Text

Exam Rept 50-348/OL-86-02 on 860707-14.Exam Results:Eight Candidates Passed Oral & Simulator Exams & Seven Candidates Passed Written Exam.Two Candidates Passed Reexams of All Categories
	ML20215N174
Person / Time
Site:	Farley
Issue date:	10/27/1986
From:	Bill Dean, Munro J; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML20215N166	List: ML20215N165; ML20215N174; ML20215N179;
References
	50-348-OL-86-02, 50-348-OL-86-2, NUDOCS 8611040397
	Download: ML20215N174 (119)
	v • d • e

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>QCEGo UNITED STATES

'o NUCLEAR REGULATORY COMMISSION O\ f[ REGION li h,(

, 101 MARIETTA STREET. N.W.

, C ATLANTA GEORGI A 30323

%.~....f ENCLOSURE 1 EXAr4INATION REPORT 348/0L-86-02 Facility Licensee: Alabama Power Company ATTN: Mr. R. P. Mcdonald Senior Vice President P. O. Box 2641 Birmingham, AL 35291 Facility Name: Farley Nuclear Plant Facility Docket No.. 50-348 Written, oral, and simulator examinations were administered at Farley Nuclear Plant near Ashfor A ab ma.

Chief Examiner:

Wil iam M. Dean

, ME& 6 Date Signed Approved by: 9 .I /N 7V SV E

/o/M7/R Date Signed JopF.Mubfo,getingSectionChief Summary:

Examinations on July 7-14, 1986 Oral examinations were administered to 9 candidates; 8 of whom passed. Simulator examinations were administered to 9 candidates, 8 of whom passed. Written examinations were administered to 9 candidates, 7 of whom passed. 2 candidates were administered written re-examinations of all categories; both candidates passed.

Based on the results described above, 1 of 1 R0's passed and 6 of 10 SR0's passed.

8611040397 861029 PDR ADOCK 05000348 V PDR

REPORT DETAILS

1. Facility Employees Contacted:

Lee Williams, Training Manager

Dave Morey, Operations Manager

Randy Wiggins, Supervisor, Operations Training

Walt Lee, Instructor Chris McLean, Instructor

Attended Exit M;eting

2. Examiners:

Bill Dean, NRC Bill Hemming, EG&G Nels Jensen, EG&G Dave Nelson, NRC

Chief Examiner

3. Examination Review Meeting At the conclusion of the written examinations, the examiners provided your training staff with a copy of the written examination and answer key for review. The comments made by the facility reviewers are included as Enclosure 3 to this repor+., and the NRC Resolutions to these comments are listed below.

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a. RO written Exam (applicable SRO exam questions are in parenthesis)

1. Question 1.05 (5.06)

NRC Resolution:

No action required. Restatem(nt of proctor's directions to candidates

2. Question 1.09 (5.10)

NRC Resolution:

Agree. The 78 F or 226 F limits are recent modifications to 2

ESP-02. 50 F and 200 F have apparently been superceded by TCN-2A (which instituted the 78 /226 F limitations), but were still lef t

2 in the procedure (step 7.5), thus providing conflicting require-ments within the same procedure. As this change was recently implemented and is more conservative, it will be accepted, but not the 50/200 F limits.

3. Question 1.11 (5.12)

NRC Resolution:

No action required. Restatement of proctor's directions to candidates.

4. Question 1.12 (5.01)

NRC Resolution:

No action required. Restatement of proctor's directions to candidates.

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5. Question 1.14 (5.14)

NRC Resolution:

Agree. This question has been deleted and the sectional point value adjusted.

6. Question 2.15(6.12)

NRC Resolution:

Agree. The question has been deleted and the sectional point value adjusted.

7. Question 2.18 NRC Resolution:

(a) Agree. The answer key has been modified to allow lead screw housing or CRDM housing as acceptable answers for Part A.

(b) Agree. Part B has been deleted based on review of reference material supplied by the facility. The sectional point value has been adjusted.

8. Question 2.22 NRC Resolution:

i Agree. The question has been deleted and the sectional point value adjusted.

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9. Question 3.07 NRC Resolution:

Agree. The answer key has been modified to allow A or D as acceptable answers.

10. Question 3.08(6.08)

NRC Resolution:

No action required. Restatement of proctor's directions to candidates.

11. Question 3.10(6.11)

NRC Resolution:

(a) Agree. The function of P-12 has been deleted from the answer

key and is not required for full credit.

. (b) Agree. The answer for Part 3 has been modified per informa-l tion in the original NRC referenced document; note, the j facility referenced document did not state these conditions.

12. Question 3.12 I

NRC Resolution:

Agree. This has been added to the answer key as an acceptable answer to Part C. Any two of the three are now required for full credit.

13. Question 3.13(6.13)

NRC Resolution:

(a) Agree. The information in parenthesis is given for purposes of clarity only and not required for full credit in Part A.

(b) Disagree. 'AUT0' position is a condition for valve closure.

The question did not ask for_ interlocks.

(c) Agree. Additional qualifications for Part C.1 have been added to the answer key. For Part C.2 the candidates are not required to state the conditions for resetting reactor trip; only those required for resetting a FWI.

(d) No action required. The coincidence for Part C.3 is given for clarity only and not required for full credit.

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14. Question 3.14(6.14)

NRC Resolution:

Agree. The answer key has been corrected to require 220 steps as the full credit answer for Part D.

15. Question 3.16(6.15)

NRC Resolution:

Disagree. The candidates were briefed that the point value of a question is an indication of the depth of answer required. The question specifically lists the topics that the candidate should address. The exact format of the answer key-is not required for full credit if all required information is present.

16. Question 4.11(7.10)

NRC Resolution:

Agree. Each response is worth 0.33. Point value for the question remains at 1.0.

17. Question 4.13(7.13)

NRC Resolution:

No action required. Restatement of proctor's directions to candidates.

18. Question 4.14(7.14)

NRC Resolution:

Disagree. The referenced procedure, EEP-3; step 23.3.4, states

" reduce make-up flow". This is a substep of 23.3 "Make-up Flow Control". The only options available by this procedure are l " throttle charging flow" or " start additional charging pumps."

The procedure states nothing about varying letdown flow. Decrease charging is recognized as being equivalent to reduce make-up flow.

19. Question 4.16 NRC Resolution:

Agree. The answer key has been modified to allow Pot Setpoint Book as an acceptable answer.

20. Question 4.17(7.15)

NRC Resolution:

5 No action required. Restatement of proctor's directions to candidates. Additional answer based on information contained in new procedure referenced in the question.

21. Question 4.20(7.17)

NRC Resolution:

Agree. Answer key corrected.

22. Question 4.21(7.18)

NRC Resolution:

Agree. The answer key has been modified to allow recommended response as an additional correct answer.

b. SRO written exam

1. Question 5.08 NRC Resolution:

No action required. Restatement of proctor's directions to candidates.

2. Question 5.15 NRC Resolution:

Agree. Part A deleted due to duplication with question 8.05.

Part B answer modified based on information supplied by facility.

3. Question 8.05 NRC Resolution:

Agree. The FNP TS definition will be used as the answer for full credit.

4. Question 8.07 NRC Resolution:

Agree. Based on additional reference material not originally supplied by the facility, Recovery Manager will alsa be an acceptable answer.

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5. Question 8.10 NRC Resolution:

Agree. .The information in parentheses is given for purposes of clarity only and not required for full credit.

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6. Question 8.14 I

! NRC Resolution:

Disagree. Section 3.0 of Technical Specifications is commonly i

referred to as the " Motherhood Statements". The question asks the j . candidate to state which specific LC0/ Action is applicable. The 1 answer " Motherhood Statement" is not acceptable.

7. Question 8.16 i

l NRC Resolution:

! Disagree. TS 3.0.5 states when a device is inoperable solely due i to loss of normal or emergency power supply, it can be considered

! operable if 1) the normal or emergency power supply is operable i and 2) the remainder of the system is operable. With the 'A' DG inoperable, AFW pump 'A' can be considered operable under TS i 3.0.5. When the TDAFW pump steam supplies are identified as i inoperable, the remainder of the AFW system is not operable as

! required by TS 3.0.5 so the action statement must ce entered. TS j 3.0.5 allows a 2-hour delay prior to initiating action to go to l Hot Standby which is more restrictive than the 72-hour action allowed by TS 3.7.la. The 6-hour action allowed by TS 3.7.1.2b. ~

is not applicable in this situation, even though it is more j- conservative, due to the fact that one AFW pump is inoperable l solely due to an inoperable emergency power supply. The j candidates should be familiar with Section 3.0 without it being i provided during the exam. The answer key is required for full 4 credit. 3 j 4. Exit Meeting .

l At the conclusion of the site visit, the examiners met with representa- :

{ tives of the plant staff to discuss the results of the exanination.

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! There were no generic weaknesses noted during the oral or simulator j examinations.

I The cooperation given to the examiners and the effort to ensure an

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1 atmosphere in the control room conducive to oral examinations was also i noted and appreciated.

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The licensee did not identify'as proprietary any of the material j provided to or r.eviewed by the examiners.

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I c-8 O 'R$MJ U. S. NUCLEAP REGULATORY COMMISSION EENIOR REACTOR OPERATOP LICENEE EXAMINr. TION CACILITY: FARLEY 182 OEACTOR TYPE: OWR _EC3 OATE ADMIPISTERED: 86/07/1c EXAMINER: NELSON, D.

APPLICANT: - - - - . _ _ - _ - _ _ _ - - - - _ - - _ - - _ - -

INSTRUCTIONE TO APPLICANT:

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

Steple question sheet on top of the answer sheets. Points for each question are :ndicated in parentheses after the question. The passine stede requires at leest 70% in ecch category end e final grade of et least 90%. E::a m in a t io n papers will be pick.ed up si: (6) hours efter the e:(etinction stetts.

0F CATEGORY ' Or Art'LICANT'E CATEG0ri UALUE TOTAL ECORE VALUE CATEGORY

_ 1 __ _ 1 ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS on or

_;;1_(_g__ __jl_oc_ ^

___________ ________ 6. L A N T SYSTEMS DEEIGN, CONTROL, AND INETRUMENTATION 9 o e.

0'.00- - _ - I .'. _O

.' n ___________ ________ 7 PROCEDUPES - NORMAL. ABNORMAL, EMERGENCY AND RADIOLOGICA_

CONTROL on-- on oc

-i+or a o w ----e nit:. n'-- ,

- - - _ _ _- ----------- -----_-- -- gnx1pT -g7: ,- vc---

CONDI' IONS. AND LIMITA' IONS 120.00 100.00 TCTALS rTNe

- - r. R+nr

-m-- - - - - - - - - - - - - - - - - -

All wa-t done on this e m ination is my own. I have neither 3:n - ;or re:e:s.cd tid.

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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS 1

During the administration of this examination the following rules apply:

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

2. Restroom trips are to be limited and only one candidate at a time may i' 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. i

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

6. Use only the paper provided for answers.

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

8. Consecutively number each answer sheet, write "End of Category as appropriate, start each category on a new page, write only 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. l r

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18. When you complete your examination, you shall:

a. Assemble your examination as follows:

(1) Exam questions on top.

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

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

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

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

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

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5. THEORY OF 'JUC' EE R COPEp r"_ % 1 OPERATIDHr FLUIDS, AND DAGE 2 00ESTION 5.01 (2.00)

2. If a s p e c. i f i c ::1. ; u n t e r ' e ; ;. t i v i t y it cdded to 2 subtritical 9 k reectors, that ere identicc: r cm c e p t for S_hutdown Mergin:_-bow'T

1. Which reactor will undergo the greatest change in count rate? (0.5)

2. Which reactor will'take e greater amount of time to reach e stable count rate ? (0.5)

b. For a reactor startup, how does the initial source count rate affect:

1. Rod height at criticality ? (0.5)

2. Count rate (power level) et criticality ? (0.5)

GUESTION 5.02 (1.00) l In order to maintain a 209 e subcooling nargin in the RCS when icducing RCS precsure a 2600 pris, sterr. generttor precture muct be reduced to appronimately; (c) 245 pris (b) 445 psis (c) 645 pris (d) 845 psis 9UESTION 5.03 ti.00)

Which of the following vill result in the largest INCREASE in the concentrction of dissolvec' srces in e quentity of urter? (Assume the changes : r. t e a:P e r a t v

e e n d p e s an e below are of equal 3:asnitude)

r. Increetins the pretsur e t v.' lower ins the temperstere.

b. Decretsing the pressure a r.d lowering the temperature.

c. Increes, ins the pretcure end raising the tempereture.

d. Decreasing the pressure and raising the tenpersture.

(rrrWu CATEGORv 05 COUTINUED ON NEXT PAGE *****)

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5. THEORY OF NUCLEAP DOWER PLANT OPERAT~00. CLUIDE. AND t' ti G E 3 QUISTION 5.04 (1.00)

'J i . i c h of the 'o: lowing equ?tler. used to perfern 3 o!'D heat b;1:nce cilculttiori it correct 7

3. G r :- = M(s) Eh(s) - h(fw)] + M(bd> Eh(bd) - h(fw)~ + Grcp

b. Gr: = M(s) Ch(s) - h(fw)3 + M(bd) Eh(bd) - h(fw)2 - Prep

c. O r :: = M(s) Ch(s) - h(fw)2 - M(bd) Eh(bd) - h ( +' u ) 2 - Grep

d. O r :-: = M(s) [h(s) - h(fw)? - M(bd) Ch(bd) - h(fw)J + Prep NOTE: Notation Key 0 = Power M = Mass Flow Rate fw = Feedwater r:: = Reactor bd = Blowdown s = Steam h = Specific Enthcipy 00ESTION 5.05 (2.00)

For the following ctatements chcose the mest correct 2n2wer:

L. The verttivity Lett* o' s e r s * : u r, et 2 5 ;. Est ^ibtivm power is (greater than/less than/or equal to) the reactivity worth at 100% equilibrium po' wet. (1.0)

b. The total power coefficient (pem/% power) at BOL is (more nesetive then/less negative thet/or equel to) the power coefficient at EOL. (1.0)

(***** CATEGORY 05 CONTINUED ON NEXT R AGE mn )

5. l EORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS. AND PAGE 4 THERMODYNAMICS 00ESTION 5.06 (2.50)

Assune one 9CP trips at '30% power r without a reactor pr ote c tiot r s y s t e n. actuation or e chense in turbine load. Indicate whe'her the following parameters will INCREASE, DECREASE, or REMAIN THE SAME. c% @ gw ,

Flow in the operating reactor coolant loops. (0.5)

s. 9 p g,,

b. The ratio of core flow compared to the total loop f l q_w . te[f c

(0.5)

(Core Flow / Total Loop Flow)

c. Reactor Vessel delta pressure.- (0.5)

d. Core delta temperature. (0.5)

e. Operating loop steam seneretor temperatures. (0.5)

QUESTION 5.07 (2.00)

An ECP is caleviated for 3 startup a hours after a shutdorn '*om 100%

s-t e e d y t t it t i power. Indicate whether the actuel eritice; p: :; ion ui.1 be GREATER THAN, LESS ' FAN or the SAME AS the ECP #cr the followins conditions.

a) The steam dump pressure setpoint is increased by 35 psis. (0,5) b) All steam senerator levels are increased by 5% five minutes before criticality is reached. (0.5) c) Condenser v a cuun: it reduced 3' Hs due te a small air leak. (0.5) d) Reactor startup is delayed for 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . (0.5)

GUESTION 5.0E (1.50)

Indicate whether the followins will INCREASE. DECREASEr or uAVE r0 ErrECT on the availtble (e tue'.' Net Positive Suttfor. Hesd (hDSM).

3. Increasing punp speed.

b. Increasing pump suction t e nip e r a t u r e ,

c. Increas:ns systen, pressure. ._sy gi.h c /o w.0 S y t (wr*** CATEGORY 05 CONTINUED ON HEXT PAGE *Wriv) l l

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 5 GUESTION 5.09 (3.00)

The Heet Clu. Hot Channel r etter (FQZ) and Nuclear Enthelpy Hot Channel Factor (FNDH) are both power distribution limits.

1. Which limit is calculeted using a Rod Bow penalty based on the core region average burnup? (0.5)

2. Which limit is defined as 'The ratio of the integral of linear power along the rod with the highest integrated power to the average power'? (0.5)

3. Technical Specification surveillance requirements using in-core detectors is infrequent provided that FOUR items are monitored and verified to be within their limits. What are these four items? (2.0)

DUESTION 5.10 (2.50)

a. If during a cooldown on natural circulation, the PCS pressure was 1200 psis, what would be the mcximum s t e p) r. generetor pressure to assure adequatetsubco Q s? @ p.u eQ (1.0)

b. During natural circulation cooldown, a steam bubble may form in the reactor vessel head area. What is the primary indication of this bubble formation? (0.5)

c. What is the maximum core Delta Temp. which would be indicative of PROPER netural recirculetion flow following a full power trip AND what is the a pp r o::im a t e loop transit time? (1.0)

QUESTION 5.11 (1.00)

' ist three significant heat transfer advantages of a counter flow heat e::c h a n g e r over a parallel flow heat exchanger,

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

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5. THEORY OF NUCLEAR POWER FLANT OPERATION, FLUIDS, AND PAGE 6 GUESTION 5.12 (2.50)

During 100?. power operations it is decided to reduce power by 20%

using control rodt only for reactivity control.

a. E:: plain HOW AND WHY the a::ial flux shape will change for the first hour efter the power reduction. (1.0)

b. Explain HOW AND WHY the flux shape will change over the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Include the effects of control rod movement to maintain power stable. f a (1.5)

,sJ ) ,p. toj a.~jw m[y a rww a po;7 {3 ep GUESTION 5.13 (2.50)

a. Stittle fracture of any carbon steel pressure vessel can occur et stresses well below yield stress if TWO other conditions are present. What are these TWO conditions? (1.0)

b. How de hestup/cooldown rete limits on the reactor coolant systen reduce the probability of brittle fracture? (0.5)

c. Why does the concern.about brittle frecture of the reactor pressure vessel increase as the plant ases? Include in your answer the specific material PROPERTY that is affected. (1.0)

GUESTION 5.14 (1.00)

Explain how the starting of a Reactor Coolant Pump in a water-solid plant can cause a pressure transient. (1.0)

QUESTION 5.15 (1.50) a) 'J h a t le the definition of Shutdown Margin (SDM)? (1.0) b) If E stuct rod exists while the recetor is at power, whet Edjustnent, if a r.y , must be made to the SDM calculation? (0.5)

(xxx*r CATEGORY 05 CONTINUED ON NEXT PAGE *xxxx) 1

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5. THEORY OF NUCLEAR POWEY PLANT OPERATION, FLUIDS, AND PAGE 7 QUESTION 5.16 (1.00) otimary syster flow rate is many tic.es greater than secondary syster flow rate while the heat transferred by the two systems is essentially the same. Explain how this is possible.

QUESTION 5.17 (2.00)

a. Steam exiting the HP turbine is at 785 psis, 90% quality. Steam entering the LP turbine is superheated to 100 F. What is the enthalpy change of the steam? '

1. 85 ETU/lbm

2. 140 E:TU/lbm

3. 156 BTU /lbm

4. 705 BTU /lbm

b. During the process in Part 'a', how much energy is added to the steen by rechan2cel moisture sePerction in the MSR c s s un:I n s that the stesm quality after separation is 100%.

1. 71 BTU /lbm

2. 85 BTU /lbm

3. 140 BTU /lbr:

4. 156 BTU /lbm

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

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 8 GUESTION 6.01 (1.00)

Which of the following is NOT a function of the P-4 permissive (trip and byrets breakers open)?

a. Allows bypassing of steam dump cooldown interlock.

b. Allows operetor block of SI sisncl after a time delay.

c. Causes feedwater isolation if low Tavs is also present.

d. Causes a turbine trip.

QUESTION 6.02 (1.00)

Which statement below regarding the pressurizer pressure control and protective system is NOT correct? -

O. The master pressure controller provides the control signal for only one of the PORVs. i

b. There is a lead /las compensation circuit for pressure inputs to the low pressure reactor trip thet veries the trip setpoint with the rate of pressure decrease,

c. The two pressuriner spray valves are controlled-by separate transmitters

d. To block SI actuation on a r. cts.a1 plant depressurization, the operator.

must operate TWO block switches to prevent inadvertant ECCS actuation.

