that NO Reactor Protection System (RPS) action occurs.
Which Plant will stabilize with the HIGHER Tavg and WHY?
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1.REhCTOR PRINCIPLES (7%)' THERMODYNAMICS Page 8 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
QUESTION 1.05 (1.00)
Assume Plant X is near the beginning of its fuel cycle, and Plant Y, with an identical fuel load, is near the end of its fuel cycle. Both Plants are being started up after 2-week shutdown periods.
After criticality data has been taken at 1 X 10 E-8 amps at both Plants the operators add small, EQUAL amounts of reactivity to continue the power ascensions. Assume that NO plant heating or cooling occurs.
Which Plant will have the HIGHER stable startup rate and WHY?
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E1. ' REICTOR PRINCIPLES (7%) THERMODYNAMICS Page 9 )
(7%) - AND COMPONENTS - (11%) (FUNDAMENTALS EXAM)
QUESTION 1.06 (1.00) p MULTIPLE CHOICE (Select the'best answer)
The plant is operating at 50 percent power with'the rod control j system in Manual. A special RCS chemistry analysis requires a 4 '
degree F reduction in Tavg to be achieved WITHOUT' changing rod position or plant power (i.e., by changing boron concentration only)
Given the following initial parameters, which one of the following final RCS boron concentrations is needed to cause'the required Tavg decrease?
(Assume turbine load stays constant.)
Initial RCS boron concentration = 600 ppm Moderator temperature coefficient = 15 pcm/ degree F' Differential boron worth = 10 pcm/ ppm
,a . .594 ppm
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~~b. 597-ppm~~
~~c. 603 ppm~~
~~d. 606 ppm 1~~
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-l'I REACTOR' PRINCIPLES (7%) THERMODYNAMICS Page 10 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM) l' l
QUESTION 1.07 (1.00)
MULTIPLE CHOICE.(Select the best answer) r After operating at 30 percent power for several days, the plant is ramped to_40 percent power over a.one hour period. Power, RCS boron concentration, and Tavg are promptly stabilized. Which one of the following describes which direction control rods will need
.to be moved during the next hour in order to maintain Tavg constant? Assume a constant power level and RCS boron
. concentration.
~~a. 'OUT, because Xenon production by direct fission yield will be temporarily dominant.~~
~~b. IN, because Xenon removal by decay will be temporarily dominant.~~
~~c. OUT, because equilibrium Samarium is proportional to power level.~~
'd . IN, because Xenon removal by neutron absorption will be temporarily dominant.
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,-5. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 11 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM) 1 QUESTION 1.08 (2.00)
List FOUR (4) metheds by which the RO ensures that the axial and radial power distributions are being maintained within Technical Specification limits. (0.50 each)
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1.' REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 12 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
QUESTION 1.09 (1.00)
MULTIPLE CHOICE (Select the best answer)
Which one of the following describes how pressurizer level indication will be affected if the associated pressurizer level reference line ruptures?
~~a. Indicated level will fail as is.~~
~~b. Indicated level will be higher than actual level.~~
~~c. Indicated level will be lower than actual level.~~
~~d. Indicated level vill be equal to actual level.~~
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I Page 13 l l '*. RECTOR A PRINCIPLES (7%) THERMODYNAMICS (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM) l,
. QUESTION 1.10 (1.00)
MULTIPLE CHOICE (select the best answer)
Which one of the following describes how critical heat flux (CHF) varies from the bottom to the top of the reactor core during normal full power operation?
~~a. Continuously decreases~~
~~b. Continuously increases~~
~~c. Increases, then decreases~~
~~d. Decreases, then increases l~~
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I.* REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 14
,7%)
( AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
QUESTION 1.11 (1.00)
MULTIPLE CHOICE (Select the best answer)
Which one of the following sets of RCS parameters is closer to the conditions which could result in Brittle Fracture of an RCS pressure boundary?
~~a. 120 degrees F, and 400 psig.~~
~~b. 120 degrees F, and 220 psig.~~
~~c. 400 degrees F, and 400 psig.~~
~~d. 400 degrees F, and 220 psig.~~
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' 1' / REACTOR PRINCIPLES (7%) THERMODYNAMICS Page'15- 'I P (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM) 1 QUESTION 1.12 (1.00)
MULTIPLE CHOICE (Select the best answer)
With the plant operating at 80% power, which one of the following parameter' changes will DECREASE the departure from nucleate boiling ratio (DNBR) (Consider each change separately)
G. An INCREASE.in coolant pressure
~~b. A DECREASE in coolant flow~~
.c. A DECREASE in average coolant temperature d '. A DECREASE in core delta-T
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ll ~ REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 16 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
QUESTION 1.13 (1.00)
MULTIPLE CHOICE (Select the best answer)
Which one of the following steam generator pressures must exist in order to maintain 200 degrees F subcooling margin in the RCS wh n RCS pressure is 1600 psig?
~~a. 235 psig~~
~~b. 250 psig~~
~~c. 265 psig~~
~~d. 280 psig~~
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' REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 17 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM) l QUESTION li14 (1.00)
MULTIPLE CHOICE ' (Select the best answer)
When control power is. removed from an electrically-operated circuit breaker, the' circuit breaker CANNOT be:
~~n. Opened-locally,~~
~~b. Closed remotely.~~
~~c. Racked in,~~
~~d. Racked out.~~
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' REACTOR PRINCIPLES-(7%) THERMODYNAMICS Page 18 (7%) AND COMPONENTS (11%) (FUNDAMENTALS ~ EXAM)
~ QUESTION 1.15- (1.00)
MULTIPLE CHOICE ~(Select the best answer)
.When a disconnect is located between a circuit breaker and its load, the disconnect MUST NOT be opened UNLESS:
1
~~a. The circuit breaker is closed.~~
~~b. The circuit breaker is open'.~~
~~l- c. The load ~is rated at greater than 300 amperes.~~
~~d. ~The load is rated at less than 300 amperes.~~
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-I.' ' REACTOR-PRINCIPLES'(7%) THERMODYNAMICS Page 19
-(7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
QUESTION 1.16 (1.00)
MULTIPLE CHOICE (Select the.best answer)
Frcquent' starting _and Stopping (several times in one hour) of a reactor coolant pump (RCP) is prohibited by operating procedures in order to:
~~a. Minimize the mechanical stresses placed on the RCP rotor.~~
~~b. Minimize the possibility of overheating the RCP motor windings.~~
~~c. Minimize the mechanical stresses placed on the'RCP impeller.~~
~~d. Minimize the possibility of seal failure.~~
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REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 20 L (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
QUESTION 1.17 (1.00)
MULTIPLE CHOICE (Select the best answer)
When' connecting the main generator to the grid, prior to closing the main generator output breaker, main generator-frequency should be:.
~~c. Slightly lower than grid frequency.~~
~~b. Matched with grid frequency.~~
~~c. Slightly higher than grid frequency.~~
~~d. Adjusted to exactly 60 Hz.~~
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_1_['IREACTOR PRINCIPLES (7%) THERMODYNAMICS Pago 21.
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(7%):AND COMPONENTS (31%)- (FUNDAMENTALS EXAM) q QUESTION 1.18 ~ (1.00) j MULTIPLE CHOICE (Select the best answer)
When a component cooling water pump is started, indicated motor current immediately goes to a value several times greater than rated current and' remains there until its circuit. breaker trips.
The pump has probably experienced:
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~~a. ,~~
Runout conditions.
b.- A shut discharge valve.