GUESTION 6.03 (1.00)

The S/G PORV's maximum capacity is limited by design to approximately 8%

! of rated steam flow. Which of the followins is the reason for this limit?

a. Maintains mass discharge rate within the' capacity of the condenser makeup system.

b. Limits plant cooldown rate if any one PORV sticks open.

c. Miriimires the possibility of S/G differential _ pressure.SI if any

one FORV sticks open,

d. Minimizes erosion of PORV valve seats.

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSlRUMENTATION PAGE 9 DUESTION 6.04 (1.00)

Which of the following statements about temperature detectors is true?

a. The thermocouple is connected to one les of a bridge circuit and as the temperature changes the output voltage across the bridge changes.

b. When a thermocouple fails open it will respond in the same manner as an RTD and will indicate a full scale readins on the meter.

c. When a thermocouple becomes shorted, a new thermocouple will exist at the point of the short and the meter will respond to the ambient temperature at the point of the short.

d. An RTD is comprised of two wires of dissimilar metals in contact with each other and generates an EMF proportional to the temperature difference between the open ends of the wires.

QUESTION 6.05 (1.00)

What type of radiation is the Gross Failed Fuel Detector monitoring for indiection of fuel frilure?

a. Gammas from iodine and cesium decay.

b. Alphas from uranium and plutonium decay.

c. Neutrons from bromine and iodine decay.

d. Betes from tritium decay.

QUESTION 6.06 (2.00)

The condenser steam dumps will not open unless certain interlocks are met, armins signals are present, and there is a demand signal.

a. What 3 interlocks (permissives and blocks) must be met? E0.62

b. How it the load reject:on erming signe; reset? E0.43

c. What determines the magnitude of the demand signal when in the Tavs - turbine trip submode?. C0.43

d. What are the three signals that can arm the steam dumps? CO.63

(*rrrr CATEGORY 06 CONTINUED ON NEXT PAGE rrrrr)

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 10 QUESTION 6.07 (1.50)

Indicate whether the OP Delta T trip setpoint will increaser decrease or rettin the same for the followins parameter chrnses. Consider each seperately and address only its affect on OP Dcita 1.

a. Increasins Tavs.

b. Tavs < Rated Thermal Power Tavs.

c. Pressuriner pressure decreasing.

QUESTION 6.08 (3.00)

Indicate what happens to the Rod Control System (rods in, rods out, no chanse) and BRIEFLY e>; plain why the change will or will not occur for the followins conditions. Rods are in auto unless otherwise specified.

a. Reactor power is 17% when the controlling turbine first stase impulse pressure transmitter fails hish.

b. Reactor jgower is 100% and loop 1 Thot fails hish.

c . Augi.actree >ower is 50%, Rod Control it in manual.

Instrument-testing is in progress on the turbine power input to rod ,

control which has turbine power at 100%.

All indications have been stable for the last hour. ,

The Bank Selector switch is then placed in AUTO.

QUESTION 6.09 (2.50)

Indicate whether the followins statements are true for DTDT, OPOT, or both OTDT and OPDT protection instruments.

1. Protects the core from DNB.

2. Protects the core from overpower. (kw/ft.)

3. Backup for the high neutron fiv:- trip.

4. Circuitry includes dynemic compensetion for pipins delays to the loop temperature detectors.

5. Requires pressure to be witFin the high end lov pressure reactor trip setpoints to be valid. (2.5)

(***** CATEGORY 06 CONTINUED ON NEXT PAGE rr.rrr)

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 11 i_

j QUESTION 6.10 (3.00)

What are the NORMAL, BACKUP, AND ALTERNATE power supply paths from

, the 600V LC's to the 120 vital AC instrument distribution pene; 1A?

j Cir1uit breaker numbers are not necessary. (3.0)

QUESTION. 6.11 (3.00)

For each of the followins, give the set point and coincidence:

. 1. P-12

2. P-14 I 3. High Main Steam Flow with LowLow Tavs 64, 44 $/ v/ i QUESTION 6.12 (1.00) under what plant conditions is the reactor vessel level indication system NOT available for the operator's use?

t QUESTION 6.13 (3.00)

a. What 2 time delay actions occur in the condensate / main feedustWr sys-tem when main feed pump suction pressure stays below 300 psis? (1.0) j b. What conditions cause automatic closure of the steam senerator

Main Feedwater Stop (Isolation) Valves (3232A,B,C)? (0.5)

c. If the stera generator Feed Regulating valves are closed by a protection signal (SSPS), the signal must be cleared to reopen the valves. What are the THREE protection signals that close the valves AND HOW is each cleared? (1.5) i

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

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6. P 1T SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 12 00ESTION 6.14 (2.00)

For each condition EXPLAIN which component (s) would be generating a rod n.o v e m e n t signal and the response of Dank D rods to this signal.

Assunie no other Rod Stop signale present, Reactor at power.

BANK SELECTOR SW. IN-0UT-HC'D LEVER PLANT PARAMETER

- 'D' POSITION

s. Manual In RIL LO-LO ALARM 180 steps

b. 'D' Hold Tavs-Tref +4 deg. 180 steps

c. Manual Out '.. Urgent Failure

200 steps j

d. Auto Hold Tavs-Tref -4 deg. 222 steps r

GUESTION 6.15 (3.00)

Describe the controls for the charging pumps. Include control location interlocks, switch positions, and auto starts.

QUESTION 6.16 (1.00)

What feature of the nain feed reg. bypass valves controller allows the controller to cnticipate rapid level fluctuations?

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

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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 13

~~~~Ed656L65fEht E60TRUL'~~~~~~~~~~~~~~~~~~~~~~~

! -QUESTION 7 01 (1.00)

Which of the following i t. e 10 CFR 20 exposure limit?

a. 5 rem / year-whole body.

b. 1-rem / quarter-whole body.

c. 18.75 rem / quarter-hands.

d. 7 rem / quarter-skin of whole body.

i OUESTION 7.02 (1.00) l Which of the followins radiation exposures would inflict the greatest l

j biological damase to man?

a. 1 Rem of GAMMA.

b. 1 Rem of ALPHA.

t

c. 1 Rem of NEUTRON.

d. NONE of the above; they are all equivalent.

QUESTION 7.03 (1.00) , ,

ESP 0.4, Natural Circulation Cooldown With Allowance For Reactor Vessel Stean Voiding is the preferred procedure over ESP 0.2, Natural Circulation I Cooldown to Prevent Reactor Vessel Steam Voidins, when

a. no CRDM fans are runnins.

b. RVLIS is in operation.

c. boron is added to meet shutdown margin requirements.

d. condensate inventory is low.

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

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7. PFDCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE 14

- ~~~~~~~~~~~~~~~~~~~~~~~~

~~~~kED56LUGEC5L 65UTR6L QUESTION 7.04 (3.00)

Match the trends fron Column B that would be indicative of conditions for Column A melfunctiont pr ior to any protective function actuetions.

There may be more than one Column B item for each Column A item. Place answers on answer sheet (e.g., c-7,Br9).

COLUMN A COLUMN B

a. Small Break LOCA Inside 1. Decreasing Pressurizer Level Containment

2. Decreasing Steam Pressure

b. Large Steam Leak Inside Containment 3. IncreasinS Containment Pressure

4. Decreasing Tave

5. Increasing Containment Radiation

6. Decreasing Pressurriner Pressure

7. Near Normal Steam Pressure (3.0)

OUESTION 7.05 (1.00)

Fill in the blanks:

ESP-1.1 (SI termination), foldout pagerstep 1, " Monitor SI criteria":

Greater than __a.__ subcooled in T/C position

{__b.__ subcooled in __c.__ Position} AND PRZR level above __d.__ {50%}.

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

7. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY ANO FAGE 15

~

~~~~RdD56L6656dL C6hTR6L~~~~~~~~~~~~~~~~~~~~~~~~

DUESTION 7.06 (3.00)

a. Prior to increasing 'cVe from Mode 5, your heatup procedure (UOo-1 1) gives you the option NOT to withdraw shutdown banks.

What condition nust exist prior to taking this option? (0.5)

b. If the heatup began with a solid Reactor Coolant System (RCS) condition, at approximately what PRESSURE will a steam bubble be formed in the pressurizer? (0.5)

c. What are the maximum allowable pressurizer HEATUP AND COOLDOWN rates? (1.0)

d. After the steam bubble is formed in the pressurizer in Mode 5, and prior to further RCS heatup, will the hot calibrated Pre 55Uriner level channels indicate HIGHER OR LOWER than actual level? EXPLAIN. (1.0)

GUESTION 7.07 (2.00)

The following refer to information found in FNP-1-SOP-22.0,

' Auxiliary Feedwater System."

A. What is the maximum feedwater flow allowed:

1. to a steam senerator whose level is 20 % Narrow Range and increasing during a level recovery transient?

2. from one motor driven AFW pump?-

3. from the turbine driven AFW pump? (1.5)

8. Why should service water supply to AFW only be used in an emergency?

(0.5)

GUESTION 7.08 (1.00)

What actions should you take upon discovering a spill involving a minor radiation hazard to personnel?

QUESTION 7.09 (1.00)

List the immediate operator actions for a loss of 1B reactor coolant pump when at 30% power.

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

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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND :G