~~c. A shut suction valve.~~
~~d. A locked rotor.~~
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o (12~ REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 22 (7%) AND COMPONENTS ( 1_1 % ) (FUNDAMENTALS EXAM)
QUESTION 1.19 .(1.00)
MULTIPLE CHOICE (Select the best answer)
'A centrifugal charging pump is operating at a low-flow condition.
The downstream flow control valve begins to open. How will each of the following parameters be affected? (INCREASE, DECREASE, OR NO-CHANGE). Assume VCT level remains constant.
~~a. Pump Discharge Pressure (0.50)~~
~~b. Motor Amps' (0.50) i 1~~
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'lY 'RFACTOR' PRINCIPLES '(7%) THERMODYNAMICS .Page 23 i
t (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
QUESTION 1.20 . (1. 00) .
MULTIPLE CHOICE (Select the best answer)
Which set of parameters best describes RUNOUT conditions-for a Centrifugal Charging Pump?
-a. High' discharge pressure, high. flow, high power. demand
~~b. High discharge' pressure, low flow, low power demand~~
~~c. Low discharge pressure, high flow, high~ power demand'~~
~~d. Low-discharge pressure, low flow, low power demand .~~
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11'[ ' REACTOR PRINCIPLES (7%) THERMODYNAMICS- Page 24 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
~ QUESTION 1.21 (1.00)
MULTIPLE CHOICE (Select the best answer)
Plant power is being ramped from 30 percent to-100 percent.
Which one of'the following describes the response of STEAM FLOW
~~indication if the associated steam generator PRESSURE' detector sticks hard at the'30 percent position?~~
a ~. Indicated steam flow will be lower than actual steam flow.
b.. Indicated steam flow will be higher than actual steam flow.
~~c. Indicated steam flow will be equal to actual steam flow.~~
~d. Indicated steam flow will remain at the 30 percent flow indication.
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1 1' .' REACTOR PRINCIPLES'(7%) THERMODYNAMICS- Page 25 l (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM) l l
QUESTION 1.22 (1.00) i MULTIPLE CHOICE ; (Select the best answer) )
)
Which one of the following describes the response of RCS cold leg temperature indication if the associated RTD (resistance temperature detector) develops an open circuit?
~~a. Fail as is.~~
~~b.- Track actual temperature, but with an offset.~~
~~c. Indicate higher than actual temperature.~~
~~d. Indicate lower than actual temperature.~~
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d. ' EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 26 (27%) e
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- 2 .' ' EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 27 (27%)
I QUESTION 2.01 (2.50)
.a. State the TWo (2) Safety Limits established by Callaway-Technical Specifications. Include setpoints if applicable.
(1.50) !
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~~b. What is the basis for establishing Safety Limits? .(1.00) l l~~
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f/ EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 28 (27%)
l QUESTION 2.02 (2.00)
Lict FOUR (4) unassociated plant parameters / instrumentation that sro monitored to determine if safety injection may be reduced in accordance with E-0, Reactor Trip or Safety Injection or E-1, Loss of Reactor or Secondary Coolant. (0.50 each) i
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, /.' 'EM2RGENCY AND' ABNORMAL PLANT EVOLUTIONS Page - 29
-(27%)
~~QUESTION 2.03 (1.50)~~
~~E-0, Reactor Trip or Safety Injection provides criteria for tripping RCPs.in the event of a loss of coolant accident (LOCA). ,~~
~~n. Why must either a CCP or SI pump be running prior to stopping all RCPs? (0.75)-~~
~~b. What'is the basis.for tripping RCPs prior to RCS pressure decreasing to 1400 psig during a LOCA? (0.75)~~
)
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~~N.' ' tMERGENCY ' AND AENORMAL PLANT EVOLUTIONS Page 30 1~~
(27%)
I c-
-QUESTION 2.04L (3.50)
.The. plant is~ operating at 100% load when a turbine trip occurs
~
WITHOUT an automatic reactor trip.
a. List'the FIVE '(5) indications / parameters required to be checked to verify a reactor trip has occurred in accordance with FR-S.1, Response to Nuclear Power Generation. (1.50)
b. List the FOUR (4) operator actions that must be performed insert the control rods if a reactor trip has not occurred '
in accordance.with FR-S.1? Assume each action taken fails to trip the reactor. (2.00) 1
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2'. EMERGENCY AND ABNORMriL PLANT EVOLUTIONS page 31 L- , (27%)
I (1.75)
~~. QUESTION 2.05~~
.A' control room fire has prompted the Shift Supervisor (SS) to
' direct evacuation of the control room.
~~a. Without further guidance from the SS, what TWO (2) control board manipulations should be performed by the BOP reactor~~
~~-operator prior to leaving the control room? (1.00)~~
~~b. Why are Train B equipment / components preferred over Train A equipment components during plant recovery from the Auxiliary Shutdown Panel? (0.75)~~
~.
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' EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 32 (27%)
L QUESTION' 2.06 (3.00)
~
The plant .si decreasing load from 100% to 50% whJn a valid ROD BANK LOLO LIMIT alarm is received.-
i' a. What adverse plant condition has occurred to actuate this alarm in accordance with OTA-RL-RK081C? (0.50).
b. (List FIVE (5)-IMMEDIATE operator actions / control board manipulations'that are required in response to: receiving this alarm in accordance with OTO-ZZ-00003, Loss of Shutdown Margin; (In addition to verify validity and notify SS).
(0.50 each)
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I I.* NMtRGENCY AND' ABNORMAL PLANT EVOLUTIONS Page 33 (27%)
QUESTION 2.07 (3.00)
The plant is operating at~100% power when a station blackout (loss of onsite and offsite power) occurs. :ECA-0.0, Loss of All AC POWER, is immediately implemented, a.- ~ List THREE (3) methods by which ECA-0.0, Step 3 " Checki.f RCS is Isolated", is verified / performed? (1.50)
~~b. Why is RCP_ seal integrity threatened, and therefore a major concern during ECA-0.07 (0.75)~~
~~c. During RCS cooldown and depressurization, why does ECA-0.0 require S/G pressure be maintained above 150 psig? (0.75)~~
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2' .' EMEF.GENCY AND ABNORMAL PLANT EVOLUTIONS Page 34 (27%)
I QUESTION 2.08 (1.50) ;
The plant is operating at steady state 75% power. What are the j
i THREE (3) automatic equipment / component actions that occur directly as a result of lowering condenser vacuum (in accordance with OTO-AD-00001, Loss of Condenser Vacuum)? Assume condenser vacuum is initially 3 inches Hg abs, and no operator action is taken. Setpoints are not required. Do not include alarms or indicating light actuations. (0.50 each)
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2' .' " EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 35 7(27%)
QUESTION 2.09 (2.50)
The plant is operating at 90% power when improper switch alignment in the reactor makeup control system (RMCS) disables VCT makeup.
c. If VCT level is allowed to decrease continuously, what are the TWO (2) automatic equipment / component actuations that will occur to prevent a loss of suction to the charging pumps? (0.50 each)
~~b. What are THREE (3) immediate operator actions required in malfunction to restore VCT accordance level? with OTO-BG-0003, E (0.50 each) f Y $~~
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.' 2 . EM2RGENCY AND ABNORMAL PLANT' EVOLUTIONS page.36 (27%)'
. QUESTION. 2.10 (1.50)
Tho' plant is operating at.75% power when MFP B trips. What are the THREE (3).immediate actions required by OTO-AE-0001, Foodwater System Malfunction? ' Assume the reactor does not trip.
(0.50 each) l I
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2 .'. bMERGENCY AND-ABNORMAL PLANT EVOLUTIONS Page 37 l l (27%) 'l QUESTION 2.11 ( 2 .' 00 )
.)
The plant is operating'at 75% load when a rupture in the instrument air-header results in a constant decrease in l
'l instrument air pressure. 1
~~a. What TWO (2) immediate operator actions must be taken in l accordance with OTO-KA-00001,-Partial Loss of Instrument~~
~~. Air? (0.50 each).~~
~~b. If instrument air pressure continues to decrease, and no operator action is taken, feedwater flow to the S/Gs will decrease. List TWO (2) causes for this-decrease.(0.50 each) i t~~
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(27%) l
~
l QUESTION' e
2.12. (2.75) ,
The plant is operating at'100% load when a loss of power to PB-123-(site.. load bus feeder to circ and-service water pump ~
house) results-in an IMMEDIATE turbine runback.
a. What caused the turbine runback? (0.75) b What FOUR (4) immediate operator actions are required in accordance with OTO-MA-00001, Load Rejection?-'(0.50 each) g.
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~~2. ~bMSRGENCY AND ABNORMAL PLANT EVOLUTIONS Page.39 (27%)~~
l
' QUESTION '2.13 (1.00)
MULTIPLE CHOICE'(Select the best answer)
Pressurizer PORV PCV-456A.is leaking to the PRT at a rate of~1
.gpm.- In accordance with Technical Specifications, this is:
~~a. Controlled leakage~~
~~b. Pressure boundary' leakage~~
~~c. Identified. leakage~~
~~d. . Unidentified leakage~~
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2.* EMERGENCY ~AND ABNORMAL PLANT EVOLUTIONS Page 40 (27%)
_' QUESTION 2.14 (1.00)
MULTIPLE CHOICE (Select the best answer)
The plant is operating at steady state 60% power. Which ONE of
'tha following conditions requires an IMMEDIATE manual reactor trip?
~~n. RCP A seal injection temperature equals 138 degrees F and is increasing at 1 degree F/hr.~~
~~b. RCP B frame vibration equals 4 mils and is increasing at .5 mil /hr.~~
~~c. RCP C No. 1 seal DP equals 192 psid and constant.~~
~~d. RCP D shaft vibration equals 17 mils and is increasing at 1.4 mil /hr.~~
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- $1 ' CANT P SYSTEMS'(38%) AND PLANT-WIDE GENERIC Page 42 RESPONSIBILITIES (10%)
QUESTION 3.01 (3.50)
a. What are the normal and alternate sources of water available to the AFW pumps. (0.50 each) b.- What TWO-(2) signals / conditions must be present for the AFW pump suctions to automatically shift from the normal source to the alternate source. (0.50 each)
~~c. Downstream of the AFW flow-control valves there are 8 orifices (one per valve). What are the THREE (3) safety design bases of these orifices? (0,50 each)~~
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g3. hLANTSYSTEMS (38%) AND PLANT-WIDE GENERIC Page 56-RESPONSIBILITIES (10%).
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l QUESTION 3.15 (3.00)
Diesel generator NEO2 has just started automatically due to loss
~~of. power to its associated 4160 volt ESF bus.~~
~~a. List FOUR (4) conditions / signals that must exist for~the diesel generator output' breaker to automatically close.~~
~~(0.50 each)~~
~~b. List the TWO (2) conditions / signals, either of which will automatically trip the diesel gernerator output breaker.~~
(0.50 each)
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! 3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 57 RESPONSIBILITIES (10%)
1 QUESTION 3.16 (2.00)
The RHR suction valves from the RCS hot legs (8701A/B) cannot be opened unless the following interlock conditions are satisfied:
~~a. Containment recirc sump isolation valve (8811A/B) must be closed~~
~~b. RWST isolation valve (8812 A/B) must be closed~~
~~c. RCS pressure must be less than 360 psig~~
~~d. RHR to SI/CCP suction valves (8804 A/B) must be closed State the reason / basis for each of the above listed conditions (i.e., what is being prevented?). (0.50 each)~~
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~~3. PLANT SYSTEMS ( 3 8 % )' AND RLANT-WIDE GENERIC' Page 58 RESPONSIBILITIES (10%)~~
QUESTION 3.17 (1.00).
The plant-is. operating at 80% power when a turbine runback occurs. The plant stabilizes at 60% power and corrective actions are completed. The Operating Supervisor notifies the Chemistry D:partment to sample the RCS for Dose Equivalent Iodine-131 in accordance with Technical Specifications. Assuming that no fuel dcmage resulted from.the runback, what phenomenon could cause a large increase in Dose Equivalent Iodine-131?
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- QUESTION ' 3.18 (1;00)-
MULTIPLE CHOICE'(Select the best answer)
R The plant.is operating at 60% steady state power when loop.3 Dmita-T indication fails LOW and loop 3 Tave. indication fails HIGH. The.most probable cause for these indications is that loop . 3 : '
a.- .Thot failed high-b.- Thot-failed low c.. Tcold failed'high
~~d. Tcold failed ~ low I~~
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~~3. PLANT SYS'iEMS (38%) AND PLANT-WIDE GENERIC Page 60 RESPONSIBILITIES (10%)~~
QUESTION 3.19 (2.00)
Tho plant is operating at 90% power with a ramped load increase in progress when an OP Delta-T rod block occurs.
~~c. What other automatic circuit actuation will occur? (0.50)~~
~~b. What was the OP Delta-T value relative to the OP Delta-T setpoint when the rod stop occurred? (0.50)~~
~~c. MULTIPLE CHOICE (Select the best answer)~~
~~Which ONE of the following will LOWER the OP Delta-T setpoint? (1.00)~~
~~1. Decreasing RCS pressure from 2235 psig to 2200 psig over 10 minutes.~~
~~2. Increasing RCS Tavg from 586 to 588 over 10 minutes.~~
~~l 3. Inserting bank D control rods from 200 steps to 170 l steps over 10 minutes.~~
~~4. Increasing RCS pressure from 2235 psig to 2260 psig over 10 minutes, i~~
l
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F
3.- PLANT SYSTEMS (38%)~AND-FLANT-WIDE GENERIC Page'61' RESPONSIBILITIES (10%)
l j: ' QUESTION-3.20~ (l'. 0 0 )
Under what TWO (2) specific conditions may an unlicensed operator-
~~manipulate reactivity controls in the control room?J (0.50 each).~~
1 l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
o .
1 l 4 . . . . ...
iPLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 62
[ 3.
RESPONSIBILITIES (10%) i
.i
'i L ' QUESTION. 3.21 (1.50)
How is the position of-the following valves verified? (0.50.each) g a. Manual locked ~open valve e j
b.- Manual locked closed valve .l
! c.- Manual locked throttled valve ;
I l
1 i
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1
3 ,
~~3. PLANT' SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 63 "~~
RESPONSIBILITIES (10%)-
l l
~~QUESTION 3.22 (1.00). .;~~
Lint TWO

(2) general operator actions required if, during implementation of an operations procedure, the Reactor Operator dstermines that equipment damage may result if.the procedure ~is continued in accordance with APA-ZZ-00100, Procedure Requirements. (0.50 each) l l
l
'l
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r. . _ - . _ -_ - _ _ - _ _ _ _ ._ _ - - - _ _ - _ _ . - _ _ _ _ _ _ . . - - _ _ _- __- __ ___ _ _ _- - - _ -
l: .