7. PAGE 16 'g UNTR5t-----~~~----------------

~

~~~~Rd656L6GICdL C #;

a D

QUESTION 7.10 (1.00) A :

List the imrediate operator actions #ct a continuous control bank D withdrawal. ,I s

GUESTION 7.11 (2.50)

List five radiation monitors that could indicate high radiation during -

p a r@ fueling accident according to AOP-30.0. (multiple channels of the same O crco monitor are not acceptable.) d QUESTION 7.12 (2.00) 1[

. i.A The following refer to FNP-1-AOP-8.0, ' Partial Loss of Condenser j Vacuum.' . if List the five immediate operator action steps which Cj should be performed to restore vacuum if it is decreasing h rapidly. (2.0) }a ,

QUESTION 7.23 (3.00) e

s. After a Residual Heat Removal (RHR) pump is started for plant !

(

cooldown, but before placing the train in service, the pump is j operated on miniflow recirevlation for a minimum of 10 min. WHY?- (0.5) y

b. How is a low boron concentration (in an RHR train to be placed j in service) corrected? (1.0) Q

c. Would starting an RHR pump, with the CVCS letdown pressure control y valve (PCV-145) in manual result in a pressure INCREASE, DECREASE, p o* NO CHANGE in the,,Rgaetor Coolant System during solid plant

operation' /n i n ahy (0.5)

d. W h e r. estEblishins a bubble in the pressurirer, WHY must both *-

RMP tra qs be valved into their respective RCS hot less? (1.0) ;

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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE~ 17

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~~~~Rd656L5siCdL E5sTR6L QUESTION 7.14 (1.00)

While performing EEP-3r Steam Generator Tube Rupture procedurer how is t

lowering both the Pressurizer level AND ruptured Steam Generator level accomplished?

QUESTION 7.15 (1.00)

When would a transition be made from AOP-1.0, Excessive RCS Leakaser to

the-EP s e r i : .- d u e t o a l o s s o f r e a c t o r coolant? Assume no reactor trip

or SI occurs. pop _3,l QUESTION 7.16 (1.00)

How is pressurizer level maintained immediately followins a loss of instrument air?

QUESTION 7.17 (1.00)

] Under what ecnditions are the adverse containment values for instrumenta-tion used in the EEPs?

QUESTION 7.18 (1.50)

List three different allowable times an FRG may be exited if it was entered by direction of an ORANGE path.

QUESTION 7.19 (2.00)

What are the four immediate operator actions of AOP-27.0, Emergency Boration?

t i

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

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 18 d

DUESTION 8.01 (1.00)

Which of the following conditions requires action according to Tech Specs ;

in less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months if in Mode 2 on Unit !?

a. The shutdown margin is 1.0.

b. One Boric Acid Transfer pump .is inoperable,

c. One Shutdown rod fully inserted with the reactor critical.

d. Primary Containment averase air temperature is 140 deg. F.

QUESTION 8.02 (1.00)

Unit 1 is operating in Mode 3 during a reactor startup with the following deficiences:

One Main Steam Line Isolation Valve inoperative One Motor Driven AFW pump inoperative

, See attached LCO's

] Which one of the followins actions most accurately details the allowances and/or limitations imposed by Tech Specs in this instance;

a. Mode 3 must be maintained (entry into Mode 4 acceptable),

i b. Startup activities may continue; Mode 2 may be entered but not exceeded.

c. Startup and power operation into Mode 1 may be accomplished provided the Mode 1 action statement for MSIV is met.

d. Startup activities may continue into Mode 2 provided subsequent-restoration of the MDAFW pump to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

I GUESTION 8.03 (1.00)

Steam senerator tube leakase falls under which Tech Spec leak classification below?

r. identified leakage

b. pressure boundry leaksse

c. controlled leakase

d. unisolable leakase

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

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B. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 19 QUESTION 8.04 ( .50)

True/ False All conditions applicable to a licensee under 10 CFR 55

' Operator License" shall be stated within the license.

QUESTION 8.05 (1.00)

Tech Specs defines Shutdown Marsin as ...' Shutdown Margin shall be the instantaneous amount of reactivity by which the Reactor is suberitical or would be suberitical from its present condition assuming ... '

STATE the assumptions made for the plant conditions which. complete the definition of Shutdown Marsin.

QUESTION 8.06 (1 00)

To prevent entering a technical specification action statement on Unit 1, the quadrant power tilt ratio shall not exceed ______ when reactor power is above 50%. If this limit is exceeded, then an extended temporary GPTR limit of ______ is allowed durins efforts to restore OPTR to normal levels.

QUESTION 8.07 (1.50)

The following refer to " Emergency Plan Implementation Procedures", -

FNP-0-EIP's. 3' 3'

A. Who, by title, is responsible for the immediate and unilateral declaration of an emergency AND the initiation of emergency response during the initial phase of an emergency? (0.5)

B. Who, by title, is the ONLY individual authorized to downgrade an emergency level once en emergency has been declared? (0.5)

C. Which channel number on the public address system is designated for use during emergencies? (0.5)

GUESTION 8.08 (2.00)

What are two provisions which must be met, according to Technical Specifications, before a temporary change can be made to an Oper-ating Procedure? Be Specific. (2.0) 4 (xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx)

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8.. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATION 5 esGE 2u GUESTION 8.09 (1.00)

. What are the Tech Spec maximum heatup rates in any one hour for the Reactor Coolant System AND the Pressurizer?

GUESTION 8.10 (1.50)

The Technical Specifications for reactor trip system instrumentation channels specifies if one channel of Power Range Nuclear Instrumenta-tion is inoperable, a Guadrant Power Tilt Ratio must be done at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months if power is at 100 %.

A. How is the Guadrant Power Tilt determined in this case? (0.5)

B. If the Guadrant Power Tilt Ratio is not determined within the '

allowable time, what must be done? (1.0)

GUESTION 8.11 (3.00)

The concentration of the boric acid solution in the RWST shall be verified once per seven days in accordance with Technical Specifications. The chemist sampled the RWST under the following schedule. (All samples taken at 1200 hours0.0139 days 0.333 hours 0.00198 weeks 4.566e-4 months .)

January 1 -- Janvery 8 -- January 16 -- January 24 --January 31 ;

A. EXPLAIN why or why not surveillance time interval requirements .

I were exceeded on January 16. (1.5)

B. Explain why or why not surveillance time interval requirements were exceeded on January 24. (1.5)

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

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B. ADMINISTRATIVE FROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 21 k

GUESTION 8.12 (2.50)

The followins concern procedures for checkins MANUAL VALVES in their proper position durins valve lineups in accordance with FNP-0-AP-16.

A. How does an operator:

1. verify a normally OPEN valve in the open position? (0,5)

2. verify a normally LOCKED OPEN valve? Be specific.

Include any additional verification requirements NOT required in 1 above. (1.0)

3. verify a normally SHUT valve in the shut position? (0.5)

4. verify a LOCKED and THROTTLED valve? (0.5)

GUESTION 8.13 (1.50)

What action must be taken if the RCS Pressure Safety Limit is' -

exceeded, in accordance with Technical Specifications? Consider ALL modes. (1.5)

GUESTION 8.14 (1.00) ;

Durins Mode 1 operation of unit 1 it is found that 2 of 3 channels for l Pressuriner Pressure high Reactor trip are inoperable due to a seneric material deficiency (repair time 14 days). Using Tech-Spec LCO's provided, determine what actions must be taken as a result of this failure? State specific LCD/ action steps which apply.

QUESTION 8.15 (1.00) uith regard to FNP-0-APS2, ' Equipment Status Control and Maintenance Authorination*, answer the followins questions:

a. A Deficiency Tag once placed on a piece of equipment is removed at what Point in the MWR process?

b. When releasins equipment for work authorization who (by position) makes the initial determination c.' operability /inoperability?

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

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 22 GUESTION 8.16 (1.00)

Unit 1 is operating in Mode 1 with Diesel Generator Set A (DG 1-2A & 1-C) inopercble when it is identified that the steer supplies te the Turbine

driven AFW pump are inoperable. Specifically which Action /LCO would be l entered and cctried out? Assume remaining-DG's surveilances completed satisfactorily.

OUESTION 8.17 (1.50)

With regard to FNP-0-AP 16, ' Conduct of Operation't answer the followins*

a. The Shift Supervisor has three basic reporting chains. List these three reporting chains.

I b. If control room verification cannot be made upon removal and return to service of safety related equipment, indepedent verification will be made. Is this verification made seperately from the initial lineup AND how is indepedent verification documented?

OUESTION B.18 (1 50) 4

a. While operating in Mode 2 a piece of Tech Spec related equipment which is only required in Mode 1 becomes inoperative. In accordance with-AP-16, WHERE and HOW is this type of LCO documented?

b. While performing the required review by AP-16 for verification'of valve -4 linup the Shift Supervisor identifies a valve which has a circle around both the 'as found' and the ' lineup initials' and an 'open' written above the 'as found' and initials above the ' lineup initials' on the system checklist. There are no System Checklist Exception Sheets for s this valve / system. Is this a satisfactory condition? (yes/no).

a 4

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

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATION 0 PAGE 23 00ESTION 8.19 (2.00)

a. In accordance with FNP-0-AP-14, " Safety Clearance and Taggins' Power Operated Relief Valves (MOVs and ADVs) with accessible manual operators should be tassed at two locations. State the two locations these valves are to be tagged at.

b. According to AP-14, which individual (by position) determines if verification may be accomplished at the Control Room?

c. If an additional clearance is requested on a system which is already

! cleared, (the additional clearance is identical to the first), HOW would the Tassing Official notate the additional clearance in the

' clearance issue to'.and ' clearance released by' blocks?

DUESTION 8.20 (i.00)

Unit 1 is operatins in Mode 6 with the following conditions present:

Water level above the top of the reactor pressure vessel flange >23 feet DG Set A is inoperative 1

RHR Train B (the inservice loop) is found to be inoperative due to flow inadequacies Specifically which LCO/ Action statement would be entered and carried out?

(LCO's attached) .

l QUESTION 8.21 (1.50)

c. According to the EIPs which Cmergency Classification level has as one of its purposes to ' initiate predetermined actions for the public'?

b. Per EIP-14 'Re-entry procedures *, if a reentry team encounters dose rates exceeding the limits established by the Emergency Director, What

'TWO actions shall they take?

i GUESTION B.22 (1.005 In accordance with FNP-0-AP-13 " Control of Temporary Alterations", EXPLAIN 1 how temporary alterations may be made on ' operable' equipment. Include in your discussion who must approve and what determination must be made.

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

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. 3/4 LIMITING C00ii:0NS FOR OPERET10?i AC SUTVE:0 D.' M OUIRE M T5 3 /4.0 APPLICABILITY 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 A condition prohibited by the Technical Specifications 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 the 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 one hour ACTION shall be initiated to i place the unit in a MODE in which the specification does not apply by placing q~

it, as applicable, in-(

1. At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

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

3. 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.

f, 4'

3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the conditions of the Limiting Condition for Operation are met without reliance on provisions contained in the ACTI0li requirements. This l provision shall not prevent passage through OPERATIONAL MODES as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual specifications.

3.0.5 When a system, subsystem, train, component or device is detemined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for.

the purp6se of satisfying the 7equirements of its applicable Limiting Condition " i for Operation, provided: (1) its corresponding normal or emergency power source l 1s OPERABLE; and (2) all of its redundant system (s), subsystem (s), train (s),

component (s) and device (s) are OPERABLE, or likewise satisfy the requirements of this specification. Unless both conditions (1) and (2) cre satisfied within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , ACTION shall be initiated to place the unit in a MODE in which the applicable Limiting Condition for Operation does not apply by placing it, as applicable , in:

FARLEY-UN:T 1 3/4 0-1 AMEl:DMENT NO.57

. PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM g LIMITING CONDITION FOR OPERATION -

3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated manual actuation switches in the control room and flow paths shall be OPERABLE with:

a. Two auxiliary"feedwater pumps, each capable of being powered from separate emergency, busses, and

b. One auxiliary feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY: MODES 1, 2 and 3.

ACTION: I-

a. With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With two auxiliary feedwater pumps inoperable be in at least HOT (

STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following '

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . -

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c. With three auxiliary feedwater pumps inoperable, immediately initiate ^

corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible.

SURVEILLANCE REQUIREMENTS 4.7.1.2.1 Each motor-driven and the turbine-driven auxiliary feedwater pump ~

- shall be demonstrated OPERABLE. pursuant to Specification 4.0.5. For the turbine-driven pump, the provisions of Specification 4.0.4 are not applicable for entry into MODE 3.

t 4.7.1.2.2 Each auxiliary. feedwater pump shall be demonstrated OPERABLE:

a. At least once per 31 days by:

1. Verifying that each non-automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

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FARLEY-UNIT 1 3/4 7-4 AMENDMENT NO. 26

. . PLANT SYSTEMS

( MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION l

3.7.1.5 Each main steam line isolation valve shall be OPERA 8LE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

M00ES 1 - With one main steam line isolation valve inoperable, POWER OPERATION any continue provided the inoperable valve is restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; otherwise, reduce power to less than or equal to 55 of RATED THERMAL POWER within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

MODES 2 - With one main steam line isolation valve inoperable, subsequent and 3 operation in M00ES 2 or 3 may proceed provided the isolation ,

valve is restored to OPERABLE status or closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entering MODE 2; otherwise, be in HDT STAND 8Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS f

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4.7.1.5 Each main steam line isolation valve shall be demonstrated OPERABLE by verifying full closure within 5 seconds when tested pursuant to Specification 4.0.5.

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FARLEY-UNIT 1 3/4 7-9 AMENDMENT NO. 26

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3/4.3 INSTRUMENTATION i. [

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3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION l

LINITING CONDITION FOR OPERATION

[N L .,

:1c1 .

3.3.1 As a minimum, the reactor trip system instrumentation channels-and '

M.g .

interlocks of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2.

jg

'Qf APPLICABILITY: As shown in Table 3.3-1. [.d'[

W;,-

ACTION: .

fi'h As shown in Table 3.3-1. -

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{F SURVEILLANCE REQUIREMENTS I ;. a 7;

f 4 r; Q 4.3.1.1 Each reactor trip system instrumentation channel shall be demonstrated (

OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and ,.,W CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown r in Table 4.3-1. -

1- [

4.3.1.2 ThelogicfortheinterlocksshallbedeednstratedOPERA8LEpriorto :p cach reactor startup unless performed during the preceding 92 days. The total 7-interlock function shall be demonstrated OPERA 8LE at least once per 18 months. ;.g 4.3.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function W;d le shall be demonstrated to be within its limit at least once per 18 months. '

Each test shall include at least one logic train such that both logic trains cre tested at least once per 36 months and one channel per function such that h '.

g;. ^

i-011 channels are tested at least once every N times 18 months where N is '

. ~

the total number of redundant channels in a sp.ecific reactor trip function as ,,

shown in the " Total No. of Channels" column of Table 3.3-1. .

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FARLEY-UNIT 1 3/4 3-l AMENDMENT NO. 26 l

TABLE 3.3-1 s .

[ . REACTOR TRIP SYSTEM INSTRt9GITATION .

l E: NINIORSE i

5 -

TOTAL NO. CHA191ELS CHAf01ELS APPLICABLE

] FUNCTIONAL UNIT . OF CHAl01ELS TO TRIP OPERABLE N00ES ACTION Manual Reactor Trip

1. 2 1 2 1, 2, and

12

, 2. PowerRange, Neutron,Fium A. High 4 2 3 1, 2 2,,

B. Low 4 -

2 . 3 2 2

3. Power Range, Neutron Flux 4 2 3 1, 2 #

4 .

2 1 High Positive Rete ,

4. Power Range, Neutron. Flux, 4 2 3 1, 2 #

. . 2 High Negative Rate -

S. Intermediate Range, Neutron. Flux 2 2 1, 2, and

8"

} , . 1 3 Y

6. Source Range, Neutron Flux A. Startup 2 1 2g 2 , and

4 B. Shutdown . 2 0 1 3, 4 and 5 5

7. Overtemperature AT Three Loop Operation 3 2 2 1, 2 /

Two Loop Operation 3 1 **

2 1, 2 9

8. Overpower AT k

g Three Loop Operation 3 2 2 1, 2 /

Two Loop Operation 3 1** 2 1, 2 9

9. Pressurizer Pressure-Low 3 2 2 1 /

! 5 10. Pressurizer Pressure--High 3 2 2 1, 2 /

l 5 l

C , . .

TABLE 3.3-1 (continued)

TABLE NOTATION With the reactor trip system breakers in the closed position, the control rod drive system capable of rod withdrawal, and fuel in the reactor vessel. ,

The channel (s) associated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped condition.

The provisions of Specification 3.0.4 are not applicable.

! " High voltage to detector may be de-energized above P-6.

~

" Indication only. ,

"The provisions of Specification 3.0.3 are not applicable if THERMAL POWER level > 105 of RATED THERMAL POWER, Ad*TTAM CTATEleckTc 1

-1 5

4 1

l I

r TABLE 3.3-1 (Continued) l ACTION 3 .With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a. Below the P-6 (Block of Source Range Reactor Trip) setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint.

b. Above the P-6 (Block of Source Range Reactor Trip) setpoint, but below 5% of RATED THERMAL POWER, restore the inoperable channel to 0PERABLE status prior to increasing THERMAL POWER above 55 of RATED THERMAL POWER.

c. Above SE of RATED THERMAL POWER, POWER OPERATION a v continue.

ACTION 4 - With the number of OPERABLE channels one less than required by .

the Minimum Channels OPERABLE requirement and with the THERMAL POWER level-

[

p

a. Below the P-6 (Block of Source Range Reactor Trip) setpoint, i restore the inoperable channel to OPERABLE status prior to . ,

increasing THERMAL POWER above the P-6 Setpoint. '

l

b. Above the P-6 (Block of Source Range Reactor Trip) setpoint, operation may continue.

ACTION 5 - With the number of OPERABLE channels one less than required by :j the Minimum Channels OPERABLE requirement, verify compliance with ;[

the SHUTDOWN MARGIN requirements of Specification 3.1.1.1~or i

. 3.1.1.2, as applicable, within I hour and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter. y ACTION 6 - With the number of OPERA 8LE 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.1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .,, *

b. The Minimum Channels OPERABLE requirement is met; however, one additional 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.

ACTION 7 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed until performance of the next required CHANNEL FUNCTIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

i l R

FARLEY-UNIT 1 3/43-7 AMENDMENT NO. 26 l

, . . =

=

REFUELING OPERATIONS

(' . .- WATER LEVEL - REACTOR VESSEL CONTROL RODS LIMITING CONDITION FOR OPERATION 3.9.10.2 At least 23 feet of water shall be maintained over the top of the irradiated fuel assemblies within the reactor pressure vessel.

APPLICABILITY: During movement of control rods within the reactor pressure vessel while in MODE 6.

ACTION:

With the requirements of the above specification not satisfied, suspend all operations involving movement of control rods within the pressure vessel.

The provisions of Specification 3.0.3 are not applicable.

i. -

SURVEILLANCE REQUIREMENTS 4.9.10.2 The water level shall be determined to be at least its minimum .

required depth within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter during movement of control rods within the reactor pressure vessel.

, FARLEY-UNIT 1 3'/4 9-12a AMENDMENT NO. 26 -

s l REFUELING OPERATIONS 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION ALL WATER LEVELS LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal loop shall be in operation.

APPLICABILITY: MODE 6.

ACTION:

a.

With less than one residual heat removal loop in operation, except as provided in b. below, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

The residual heat removal loop may be removed from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performance of COP.E ALTERATIONS in the vicinity of the reactor pressure vessel hot legs.

l f

c. The provisions of Specification 3.0.3 are not applicable.

\

}4 SURVEILLANCE REQUIREMENTS ,

4 4.9.8.1 A residual heat removal loop shall be determined to be in oper~ation and circulating reactor coolant at a flow rate of greater than or equal to 3000 gpm at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

1 l

i. .

FARLEY-UNIT 1 3/4 9-9 AMENDMENT N0. 26 l

1

e ..

3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES i

0PERATING LIMITING CONDITION FOR OPERATION i 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be ,

OPERABLE: 1

a. Two physically independent circuits from the offsite transmission network to the switchyard and two physically independent circuits from the switchyard to the onsite Class 1E distribution system, and

b. Two separate and independent diesel generator sets (Set A: DG 1-2A and DG-IC, Set B: DG-18 and DG-2C) each with:

1. Separatedaytankscontainingaminiisumvolumeof900 gallons of fuel for the 4075 kw diesel generators and 700 gallons of fuel for the 2850 kw diesel generators.

2. A separate fuel transfer pump for each diesel.

. c. A fuel storage system consisting of four, independent storage tanks each containing a minimum of 25,000 gallons of fuel.* '

APPLICABILITY: MODES 1, 2, 3 and 4.

v ACTION:

a. With an offsite circuit inoperable, demonstrate the OPERABILITY of the remaining offsite A.C. source by performing Surveillance Require-ment 4.8.1.1.1.a within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, and performing Surveillance Requirement 4.8.1.1.2.a, items 1, 2, 3, 4, and 6 on diesel generators 1-2A and IB within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months unless such surveillance has been performed within the pre-vious 7 days. Restore at least two offsite circuits to OPERABLE status within,7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The provisions of Sp'ecification 3.0.4 are not applicable.

b. With one diesel generator set inoperable for reasons other than the yearly scheduled maintenance ** demonstrate the operability of the remaining A.C. sources by performing Surveillance Requirement 4.8.1.1.1.a within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months thereafter, and per-forming Surveillance Requirement 4.8.1.1.2.a items 1, 2, 3, 4, and "One operable fuel storage tank must be available for each required diesel generator. l

' **If this scheduled maintenance exceeds 10 days,. the diesel generator set must be

declared inoperable. j i

i l

FARLEY-UNIT 1 3/4 8-1 AMENDMENT NO. 26 ,,

l ELECTRICAL POWER SYSTEMS r-ACTION (Continued) 6 on two* diesel generators within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Restore the diesel

- generator s' et to CPERABLE status within 18 days or be in at least HDT SHUTDOWN following30 hours. within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the The provisions of Specification 3.0.4 are not applicab e if only one of the four diesel generator units is inoperable.

c. With one offsite circuit and one diesel generator set of the.above required A.C. electrical power sources inoperable for reasons other than the yearly scheduled maintenance,** demonstrate the OPERABILITY of the remaining offsite A.C. source by perfoming. Surveillance Require-ment 4.8.1.1.1.a within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and performing Surveillance Requirement 4.8.L1.2.a. itans 1, 2, 3, 4, and 6 on two* diesel generators within 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months . Restore at least one of the inoperable sources to OPERA 8LE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HDT STAND following8Y 30within hours.the Restore next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the i the other AC power source (offsite 1 circuit or diesel generator set) to OPERA 8LE status in accordance

- with the provisions of Section 3.8.1.1 Action Statements a or b, as appropriate.

d. I With two of the above required offsite A.C. circuits inoperable, i demonstrate the OPERABILITY of both diesel generator sets by per-forming Surveillance Requirement 4.8.1.1.2.a within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ; unless the diesel generators are already operating; restore at least one of .s the inoperable offsite sources to OPERA 8LE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or j

be in et least NOT STA MBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . With only one *Q offsite source restored, restore both offsite circuits to OPERA 8LE status within 7 days from time of initial loss or be in at least HDT STAMBY following 30 within hours. the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the

~

e.

With both of the above required diesel generator sets inoperable, demonstrate the OPERASILITY of two offsite A.C. circuits by per-forming Surveillance Requirement 4.8.1.1.1.a within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and performing Surveillance Requirement 4.8.1.1.2.a, items 1, 2, 3, 4, 1 and 6 on ene diesel generator in a diesel set on the other' Unit -

d within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; restore at least one of the inoperable diesel gen- !

erator sets to OPERA 8LE status:

i "Ine two diesel generators chosen to be tested shall verify that at least one train of LOCA/ shutdown loads is capable of being powered at each Unit.

- If this scheduled maintenance exceeds 10 days, the diesel generator set must be declared inoperable. -

. k FARLEY-UNIT 1 3/4 8-2 AMENDMENT NO. 26

/7/ 65 b >

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 24 ANSWERS -- FARLEY 1&2 -86/07/14-NELSON, D.

ANSWER 5.01 (2.00)

a. 1. The reactor with less FDM. (0.5)

2. The reactor with less 50M. (0.5)

b. 1. Does not affect critical rod height. (0.5)

2. The larger the initial count. rate, the higher the power level at criticality. (0.5)

REFERENCE NUS Nuclear Energy Trns. ,

ANSWER 5.02 (1.00)

A l REFERENCE !

Thermo LP and steam tables i ANSWER 5.03 (1.00) i a

I

. REFERENCE l General Physics, HT & FF, Chapter 1 004/020; K5.08 (2.3/2.6) i ANSWER 5.04 (1.00) b REFERENCE General Physics, HT&FF l

L

s. .

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 25 ANSWERS -- FARLEY 182 -86/07/14-NELSON, D.

ANSWER 5.05 (2.00)

a. Equal to (1,0)

b. Less negative than (1.0)

REFERENCE Reactor Theory Manual ANSWER 5.06 (2.50)

a. Increase

b. Decrease

c. Decrease

d. Increase f 2.5

e. Decrease CO.4 each] 4GT4a REFERENCE General Physics PT&FF, Part B, Chapter 1, pp 324-332 ANSWER 5.07 (2.00)

c. GREATER THAN (0.5)

b. LESS THAN (0.5)

c. SAME AS (0.5)

d. GREATER THAN (0.5)

REFERENCE NUS, Vol 3, Unit 11 001/010-K5.13 (3.1/3.6)

ANSWER 5.0E (1.50)

a. DECREASE (0,5)

b. DECREASE (0.5)

c. INCREASE (0.5)

6 ', s

5. THEORY OF NUCLEAR' POWER PLANT OPERATION, FLUIDS, AND PAGE 26

' ANSWERS -- FARLEY 182 -86/07/14-NELSON, D.

REFERENCE General Physics, HTEFF, p. 320 ANSWER 5.09 (3.00)

1. FNDH. (0.5)

2. FNDH. (0.5)

3. . Rods within a group are maintained within +/- 12 steps.

Control rod banks are sequenced and overlapped.

Rod insertion limits are maintained.

AFD limits are mainteined. CO.5 ea.] (2.0)

REFERENCE FNP T/S 3/4.2 and Basis.

ANSWER 5.10 (2.50)

a. Tsat for 1200 psis is 567 F (from steam tab 1cs). '

567 - 5D = 517 F (subcooling of 50 r' J 76 au 1%'

Pset for 517 F is about 800 psis. .c, 4 (1.0)-

uti nat'

b. Erratic pressuriner level indication. ~.

~ .

n.

(0.5)

c. 65 F. . (+/- 3 F.) 6'r/"5 6* pro 10 minutes. (1.0)

REFERENCE FNP EOP 7-1, pp.12 & EOP 7-2 pp. 8.

ANSWER 5.11 (1.00)

-Minimizes thermal stress due to more uniform tem; difference of fluids

-The outlet temp of the colder fluid approccher the inlet temp of the hotter fluid

-A more uniform heat transfer rate is achieved throvshout the heat er:c h e nser (+.33 ea)

(more efficient is an acce able response also)'

REFERENCE #L S" CNTO, ' Thermal / Hydraulic Principles and Applications *, pp 5-10 004/020; K5.02(2.5/2.9)

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 27 ANSWERS -- FARLEY 1&2 -86/07/14-NELSON, D.

ANSWER 5.12 (2.50)

a. Flux will be depressed toward the bottom of the core E0.52 due to:

1. Lower control rod level E0.25] and

2. Xenon buildup in top of core E0.25] (1.0)

b. Flux continues to be depressed more and more towards the bottom as Xenon builds in top, then reverses as it decays off E0.75].

Control rod movement to compensate for Xenon changes reduces the flux shift E0.75]. (1.5)

REFERENCE Farley Reactor Theory Manual, pp I-3.15,3.16,2.10 GLJ 126 ANSWER 5.13 (2.50)

a. 1) Presence of a flaw (or crack of sufficient sire). E0.52

2) Low temperature E0.5:, (1.0)

b. Peduces the therral stress. (Reduced DT act ._ the RV wsil reduces t o t a l / t h e r r..a l / t e n s i l e stress.) (0.5)

c. Neutron exposure (integrated) E0.5_ makes the material more brittle (raises NDT) (Reduces duc.ility.) [0.52 (1.0)

REFERENCE WNTC Ther r:odynamics , Volume II, Chapter 13, pp 58-68.

JMF Nucleer Flant Technical specifications pp. B3/4 - 3/4 4-12.

ANSWER 5.14 (1.00)

The idle RCP can develop temperatures in the seal area that are less ther. stee" s e n e r e foi*Ife n p e r c t u r e s E0.53. When 1. h e cold slu: coer

+hrough the stear- generstoft M pic a s up. heat and expands. 'The ~

thermal expensicr. :n E solid plent cavers a pressure Increcte EO 53. (1.0)

REFERENCE A0D-Za GLJ 133

(

5. THEORY OF NUCLEAR POWER DLANT OPERATION, FLUIDS, AND PAGE 28 ANSWERS -- FARLEY 1&2 -86/07/14-NELSON, D.

ANSWER _> :. 59" jk a) 5. vnt by which core whould be suberitical (+.25) at hot shutdown conditions (540 des F) (+.25) if all control rods were tripped assuming highest worth rod fully withdrawn (+.25.) and no changes in Luenon or~ boron concentration. (+.25)-

b) sair^'d" Oceconted far ir SDM calc'.'Istienb (+.5)

FE bd Yt * # *

Q g[*

FNP T/S, pp 3.1.1 - 3.1.3 ANSWER 5.16 (1.00)

In the secondary system there.is a phase change (0.5 pts). A phase change requires a lerse delta h. With the larger delta h of the secondary, the same heat can be transferred with a lower flow rate (0.5 pts).

REFERENCE General Physics, HT & FF, Section 3.2 002/000-K5.01 (3.1/3.4)

ANSWER 5.17 (2.00)

a. 3

b. 1 ti.0 each3 REFERENCE Westinghouse Thermal Science, Chapter 7, Pp 32-46.

6

.l l

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 29 ANSWERS -- FARLEY 1&2 -86/07/14-NELSON, D.

ANSWER 6.01 (1.00) a REFERENCE FNP, Reactor Protection Sys, p. 50 ANSWER 6.02 (1.00) c.

REFERENCE FNP, LP' PZR Press. and LVL Cont.

ANSWER 6.03 (1.00) b REFERENCE FNP, Main and Reheat Steam Sys. p.8 ANSWER 6.04 (1.00) e REFERENCE Nuclear Power Plant Instrumentation Sys. Manual, Ch. 4' ANSWER 6.05 (1.00) c REFERENCE FNP, Gross Csiled Fuel Detector, p. 7 s

~

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 30-ANSWERS -- FARLEY 1&2 -86/07/14-NELSON, D.

ANSWER 6.06 (2.00)

a. 1. No low-low Tavs (P-12) .

E0.2:

2. Steam-dump bypass interlock switch not in OFF RESET E0.23

3. Conderise r available'(C-9) . (Condenser vacuum and CW pump breaker _

at 0.1. pts. each also acceptable) E0.23

b. Positioning the MODE select switch E0.23 to RESET. CO.23

c. Difference between auct. hish Tavs and Tno-load (547 F) CO.43

d. 1.-Load rejection (C-7) CO.23 2 Turbine trip (C-8) [0.23

3. Mode switch in STEAM PRESSURE 'CO.23 REFERENCE FNP, Steam Dump Sys. pp. 17,22,25,27,28 ANSWER 6.07 (1.50)

a. decrease

b. remain the sene

c. remain the same REFERENCE FNP, LP: RPS i

i ANSWER 6.08 (3.00)

a. rods out CO 253 Tref will be ma:: so Tave/ Tref mismatch and NI/ Turbine Power mismatch will both give a rods out signal E0.752

b. rods in E0.253 Loop 1 Tave increases and auctioneered'high Tave also increases. Teve/ Tref misnatch gives a rods in signal CO.753

c. rods out E0.253 the power mismatch circuit of the reactor control unit responds only to rete of change of deviction between turbine anC nucleer power but rod motion will occur due to the Tave - Tref difference. E0.753 REFERENCE FNP-, LP: Rod Cont. Sys.

i i

h Y __ . - - _ - _ _ _ - . - _ - - - _ _ _ - _ - _ - _ - - _ - _ _ - - - - - - - - - - - - .

C

,& 'O 6, PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION- PAGE 31

! ANSWERS --'FARLEY 1&2 -86/07/14-NELSON, D.

1 ANSWER 6.09 (2.50)

1. DTDT.

l 2. OPDT. ,

i

3. .0PDT. ;

4. Both.

5. OTDT. .EO.5 ea.] (2.5)

REFERENCE

[

FNP Lesson Plans, Vol. 7, pp. 5 & 6.

FNP Technical Specifications, pp. B2-4'& B2-5.

1 ANSWER 6.10 (3.00)

Normal - Emergency 600V LC-1D to eOOV MCC-1A to Inverter.1A which

supplies 120V to Vital AC-1A instrument distribution panel Backup - Emer3ency 600V LC-1D to Battery Charger 1A (or 1C) which supplies 125 V to DC bus 1A to Inverter 1A (and as above)

Alternste - Emergency 600V LC-1D to 600y MCC1A to 600V/208V transformer to SOLA 1A which supplies 120V to Reg AC-1A to Vital AC-1A instrument distribution. panel E1.0 each] (3.0) g(. ,r ,p ~ 5 5 y N s (b c .h s lk ) k tu d r 1 4 k, v'Jc3 A t -n e a REFERENCE Farley Lesson Plans Volume 4, Tab 1, Fig. 2 & Tab 3 Fig. 2 GLJ 143-4 ANSWER 6.11 (3.00)

1. 543. des F 2/3 < S.P.

-L i t ' S . "' . F-_r-iss4cn-te blecP C I-- '

2. 75% 2/3 on 1/2 SG's S.P.

} 3. Variable: 1/2 >= S.o._in 2/3 loops 40% for P<20%

increases linearly 40 - il0%'for P>20% 3 tro7.

I Tavs < P-12 EO.f' each st. Pt. ; E0.53 each coint.

REFERE,LE j FNP, License Retrang, SG Protect.

J i

i l

i !

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE- 32 ANSWERS -- FARLEY 1&2 -86/07/14-NELSON, D.

' ANSWER 6.12 (1.00)

'Available in normc1 and abn_ormal conditions in all modes e:: cept refueling.

REFERENCE FHP, License Retrns, ICCMS ANSWER 6.13 (3.00)

a. Low FWP suction pressure (300 psis -for 10 sec)-starts the standby condensate pump E0.53. Low FWP suction pressure pressure (300 psis for 30 see) causes FWP trip E0.53. (1.0)

b. Valve auto closes on FWP trip.sj,gnal from both pumps.with control switch e T (0.5) c.

in AUTOGCMTs.' s\d'lm A kLL %el c%s } k> 172I 1.-Hi-Hi S/G level - cleared by; closing re' actor trip breakers

2. SI - cleared by closing reactor trip breakers

3. Lo Teve E P cleered by manual reset button (1/2) on the MCB to reset La Tave signal EO.5-each] (1.5)

REFERENCE Farley Lesson Plans Volume 4, Tab 7, pp 28,29,36 & Volume 6, Tab 11, pp 18,19 GLJ 149 ANSWER 6.14 (2.00)

a. Manual signal calling for rod movement Rods move IN. [0.5oea.3

b. Tave-Tref deviation calling for rod movement however.with 'D selected rods DO NOT move.

c. Manual signal calling for rod movement however rods DO NOT cove

d. Tave-Tref deviation calling f or rod movement hovever 181P steps blockt m c v e n.e n t , reds DO NOT move. Ecaf] t40 RErERENCE FNP, LP: Rod Cont. S y *4 .

pr-i PLANT SYSTEMS DESIGNr CONTROL, AND INSTRUMENTATION PAGE 33 6.

ANSWEOS -- CARLEY 1&2 -P6/07/14-NELSON, D.

ANSWER 6.15 (3.00)

' ump s A & C:

controlled f r o r. MCB or HSP E0.25:

MCE' - STOP/AUT0/ START E0.252

- Opersble.only when ten ote selected at HSP E0.252 Avtc stcet:

1. ESF sequencer if B pump bkr open CO.252

2. LOSP tequencer E0.252

3. 9 pump fault trip :C.252 Per:p B controlled from MCB or HSP E0.252 MCP - ST0r/AUT0/ START E0.252 ene for each train CO.252

- Operable only when remote selected at HSP E0.252 Auto stret:

. Selected train charging pump trips E0.252 2 ' O S ;' o r ESP sequencer with selected train charsing pump bkr racked out E0.252 REFERENCE FNP, License Retrns, ECCS ANSWER 6.16 (1.00)

A de*:vative function in the control circuit provides anticipatory valve response.

or o1D controller with shapirrs function senerator.

REFERENCE FNPr License Rettns. RCN's

b f: 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 34

--- EE5i5E55i5AE 56siE5E------------------------

ANSWERS -- FARLEY 1&2 -86/07/14-NELSON, D. ;

I 4

ANSWER 7.01 (1.00) i C -j REFERENCE 10 CFR 20.101 ANSWER 7.02 (1.00) d REFERENCE 10 CFR 20. i ANSWER 7.03 (1.00) d REFERENCE FNP 1-ESP-0.2.

ANSWER 7.04 (3.00)

a. 1,3,5,6,7

b. 1,2,3,4r6 CO.3 each3 (3.0)

REFERENCE EDP-0 GLJ 160 ANSWER 7.05 (1.00)

a. 28 des F

b. 88 des F

c. RTD

d. 7% CO.253 each REFERENCE FNP-1-ESP-1.1 L

I

. . . I

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 35

~ ~~~~~~~~~~~~~~~~~~~~~~~~

~~~~Rd656EE55CdL U6hTRUL ANSWERS -- FARLEY 182 -86/07/14-NELSON, D.

l ANSWER 7.06 (3.00)

a. RCS borated to at least the cold shutdown concentration (0.5)

b. 350 to 425 psis (0.5)

c. Heatup 100 F/hri cooldown 200 F/hr (1.0)

d. High E0.5J due to measured les density greater than when hot (1.0)

REFERENCE FNP UDP 1.1 i TS ANSWER 7.07 (2.00)

A. 1. 150 spm.

2. 350 spm.

3. 700 spm. [0.5 ea.] (1.5)

B. Service water does not nect secondtry makeup specifications. (0.5)

REFERENCE FNP-1-SOP-22.0 ANSWER 7.08 (1.00) 0 0.25 points each'

1. Stop/ confine spill.

2. Clear unnecessary personnel from stes.

3. Notify Control Room.'

4. Notify HP Office.

REFERENCE rNP-0-h-001.

ANSWER 7.09 (1.00)

1. Place n.ain feed res valve in r.anual and maintain SG program level.[0.52

2. Place 1E spray valve in nenucl and ensure it's fully closed. CO.52 l

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-________-_____________=_-____-____-____-

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7.- PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 36 I

ANSWERS -- FARLEY 1&2 -86/07/14-NELSON, D.

i i REFERENCE 4

FNP-ADP-4.0. -

! .l1

ANSWER 7.10 (1.00) 5 O'3 t' U W " U ' N .f m), Q,l S& br: q N 4 56,0 4

1. Place rods in nanus 1. ,

! 2. Place turbine DEH in hold.

i 3. Insert control bank D to return Tavs to prostem. l REFERENCE !

FNP-1-AOP-19.0. I 1,

! ANSWER 7.11 (2.50) t R-2-(enmnt area monitor)

! R-5 (spent fuel bids. area nonitor)

R-11 E R-12 (enmnt seseous and particulate monitors)

R-24 A E E (cnmnt purse moni. tors)

! R-25 A tE (fuel handlins monitors)' [0.52 each

, REFERENCE FNP-AOP-30.0 ANSWER 7.12 (2.00)

-Start standby condensate pump.-

j -Reduce turbine load.

j -Place standby SJAE in service.

-Start additional circulating water pumps. . .

-Stert additional cooling tower fans.- ES e 0.4'es.2 (2.0) j' RECERENCE FNP-1-AOP-8.0, pp.1, 2.

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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 37 'I 1

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~~~~Rd656L6656dL"665TR6L ANSWERS -- FARLEY 1&2 -86/07/14-NELSON, D.

i ANSHLR 7.13 (3,.00)

I

a. To mi:: water for samplins (0.5)

b. Flowpath aligned from RWST to CVCS letdown (without exceeedins i 130 spm thru LTDN HX) until baron concentration is equal to or greater than RCS boron concentration. (1.0)

I c. Increase (0,5) i d. RCS overpressure protection provided by RHR inlet relief valves. (1.0) l REFERENCE FNP SDP 7.0 GLJ 163 ANSWER 7.14 (1.00) .

Reduce nakeup flow. q 4g REFERENCE ,

4 FNP-EEP-3.

1

) ANSWER 7.15 (1.00) l When charsins pumps cannot maintain pressurizer level.

REFERENCE SL /a#hy 'y T[f /j,.',k FNP-1-ADP-1.0.

ANSWER 7.16 (1.00)

Start /Stop charging pumps.

! REFERENCE i FNP-1-ADP-6.0.

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

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 28

~~~~~~~~~~~~~~~~~~~~~~~~

~~~~R5656L6G565[~56UTR6L ANSWERS -- FARLEY 1&2 -86/07/14-NELSON, D.

ANSWER 7.17 (1.00) p.[ks.

> 4psis in containment CO.53 OR > or = 10E5 din containment E0.53.

REFERENCE FNP-EEP-Or Fold Out Pase.

ANSWER 7.18 (1.50) t 0 0.5 points each'

1. By the FRG direction.

2. Any red path occurs.

3. An higher priority orange path occurs.

REFER 33 CE I 4 po++s FNP ERP Requal Training.

ANSWER 7.19 (2.00)

1. start boric acid transfer pump

2. open emergency borate to charging pump valve (MOV-8104)

3. verify one chargin3 Pump running and charging pump suction header isol.

valves are open. (MOV-8131A & B)

4. Increase letdown flow (to 120 spm).

REFERENCE FNP ADP-27.0 t

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 39 I' ANSWERS -- FARLEY 1&2 -86/07/14-NELSON, D.

I

ANSWER 8.01 (1.00) c.

REFERENCE North Anna TS 3/4 1-1,-9,-12,-22

-Farley TS 3//4 i

ANSWER 8.02 (1.00) a.

REFERENCE FNP TS 3.7.1.5/3.7.1.2 ANSWER 8.03 (1.00) a.

1 REFERENCE FNP, T5 A N S W'_R 8.04 ( .50)

False i

! REFERENCE 10 CFR 55-31 ANSWER 8.05 (1.00) j All full length control element assemblies shutdown and res. are fully

! inserted [0.52 except for the single assembly of hishest reactivity worth which is assumed to be fully withdrawn.[0.5]

' REFERENCE

, St. Lucie Tech Spec def. 1.29 North Anns TS def.

FNP-def, a

fN * $ C$$ N / b Ce l c~r i;nM b.g} wd L. A ski. rd LA .

l asseJy g 6 % c e m A 4 % + c h is m-.2 A L % AG.

[ , ..

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 40 ANSWERS -- FARLEY 1&2 -86/07/14-NELSON, D.

ANSWER 8.06 (1.00) 1.02 (+.5 ea)-

1.09 REFERENCE NA U1 TS 3.2.4 FNP TS U1 001/050; PNG-5 (2.9/4.3)

ANSWER 8.07 (1.50.)

A. Shift Supervisor.

B. Emergency Director.

C. Channel 5. E3 0 0.5 ea.] (1.5)

REFERENCE FNP-0-EIP 2r p.2; EIP 3, p.1, EIP 8, p. 1.

ANSWER 8.08 (2.00)

1. The intent of the procedure is not altered. E1.03

2. Must be approved by two members of the plant staff E0.5]

at least one of whom holds a Senior Reactor Operators licerise. E0.53 (2.0)

REFERENCE FNP Technical Specifications, Sect. 6, pp. 6-12, 6-13.

ANSWER 8.09 (1.00) 100 des F RCS 100 des F Pr:r REFERENCE FNP U-1 TS RCS

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE- 41 ANSWERS -- FARLEY 1&2 -86/07/14-NELSON, D.

ANSWER 8.10 (1.50)

A. The GPTR is deternined by using the incore moveable detectors. (0,5)

B. Reactor power must be reduced (to less than 75 %) [0.53 and the Power Range high neutron flux trip setpoint must be reduced (to < 85 % within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .) [0.53 (1.0)

NEFERENCE FNP Technical Specificationti pp. 3/4 2-13, 3-6.

ANSWER 8.11 (3.00)

A. Interval requirement not exceeded E0.53. Eight days does not exceed 1.25 times the specified interval C1.03. (1.5)

B. Interval requirement exceeded E0.53. The last 3 consecutive intervals exceed 3.25 times the specified interval [1.03. (1.5)

REFERENCE CNP Technical Specifications, pp. 3/4 0-2, 5-11.

ANSWER 8.12 (2.50)

A. 1. Move the valve handwheel in closed direction and return valve to original position. (0.5)

2. Remove the locking device and move the valve in the closed direction, then return to original position.

Re-install the locking device. [0,53 A second person verification is required to verify proper installation of the locking device. [0.53 (1.0)

3. Attempt to close. (If velve is in the correct position. no motion will occur.) (0.5)

4. Verify locking device locked and visually verify vcive is in the specified position. (0.5)

REFERENCE FNP-0-AP-16, pp.-13, 4.

c s

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B. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITA110NS PAGE 42 ANSWERS -- FARLEY 122 -86/07/14-NELSON. D.

ANSWER B.13 (1.50)

-Modes 1,2 9e in HSB with pressure within its limit (in one hour) [0.52

-Modes 3,4,5 Reduce pressure within its limit (within 5 minutes)

E0.52

-All Modes hetify the NRC Operations Center immediatly (within one hour.) OR (comply with Admin. T.S. 6.7.1. [0 52 (1.5)

REFERENCE FNP Technical Specifications, pp. 2-1, 6-14.

ANSWER G.14 (1.00)

LCD 3.0.3 applies REFERENCE North Anna LCO 3.0.3.

F N :' 3.0.3 ANSWER 8.15 (1.00)

a. upon functional acceptance of the work requested.

b. Shift Foreman E0 5 es.2 REFERENCE FNP-0-AP52 p.p. 6 27 ANSWER 8.16 (1.00)

FN'-TS 3.0.5 REFERENCE FNP TS 2.0.5 4

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 43 ANSWERS -- FARLEY 182 -86/07/14-NELSON, D.

ANSWER 8.17 (1.50)

a. Operations Superintendent Operations Manager [0.25 es.]

Emergency Director

b. Yes E0.253, (however it does not preclude both personnel.from being at the same general area at the same time)

The resultant verification will be documented on the work request, work request authorization, or on the controlling procedure (one required) [0.53 REFERENCE FNP-0-AP-16 p. 43 ANSWER 8.18 (1.50)

a. On an LCD status sheet E0.53 by noting on the top of the form by marking the Administrative space [0.53.

b. Yes [0.53.

REFERENCE FNP-0-AP-16 p. c-2 ANSWER 8.19 (2.00)

a. At the valve and at the supply breaker / air supply E0.5 ea.]

b. Shift Supervisor [0.53

c. A notation of 'See Attached * [0.53 REFERENCE FNP-0-AP-14 p.p. 4,7 ANSWER B.20 (1.00)

TS 3.9.C.1.b e9 ".C..z .: 1,20.I.5' REFERENCE FNP U1 TS 3 9.8.1.b l

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 4c ANSWERS -- FARLEY 1&2 -86/07/14-NELSON, D.

ANSWER B.21 (1.50)

3. General Enersency [0.5]

b. Return to a safe area contact the Emersency Director for further instructions E0.5 ea.]

REFERENCE FNP EIP 19 EIP 14 ANSWER 8.22 (1.00)

The Shift Supervisor fo- each specific alteration must approve E0.5]

The individual performin the alteration and the Shift Sup. [0.25] must datormine the the alteration will not prevent the system / component from Performins its designed function E0.25]

REFERENCE FNP-0-AD-13 p. 3

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3 M 5%

U. S. NUCLEAR o E G U '.. A T O R Y COMMISSION REACTOR OPEPATO:' LICENSE EX6MINATIOb EACIL11Y: FARLEY 1E2 REACTOR TYDE: PL'9-LEC3 OATE c0 MINISTERED' 86/07/14 EXAMINER: NE EUN. D.

A o t" 1 C A N 1 : _________________________

INSTRUCTIONS TO Act'LICANT:

Use separate peper for the answers. Write answers on ene side only.

Staple question ' sheet on top of the answer sheets. Points for each question are indicated in parantheses after the question. The passing grade r e gt*: r e s Et letst 70% in each category end e final stade of et least SO%. E::anination papers will be picked up s i >' (6). hours after the e >: r m i n s t i o n stetts.

v. nr CATEGORY %0 A D F L I C A L"' ' E CATEGOPY VALUE T07AL SCDCE VALUE CATEGORY 30.00 RINCIFLES OF NUCLEAo 00WEP

________ _'5.00_____ ___________ ________

1.

PLANT OPERATION, THERMODYNAMICS.

HEAT TRANSFER AND CLUID F'_0W

'E 0

[o

_'_1_00____::1__0 ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 30 F

___100_-_ 2'11_00_ _-_--_____- _--_____

3. INSTPUMENTS AMD CONTROLS on or 1_00___ _[ 1_o;n ___________ ________

c. ROCEDURES - NORM A' . A9NORMALr EMERGENCY AND RADIOLOGICAL CCNTROL 120.00 100.00 TOTALE cTNa- GeA9c- _________________~ w All wort done on this c:::nination is ny own. I have neithet siven ner received s. t .

AotLICANE EIGNATURE k_

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I NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS t

During the administration of this examination the following rules apply:

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

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2. Restroom trips are to be limited and only one candidate at a time may l 1 eave. 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. l S. Fill in the date on the cover sheet of the examination (if necessary).

6. Use only the paper provided for answers.

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

section of the answer sheet.

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

5

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

10. Skip at least three lines between each answer.

j 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 I

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 i AND DO NOT LEAVE ANY ANSWER BLANK.

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

the examiner only.

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

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

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18. When you complete your examination, you shall:

a. Assemble your examination as follows:

(1) Exam questions on top.

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

(3) Answer pages including figures which are a 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 1 in progress, your license may be denied or revoked.

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. t :3 NCI;'LES D HUCLEAR t OvE - r L AfD O r g c:g 7 3 e y , p499 7

~'~~IU E E5665UdE5CEI~UEET ~TE5U5EE E ZG5 ECUi5 FL60 90ES' ION 1.01 (1.00)

Wh;ch of +5c +o .owin3 .ta*_ement' concerr:ng s a r.a r i v t r e. s e t i v i t *,

effects :s cor eM ?

2 The equilibrivn (at power) value o# t e n.s r i u m is dependent upon powe* .L e v e l . The peek velve of s e n e r d u c- following c shutdown is dependent upon power level prior to shutdown.

b. The equilibriun (et power) vc1ve of senerium it dependent upon power level. The peak value o# samarium following a shutdowr i t- :ndependent of powe* 1evel prior to shutdown.

c. The equilibrium (at power) value of samarium is independent of power level. The peak value of samcriun following a shutdcwn is dependent upon power level prior to shutdown.

d. T% e :;ui : i b t i v e- (Et power) value o# semer:un it Independent o' oeuct leve'. The peak value of samarium fallov qs a s5M dow- :s independent of power leve: prior to shutdown.