-3. PLANT SYSTEMSE(38%) AND PLANT-WIDE GENERIC Page 64 A
. RESPONSIBILITIES (10%)
e
-QUESTION 3.23- (1.00) ,
MULTIPLE CHOICE (Select the best answer)
Select'the Callaway administrative WEEKLY wholebody exposure limit for a 27-year old male Reactor Operator during NORMAL power opnrations.
~~a. 300 mrem b.-- 600 mrem~~
~~c. -900 mrem~~
~~d. 1200 mrem~~
'4 L
I l *****)
j (***** CATEGORY 3 CONTINUED ON NEXT PAGE j L
1 i
- _ - - - _ _ _ _ - - _ _ _ _ - . - - _ - _ _ _ - . _ _ - _ _ _ _ - _ - _ . _ _ _ . _ _ _ _ _ _ _ - _ = _ - _ _ _ - _ _. _ _ - . - _ - _ _ _ - . - _ _ . - - _ - _ - .________-_--.__-___-__________.a
1
-3 -
IINTSYSTEMS (38%) AND PLANT-WIDE GENERIC Page 65 RESPONSIBILITIES (10%)
QUESTION .3.24 (1.00)
MULTIPLE CHOICE (Select the best answer)
Select the lowest reading on a personnel ion chamber (PIC) that requires exiting a radiologically controlled. area (RCA).
~~a. 25%~~
~~b. 50%-~~
~~.c. 75%~~
~~d. '100%~~
l I
l l
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-3.- PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 66 RESPONSIBILITIES (10%)
QUESTION 3.25 (1.00)
MULTIPLE CHOICE (Select the best answer)
A " Working File" copy of a diesel generatorLsurviellance procedure is to be.used by a Reactor Operator in the control room for several days. How often must this pocedure be verified to be a " Controlled" copy in accordance with ODP-ZZ-00009, Operations Dtpartment - Control of Documents?
~~a. -Every shift~~
b. .Every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months c.- Every 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months 
~~d. Every day I~~
L
(***** END OF CATEGORY 3 *****)
(********** END OF EXAMINATION **********) .
i o
P 1. ~ REACTOR' PRINCIPLES (7%)-THERMODYNAMICS Page 67 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)-
r
' ANSWER' 1.01 (1.00) b.
REFERENCE LP, Neutron Kinetics, Pg. 22 LP, Neutron Kinetics, LO-Q r- 3.3, 3.0, 2.9 192008K123 192003K107 192004K106- ..(KA's)
ANSWER- 1.02 (1.00)
~~d. ,.~~
REFERENCE LP,' Control Rod Reactivity Effects, LO-F L
..P, Control Rod Reactivity Effects, Pg. 7 2.8 192005K107 ..(KA's)
ANSWER 1.03 (1.00)
G.
REFF RENCE LP, Subcritical Multiplication, LO-G LP, Subcritical Multiplication, Pg. 7 3.8 192008K104 ...(KA's)
ANSWER 1.04 (1.00)
Plant Y (0.50), because it has the more negative MTC value (so it will require a smaller temperature drop to offset the control rod reactivity). (0.50)
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
.... .. x i -it ' REACTOR' PRINCIPLES ~(7%) THERMODYNAMICS . ..
Pago.68 b -(7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
!=
. REFERENCE LP, MTC,:and Total. Power Defect,JLO-J.1-E LP,'MTC,.and Total 1 Power Defect,-Pg.215
-3.5,-3.1, 3.1 . .
..(KA's) 192004K106 192008K124 . 192008K114 b ANSWER 1.05 .(1.00)
Plent if (0.50) , - because it has '-the smaller Bef f value (so its response =to reactivity will'be. quicker). (0.50)
REFERENCE LP, Neutron Kinetics, LO-H LP, Neutron Kinetics, Pg. 16 3.5 192008K124. ..(KA's) l ANSWER 1.06' (1.00) d.
REFERENCE j l
'SH, Reactivity: Variations, LO-P,Z.
i SH, Reactivity Variations, Pg. 6-21 3.6, 3.1 !
192004K108 004000A404 ..(KA's) l I
ANSWER 1.07 (1.00)
~~d. ,~~
REFERENCE LP, Fission Product Poisoning Effects, LO-F !
LP,. Fission Product. Poisoning Effects, Pg. 8 j 3.4 .i 192006K106 ..(KA's) lj
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
i I
h' _ _ _ _ _ _ _
-l . REACTOR PRINCI5LES'(7%) THERMODYNAMICS: Page 69' (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM) l
' ANSWER 1.08 (2.00)
~~0. - 1. All control rods in group. move together~~
~~2. Control rod groups are properly sequenced~~
~~3. Control rod groups are properly overlapped~~
~~4. Control rod insertion limits are' observed~~
~~5. AFD maintained within limits (any 4 at 0.50 each).~~
REFERENCE LP, Nuclear Power Distribution in a PWR Core, LO-H LP, Nuclear Power Distribution in a PWR Core, Pg. 13, 14 Technical Specifications, Bases, 3/4.2.2, 3/4.2.3 2.8, 2.8, 2.9 t
193009K107 192005K116 192005K109 ..(KA's)
ANSWER 1.09 (1.00)
'b .
REFERENCE LP, RCS Instrumentation, II.b.11 3.0, 2.6 193001K103- 191002K109 ..(KA's) m ANSWER 1.10 (1.00) a.
REFERENCE LP, Power. Distribution Limits, Pg. 10 2.9 193008K106 ..(KA's)
ANSWER 1.11 (1.00) !
a.
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
?E 2s.
' ; ... REACTOR' PRINCIPLES'(7%) THERMODYNAMICS .
Pago 70
-(7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
REFERENCE LP, Materials, Pg.-69, 35
~~LP, Materials, LO-I 3.6,-3.3~~
-193010K104 002000K518 ..(KA's)
ANSWER .'1.12 (1.00)'
b.
REFERENCE'-
LP, Power Distribution Limits,'LO-B ELP,. Power Distribution. Limits, Pg. 3,4 3.4 193008K105 ..(KA's)
ANSWER. 1.13 (1.00) b..
REFERENCE
~~Steam Tables 3.4 193003K125 ..(KA's)~~
ANSWER 1.14 (1.00) b.
n ~ REFERENCE LP, T61015C6, Circuit Breakers, Pg. 12
' 2. 8 191008K109 ..(KA'ri
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
Y
-1. ' ' REACTOR ~ PRINCIPLES ~-'(7%) : THERMODYNAMICS
~~Pago 71'-~~
(7%) AND COMPONENTS-(11%)-(FUNDAMENTALS EXAM)
L L
' ANSWER- 1.15- '(1.00) b ~. ..
[- REFERENCE-
_LP.T610615C,; Electrical Components, Pg. 26 3.0.
191008K107' ..(KA's)
~ ANSWER- . 1.16.) ( 1.' 0 0 ) -
'b.
REFERENCE'
.OTN-BB-00003,.RCPs,' Pg. 3 -
~3.0
.191005K106~ ...(KA's)-
' ANSWER' . 1.17 (l'. 0 0) c.
.' REFERENCE f.LP, . Electrical Distribution System, Pg. 18
'.LP, Electrical Distribution System, LO-M
.. :3.3
.191008K108 . .- ( KA 's ) '
ANSWER - 1.18 (1.00) d.
i 1
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L___=___=____-___--_
i
1.
-REACTOR PRINCIPLES'(7%) THERMODYNAMICS Page 72 (7%) AND COMPONENTS (11%) (FUNDAMENTALS' EXAM)
REFERENCE-1Azaociated= reference not provided by facility.
2.8 191005K101; ..(KA's)
ANSWER 1.19 (l'. 00 )
~~a. Decrease (0.50)~~
'b. . Increase. (0.50)
REFERENCE LP,1 Pumps, LO-D.1 LP, Pumps, Pg. 12
-2.5 193004K112 ..(KA's)
INSWER 1.20 (1.00) c.
" REFERENCE LP,. Pumps, LO-D.3 LP, Pumps, Pg. 13 2.7 191004K112 ..(KA's)
. ANSWER 1.21 (1.00) b.
REFERENCE SH, MFW System, LO-F SH, MFW System, Pg. 18 2.9 191002K102 ..(KA's) i
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
~~1. REACTOR PRINCIPLES (7%) THERMODYNAMICS Page'.73 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)~~
.1
' ANSWER. 1.22' (1.00) a.
.c.
REFERENCE L SH,' Reactor; Instrumentation, LO-B
'SH,: Reactor Instrumentation, Pg. 5
~~2. 9-L 191002K114 ..(KA's) l l:~~
t
(***** END OF CATEGORY 1 *****)
._-_ _ -___ - __ _ _ _ a
~~2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 74 (27%)~~
~~I ANSWER 2.01 (2.50)~~
~~a. 1. Reactor Core Safety Limit - (limit on combination of T-avg, thermal power, & pressurizer pressure for 4 -~~
~~loop operation) (0.50)~~
~~2. RCS Pressure Safety Limit - (RCS pressure shall be limited to) (0.50) < or = to 2735 psig (2750 psia)~~
~~(0.50)~~
~~b. To protect the integrity of barriers (fuel clad and RCS)~~
(0.50) and prevent uncontrolled release of radioactivity.
(0.50)
REFERENCE LP, Tech. Specs., LO-D LP, Tech. Specs. Pg 17 - 19 3.2, 2.6 010000G006 010000G005 ..(KA's)
ANSWER 2.02 (2.00)
~~1. RCS (wide range) pressure~~
~~2. In-core thermocouple (or RCS hot leg wide range indicators)~~
~~3. AFW flow (or MFW flow)~~
~~4. SG (narrow range) level~~
~~5. Pressurizer level (any 4 at 0.50 each)~~
REFERENCE LP, E-1 Series, LO-B.3 E-0, Step 25, and E-1, Step 6 3.9 000011A211 ..(KA's)
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
h.
-2. EkERGENCYANDABNORMALPLANTEVOLUTIONS Page 75 (27%)
Ic l ANSWER 2.03 (1.50) !
~~a. To ensure core heat removal. (0.75)~~
.b. To prevent excessive mass loss (increased fuel temperature, or deeper / prolonged core uncovery). ( O '. 7 5 )
REFERENCE LP, E-0 Series, LO-B.3, B.5 LP, E-0 Series, Pg. 9, 10, 11 ,
e 4.4, 4.1 000011K312 ..(KA's) 000011K314 ;
i
~~ANSWER 2.04 (3.50) j~~
~~c. 1. Rod bottom lights~~
~~2. Reactor trip breakers l~~
~~3. Reactor trip-bypass breakers i~~
~~4. Digital rod position indicators {~~
~~5. Neutron flux (0.30 each) j~~
~~b. 1. Manually trip reactor from RLO3~~
~~2. Manually trip reactor from RLO6 (BOP-console) f~~
~~3. Open feeder breakers to PG19 and PG20~~
~~4. Manually insert RCCAs (0.50 each) {~~
REFERENCE' FR-S.1, Pg. 2 LP, FR-S.1, LO-A.6
'4.5 000029G010 ..(KA's)
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
I
^ 2 -. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 76 ;
.127%) l l
l k
ANSWER 2.05 (1.75) ,
c 1. Manually- trip the reactor (at RLOO3 or RLOO6)
~~2. Close MSIVs (from RLOO6 or RLOO25)~~
~~(0.50 each)~~
~~b. Train B equipment / components can be electrically isolated~~
< from the control room (Train A equipment / components cannot be electrically isolated from control room and may be inoperable or operate spuriously). (0.75)
REFERENCE OTO-ZZ-00001, Control Room Inaccessibility, Att. 2, Pg. 1 OTO-ZZ-00001, Control Room Inaccessibility, Att. 5, Pg. 1 4.1 000068G010 ..(KA's)
ANSWER 2.06 (3.00)
~~c. A rod bank is (at or) below the (Lo-Lo) rod insertion limit (0.50)~~
~~b. 1. Open the Emerg. Borate to Charging Pump Suction Valve, BG-HV-81-04.~~
~~2. Start both BA transfer' pumps.~~
~~3. Verify Boric Acid flow (on "G-FI-183A).~~
~~4. Place the RCS M/U CTRL switch (BG-HIS-26) in STOP.~~
~~5. Ensure a CCP is running.~~
~~6. Establish 120 gpm letdown flow. (any 5 at 0.50 each)~~
REFERENCE ARP, OTA-RL-RK081C, ROD BANK LOLO LIMIT, Pg. 1 OTO-ZZ-00003, Loss of Shutdown Margin, Pg.2 LP, Off-Normal Procedures, LO-VV.1 and VV.3 4.1, 4.0, 4.3 001000K504 000024G010 000024K301 ..(KA's)
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
2 EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 77 ]
(27%)'
1
-ANSWER 2.07 (3.00)
~~a. Check Pzr PORVs closed (0.50 each)~~
~~Check letdown isolation valves closed Check excess letdown isolation valves closed~~
~~b. Seal cooling is lost (i.e. seal injection and thermal barrier CCW flow are lost). (0.75)~~
~~c. To prevent accumlator N2 injection. (0.75)~~
REFERENCE LP-ECA-0.0, LO-D, H, C LP-ECA-0.0, Pg. 4, 5,.18 ECA-0.0, Pg. 3 4.3, 4.1 000055G010 000055K302 ..(KA's)
ANSWER 2.08 (1.50) 1.
2.
Standby vacuum pump auto start (at 5" Hga) q gy, 3 gp Steam dump blocked (C-9) (at 7" Hga)
3. Main turbine trip (at 8.4" Hga) (0.50 each) 4, /4 F W pctm/* fJrks'ov6 hof REFERENCE LP, Off-Normal Procedures, LO-I.3 OTO-AD-00001, Pg. 1 2.8, 3.9 000051A202 000051K301 ..(KA's) l l
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
l:
,. .4 L- 2. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 78 (27%)
l:
. ANSWER. 2.09 (2.50)
~~c. 1. RWST supply valves (LCV-112 D&E) to che.rging pump suction open.~~
~~2. VCT outlet valves (LCV-112 B&C) close. (0.50 each)~~
~~b. 1. Ensure RMCS (BG-H15-25) Mode Sel. Switch is in AUTO.~~
~~2. Ensure RMCS (BG-H15-26) Control Switch has been turned to RUN.~~
~~3. Verify RMCS flow with the BA Blended Flow Rcdr (BG-FR-110) (0.50 each)~~
REFERENCE
-0TO-BG-00003, Pg. 2 LP, Off-Normal Procedures, LO-W . 3 ., W.4 2.8, 3.5, 3.1 004010K404 000022G010 004020K402 ..(KA's)
ANSWER 2.10 (1.50) 1.- Quickly run back turbine load to < 60%.