CUEE' ION 1.02 (1.00)

Which of the following will result in the largest INCREASE in the concentrat:on of dissolved gases in a quantity of water? (Assume the changes in temperature and pressure below are of equal magnitude)

c. Incressing the pressute end lowering the temperature.

b. Decreasing the pressure and lowering the temperature.

c. Ine'-etting the pressure end reising the ten.perrture,

d. Decreasing the pressure end raising the t e n.p e r a t u r e ,

tzrrr2 CATEGORY 02 CONTINUED ON NEXT PAGE 2>z'**)

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1. PRINCIPLES OF UUCLEAR DOWER PLANT OPERATIDri, PAGE 3

~

~~~~IEEEUUbYU5kEC57~UEdT ~I E53EEE~5UU~ELU59 R EEUU DUESTION 1.03 (1.00) c WSich of *be ollow:ng equatio^2 used to p e r o r n 3 RWR heat balance erleulttion is correct?

2. O r :' = M(s) Eh(s) - h(fw)2 + M(bd) Eh(bd) - h(fu)2 9 Orcp

b. Orr = h(s) th(s) - h'fw)2 + "(bd) Ch(bd) - h(fw)2 - 9tep

c. O r :* = M(s) Ch(s) - h(fw)2 - M(bd) Eh(bd) - h(fw)2 - Orcp

d. 0 7 :- = M(s) Ch(s) - h(fw)2 - M(bd) [h(bd) - h(fw)] + Grep NOTE: Notation Key 0 = Pcwer M '= Mass Flow Rate fw = Feedwater r :: = Resctor bd = 910wdown s = Steam h =. Specific Enthelpy GUESTION 1.04 (1.00)

T"e +rospheric D00V for 'O' E/G partially opens to : throttling position dur in g opere".cnt it 8 5 ). power. Whier o' the following descr:bes the

t s '. e of the stean on the downst ear. side of the PORV?

e. Het Steen

b. Superheated Steam

c. Setursted Steen,

d. Ecturated Liquid

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

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1 ORIrJCli L ES OF NUCLEAR POWL R o L A tJ T 3PERMION, PAGE a

~~~~IEEE5655U5EICSI~UE5T IIE55EIEE'5U6~ELUIU~EEUU DUESTION 1.05 (2.50)

Assume one 60' tvios at 1% power , without a reactor protectiore syster cetuti:on or e chenge in turbine I ti e d . Indierte whether the f ollow:ng par ar.eters will INCREASE, DECREASE, or REMAIN THE SAME.

a. clow in the operatino reactor coolsnt loops. (0.5)

b. The rrtie of core flow compered to the totel loop flow. $ Y (Core clow/ Total Loop Flow) lc-f2 (0.5)

c. Retctor Vessel deltr pretture. (0.5)

d. Core delta temperature. (0.5)

e. Operrting loop steet generator temperstores. (0.5)

GUES~~ ION 1.06 (2.00) an EC ' :s c21culated #or I ctartup a hours s'ter a shutdown ""om 100?.

s.src; sttte power. ' Int: c E te wh( the r the ectuel cr.ticri position w;12

t 0: :EATEP TuAN7 LEEE THAN or the SME AS the ECF for the follcuing conditier.s.

a) The steam dump pressure setpoint is increased by 35 psig. (0.5) b) All steam generator levels are increased by 5% five minutes before criticality is aeached. (0,5) c' Co. denser v a c u u n, is reduced 3' He due to e smell cir leek (0.5) d) Peactor startup is delayed #or 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . (0.5)

(

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CATEGORY 01 CONTINUED ON NEXT PAGE z~zit*)

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1. PRINC]PLES OF NUCLEA 'OWER PLANT UPERA710N, PAGE

TRERMODYNAMICE, HEAT TRANSFER AND FLUID FLOW DUESTION '.07 (1.50) 2nd::s whethe the fo~1ouir.s st t t r, $: its concernina delayed neetr m tre TRUE o* CALSE.

a. If the reactor is supercriticair then the fraction of delayed neutrons shifts to the shorter lived precursors end the velve of the effective decay constant (lanbda) decreases. (0.5)

b. Due to the significent decrease in the percentage of fact fission occuring over core life, the value of the effective de: eyed neutror. freetion decreases over core life. (0.5)

c. Delayed neutrons are produced at some time after fission as E result o' the radioective deccy of fission products. (0,5)

OUES'!ON 1.09 (3.004 N 'tr- C '. u ' Het C5tnnel r c et e' (FGZ) enc' Nucl ee r Enthc1py

'^*

~_+ s m . e l Czcter (TN0H' ere both power distribution limits.

2. W h : c .- . i - i *. it etico rict usins t Rod Bow per.tity besec~

on che core recion average burnup0 (0.5)

2. Which limit is defined es 'The retic of the integral of linear power along the rod with the highest integrated power to the everege power'? (0.5)

2. Tee"nical Specification surveillance requirements using in-cere detectort it infrequent provided that FOUR itemt are onitored a rid verified to be within their limits. What are these fou- itette ( 2 . 0 '-

irn** CATEGORY 01 r.;0NTINUED ON NEXT PAGE **u a) i i

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. tci v icLES or 90ctE p powEP ptAN1 OtEcATION, PACE 6

~~~~T55EEU655555CEI~5E5I'5E505EEE'556~EEU5D'E[6U GUESTION 1.09 (2.50)

2. I' -- i ; coolecm en , . ' tv s ' ir:vistion- ttw :T S pres. sir-c ':s

200 ps:s, v5-- vov1d be t h e n e r i m o ni s t e e n. sener etor precs er e te assvae a d e r,v ? t e subcoolinci 9% jLrgM (1.0)

6. During nrtur-c1 c : v e ul e t i cr- cooldown, e s'tecr. bubble mey #ctm in the reactor vessel head area. What is the primary indication c' this bubble fornatione (0.5)

c. What is the .3: ir.um cor e Delta T enp. which would be indicative of PROPER netveci recircultiion flow following a full power trip AND what is the approximate loop transit t i n. e ? (1.0}

Q U E S T I P '.' 2.10 (1.00)

List three significant heat transfer advantages of 2 counter #1cw heat exchanger ever t perellel flow heet exchanger.

QUESTION 1 11 (2.50)

During 100% power operatione it is decided to reduce power by 20%

vsing control rods only for recetivity c.ontrol.

a. Explain HOW AND WHY the axial flux shape will change

'or the first hour sfter the power reduction. (1.0)

5. E::p l a in F0W AND WHY the fivr shape will change over the ne::t 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Include the effects of control rod movement to maintcin power stable. (1.5)

Crp cLS% tok Ny AS0m OCY -

(*arra CATEGOc? 01 CONTINUED ON NEXT PAGE *n n i 1

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1. PU NCl* LEG Or NULLEAR POWER DLAN1 OPE 0ATION, PAGE 7

~~~~I55EC 06YU5h5C57~5E5I~5EIU5 FEE ~5U6~EE656~EL6E GUESTION 1.12 (2.00) h le'ly E y r.L A I S hov the eddit.e* e' O.b% positive reectivity i to a suber:tical reactor would affect the following: (No crievicticer i ett receited.)

a. THE CHANGE IN THE COUNT PATE: (if the reactor was slightly soberitical t hetdown m er gi ri = 1%? et compar ed to grectly suberitical Eshutdown margin = 5%2). (1.0) l'

b. THE TIME TO REACF A STAPLE COUNT RATE: (for the different j shutdown nargin eqnditions in (a) above.) (1.0) t I

GUESTION 1.13 (2.50)

a. Brittle fracture o' any carbon steel pressure vessel can occur et riresses well below yield strest if TWO other conditions are 1 present. What re these TWO conditicns? (1.0)

b. How de hettup/cooldown irte limits on the recetor coolent i system reduce the probability of brittle frscture? (0,5)

c. Why does the concer n ebout bt ittle frecture of the reactor Pressure vessel increase as the plant ages? Include in i your answer the specific meterial PROPERTY that is effected. (1.0)

QUESTION 1.14 (1.00)

Explain how the starting of a Reactor Coolant Punp in a water-solid plant can cause e pressure tr aris i ent . (1.0) 9UESTION 1.15 '2,50) i

2. Define ONBR (0.5)

1. Whct is the limit on DNPP" (0.5) ;

c. Since the DNDR :: not a directly observable parameter, name SIX p risnetet s the opetetor monitors and/or controls te ensure the DNBR is not violated. (1.5) l (rirrr CATE00:!Y 01 COUTINUE0 ON NEXT PAGE rrrst I

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. PRItlCIFLES Or NUttEAR 'UWER PLANT OPERM IDN . r- (,G E E

~~~~I5ERU66YU 55C5I~EE5T'YU505EEI506'f[U 6~fE60 00ESTION 1.16 /1.00) o ra - cyste + : ., s t e. 3 -.1 9 icOS Stectef thi' tecondasy s yst e .

flow tite wh:2e the hett ticetferred by the tro s y t. t e n. t is et t er ti elly the sanc. E:: Plain how this is possible.

QUESTION 1.17 (2.00)

a. USy does nucleate boilins heat transfer renove note heat than non-boilins heet trent'er? E1.02

b. Why does file. boilins remove less heat than nucleate boilins? El.02 (rrrrr DD Or CATEGORY 0: rur r )

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2. PLAT' OESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMC DAGE t

"ESTION 2.01 (l.00)

Which c' the followin is NOT a design 'rature of the steam dump vr,1ve: S

t. De the tvo .a.vc. F. e

brni., one :t tt. 3ntc' to condent.c* A sne c.c 15 ass:gne to condenser 9

b. The usiver in t bato operrte (open cnd clote) togethet ,

c. Together the stest dump valves will pass 85% of rated stesn flav.

d. ThE vrives it one bank so f ully oper, bec ore velves in the next b e rd leave their shut seats.

QUESTION :.0; (1.00)

Accord:n3 to 10CFR50.c6, which of the following is NO- 2 decign criteria e' the Energency Cote Cc#.ing System subsys.ets;

a. The calcul:ted peat c e r.t e r li n e fuel temperature shall not s ::c e e d 2000 deg. F.

S. ~ ~. e . 2 :: I m u m civ_s:ns oxidation shall not exceed . T '. o f the total

~

c . e t' t'icknett.

c. The calculated total ancunt of hydrogen generated. f r c n. the cladding reaction with water shall not e::ceed 1% of the EBount that would be generated if all cladding around the fuel aeacted.

QUESTION 2.03 (1.00)

The S/G PORV's maximuni capacity i lirr.ited by design to apprc::liately 8%

of rated tie a flow. Which of the following it the retton for this lin.iti

a. Maintains nass discherse rate within the capacity of the condenser nci.eup s.y s t e n ,

b. Litits plani cooldown rate if any one FORV s t i Fs.s open.

c. t'.i n i m i r e t the pe t.ibility of S/C -Jif*erentiel pretsure SI i' eny one 20RV sticks open,

d. Kin r:::et erosion c# FORV valve seats.

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

2. PLAN 1 DESIGN INCLUDING W ETY AND EMERGENCY SYSTEMS PAGE 10 GUESTION 2.04 (1.00)

Which of the following signals, acting independently, will automatically CLOSE ths hei- Feedvete: E e _r v . E t i n g V e 2 v e s '-

a. Pi ui S/G-level in any S/G (2 out of 3 detectors)

b. Low Tevs (2 out of 3)

c. Reactor Trip

d. Phase B Isolation GUESTION 2.05 (1.00)

Choose the correct stetement concerning AFW:

a. TDAFP. low speed (coinc. with TDAFP star t signal) WILL NOT result in a common alerr, on the MCB.

b. TDACP AND MDAFF's in unit one have their miniflow flow control velves failed open.

c. Loss of instrument ai' will result in fail-safe openir.: :# TDACP ster supply velves 3' 35 A & B. -

d. Each AFP has discharge flow indication on the MCB.

QUESTION 2.06 (1.00)

The purpose of the CVCS demineralizers is to:

a. remove rll chemicals froe the RCS fividt

b. r e n.o v e solvable and insolvable material from the SCS.

c. rep:ece insolvable n.t tet : r; with solueble ions.

d. Provide a ethod for toro- control during eactor operaticrs.

(tzErr CATEGORY 02 CONTINUED ON NEXT PAGE ****.x) l l

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2. "La m DELIGN I N t '_ U' ING S AF El Y AND '_ MEPGENCY SYSlEMS t0' 0:JESTIGN 2.07 '1.50)

Indicate whether the #ellow:ng statemen+.- regarding PCP seals sre 19t!

e- CA'9E.

) 1he floating r ng seele loc 3ted between the pur.p radial bes:ir c enc the t'. s e c.1 c.11 lim:t l e c k.c a c to 50 spr. 11 the il seci ft:Is, (.5-b) 43 seal is designed to withstand ? v 1 '. RCS pressure. (,5) c) Seel weter inject:en ' rom CVC" entert the RCP between the tec1 ptekese and the p u r,p rad:a1 bearing. (.51 QUESTION 2.09 (1.00)

With tut reactor cooltre purps operating, indicate if the flow in the given loop sesnert wil: be in the NORMAL or REVERSED direction in the loop with t*e non-operating pump,

e. T-t t'D ntnifold

b. '" I' D manifold DUE E'IC- 2.09 (3.00)

What are ,he NORMAL, BACMUPr AND ALTERNATE power supply paths from the 600V LC's to the 120 vitel AC instrument distribution penel 1A?

Circuit breaker numbers are not necessary. (2.0)

GUESTION 2.10 (1.50)

List the i n t e r l o c .s and autoratic functions assoc:ated with the letdown isoletion v:Ives (LCV-45C end -460).

QUES'I'? 2 . '.1 '2.00?

The VC' el:ef valve design ' low ryecity is b sed ,n the sur u n :t 4 o# '1:'

r o n- whr 1our rov!cesi

(***** CATEGORY 02 CONTINL'ED ON NEXT PAGE *****)

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2. PLAN 1 ZEIGN INCLUDING SAFETY AND E :ERGENCY SYSTEMF rACL 12 00EST10N 2.12 (1.50)

List the pows- supplies for the fo11cuins RHR comp orient s

. RWS1 to ;'? pump A (h0V-E!?O?A)

b. CTMT sump tc RW pump B ( M O V- 8 E(11 B )

c. RHR purp B 00ESTION 2.13 (1.00)

How is a dedicated supply to AFW from the condensate storage tank assured?

What eutomrtie cetions occur, if eny, upon reaching lov level in the condensate storage tank-(while supplying AFW)?

DUESTION 2.14 (1.50)

List the inputs to the core subcooling .onitor.

QUESTION 2.15 (2.00)

Unde: what plant condit: ens is the reccior vessel im :. Indication systen NC' evrilable.for the operrtor's use?

GUESTION 2.16 (1.00)

Name the plant temperature (HOT or COLD) and the RCP stetus (sterting CIRST or LAST reactor coolant pump) for'which you would expect the highest pump starting current.

(r**** CATEGDRY 02 CONTINUED ON NEXT PAGE trrrr)

2. PLAN 1 DES 2 0N INC'UDINC SAFETY AND EMERGENCY SYSTEMS PAGE 12 90ESTION 2.17 (1.00)

Arrange the #cllowing in the correct sequence for rod v thdrawal (one ster). (Use itert rom ther once i' requ2 red.)

c. Moveable gripper coil 0FF.

b. Sittiontry gripper coil DN.

c. Lift coil 0FF.

d. Moveabit gripper coil DN.

e. Stationary grippar coil 0FF.

f. Lift co:1 ON.

QUESTION 2.18 (2.00)

a. Which component (s) of the Control Rod Drive Mechanisn act as the pres-sure bounde*y between the RCS and the Contcinment etmosphere? (0.5)

b. When perrlleling the output of the tut Rod Drive FO c- e t s automatically, the ' speed netcher* evton.etictlly changes the tpeed on which of the two MG sets ~ (:nconing/ running) (0.5)

c. Expirin why the statiencty coils for the CRDh are supplied with TWO DC voltages. (1.0)

QUESTION 2.19 (2.00)

Describe the inputs to the voltage regulator null meter and how a bumpless transfer can be nede by its use.

QUESTION 2.20 (1.00)

What is compared in the power range detector current ccmperator?

OUESTION 2.21 (1.00)

Following a RCS-boron dilution at power, how is the water in the precsuvirer trought to the sane concentration?

(***** CATEGORY 02 CONTINUED ON NEXT DAGE ***.**)

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".. PL A t:1 DESIGN INCLUDING SAFETY AND EMEcr,ENCY SYSlEMS 5'GE 14 GUESTION 2.22 (2.00)

Concerni>3 HJTC, state two purposes of the heater power controlle" logic.

(***** END OF CATEGORY 02 r*xxx)

2. INSTRUMEk11 AND CONTROLS ' AGE :1 DUESTION 3.01 (1.00)

Which o' the following is NOT a function of the P-4 permissive M. rip and byptss brecie s open)'

s. Allows bypass:ng of steam dur.P cooldown interlock,

b. Allows operstor bloci of SI signol efter e time delry.

c. Causes feedwater isolation if low Tavs is also present.

d. Cevses e turbine tiip.

90ESTION 3.02 (1.00)

Which statement below regarding the pressuriner pressure control and protective system is NOT correct?

a. The naster pressure controller provides ^*1e control signal for only cne c# the PORVs.

5. There is a le a d/1 s g coc:pe.ns s t i on circuit for pressure inputs to the low presserc resetor trip that varies the trip .etpoint with the *cte of

' pressors decr ease.

c. The tuo pressuriner spray valves are controlled by separate transnitters

d. To block S! actuation on a normal plent depressurination. the-operator must operate TWO block switches to prevent inadvertant ECCS actuation.

(u'** CATEGORY 03 C0fTINUE? U- NEVT PAGE rrnri

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3. INP ;'t'ME NT S AND CON 7 POLS P A C E. 16 OUESTION 3.03 (1.00)

Which of the follouir.s statements about temperature detectors is true?

a. The thermoceuple :s connected to ont les of s bridge c i T c i. 21 and it the temperature changes the output voltage across the bridse chanses,

b. When a thermocouple fails open it will respond in the same manner as an RTD and will indicate a full scale readins on the meter.

c. When a thermocouple becomes shorted, a new thermocouple will e::is t at the point of the short and the meter will respond to the ambient temperature at the point of the short.

d. An RTD is comprised of two wires of dissimilar metals in contact with each other and senerates an EMF proportional to the temperature difference between the open ends of the wires.

00ESTION 3 0c (1.00)

What type of radiation is the Gross Isiled ruel Detecter- monitoring "or

r.dication of fue f t: 2 '.' r e ?

a. Gammas frcm iodine and cesium decay.

b. Alphes from uranium and plutonium decay.

c. Neutrons from bromine and iodine decay.

d. Betas fran tritium decay.

OUESTION 2.05 (2.00)

The condenser steam dumps will not open unless certain int er 1c e'.s are netr crmins sisnelt are present, cnd there is a demand sisr.cl.

1. Ahat 3 intericc!<.s (permissives a d blocks) rac : ~ be met? EO.6:

b. How is the load Tejection erning s:snel reset? EO.42

c. What determines the magnitude of the demand signal wher ir the Tavs - turbine trip submode? E0.4:

d. What are the three signals that can arm the steam dumps? E0.62 (xEr** CATEGORY 03 C0r'TINUED ON NEXT cACE ny r > >

1. INSTRUMENTS AND CONTROLS PAGE 17 OUESTION 3.06 (1.00)

Which one of the #ollowing malfunctions cculd cause one of the c.'er temperrtete delte ' to:p bitteb:et to tr:pi

s. Controlling turbine inpulse pressure channel feiling low,

b. Powe- renge N43 lower detector failins low.

c. Reactor coolant # low detector failins low.

d. Contiollins pressuricer level channel frilins low.

GUESTION 3.07 (1.00)

With the reactor at 100% power and the steam dump control syster: in the Tavs mode, e 15% step lost of load occurs. Assuming no reactor trip occurs the condenser is availabler and the reactor operator manually OPERATES the control rods, which of thE folloWing WDuld occu" if Cank 1 steE! dunp valves failed to open?

e. Bent 2 w o u : c' open,

b. AticspH-ric dumps would open.

c. S/G safeties would open.

d. No other steam valves would open -

GUESTION 3.0E (3.00)

Indicate what happens to the Rod Control Systen (rods in, rods out, no chense) cnd BRIEFLY explain.why the chenge will or will not occur for the following conditions. Rods are in auto unless otherwise spec: # Ed.

2. Reactor power is 17% when the controlling turbine fi r' siege mpulse pressure transnitter fails 5:gh.

b. Recc;or oo h.perll L/

is 100*' ent 200: 1 Thot fe:Is ' v'a 5 .

c. W ,no - d :co nue.ef m_wr ic 50". Red Control is ir Enual.

Instrument testing is in progress en the turbine power input to rod control which has turbine power at 100%.

A12 indications have been stable for the last hour.

The Bank. Selector swi.tch is then placed :n AUTO.

(rarr* CATEGORY 03 CONTINUE 2 DN NEXT PAGE D' '4 m -

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3. INSTRUMEN1S AND CONTROLS PAGE 18 DUESTION 3.09 (1.00)

What input signal is used by the Rod Insertion Linit ccleulator as e dirett

.d_:ttien of reecto- pourr0 GUESTION 3.10 (3.00)

For each of the follouins, give the set point and coincidence:

. P-12

2. P-14

3. High Main Steam F l o w w i t h L o w L o w T a v s +1-- M5;)

QUESTION 3.11 (2.00)

The 'ollowing failures occur causing a subsequent autonatic reactor trip.

Whrt protection signel would cause the trip? Assume the reactor is initially at 1001 power and steady state conditions, all systens in auto-

n. s t i c cnd no operator action. Treet each independently.

a. CVCS flow rate drops to a n. i n i n. u n of 30 spm.

b. A narrow range (controlling) cold leg RTO frilt high.

QUESTION 3.12 (3.00)

List the following;

a. 'our protective functions provided by.the pressuriner pressure DRO-TECTION chenrels (PT- 455, -456, -457) other then MCB indications and a l a r n. s .

b. three control functions provided by pressuriner pressure CONTROL channel PT-444, othe- than MCB indications and alarne

. two protection / control functions provided by the pressuricer leve control and protection chtnnels (LT-459, -460r -461) other nen charging # low control and "CP indicstions and alarns.

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

3. INSTRUMENTS AND CONTROLS " AGE 19 GUESTION 3.13 (3.00)

a. What 2 time delay actions occur in the concensate/nain feeduster sys-ten wSc- r. r l m f e e t' F or - suetion prestute i t c y t- belov "'O pt:g? (;.0)

b. What conditions cause automatic closure of the stean generator Main Feeduster Stop (Isoletion) Valves (3232A,B,C)? (0.S)

c. If the stean generator Feed Regulating valves are closed'by c protection signc1 (SSFS)r the signal nust be cleared to reopen the valves. What are the THREE protection signals that close the valves AND HOW is each cleared? (1.5) 00ESTION 3.14 (2.00)

For each condition EXPLAIN which component (s) would be senerating a red movenent signal and the response of Bank D rods to this signal.

Assume no other Red Stop signals present, Reactor at power.

BANK SELECTOR -SW. IN-007-40LD LEVER DLANT PARAMETER 'O' POSITION t, "anuel In ":! L LO-LO ALAR- :PO steps

b. 'D' 'cid lav3 Tref +4 des. 186 steps

c. Mznual Out '..Ursent Failure' 200 steps

d. Auto Hold Tavg-Tref -4 deg. 222 steps DUESTION 3.15 (1.00)

A pressuricer level transnitter nas a Differential Pressure Cell which has ruptured. EXPLAIN how the level INDICATION in the Control Roon would respond.

PUESTION 3.16 (3.00)

Describe the controls for the : :' = g i n s p i_in i s . Ir li_'de centra: _ocation interlocke, switch positions. End euto sterit.

90ES' ION 3.17 (1.00)

What feature of the main feed 99 bypass valves controller 2110ws the controller to anticipate r e p i c' level fluctuetions?

(***** END OF CATEGD;:Y 03 *****)

3 PROCEDURES - NORMAL, ABNDPPAL, EMERGENCY AND PAGE 20

~

~~'~~555ELUU5bkL UUUTPOE~~~~~~~~~~~~~~~~~~~~~~~~

9UESTION 4.01 (2.00)

The fe;;ov tig refer tt : r:' c

ut ion found.:r: Fyr-1-DDR-22.0,

' Au::iliary Feeduster System.'

A. Whtt is the h.c M i nu n feedweter ' low allowed'

1. to a steam generator whose level is 20 % Narrow Range and increasins during a level recovery transient?

2. from one totor driven AFW pump?

3. from the turbine driven AFM pump? (1.5)

9. Why should service water supply to ACW only be used in an emergency?

(0.5)

GUESTION 4.02 (1.00)

Which of the following is a 10 CFR 20 exposure. limit?

E. E rem / year-whole body.

4

b. : r e- / quarter-whole oody.

c. 10.75 rem / quarter-hands.

d. 7 ren/ quarter-skin of whole body.

9,UESTION 4.03 (1.00)

Which of the following radiation exposures would inflict the greatest

_biolo3ical damage to man?

s. 1 Rem of GAMMA.

b. 1 Rem of A' PM A . -

c. 1 Rem of NEUTRON 4

d. NONE c# the ebove; they tr e ell equivalent.

QUESTION 4.04 ( .50)

The Reactor Coolant ourp seal bypass valve should be opened during a lott cf seal injection water. TRUE or FALSE?

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

a. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND t' AGE 21

--- EE5i6t55f C At E 6sT R O E--~ ~--~ ~ ~ ~---~ ~-----~~ ~ ~

QUESTION 4.05 (3.00)

Match the trerds from Cciv-m E: that wou'd be indicat2vn e' -v citiorc-for C o l u m r: A nelfunctions prior te any protactive funct:en returtions.

There may be more than one Colutn 5 item for each Column A itet. Place answers on answer sheet (e.g., e-7,8,9).

COLUMN A COLUMN B

a. Smell Breck LOCA Inside 1. Decreasing Pressur::et Level Containment

2. Decreasing Steam Pressure

b. Large Steam Leak Insida Containment 3. Increasing Containment Pressure

4. Decreasing Tave

5. Increasing Conte:nrent Radiation

6. Decreasing t'rersur*iner Presso?e 7 Nevr Normal Steve essure (3.04 OUESTION 4.06 (3.00)

a. Prior to increasing Tsve from Mode 5, your heatup procedure (UDP-1,1) gives you the option NOT to withdraw shutdown benks.

What condition must e::i s t prior to taking this option? (0.5)

b. If the hectup begen with a solid Recetor Coolant Syster (RCS) condition, at c,oproximately what DRESSURE will s steam bubble be formed in the pressuriner? (0.5)

c. What sre the maximur, sllowable ptessurizer HEATU: AMD COOLDOWN reteco (1.0)

d. After the steam bubble .s 'oried in the pressuriner in "cde 5, and prier to further RCE 5ectup, wil: the hot ec2ibr.ete1 pressuriner level chsnnels indicate uIgqER OR LOWER thsr s c +, v e l levelo EXPLAIN. (1.0i

(***** CATEGORY 04 CONTINUED ON NEXT D A EG *n**)

C. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 22