~~2. Start AFW pumps as required. (0.50 each)~~
~~3. Restore S/G levels to program.~~
REFERENCE OTO-AE-00001, Pg. 2 LP, Off-Normal Procedures, LO-K.4 3.2 000054G010 ..(KA's)
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
~
F., -j 4 e c. . ;
~~2. EMERGENCY AND ABNORMAL' PLANT EVOLUTIONS Page.79- {~~
~~-(27%)~~
~~ANSWER l 2.11 (2.00)~~
~~a. 1. Ensure all' compressors running~~
!=
~~2. Ensure service air hdr Iso Vlv-(KA-PV-11) ~ closes at 110 psig. (Manually close if required.)~~
(0.50 each)
( -h. "1. Condensate pump recirc. valves fail open
~~2. HDT pump discharge valves fail closed~~
~~3. HDT pump recirc valves fail open~~
~~4. Feed regulating valves fail closed (after N2 pressure depletes)~~
(any,2 at 0.50 each)
REFERENCE OTO-KA-0001, Partial Loss of Instrument Air, Pg. 2.
'LP, Off-Normal Procedures,'LO-AA.4 3.2, 2.9 000065K303 000065G010 ..(KA's)
ANSWER 2.12 (2.75)
~~c. ' Loss of circulating water pump (with turbine load > 75%)~~
~~(0.75)~~
~~b. 1. Ensure control rods are in auto and that Tavg is being reduced to match Tref.~~
~~2. Verify S/Gs are returning to programmed level.~~
~~3. Ensure pressurizer level and pressure are returning to programmed values.~~
~~4. Ensure a stator cooling water pump is running.~~
(0.50 each)
. REFERENCE l
OTO-MA-00001, Load Rejection, Pg. 2
- LP, S-2, Service Power, Pg. 8 3.7, 3.3 045000K412 000056G010 ..(KA's)
ANSWER 2.13 (1.00) c.
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
[g .2
, , ,. i .n . _
-2 ' EMERGENCY AND ABNORMAL PLANT EVOLUTIONS: Page_80 (27%)_
h, <
' REFERENCE.
LP,. Tech.~' Specs., Definitions ..
LP, Tech.--Specs., LO-C.10-32 000009G008~ ..'(KA's) t
' ANSWER' 2.14 '( 1. 0 0 )'
O.
- REFERENCE OTO-BB-00002, Pg. 3, 4
'> .LP,t Off-Normal Procedures, LO-0.4
~~3.4 000015G010 ..(KA's) i l-1' l:~~
l i
ll
(***** END OF CATEGORY 2 *****)
l ..
1
- _ - - _ _ _ _ _ _ _ _ _ = _ _ _ _ _ _ _ _ _
~~i. . "~~
~~3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 81 RESPONSIBILITIES (10%)~~
~~ANSWER 3.01 (3.50)~~
~~a. . 1. CST (0.50)~~
~~2. ESW (0.50)~~
~~b. 1. Low suction pressure (0.50) .,_~~
~'
j 2. .
AFAS present (0.50)
L Jc. 1. Limit'ctmt pressure increase from a'steamline rupture.
~~2. Limit pressure drop on AFW headers in event of feedline break.~~
~~3. Limit AFW pump runout. (0.50 each)~~
REFERENCE SH, AFW System, Pg. 7 LP, S-25, AFW System, L.O.B.G., I 3.9, 4.5, 3.1 061000K404 000054A101 061000K401 ..(KA's)
ANSWER 3.02 (2.75)
~~a. -1. MCB step counters~~
~~2. P/A converter~~
~~3. Bank Overlap Unit~~
~~4. Slave Cyclers- (any 4 at 0.50 each)~~
~~5. Master Cyclers~~
~~6. Internal memory~~
~~7. Alarm circuits~~
~~b. Ensures that control rod sequencing will be proper. (0.75)~~
REFERENCE LP S-26,-L.O.E.
LP S-26, Rod Control, Pg. 31 2.5, 3.2 001000G007 001010K402 ..(KA's)
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
1 j
s 't< 'j #
rv
.ia e.
~~3. PLANT SYSTEMS'(38%) AND PLANT-WIDE GENERIC Page 82 RESPONSIBILITIES (10%)~~
ANSWER 3.03- (3.00).
a.' 1.. Source range high flux l2 . Pressurizer low pressure
~~3. Pressurizer high level 4.- RCS low flow,-1-loop~~
~~5. RCS low flow, >1-loop~~
~~6. RCP UV~~
~~7. RCP UF-~~
~~8. Turbine trip (any 6 at 0.50 each)~~
REFERENCE SH, S-27, Rx Protection, Figures 7 & 8
-LP, S-27, Rx Protection, LO-O 3.2 012000K406 ..(KA's)
A'NSWER 3.04 (1.50) 1.- Low pressurizer pressure (0.50)
~~2. Low steamline pressure (0.50)~~
Blocked at 1970.psig (+/- 10 psig) (pressurizer pressure) (or P-11) (0.50)
REFERENCE LSH, S-27, Rx Protection, Pg. 23 SH, S-27, Rx Protection, LO-E 3.7 013000K412 ..(KA's)
ANSWER 3'.05 (1.50)
.a. 1. Pulse discriminator (0.50)
~~2. Compensating voltage (0.50)~~
'b. Gamma current is relatively negligible (or is proportional to~ power level) (0.50)
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) i I
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.. ~ )
3.- PLANT SYSTEMS'(38%) AND PLANT-WIDE GENERIC Page H
. RESPONSIBILITIES (10%)
k l l- REFERENCE SH, S-28, Excore NIS, Pg. 4, 9, 11 SH, S-28, LO-B,-C 2.7,-2.6 015000K602 015000K502' ..(KA's)
ANSWER 3.06 ( l .' 7 5 )
~~0. . 550 (+/- 1) degrees F. (0.75)~~
b. When Tave decreases,to the P-12 (low-low Tavg) setpoint, all steam dump valves will close. (0.50) As the RCS heats up above P-12 the 3 steam dump valves will reopen and shut to maintain RCS at P-12 setpoint (550 degrees F). (0.50)
REFERENCE 3 SH, Main Steam, LO-J, K
.SH, Main Steam, Pg. 27 3.0.