~~~~kEDist6EiEEL 55GTE6E----~~~~~~---~~~--------

QUESTION 4.07 (1.00)

List the preferred order of re-starting Unit I reacter c 3 c l a ret pumps when recovering from a nrturri circulstion coeldoun<

QUESTION 4.08 ( .50)

What is the lowest whole body administrative e::posure limit (no ex-tensions)?

QUESTION 4.09 (1.00)

What actions should you take upon discovering a spill involving a a;inor radietion hacerd to personnel?

QUESTION 4.10 (1.00)

List the immediate operator actions for a loss of 1B *enctor coolant p u r.p when et 30% power.

GUESTION 4.11 (1.00)

List the immediate operator actions for a continuous control bank D withdrawal.

QUESTION 4.12 (1.00)

List the immediate operator actions for a Loss of All AC Power (ECP-0.0).

(***** CATEGORY 04 CONTINUED CN NEXT PAGE xx***)

4. PROCEDURES - NOPMAL, ABNORMAL, EMERGENCY AND PAGE 23

~~~~EE5i5E55iCEE C5sTR5E---------------~~~~~~-~~

GUESTION 4.13 (3.00)

a. After a Res: dual uest Recovel c;"R) p u r.p it st;t tec t ot 7 -

cooldown, but before plceing the train in tervice. the tenp is operated on miniflow rec:rculation for a minimum of 10 n n. WHY? (0.5)

b. How is e low boron concentration (in an RHR trein to be placed in service) corrected? (1.0)

c. Would sterting an RHR pump, with the CVCS letdown pressure control valve (PCV-145) in manual result in a pressure INCREASE, DECREASE, or NO CHANGE in the Recetor Coolant Syste:n during solid plant i, operation? (0.5)

d. When establishing e bubble in the pressurizer, WHY must both RHR trains be valved into their respective RCS hot less? (1.0)

DUESTION 4.14 (1.00)

While per f orming EEP-3 Sten Ge wrator Tube Rupture procedure. hou is 2 ewer ins bce the Pressu*2:e' revel AND ruptured Stean Ocncirtm level accocplished?

QUESTION 4.15 (1.00)

What is the maximum allowable total Reactor Coolant Pump seal injection flow?

DUESTION 4.16 (1.00)

What reference is used to verify the settings on PCB Manus 1/ Auto station Potentioneters are correct when s tar ti n,: up the unit fror ceid shutdown?

00ESTICN 4.17 '1.00)

When would a t r e n s i t i o r, be nede #ror ADP-1.0, Excessive RCE _tckage, to the fE: ter++s due to t loss of reactor coolanto A s t o n. e no recetor trip or EI occurs. \

4 u e P -3.1

(***** CA'EGORY 04 CONTINUED ON NEXT DAGE wr*r*)

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4 PROCEDURES NORMAL, ADNORMAL, EMERGENCY AND PACE 24

~~~~EED56[65EC5[~66UTEUL~~~~~~~~~~~~~~~~~~~~~~~~

GUESTION 4.18 (1.00)

If you are the Unit Operator s i.d you have a vis tor vou s; e e ::- c c.r t i n g when e Generel Evecurtion is declered, whrt should you % ent whet will you have the visitor co?

DUESTION 4.10 (1.00)

How is pressuricer level naintained immediately following a icss of instrument air?

GUESTION 4.20 (1.00)

Under what conditions are the adverse containn.ent values for instrumenta-tion used in the EEPr?

OUESTION 4.21 (1.50)

List tb ee dif f er er;* 211ovab2m tires an CRG try Sc e:a i t e d 1 #

' uss entered by direction of an ORA'OE - *-

GUESTION 4.22 ( .50)

How far from the ECP are you allowed to withdraw rods before you must terminate e reactor stcatop when en inverse count rete plo*. it not beine nade?

DUESTION 4.23 (2.00)

What are the four 1r.r.ediate operator actions cf AOP-27.0, Emergency Boretion?

(***** END OF CATEGORY Of *****)

(vx rrvi ru rtr> END OF EXAMINATION rrr):rrrrw x x i rr * )

. 1 o ma vc s/t CycBe e79iciency o (Net tork cut)/(Energy in)

, . mg sa V,t + 1/2 at 2

[s mC"

<E = 1/2 mv a = (Vf - 13 )/t A = AN A=Ae3 PE = mgn vf = V, + a t *= e/t x = zn2/t1/2 = 0.693/ti/2 n0 2 y , , ,p 1/2 A= 4 [(t1/2)

5td))

.E = 931 sn -

. m = V,VAo - T.x Q.=ph I=Iec Q = mCpat h = UA A T I=Iec Pwr = Wfah I = I,10-*/ M TVL = 1.3/u -

sur(t) HVL = -0.693/u P = P*10t P = Po e /

SUR = 26.06/T SCR = S/(1 - K,ff)

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

SUR = 26c/t= + (a - o)T CR j (1 - K,ff)) = CR2 II ~ keff2)

T = (t=/s) + [(a - oV Io] M = 1/(1 - K,ff) = CR j/CR, T = t/(p - a) M = (1 - K,ff,)/(1 - K,ff))

T = (a - o)/(Io) SDM = ( - K ,ff)/K,ff a = (Keff-III eff * # effIK eff t' = 0 seconos I = 0.1 seconds-I o = [(1*/(T K,ff)] + [I,ff /(1 + IT)]

Idjj=Id P = (rov)/(3 x 1010) I jd) 2 ,2gd2 22 2 I = eN R/hr = (0.5 CE)/d (meters)

R/hr = 6 CE/d2 (f,,g) _

water Parameters Miscellaaeous Conversions 1 gal. = 8.345 lbm. 1 curie = 3.7 x 1010 aps 1 ga:. = 3.78 liters 1 kg = 2.21 lom 1 fte = 7.48 gal 1 np = 2.54 x 103 Stu/nr Density = 62.4 1 /ft3 1 mw = 3.41 x 100 5tu/hr Density = 1 gm/c. lin = 2.54 cm Heat of vaporization = 970 Stu/lem *F = 9/5'C + 32 Heat of fusion = 144 Btu /lem 'C = 5/9 ('F-32) 1 Atm = 14.7 psi = 29.9 in. Hg. 1 BTU = 778 ft-lbf 1 ft. H 2O = 0.4335 lbf/in.

e = 2. 71-

= " - N N. -

o

Volume, ft'/lb Enthalpy, Blu/lb Entropy, Str/lb a F P

P ss. Evap Steam Water Evap Steam Water Evep Steam Water V hg hq hy sg s eg s, VI Vse e

-0.02 1075.5 1075.5 0.0000 2.1873 2.1873 32 32 0.08859 0.01602 3305 3305 3.00 1073.8 1076.8 0.0061 2.1706 2.1767 35 35 0.09991 0.01602 2948 2948 8 03 1071.0 1079.0 0.0162 2.1432 2.1594 40 40 0.12163 0 01602 2446 2446 13.04 1068.1 1081.2 0 0262 2.1164 2.1426 45 45 0.14744 0.01602 2037.7 2037.8 18.05 1065.3 1083.4 0.0361 2.0901 2.1262 50 50 0.17795 0.01602 1704.8 1704.8 1207.6 28.06 1059.7 1087.7 0.0555 2.0391 2.0946 60 60 0 2561 0.01603 1207.6 868.4 38.05 1054.0 1092.1 0.0745 1.9900 2.0645 70 70 0.3629 0.01605 868.3 633.3 48.04 1048.4 1096.4 0.0932 1.9426 2.0359 80 80 0.5068 0.01607 633.3 0.01610 468.1 4t>8.1 58.02 1042.7 1100.8 0.1115 1.8970 '2.0086 90 90 0.6981 1105.1 0.1295 1.8530 1.9825 100 0.9492 0.01613 350.4 350.4 68 00 1037.1 100 0.1472 1.8105 1.9577 110 0.01617 265.4 265.4 77.98 1031.4 1109.3 110 1.2750 203.26 87.97 1025.6 1113.6 0.1646 1.7693 1.9339 120 120 1.6927 0.01620 203.25 15733 97.96 1019.8 1117.8 0.1817 1.7295 1.9112 130 130 2.2230 0.01625 157.32 123.00 107.95 1014.0 1122.0 0.1985 1.6910 1.8895 140 140 2.8892 0.01629 122.98 97.07 117.95 1008.2 1126.1 0.2150 1.6536 1.8t>86 150 t 150 3.718 0.01634 97.05 160 77.29 127.96 1002.2 1130.2 0.2313 1.6174 1.8487 160 4.741 0.01640 77.27 62.04 62.06 137.97 996.2 1134.2 0.2473 1.5822 1.8295 170' 170 5.993 0.01645 1.8111 ISO 50.21 50.22 148.00 990.2 1138.2 0.2631 1.5480 180 7.511 0.01651 190 40.94 40.96 158.04 984.1 1142.1 0.2787 1.514S 1.7934 190 9.340 0.01657 200 33.64 168.09 977.9 1146.0 0.2940 1.4824 1.7764 200 11.526 0.01664 33.62 27.80 27.82 178.15 971.6 1149.7 0.3091 1.4509 1.7600 210 210 14.123 0.01671 26.80 180.17 9703 1150.5 0.3121 1.4447 1.7568 212 212 14.696 0.01672 26.78 23.15 188.23 965.2 1153.4 0.3241 1.4201 1.7442 220 i 220 17.186 0.01678 23.13 19364 19381 19833 958.7 1157.1 03388 13902 1.7290 230 230 20.779 0.01685 16.304 16.321 208.45 052.1 1160.6 0.3533 1.3609 1.7142 240 240 24.968 0.01693 13.802 13.819 218.59 945.4 1164.0 0.3677 1.3323 1.7000 250 250 29.825 0.01701 11.745 11.762 228.76 938.6 1167.4 0.3819 1.3043 1.6862 260 260 35.427 0.01709 10.042 10.060 238.95 931.7 1170.6 03960 1.2769 1.6729 270 270 41.856 0.01718 8.627 8.644 249.17 924.6 1173.8 0.4098 1.2501 1.6599 280 280 49.200 0.01726 7.443 7.460 259.4 917.4 1176.8 0.4236 1.2238 1.6473 290 290 57.550 0.01736 6.448 6.466 269.7 910.0 1179.7 0.4372 1.1979 1.6351 300 300 67.005 0.01745 5.609 5.626 280.0 902.5 1182.5 0.4506 1.1726 1.6232 310 310 77.67 0.01755 4.896 4.914 290.4 894.8 1185.2 0.4640 1.1477 1.6116 320 320 89.64 0.01766 3.770 3.788 311 3 878.8 1190.1 0.4902 1.0990 1.5892 340 340 117.99 0.01787 2.939 2.957 332.3 862.1 1194.4 0.5161 1.0517 1.5678 360 360 153.01 0.01811 2.317 2335 353.6 844.5 1198.0 0.5416 1.0057 1.5473 380 380 195.73 0.01836 1.8444 1.8630 375.1 825.9 1201.0 0.5667 0.9607 1.5274 400 400 247.26 0.01864 1.4808 1.4997 396.9 806.2 1203.1 0.5915 0.9165 1.5080 420 420 30S.78 0.01694 1.1976 1.2169 419.0 785.4 1204.4 0.6161 0.8729 1.4890 440 440 381.54 0.01926 0.9746 0.9942 441.5 753.2 1204.8 0.6405 0.8299 1.4704 460 460 466.9 0.0196 0.7972 0.8172 464.5 739.6 1204.1 0.6648 0.7871 1.4516 480 450 566.2 0.0200 0.6749 487.9 714 3 1202.2 0.6890 0.7443 1.4333 500 500 680.9 0.0204 0.6545 0.5596 512.0 687.0 1199.0 0.7133 0.7013 1 4146 520 520 812.5 0.0209 0.5386 0.4437 0.4651 536.8 657.5 1194.3 0.7378 0.6577 13954 540 540 962.8 0.0215 560 03651 0.3871 562.4 6253 1187.7 0.7625 0.6132 1.3757 SED 1133.4 0.0221 0.2994 03222 589.1 fr39.9 1179.0 0.7876 0.5673 13550 580 550 1326.2 0.0228 0.2675 617.1 550 6 1167.7 0.8134 0.5196 1.3330 500 600 1543.2 0 0236 0.2438 0.2208 646.9 506.3 1153.2 0.8403 0.46S9 1.3092 620 620 1786.9 0.0247 0.1962 0.1802 679.1 454.6 1133.7 0.8666 0.4134 1.2821 640 640 2059 9 0.0260 0.1543 0.1443 714.9 392.1 1107.0 0.8995 0.3502 1.2458 660 660 2365.7 0.0277 0.1166 0.1112 758.5 310.1 1068.5 0.9365 0.2720 1.2086 680

&&O 2708.6 0 0304 0.0608 0.0752 822.4' 172.7 995.2 0.9901 0.1490 1.1390 700 700 30943 0.0366 0.0386 0.0508 0 0.0508 906.0 0 S06.0 1.0612 0 1.0612 705.5 705.5 3208.2 TABLE A.2 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (TEMPERATURE)