041020K409 ..(KA's)
ANSWER 3.07 (2.00)
O. 1. CCW surge tank level increasing. (O. )( A h 6f# )
~~2. CCW rad level increasing. , ( 0) 3 CC W fc1Mn Y2nf . < 1 VM Mf~~
~~b. 1. Individual thermal barrier HXer outlet isolation valve will close (at 50 GPM). (HV-13, 14, 15 or 16) (0.50)~~
~~2. Common thermal barrier heat exchanger outlet isolation valve (HV-62) will close (at 200 gpm). (0.50)~~
REFERENCE SH,.CCW, TP-2 P&ID # 22BB03 OTO-BB-00003, RCS Excessive Leakage, Pg. 2 2.9, 2.8 003000K404 000026A201 ..(KA's)
I
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
~~. .- ~~~
~~3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC- Page 84 RESPONSIBILITIES (10%)~~
~~ANSWER 3.08 (2.50)~~
~~n. . 1. Reduces thermal stress / shock (when spray valves open)~~
~~(0.50)~~
~~2. Helps maintain uniform water chemistry.(and/or temperature in the pressurizer). (0.50) b.. 1. D/P between spra line and-surge line. (0.50)~~
~~2. Velocity head of RCS loop flow (1&2) (acting on spray scoops). (0.50)~~
~~c. RCP D (0.50)'~~
REFERENCE SH, RCS, Pg. 26 SH, RCS, LO-D, C.7 2.7, 3.6 010000K103 010000K401 ..(KA's)
ANSWER 3.09 (1.50) lock rate = net charging - net letdown lesk rate =-(95-12) - 75 leak rate = 83 - 75 leak rate = 8 gpm- (0.50 for correct answer)
(1.00 for methodology)
REFERENCE
_LP, CVCS, Pg. 26 3.2 004020A203 ..(KA's)
ANSWER 3.10 (2.25) yt,(([ 9dW
~~a. 1. Charging flow control valve closesg(charging flow decreases).~~
~~2. O r '^tf; -i 11;.c ir^' *4nn "-'"r (LOV '.59) c' ::: 'c//d M,j~~
^
. i.11 p essu m m '.::t:r- *"*" ^** A gg ,
W each)
~~b. High pressurizer level (0.75) J. /5~~
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7 f h '4 N 3 PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 85 RESPONSIBILITIES (10%)
REFERENCE SH, Rx Inst.,fLO-K l' SH, Rx Inst., SNP-LOG-11-2 3.4 011000A210 ..(KA's)
ANSWER 3.11 (3.00) >
~~o. IN (0.50), because' rate of (sensed) nuclear power increase i exceeds rate of load increase. (0.25)~~
~~b. NO MOTION (0.50), because instrument is not'used as an input. (0.25)~~
~~c. IN (0.50), because auctioneered high Tavg exceeds Tref.~~
~~(0.25)~~
~~d. OUT (0.50), because rate of nuclear power decrease exceeds rate of load decrease. (0.25) , f, f#"dh.~~
" dl,,
REFERENCE '4 _f C t
3.2 * $* W VN
~~001010K507 ..(KA's) 1 ANSWER 3.12 (2.25)~~
~~2. Low RWST level (0.50) at 36%. (0.25) l~~
.b. To prevent contaminating the RWST. (0.50)
I
~~c. 1. Flush concentrated boron from core. (0.50) l~~
~~2. Quench void in Px vessel head. (0.50) ]~~
1 i
REFERENCE LP, SIS, Pg. 7 LP, SIS, LO-H SH, SIS, Pg. 6, 2 4.2, 4.4, 3.8 ,
000011K313 000011K312 013000K101 ..(KA's) {
j
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L' u
, 3Y ~ PLANT
SYSTEMS'(38%) AND PLANT-WIDE GENFJIC 'Page 86 p RESPONSIBILITIES'(10%)
L.
I ANSWER 3.13 (1.50)
~~1. Main,steamline' isolation signal (0.75)~~
~~2. SIS (0.75)~~
L r
REFERENCE
? 'SNUPPS Reference Block Diagram 8756D37, Sheet 22 SNUPPS Funct. Diagram, 7250D64, Sheet B 4.2 013000K101 ..(KA's)
ANSWER 3.14 (3.00)
~~a. High ctat. press (0.50) 2/4 (0.25), > or = 27 psig (0.25)~~
,b. 1. RWST isolation valves (HV-3 & HV-4) (open)
.2, . CSPs (start)
~~3. CSP isolation MOVs (HV-6 & HV-12) (open)~~
~~4. Spray additive isol. valves (HV-15 & HV-16) (open)~~
(0.50 each)
REFERENCE LP, Containment Spray System, Pg. 14, 13 LP, Containment Spray 5;' stem, LO-G, H 4.2, 4.1, 4.3 026000A301 026000A*c*)3 013000K101 ..(KA's)
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a_-____-__-_---_-___--____--______-_____-_-__--__.__-_____---_--_____ . _ _ _ - _ _ _ _ _ _ _ - - - _ _ _ _ _ _ _ -
a- u
~~3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 87 i RESPONSIBILITIES (10%1 ANSWER 3.15 (3.00)~~
I
~~a. 1. Both feeder breakers to ESF bus open~~
~~2. Master transfer switch in AUTO~~
~~3. Generator up to voltage (4.16 Ky)~~
~~4. Generator up to speed ( 471 rpm)~~
~~5. No lockout on.ESF bus (any 4 at 0.50 each)~~
~~b. 1. Engine shutdown relay, or, a) Low lube oil pressure b) High crankcase pressure c) High jacket water temperature d) Overspeed~~
~~2. Differential overcurrent relay (any 2 at 0.50 each)~~
REFERENCE LP, Standby Generation, LO - G, H
.LP, Standby Generation, Pg. 43, 44 3.9 064000K402 ..(KA's)
ANSWER 3.16 (2.00)
~~a. Prevent depressurizing RCS (or draining RCS to containment sump)~~
~~b. Prevent depressurizing RCS (or draining RCS to RWST)~~
~~c. Prevent lifting of RHR suction relief (prevent overpressurizing RHR suction piping)~~
~~d. Prevent overpressurizing SI suction piping (0.50 each)~~
REFERENCE LP, RHR, Pg. 5, 6 LP, RHR, LO-B.3 3.2 005000K407 ..(KA's)
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a- s Page 89
~~3. P:LANT SYSTEMS (38%) AND PLANT-WIDE GENERIC RESPONSIBILITIES (10%)~~
l ANSWER 3.17 (1.00)
Iodine spiking (phenomenon)
REFERENCE Technical Specification 3.4.8, Bases SH, Power Operations, LO-A.2 SH, Power Operations, Pg. 3 3.4 002000G010 ..(KA's) 1 ANSWER 3.18 (1.00) c.
REFERENCE
-RTD circuit analysis SH, Reactor Instrumentation, LO-A, B, D SH,. Reactor Instrumentation, Pg. 7, 8 3.0 016000A202 ..(KA's)
ANSWER 3.19 (2.00) a Turbine runback (0.50)
~~b. 3% below setpoint (0.50)~~
~~c. 2. (1.00)~~
REFERENCE SH, Rx Inst, Pg. 15, 16 SH, Rx Inst, LO-F,L 2.9, 3.3 2.9, 3.3 045000K412 012000K611 ..(KA's)
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r- 'v Page 89
~~3. P:LANT SYSTEMS (38%) AND PLANT-WIDE GENERIC RESPONSIBILITIES (10%)~~
~~ANSWER 3.20 (1.00)~~
~~1. Individual in training for a license.~~
~~2. Individual directly supervised by a licensed operator.~~
(0.50 each)
REFERENCE LP, Admin procedures, LO-B.3.c LP, Admin procedures, Pg. 12 2.5, 2.8 194001A111 194001A103 ..(KA's)
ANSWER 3.21 (1.50)
~~a. Attempt to move handwheel in close direction (expect movement, return to open) (0.50)~~
~~'b. Attempt to move handwheel in close direction (expect no movement) (0.50)~~
~~c. Observe indicator on valve (or locking device installed, or may be closed and reopened while counting turns - only with SS permission) (0.50)~~
REFERENCE LP, Operations Department Procedures, LO-D.4.a, b, c LP, Operations Department Procedures, Pg. 20, 21, 22 3.6 194001K101 ..(KA's)
ANSWER 3.22 (1.00)
~~1. Place system / component in a safe / stable condition (0.50)~~
~~2. Notify supervisor (0.50)~~
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s .
l4h*v Page 90
.3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC RESPONSIBILITIES (10%)
REFERENCE l
LP, Admin Procedures, Pg. 18 LP, Admin Procedures, LO-C.3 l 4.1 194001A102 ..(KA's)
ANSWER 3.23 (1.00) b.