A.3

Volume. It'/ib Enthatpy. Stu/ib Entripy. Sto/ib a F Energy. Stu/lb

' Pre s. T p Cat;r Erep Steam Water Evep Steam Water Evep St:em Ca> St:em vg by hg h, sg 8,, s, og u, vg v, 0.01602 3302.4 3302.4 0 00 1075.5 1075.5 0 2.1872 2.1872 0 1021.3 0.0845 0.0446 32.018 0.10 35.023 0.01602 2945.5 2945.5 3 03 1073 8 1076.8 0 0061 2.1705 2.1766 3A3 10223 0.10 0.15 45.453 0 01602 2004.7 2004.7 13.50 1067.9 1081 4 0 0271 2.1140 2.1411 13.50 1025.7 0.15 0.20 53.160 0 01603 1526.3 1526 3 21.22 1063 5 1084.7 0.0422 2.07?S 2.1160 2122 1028 3 0.20 0.30 64 484 0 01604 1039.7 1039.7 32.54 1057.1 1089.7 0.0641 2.016S 2.0809 32.54 1032 0 0.30 0.40 72.869 0.01606 792.0 792.1 40.92 1052.4 10933 0.0799 1.9762 2.0562 40.92 1034.7 0.40 0.5 79.586 0.01607 641.5 641.5 47.62 1048.6 1096 3 0.0925 1.9446 2.0370 47.62 1036.9 0.5 0.6 85.218 0.01609 540.0 540.1 53.25 1045.5 1093.7 0.1028 1.9186 2.0215 53.24 1038.7 0.6 0.7 90 09 0.01610 466.93 466.94 58 10 1042 7 11008 0.3 1.8966 2.0083 58.10 1040.3 0.7 0.8 94.38 0.01611 411.67 411.69 62.39 1040.3 1102.6 0.1117 1.8775 1.9970 6239 1041.7 0.8 0.9 98.24 0.01612 368.41 368.43 66.24 1038.1 1104 3 0.1264 1.8606 1.9870 66.24 1042.9 0.9 1.0 101.74 0.01614 333.59 333 60 69.73 1036.1 1105.8 0.1326 1.8455 1.9781 69.73 1044.1 1.0 2.0 126.07 0.01623 173.74 173.76 94.03 1022.1 1116.2 0.1750 1.7450 1.9200 94.03 1051A 2.0 3.0 141 47 0.01630 118.71 118.73 109.42 1013.2 1122.6 0.2009 1.E854 1.8864 10941 1056.7 2.0 4.0 152.96 90 63 90.64 120.92 1006.4 1127.3 0.2199 1.6428 1.8626 120.90 1060.2 4.0 0.01636 5.0 162.24 0.01641 73.515 73.53 130 20 1000.9 1131.1 0.2349 1.6094 1A443 130.18 1063.1 5.0 6.0 0.01645 61.967 61.98 138.03 996.2 1134.2 0.2474 1.5820 12294 138.01 1065.4' 6.0 170.05 7.0 176.84 0.01649 53.634 53.65 144.83 992.1 1136 9 0.2581 1.5587 1A168 14431 1067.4 7.0 ~

8.0 182.86 0.01653 47.328 47.35 150.87 988.5 1139.3 0 2676 1.5384 1A060 15034 1069.2 S.0 9.0 188.27 0.01656 42385 42.40 156.30 985.1 1141.4 0.2760 1.5204 1.7964 156.28 1070.8 9.G 10 193.21 0.01659 38.404 38 42 161.26 982.1 11433 0.2836 1.5043 1.7879 161.23 1072.3 10 0.01672 26.782 26.80 180.17 9703 1150.5 0.3121 1.4447 1.7568 180.12 1077.6 14.696 14.696 212.00 15 213.03 0.01673 26.274 26.29 181.21 969.7 1150.9 0.3137 1.4415 1.7552 181.16 1077.9 15 20 227.96 0.01683 20.070 20 087 196.27 960.1 1156.3 03358 1.3962 1.7320 196.21 1082.0 to 30 250.34 0.01701 13.7266 13.744 218.9 945.2 1164.1 0.36S2 1.3313 1.6995 2183 1087.9 30 !

40 267.25 0.01715 10.4794 10.497 236.1 933.6 1169.8 0.3921 1.2844 1.6765 236 0 1092.1 40 ,

50 281.02 0.01727 8.4967 8.514 250.2 923.9 1174.1 0.4112 1.2474 J.6585 250.1 10953 50 60 292.71 0.01738 7.1562 7.174 262.2 915.4 1177.6 0.4273 1.2167 1.6440 262.0 1098.0 60 l 70 302.93 0.01748 6.1875 6 205 271.7 907.8 1180.6 0 4411 1.1905 1.6316 272.5 1100.2 70 80 312.04 0.01757 5 4536 5 471 232.1 900.9 1183.1 0.4534 1.1675 1.6208 281.9 1102.1 30 90 320.28 0.01766 4.8777 4 895 290.7 894.6 1185.3 0.4643 1.1470 1.6113 290.4 1103.7 90 100 327.82 0.01774 4.4133 4.431 298.5 888.6 1187.2 0.4743 1.1284 1.6027 298.2 1105.2 100 120 341.27 0.01789 3 7097 3.728 312.6 877A 1193 4 0.4919 1.0960 1.5879 312.2 1107.6 120 140 353 04 0 01803 3.2010 3 219 325.0 868.0 1193 0 0.5071 1.0681 1.5752 324 5 1109.6 140 160 363 55 0.01815 2.6155 2.834 336.1 859.0 1195.1 0.5205 1.0435 1.5641 335.5 1111.2 160 180 373 08 0.01827 2.5129 2.531 346.2 350.7 1196.9 0.5328 1.0715 1.5543 345.6 1112.5 180 200 35180 0 01839 2.2689 2.287 355.5 842.8 11983 0.5438 1.0016 1.5454 3543 1113.7 200 250 400 97 0.01865 1.8245 1.8432 376.1 825 0 1201.1 0.5679 0 9585 1.5264 3753 1115.8 250 l 300 417 35 0 01859 1.5233 1.5427 394.0 808.9 1202.9 0.5882 0.9223 1.5105 392.9 1117.2 300 ,

350 411.73 0 01913 1.3064 1.3255 409.8 794 2 12040 0 6055 08909 1.4968 409 6 111B ! 350 '

400 444 60 0.0193 1.14162 1.1610 424.2 780 4 1204 6 0 6217 0 8630 1.4847 422.7 111E 7 400 450 45528 0 0195 1.01224 1.0318 437.3 767.5 1204.8 06360 0 8378 1.4738 435.7 1118.9 450 500 467 01 0 0198 0 90787 C.9276 449.5 755.1 1204.7 0.6490 0.8143 1.4639 447.7 1118 8 500 ;

550 476 94 0 0199 0 82183 0.8418 460.9 743.3 1204 3 0.6611 0.7936 1.4547 456.9 !!!8 6 550 i 400 48620 0 0201 0.74962 0.7699 471.7 732.0 1203 7 0.6723 0 7738 1.4461 469.5 t i lE.2 600 700 .503 08 0.0205 0.63505 0.6556 491.6 710.2 1201.8 0.6928 0.7377 1.4304 488.9 1116.9 700 803 51a 21 0 0209 0.54809 0.5690 509.8 689 6 11994 0.7111 0.7051 1.4163 506 7 I115.2 300 900 ' E 3195 0 0217 0 4796S 0 5009 526 7 659 7 1196 4 0 7279 0 6753 1.4032 5232 1113.0 900 1000 544.5i 0.0216 042436 0 4460 542.6 650 4 1192 9 0.7434 0.6476 13910 5306 1110.4 1000 1100 SEE 2d 0.0220 0 378f.3 0 4005 557.5 631.5 1189 1 0 7573 0 6216 1.3794 5531 1107.5 1100 1200 i:67.19 0 0223 0 34013 0.3625 571.9 613.0 1164 8 0 7714 0.5969 1.3683 5569 1104 3 1200 1300 177 42 0 0227 0 30722 03299 585 6 544.6 1180 2 0.7843 05733 1.3577 530.1 1100 9 1300 1 ECD 53707 0 0231 0 278/1 0 3018 598 8 576 5 1175 3 0 7966 05507 13474 592.9 1097.1 1400 1%0 5 % 20 0 0235 0 2b372 0 27/2 611.7 550 4 117C 1 0.8035 05253 1.3373 605 2 1093.1 1500 2000 635 80 0.0257 0 16? % 0.1883 672.1 465 2 1133.3 0 Bf.?i 04256 1.?B81 662 6 10GS6 2000 2500 66111 0 02c,0 0 10209 01307 731.7 361.6 1093 3 09139 03206 1.2345 118.5 1032.9 2500 3000 695 33 0 0343 0 050/3 0.0850 801 8 218.4 1070 3 0 9723 01891 1.1619 782.8 973.1 3000 3298.2 701 47 0 0503 0 0 050d 906 0 0 9060 1.0512 0 1.0612 875.9 875.9 3708.2 1

TABLE A.3 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (PRESSURE) l l

A.4 l

t e

! Abs pesos.

Tempeeewe, F ]I D/egla.

200 300 400 900 000 700 900 900 1000 1100 1200 1300 1400 3500 (set lemp) 100 v 0.0161 392.5 452.3 511.9 571.5 631.1 690.7 1 6 68 00 1150 2 1195.7 1241.8 1288 6 13361 1384 5 (101.74) s 0.1295 2.0509 2.1152 2.1722 2.2237 2.2708 2J144 e 0.0161 78 14 90.24 102.24 114.21 126 15 138 08 150 01 161.94 173 86 185 78 197.70 209 62 221.53 233 45 6 6 68 01 1148.6 1194 8 1241.3 1288 2 1335.9 1384 3 1433 6 1483 7 1534.7 1586 7 1639 6 1693 3 1748 0 1803.5 (162.24) s 0.1795 1.8716 1.9369 1.9943 2.0460 2.0932 2.1369 2.1776 2 2159 2 2521 2.2866 2.3194 2.3509 2.3811 2.4101 e 0 0161 38 84 44 93 51.03 57.04 63.03 69 00 74 98 80.94 86 91 92 87 98 84 104 80 110.76 116.72 to 6 68 02 1146 6 1193 7 1240 6 1287A 1335.5 13840 1433 4 14835 1534 6 1586 6 16395 1693.3 1747.9 1803 4 (192.21) s 0.1295 1.7928 1.8593 1.9173 1.9692 2.0166 2.0603 2.1011 2.1394 2.1757 2.2101 2.2430 2.2744 2.3046 2.3337 v 00161 0.0166 29.899 33.963 37.985 41.986 45.978 49 964 b3 946 57.926 61.905 65282 69A58 73.833 77.807 15 6 68.04 168 09 1192.5 1239.9 1287.3 ~ 1335 2 1383 8 1433.2 1483 4 1534.5 1586 5 1639 4 1693.2 17473 1803 4 (213.03) s 0.1295 0.2940 1.8134 1.8720 1.9242 1.9717 2.0155 2.0563 2.0946 2.1309 2.1653 2.1982 2.2297 2.2599 2.2890 v 0.0161 0.0166 22.356 25.428 28.457 31.466 34.465 37.458 40 447 43.435 46 420 49.405 52.388 55.370 58.352 20 4 68.05 168 11 1191.4 1239.2 1286.9 1334.9 1383 5 1432.9 1483.2 1534.3 1586.3 1639.3 1693.1 17472 1803.3 (227.96) s 0.1295 0.2940 1.7805 1.8397 13921 1.9397 1.9836 2.0244 2.0628 2.0991 2.1336 2.1665 2.1979 2.2282 2.2572 v 0.0161 0 0166 11.036 12.624 14.165 15 685 17.195 18.699 20 199 21.697 23.194 24.689 26.183 27.676 29.168 40 h 68.10 168.15 1186 6 1236.4 1285.0 1333.6 1382.5 1432.1 1482.5 1533.7 1585.8 1638 8 1992.7 1747.5 1803.0 (26725) s 0.1295 0.2940 1.6992 1.7608 1A143 1A624 1.9065 1.9476 1.9860 2.0224 2.0569 2.0899 2.1224 2.1516 1.1807 ,

e 0.0161 0.0156 7.257 8 354 9.400 10.425 11.438 12.446 13.450 14.452 15.452 16.450 17.448 18.445 19.441 60 6 68.15 168 20 1181 6 1233.5 1283.2 1332.3 1381.5 1431 3 1481.8 1533t2 15853 1638.4 1692.4 1747.1 1802.8 (292.71) s 0.1295 02939 1.6492 1.7134 1.7681 1A168 1.8612 1.9024 1.9410 1.9774 2.0120 2.0450 2.0765 2.1068 2.1359 e 0.01'61 0.0166 0.0175 6.218 7.018 7.794 8.560 9.319 10.075 10.829 11.581 12.331 13.081 13.829 14.577 80 6 68.21 168.24 269.74 1230.5 1281.3 1330.9 1380.5 1430.5 1481.1 1532.6 1584.9 1638.0 1692.0 1746A 1802.5 (312.04) s 0.1295 0 2939 0.4371 1.6790 1.7349 1.7842 13289 1.8702 1.9089 1.9454 1.9800 2.0131 2.0446 2.0750 2.1041 e 0.0161 0.0166 0.0175 4 935 5.588 6.216 6.833 7.443 8050 8.655 9.258 9A60 10.460 11.060 11.659 100 h 68.26 168.29 269 77 1227.4 1279.3 1329.6 1379.5 1429.7 1480.4 1532.0 1584 4 1637.6 1691.6 1746.5 1802.2 (327.82) s 0.1295 0.2939 0.4371 1.6516 1.7088 1.7586 1.8036 1.8451 1A839 1.9205 1.9552 1.9883 2A199 2.0502 2.0794

, 0 0161 0 0166 0 0175 4 0786 4.6341 5.1637 5.6831 6.1929 6.7006 7.2060 7.7096 8.2119 8.7130 9.2134 9.7130 120 6 68.31 168 33 269 81 1224.1 1277.4 1328.1 1378 4 1428.8 1479.8 1531.4 1583.9 1637.1 1991.3 17462 1802.0 (341.27) s 0.1295 0 2939 0 4371 1.6286 1.6872 1.7376 1.7829 1A246 1.8635 1.9001 1.9349 1.9600 1.9996 2.0300 2.0592 v 00161 0 0166 0.0175 3 4651 3.9526 4 4119 43585 5.2995 5.7364 6.1709 6.6036 7.0349 7.4652 7.8946 8.3233 140 6 68.37 168 38 269 85 1220 8 1275.3 1326.8 1377.4 14280 1479.1 1530 8 1583 4 1636.7 1890.9 1745.9 1801.7 (353 04) s 0.1295 0 2939 0 4370 1.6095 1.6686 1.7196 1.7652 1.8071 13461 1.8828 1.9176 1.9508 1.9825 2.0129 2.0421 e 0.0161 0 0166 0 0175 3 0060 3.4413 3.8480 4.2420 4 6295 5.0132 5.3945 5.7741 6.1522 6 5293 ~6.9055 7.2811 160 6 68 42 1EB 42 269.89 1217.4 1273 3 1325 4 1376 4 1427.2 1478.4 1530.3 1582.9 1636.3 1690.5 1745.6 1801.4 (363 55) s 0.1294 0 2938 0 4370 1.5906 1.6522 1.7039 1.7499 1.7919 1.8310 1.8678 1.9027 1.9359 1.9676 1.9980 2.0273 e 0 016! 0 0166 0 0174 2 6474 3 0433 3.4093 3.7621 4.1084 4.4505 4.7907 5.1289 5.4657 53014 6.1363 6.4704 180 6 68 47 168 47 269 92 1213 8 1271.2 1324.0 1375.3 1426.3 1477.7 1529.7 1582.4 1635.9 1640 2 17d53 1801.2 (373.C81 s C 1294 0.2938 04370 1 5743 1.6376 1.6900 1.7362 1.7784 1A176 1.8345 1.8894 1.9227 1.9545 1.9849 2.0142 e 0 0161 0.0166 0 0174 2 3598 2.7247. 3.0583 3.3783 3 6915 4.0008 4.3077 4.6128 4.9165 5.2191 5.5209 5A219 200 h 68.52 168 51 269 96 1210 1 1269.0 1322.6 1374.3 1425.5 1477.0 15291 1581.9 1635.4 1689 8 1745.0 1800.9 (331.60) s 01294 02938 0 4359 1.5593 1.6242 1.677G 1.7239 1.7663 -1.8057 1.8426 1.8776 1.9109 1.9427 1.9732 2.0025 e 0 0161 0 0166 0 0174 0 0186 2.1504 2 4662 2.6872 2.9410 3.1909 3 4382 3 6837 3.9278 4.1709 4.4131 4.6546 250 6 68.66 168 63 270 05 3/5.10 1263.5 1319 0 1371.6 1423 4 1475.3 1527.6 1580.6 1634.4 1688.9 1744.2 1800.2 (400 ??) s 01294 0.2937 0 4368 0.5567 1.5951 1.6502 1.6976 1.7405 1.7601 1.8173 1.8524 1.8d58 1.9177 1.9482 1.9776

, 0 0161 0 0165 0 3174 0 0186 1.7655 2.0044 2 2263 2.4407 2.6509 2 55S5 3 0643 32688 3.4721 3.6746 3.8764 300 h 68 79 1%74 27014 375.15 1257 7 1315 2 1368 9 1421.3 1473 E 1526.2 1579 4 1633 3 1688 0 1743 4 1799.6 (417.35) s 0.1294 02937 0 4337 C5665 1.5703 1.6274 1.6758 1.7192 1.7591 1.7964 1.8317 1.8652 12972 1.9278 1.9572 e 0 0161 0 0166 0 0174 0 0186 1.4913 1.7028 1.8970 2 0332 2.2652 2 4445 2.6219 2.7980 2.9730 3.1471 33205 350 6 68 92 16385 270 24 375 21 1251 5 1311.4 1366 2. 1419 2 14718 1524 7 1578.2 16323 1687.1 1742.6 1798.9

' (431.73) . 0 1293 0 2936 043G7 0.5664 1.5483 1.6077 1.6571 1.7009 1.7411 1.7787 1.8141 1A477 1A79S 1.9105 1.9400 v 0 0161 0 0166 0 0174 0 0162 1 2841 1.4763 1.6499 1.8151 1 9759 2.1339 2.2901' 2.4450 2.5987 2.7515 2.903 400 a 69 05 168 97 270 33 375 27 12451 1307.4 1363 4 1417.0 1470 1 1523 3 1576.9 1631.2 1686 2 1741.9 1793 2 (444.60) s 01293 0 2935 0 4366 0 56G3 1.5282 1.5901 1 6406 1.6850 1.7255 1.7632 1.7988 1.8325 1A647 1.8955 1.9250 v 0 0161 0 0166 0 0174 0 0186 0 9919 1.1584 13037 1.4397 1.5708 1 6992 1.8256 1.9507 2.0746 2.1977 2.3200 500 h 69 32 IE9 le 270 51 375 38 12312 1299.1 1357.7 14!? 7 1466 6 1520 3 1574 4 1629.1 1684 4 1740 3 1796.9 (457.01) s 0 1292 0 2934 0 4364 O M60 1 4971 1 5595 1.6'23 1 65/8 1 6990 1.7373 1.7730 1.8069 1 A393 1.8702 1.8998 TABLE A.4 PROPERTIES OF SUPERHEATED STEAM AND COMPRESSED WATER (TEMPERATURE AND PRESSURE)

A.5

l Aho pms. Timper: Int. F

.

C/sg la.

. (sat.tomp) 100 200 300 400 500 600 700 800 900 1000 1100 3200 3300 1400 1500 v 00161 0 0166 0.0174 0.0186 0 7944 0 9456 1 0726 1.1892 1.3006 14093 1.5160 16711 1.7252 1 8284 1.9309 600 6 69.58 169 42 270.70 375 49 1215 9 1290 3 1351.8 1408 3 1463 0 1517.4 1571 9 1627.0 1682 6 1738 8 !?95 6 (486.20) s 0.1292 0.2933 0 4362 0.% 57 14590 1.5329 1.5844 1 6351 1 6769 1.7355 1.7517 1.78b9 1.8184 3.8494 18792 e 0 0161 0 0166 00174 0 0186 0 0204 0 7928 0 9072 1.0102 1.1078 12023 12948 1.3858 1.4757 1. % 47 1.6530 700 6 69.04 169 65 270 89 375 61 487.93 1281 0 1345 6 1403.7 1459 4 1514 4 1%94 1624.8 1207 1737 2 1794.3 (503.08) s 0 1291 0.2932 0 4360 0555 0 6889 1.5090 1.5673 16154 1.6580 3.6970 1 7335 1.7679 1 80 % 18318 1 8617 e 0.0161 0 0166 0 0174 0 0186 0 0204 0 6774 0 7825 0.8759 0 9631 1.0470 1.1289 1.2093 1.2825 13669 1.4446 l 800 4 70.11 169 88 271.07 375 73 487.88 1273.1 1339 2 1399.1 14558 1511 4 1566 9 1622 7 167E9 1735 0 1792.9 i (518.2.)