REFERENCE HDP-ZZ-01400, External Dosimetry Program, Attachment 1 2.8, 3.3 194001K104 194001K103 ..(KA's)
ANSWER 3.24 (1.00) c.
REFERENCE HDP-ZZ-01400, External Dosimetry Program, Pg. 4 2.8, 3.3 194001K104 194001K103 ..(KA's)
ANSWER 3.25 (1.00) d.
REFERENCE LP, Operations Department Procedures, Pg. 27 LP, Operations Department Procedures, LO-G.2.d 3.3 194001A101 ..(KA's)
(***** END OF CATEGORY 3 *****)
(********** END OF EXAMINATION **********) .
1 Page 43
~~3. PL'NT A SYSTEMS (38%1 AND PLANT-WIDE GENERIC RESPONSIBILITIES (10%)~~
l I
QUESTION 3.02 (2.75)
Prior to commencing a reactor startup, the Startup Reset Pushbutton is depressed.
~~a. List FOUR (4) components / circuits that are reset when the Startup Reset Pushbutton is depressed. (0.50 each)~~
~~b. What is the purpose of depresssing the Startup Reset Pushbutton, thereby resetting the above components / circuits (0.75)~~
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3'. ' PIdNT SYSTEMS.(38%) AND PLANT-WIDE GENERIC Page'44-RESPONSIBILITIES'(10%)
QUESTION. 3.03 (3. 00).
-The plant is at 8% power during a normal reactor startup.
~~a. List 6' reactor trips which are/have been DEFEATED at.this-power. level. (0.50 each) c l~~
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L-____--_____--__-__-____-_-_____---_______-_____________________________
L 4 3*.
PIANT-' SYSTEMS'(38%)-AND PLANT-WIDE GENERIC Page 45 RESPONSIBILITIES (10%)
l QUESTION 3.04 (1.50)-
Wh'at TWO.(2) signals mur>t be BLOCKED in order to complete a plant cooldown WITHOUT receiving a safety injection signal.and at what
. pressurizer. PRESSURE can they FIRST be blocked? (0.50 for each signal, 0.50 for pressure)
~
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j
3 .' DIANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 46 RESPONSIBILITIES (10%)
QUESTION 3.05 (1.50)
~~a. How are the effects of gamma radiation minimized in the:~~
~~1. Source Range NIS (0.50)~~
~~2. Intermediate Range NIS (0.50)~~
~~b. Why isn't gamma radiation compensation necessary in the Power Range NIS? (0.50)~~
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3'.' PLhNT SYSTEMS (38%) AND PLANT-WIDE GENERIC i Page 47 RESPONSIBILITIES (10%)
QUESTION 3.06 (1.75)
a. The reactor is critical at 1 X 10 E-8' amps during a reactor startup. RCS pressure is 2235 psig and Tavg is 557 degrees F. . A malfunctioning steam dump steam header pressure controller.causes THREE.(3) steam. dump valves to fully open.
~~Assuming the. reactor does not trip, at what Tavg will the RCS stabilize. (0.75)~~
~~b. Describe the Steam Dump System action / response which causes Tavg to stabilize at this value. (1.00)~~
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r 3'. ' NLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 48 RESPONSIBILITIES -(10%) -
QUESTION 3.07 (2.00)
~~a. List TWO-(2) indications / symptoms which will be observed if a' tube leak occurs in a RCP thermal barrier heat exchanger.~~
~~Assume no alarm setpoints are reached and no automatic-actuations occur. (0.50 each)~~
b. If the tube leak continues to increase in magnitude, what are the TWO AUTOMATIC component actuations that will occur to stop the leak? (Assume no RPS or ESFAS actuations)
(0.50 each)
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I
(_ _ - - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _____
~~3. PLhNT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 49 RESPONSIBILITIES (10%)~~
~~k QUESTION 3.08 (2.50)~~
~~a. Describe TWO (2) purposes for the 1 gpm continuous spray flow through the pressurizer spray nozzle. (0.50 each)~~
~~b. Describe TWO (2) sources of the driving force for NORMAL pressurizer spray. (Do NOT consider auxiliary spray).~~
~~(0.50 each)~~
~~c. Which RCP (if the only RCP running) provides the MAXIMUM normal pressurizer spray flow? (0.50) l l~~
l l
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3 '. ' P'LANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 50 RESPONSIBILITIES (10%)
l l
_ QUESTION 3.09 (1.50)
The plant is operating at steady state 50% power. Pressurizer level is constant.
Indicated CVCS letdown flow is 75 gpm. Indicated CVCS charging flow is 95 gpm. Total RCP seal return flow is 12 gpm.
Considering only the aforementioned flow values, determine the l
not RCS leak rate in gpm. SHOW ALL WORK!
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i 1 3') PLANT SYSTEMS'(38%) AND PLANT-WIDE GENERIC Page 51-RESPONSIBILITIES (10%)-
L QUESTION 3.10 (2.25)
'The plant is operating at 100% power. The normally selected
. channel for the pressurizer level controller fails high.
~~c. List the THREE-(3) automatic component actuations or equipment changes that will be-initiated DIRECTLY by the channel failure. (Do not include alarm actuations).~~
(0.50 each)
.b. If no operator. action is taken, what reactor trip condition will eventually be reached DIRECTLY as a result of the failure. (0.75) i
~
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1
.: 3' . -*' PLINT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 52 RESPONSIBILITIES (10%)
QUESTION- 3.11 (3.00)
.The. plant is operating at steady state 80% load with all~ control cystems in automatic. Bank D control rods are at 190 steps.
Given the following conditions / situations, describe the INITIAL rod motion which occurs. Your answer should contain IN, OUT, or NO MOTION with a brief explanation (referring.to input effects).
'. Assume no operator action and the reactor does not trip.
Consider each case separately. (0.75 each)
~~a. Power Range channel 42 upper detector fails high.~~
~~b. Loop A wide-range Tcold fails low.~~
~~c. Loop D narrow-range Thot fails high.~~
~~d. One group D control rod drops into the core.~~
~~(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)~~
3. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page.53 RESPONSIBILITIES (10%)
QUESTION 3.12 (2.25)
!- The plant is operating at 100% load when a rupture of the pressurizer surge line. occurs. All RPS and ESFAS actuations occur as' expected.
~~c. What plant condition initiates automatic equipment / component' actuations for cold leg recirculation (include setpoint).~~
~~(0.75)~~
~~b. Why must the SI pump miniflow valves be closed.(interlocked) before supplying the suction of the SI pumps from RHR pumps discharge. (0.50)~~
~~' c. List TWO (2) reasons for recirculating via the hot leg following a LOCA? (0.50 each) i I~~
i l
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3. PLANT SYSTEMS {38%) AND PLANT-WIDE GENERIC Page 54 j RESPONSIBILITIES (10%) j I
QUESTION 3.13 (1.50)
The plant is operating at 100% steady state power with containment pressure protection channel IV failed high. A technician troubleshooting the failed channel inadvertently d anergizes the instrument power to containment pressure protection channel II. List TWO (2) ESFAS signals that will be gnnerated. (0.75 each)
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. 3. : PIANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 55 RESPONSIBILITIES (10%)-
L y
(3.00)
QUESTION 3.14
~~a. What AUTOMATIC signal / condition will cause a Containment Spray Actuation Signal (CSAS)? Include setpoint and coincidence. (1.00)~~
~~b. What FOUR (4) NONREDUNDANT equipment / components receive an actuation signal when a CSAS is present? (0.50 each) c l~~
l 1
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3

ML20235U154

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