0.1290'02930 0.4358 0.%52 0.6885 1.4869 1.5484 1.5980 1.6413 1.6807 1.7175 17522 1.7851 38164 1.8464 l e 0.0161 0.0166 0 0174 0 0186 0 C234 0 5869 0 6858 0.7713 0 8504 0.9262 0 9998 1.0720 1.1430 1.2131 1.2825 !

900 4 70.37 170.10 271.26 375.84 487.83 1260 6 1332.7 1394.4 1452.2 1508 5 1564 4 1620 6 1677.1 1734.1 1791.6 (531.95) s 0.1290 0.2929 0.4357 0.5649 0.6881 1.4659 .1.5311 1.5822 1.6263 1.6662 1.7033 1.7382 1.7713 1.6028 18329 e 0.0161 0.0166 0.0174 0 0186 0 0204 0 5137 0 6080 0 6875 0.7603 0 8295 0.8966 0 9622 1.0766 1.0901 1.1529 1000 6 70.63 170.33 271.44 375.96 487.79 1249.3 1325.9 1389.6 1448.5 1504.4 1561.9 1618 4 1675.3 1732.5 1790.3 (544.58) s 0.1269 0.2928 0.4355 0.5647 0.6876 1A457 1.5149 1.5677 1.6126 16530 1.6905 1.7256 1.7589 1.7905 1.8207 e 0.0161 0.0166 0.0174 0.0185 0.0203 0 4531 0 5440 0.6188 0 6865 0.7505 0 8121 0 8723 0 9313 0 9094 1.0468 1100 t 70.90 170.56 271.63 376 08 487.75 1237.3 1318 8 1384.7 1444.7 1502.4 1559.4 1616 3 1673.5 1731.0 !?89.0 (556J8) s 0.1269 0.2927 0.4353 0.5644 0.6872 1A259 1.4396 1.5542 1.6000 1.6410 1.6787 1.7141 1.7475 1.7793 1309,7 e 0 0161 0.0166 0.0174 0.0185 0 0203 0 4016 0 4905 0.5615 0.6250 0 6845 0 7418 C.7974 0 8519 0.9055 0.9584 1200 6 71.16 170.78 271.82 376.20 487.72 1224.2 1311.5 1379.7 1440 9 1449 4 1556 9 1614.2 1671.6 1729.4 1787.6 (567.19) s 0.1288 0.2926 0.4351 0.5642 0.6868 1.4061 1.4851 1.5415 1.5883 1.6298 1.6679 1.7035 1.7371 1.7691 1.7996 e 0.0161 0 0166 0 0174 0 0185 0 0203 0.3176 0.4059 '0.4712 0 5282 0 5809 0.6311 0 6798 0.7272 0.7737 0.8195 1400 6 71.68 171.24 272.19 376 44 487.65 1194.1 1296.1 1369.3 1433 2 1493 2 1551.8 1609.9 1668.0 1726.3' 1785.0 (587.07) s 0.1287 0.2923 0.4348 0.5636 0.6859 1.3652 1.4575 1.5182 1.5670 1.6096 1.6484 1.6845 1.7185 1.7508 1.7815 e 0.0161 0.0166 0.0173 0.0185 0 0202 0.0236 0.3415 0.4032 0.4555 0.5031 0 5482 0 5915 0.6336 0.6748 0.7153 1600 A 72.21 171.69 272.57 376 69 487.60 616.77 1279.4 1358 5 1425.2 1486.9 1546.6 1605.6 1664.3 1723.2 1782.3 (604.87) s 0.1286 0 2921 0.4344 0.5631 0.6851 0.8129 1.4312 1.4968 1.5478 1.5916 1.6312 1.6678 1.7022 1.7344 1.7657 e 0.0160 0.0165 0 0173 0.0185 0.0202 0.0235 0 2906 0.3500 0.3988 0A426 0 4836 0.5229 0.5609 0.5900 0 6?43 1800 a 72.73 172.15 272.95 376.93 487.56 615.58 1261.1 1347.2 1417.1 1480.6 1541.1 1601.2 1660.7 1720.1 1779.7 (621.02) s 0.1284 0.2918 0.4341 0.5626 0.68*.3 0.8109 1.4054 1.4768 1.5302 1.5753 1.6156 1.6528 1.6876 1.7204 1.7516 e 0 0160 0.0165 0.0173 0.0184 0.0201 0.0233 0.2488 0.3072 0.3534 0.3942 0 4320 0.4600 0.5027 0.5365 0.5695 2000 6 73.26 172.60 273.32 377.19 487.53 614.48 1240.9 1353.4 1408.7 1447.1 1536.2 1596.9 1657.0 1717.0 1777.1 (635.80) s 0.1263 0.2916 0.4337 05621 0.6834 0.8091 1.3794 1.4578 1.5138 1.5603 1.6014 1.6391 1.6743 1.7075 1.7389 e 0.0160 0.0165 0.0173 0.0184 0.0200 0.0230 0.1681 0.2293 0.2712 0.3068 0.3390 0.3692 0.3980 0.4259 0.4529 1 2500 4 74.57 173.74 274.27 377.82 487.50 612.08 1176.7 1303 4 1386.7 1457.5 1522.9 1585.9 1647.8 1709.2 1770.4 J (668.11) s 0.1280 0.2910 0.4329 0.5609 0.6815 0.8048 1.3076 1.4129 1A766 1.5269 1.5703 1.6094 1.6456 1.67 % 1.7136 e 0 0160 0.0165 0.0172 0 0183 0 0200 0.0228 0.0982 0.1759 0.2161 02484 0.2770 0.3033 0.3282 0.3522 0.3753 3000 A 7583 17e88 27522 378.47 487.52 610.08 1060.5 1267.0 13632 1440.2 1503.4 1574.8 1635.5 1701.4 17tl.8 (695.33) s 0.1277 0.29G4 0.4320 0.5597 0.6796 0 8009 1.1966 1.3692 1.4429 1.4975 1.5434 1.5641 1.621d I.0561 1.6688 y 0.0160 0 0165 0.0172 0.0183 0.0199 0.0227 0.0335 0.1588 0.1987 0.2301 0.2576 0.2827 0.3065 0.3291 0.3510 3200 A 76 4 175.3 275 6 378 7 487.5 609.f 800.8 1250 9 1353.4 1433.1 1503.8 1570.3 1634A 1698.3 1761.2 (705.08) s 01276 0 2902 0.4317 0.5592 0.6768 0.1994 0.9708 1.3515 1.4300 1A866 1.5335 1.5749 1E26 1.6477 1.6806 e 0 0160 0 0164 0.0172 0.01E3 0 0199 0.0225 0.0307 0.1364 0.1764 0.2066 0 2326 0.2563 0.2784 0.2995 0.319P.

3500 A 77.2 176.0 276.2 379.1 487.6 608 4' 779.4 1224.6 13382 1422.2 1495.5 1563.3 1629.2 1693 6 1757.2 s 0.127a 0.2899 0.4312 0 5585 0.6777 0.7973 0.9508 1.3242 1.4112 1.4709 1.5194 1.5618 1.6002 1.6358 1.6691

, 0 0159 0.0164 0 0172 0.0182 0.0198 0 0223 0 0287 0.1052 0.1463 0.1752 01994 0.2210 0.2411 0.2601 0 2783 4000 A 78.5 1772 277.1 379.8 487.7 606 5 763 0 1174.3 1311.6 1403 G 1431.3 1552.2 1619 8 1685.7 1750.6 s 01271 0 2993 0.4304 0.5573 0 6760 0 7940 0.9343 12754 1.3807 1A361 1.4976 1.5417 1.5812 1.6177 1.6516 r 0 0159 0.0164 0 0171 0 0181 0.0196 0 0219 0.0268 0.0591 0.1038 0.1312 0.1529 0.1718 0 16*0 0 2050 0.2203 5000 A 81 1 179 5 2791 381.2 488.1 604.6 746.0 1042.9 1252.9 1364 6 1852.1 1529.1 16009 1670 0 1737.4 s 0.1265 0.2861 0 4287 0.5550 0 6726 0.7880 0.9153 1.1593 1.3207 1.4001 1.4582 1.5061 1.5481 1.5663 1.6216

0 0159 0.0163 0 0170 0 0160 0 0195 0.0216 0 0256 0 0397 0.0757 0.1020 0.1221 0.1391 0.1544 0.1684 0.1817 6000 4 83.7 181.7 281.0 382.7 4P8 6 602 9 73G 1 945.1 1168.8 1323 6 1422.3 1505.9 1562 0 1654.2 17242 s 0.1258 0.2670 0.4271 0 5528 0 6693 0.7826 0.9026 1.0176 1.2615 1.3574 1.4229 1.4745 1.5194 1.5593 15 % .2 v 0.0158 0.0163 0 0170 0.0180 0.0193 0.0713 0 0248 0.0334 0 0573 0 0316 0.1004 0.1160 0.1296 0 1424 0.1542 7000 A 86.2 184 4 283 0 384 2 489.3 601.7 729 3 901.8 1124.9 12813 1392 2 1482.6 15631 1639 6 1711.1 s 0 1252 0.2859 0 4256 0 5507 0 6563 0.7/77 0 8926 1.0350 1 2055 1.31 > l 1 1904 1.44u6 1.4938 1.53'5 1.5735 TABLE A.4 PROPERTIES OF SUPERHEATED STEAM AND COMPRESSED WATER (TEMPERATURE AND PRESSURE) (CONTINUED)

A.6

o ,, .? , .3 ,4 .5 , ,e , .9 20 f. 37 32

,. A kl hull' ,_

A /N^TA/ N // N Y/ /\ I I IX / ,100/

I/1D8I['N/I/T

,. / //%/ / soo ;

V/ /7s /

,. l/)Vl))%% f 7%

/ N/ N///%J

/ f I nob'

,_ /ll$5f N// M lH%J

/ NI I /&

,,,, )fflbWNiK I Nr NI//7%l 2 I NI

,, l)V/b0'$7 N f*%1/ ,,

N$

lN h!QW ~l 1 M 215?% M 7 7 s I .

f/ENL/OYUt %j75 ~l ~

,. IMHK h%'dW/ ICY ~

. MR77 W Y H/M /7 '"

_ K/MWM7 /7 '

'MMNMR< N ~

. #bV%W)5cM ~

M X/X67x'/W J.Et M/W

~

R. . ,.1 YM l

FIGURE A.5 MOLLIER ENTHALPY-ENTROPY DIAGRAM A.7

i PROPEHTIES OF WATER Density e

@bstit*)

PSIA Temp Saturated 2400 2500 3000 1000 2000 2100 2200 2300

(*F) Liquid l

62.888 62.909 62.93 62.951 63.056 32 62.414 62.637 62.846 62.867 62.822 62.846 62.87 62.99 62.55 62.75 62.774 62.798 50 62.38 62.446 62.465 62.559 62.371 62.390 62.409 62.427 100 61.989 62.185 60.702 80.549 60.568 60.587 60.606 60.118 60.314 60.511 60.53 200 57.859 57.882 57.998 57.537 57.767 57.79 57.813 57.836 ~

300 57.310 54.342 54.373 54.529 54.218 54.249 54.28 54.311 400 53.651 53.903 53.95 54.11 53.825 53.86 53.89 63.925 410 53.248 53.475 53.79 53.46 53.50 53.53 53A9 53.025 53.36 53.40 53.425 420 52.798 53.09 53.265 52.95 52.99 53.02 53.065 430 52.356 52.575 52.925 52.51 52.54 52.56 52.275 51.921 52.125 52.42 52.45 52.475 440 52.175 52.21 52.41 51.66 52.025 52.065 52.10 52.14 450 ' 51.546 51.76 51.96 51.61 51.64 51.68 51.725 460 51.020 51.175 51.56 51.22 51.25 51.30 51.50 50.505 50.70 51.1 51.14 51.175 470 50.78 50.825 51.035 50.20 50.62 50.66 50.7 50.74 480 50.00 50.35 50 575 50'.13 50.175 50.22 50.265 50.31 400 49.505 49.663 50.098 49.714 49.762 49.81 49.858 500 48.943 49.097 49.618 49.666 49.203 49.254 49.305 49.56 48.31 48.51 49.05 49.101 49.152 510 48.68 48.735 49.01 47.91 48.46 48.515 48.57 48.625 520 - 47.85 48.155 48.45 47.86 47.919 47.978 48.037 48.096 530 47.17' 47.29 47.89 47.362 47.428 47.494 47.56 540 46.51 47.23 47.296 46.794 46.862 46.93 47.27 45.87 46.59 46.658 46.726 550 46.216 46.29 .46.66 45.92 45.994 46.068 46.142 550 45.25 45.62 46.02 45.30 45.38 45.46 45.54 570 44.64 45.22 44.672 44.758 44.844 44.93 45.36 530 43 66 44.50 44.586 44.015 44.11 44.205 44.68 43.10 -43.73 43.825 43.92 530 43.33 43.434 43.956 42.913 43.017 43.122 43.226 600 42.321 42.55 43.14 42.08 42.196 42.314 42.432 610 41.49 41.96 41.35 41.483 41.816 42.283 40.552 40.950 41.083 41.217 620 41.44 630 39.53 40.388 '

640 38.491 39.26 650 37.31 38.000 660 36.01 36.52 670 34.48 34.638 630 32.744 32.144 690 30.516 TABLE A.6 PROPERTIES OF WATER, DENSITY

A.8

r 0

t

/ (aS h

1. PRINCI:'LES OF NUCLEAR DOWER PLANT OPERATION, PAGE 25

~~~~

TH EkEUUEU55ECEI~5E57~EEds5EER~d5b~f5UEb~FLUU ANSWERS -- FARLEY 12.2 -86/07/14-NELEON, D.

ANSWER 1.01 (1.00' C

REFERENCE DPCr Funderentals of Nuclear Reactor Engineering, p. 170 001/000-K5.12 (3.7/4.0)

ANSUER 1.02 (1.00) t REFERENCE General Physicsr HT F FFr Chapter 1 004/020' K5.08 (2.3/2.6)

ANSWEP 1.03 (1.00) b REFERENCE General Physics, HTE.FF ANSWER 1.04 (1.00) b REFERENCE l

\

S t e e n, Tables '

Mollier Diaster:

1 1

1. - 531NCIPLES OF MUC'I AR POWER PLAN 1 OPEPA110N, PAGE 26

~

~~~~555R55DYU555C5I~H5ET T 5kU555R~dkD FLU 56~FL50 ANSWERS -- FARLEY 1E2 -86/07/14-NELSON, D.

At45WEP : 05 (2.50)

a. Increase

b. Decrease

c. Decrease

d. Increase

e. Decrease CO.4 each] (2.0)

REFERENCE General Physics HT&FF, Part B, Chapter 1, pp 324-332 ANSWER 1.06 (2 00)

a. GREATER-THAN (0.5)

b. LESS THAN (0.5)

c. SAME AS (0.5)

d. GREATEP THAN (0.5)

REFERENCT NUS, Vol 3, Unit 11 001/010-K5.13 (3.1/3.6)

ANSWEP 1.07 (1.50)

a. FALSE (0.5)

b. CALSE (0.5)

c. TRUE (0.5)

REFERENCE Westinghouse Reactor Physics, pp. 1-3.9 and I-3.4 D +

6

1. P R I N C I L E S 'J F NUCLEAR POWER PLAN 1 OPERATION, PAGE 27

7 EEE66Ys5h5CE H I5E5T'IE5U5FEk'dU6'EEUEb'FLEU ANSWERS -- FAPLEY 182 -86/07/14-NELSON, D.

A N 5 W Ei: 1.0E (2.00)

1. FNDH. (0.5)

2. FNDP. (0.5)

3. Rods within a stot_ip are maintained within +/- 12 steps.

Control rod banks are sequenced and. overlapped.

Rod insertion limits are maintained.

ArD limits are maintained. [ 0 . 5 e r . (2.01 REFERENCE FND T/S 3/4.2 ond E:e s i s .

ANSWER 1.09 (2.50)

e. Tsrt fer 1200 otic is 567 (from stean tab 2er).

567 - 53 = 51- F [subcooling of "^ "'

c. . iv <- 7f . q g * #f f.C Pset for 5: 7 e is about 800 psis. 4 4 (1.0) 4t1 2. v t

5. Erra ic pressuriner .

level indication. ~

v (O.i:

c. 65 . (+/- 3 F.) 6's M 6 D

J 10 minutes. (1.0)

REFERENCE FNP EOF 7-1, pp.12 & EOP 7-2 pp. B.

ANSWER 1.10 (1.00)

-Minin.izes thermal stress due to more uniform ten.p dif#erence of fluids

-The outlet temp o' the colder fluid spFroacher the inlet temp of the hotter fluid

-A more uni'orm heat transfer rate is ach:evc; r throughout tw heat exchense- ' .?? ee)

(nore eff icient is an acceptable *esponse also)

RErERENCE CNTO, ' Thermal / Hydraulic Principles and Applications'. pp 5-10 004/020; K5.02(2.5/2.o) 1 i

1. PR3NC1FLES 0" NUCLEAR POWER PLAN 1 OPERAlIDH- PACT 2E

~~~~ ~

TUE55UEYU555C5 EEdi TEdU5 FEE d 6 7EUID FLUU ANSWERS -- FARLEY 1&2 -8 6 / 07 /14 -N E '_ S O N , O.

ANSWEP . 11 (2.50)

2. Clur will be depressed toward the bottom of the core [0,53 due to.

1. Lower control rod level E0.25] and

2. Xenon buildup in top of-core to.25] (1.0)

b. Flun centinues to be depressed more and more towardt the bottom as Xenon builds in top, then reverses as it decays off E0.75].

Control rod movement to compensate for Xenon changes r educes the flon shift E0.75]. (1.5)

REFERENCE rarley Reactor Theory-Manual, pp I-3.15,3.16,2.10 GLJ 126 ANSWER ; (2.00)

3. The s ; ; p *,1 y (greatly) suberitical reactor will '.c ve :

r larger i:- .ler ) ineretse in coun. rete. (1.0)

b. The.sligh.1 7 (greatly) suberitical reactor us.1 take 3 l orise r (t tetter) time te reach a sitble count rete. (1.0)

REFERENCE FNP Training Reactor Theory Manual, pp. H-4-20, H-4-21.

ANSWEP .3 ( 2 . 5 0 '.

a. 1) Presence of a +1ew (or crack of sufficient size). '. 0 . 5 2

2) Low tempersture 50.52 (1.0)

b. Reduces the thermal s tr ess . Reduced D' cross the RV u;11 reducet to.-el/thernel/ttnt; e strett 1 (O.5)

c. Neutt:n e"posure ( i n t e r,t s t e d 0: 0 .st.I the ccteri:1 note br: 4. t i c (iE Ees ' O ' '.' (Reducer duct:1: y.' D:.52 (1.0)

CE7ERENCE WNTC Thernodynan.ics- Volume II, Chepter 13, pp 58-68.

JMC Nuclest Plent Technical s p e c i ' i c r +.. : o n s _ p p . E: 3 / 4 - 3/4 4-12.

1

. l 1

I

v

1. oRINCIPLES Cr NUCL CM ~C- '

% AN7 GoERA110Nr PACE 29 THERMODYNAMICE. uEA1 W NEFER AND F'.UID F LOW AN5WERS -- FARLE'i !E2 -86/07/14-NELSON, D.

ANFREF :,1; ..O -

The idle RCF can den ~m t e r p e r a t e r e s. i. the ses: srez . hat are les.

thrn steen m e n e r r s 8 MI'r o e t e t.o.w

- res 50.52 When the cold sive- ~ coes t h r oi.ig h t h e stear. ; en e r a t o r eefo n n i c 6.s up heat s.nd e::p a nc;s . The thern,e expens:en in r solid p2cnt causer r pressure increase :0.52. (1.0)

REFEREi CE A0o-24 GLJ 133 ANSWER 1.15 (2.50)

s. DNBR = Meat +1c> (power) to cause DNS / actual heat flux (power) (0,5)

5. Greater then or equr: (0.2 pts) to 1.3 (0.3 pts) (0.5)

c. (any 6 at 0.25 pts each) (1.5)

1. RCE pressure 2. RCE t e nt p e r a t u r e

2. RCE f cu a. :P power E Ar3 6. 9 :' T R 7 EC 'i k ( s 5: q ' ' e P! C i 713

C V e P ; -3 p . positiCn)

C O N SIDE F: OTHERS ON CASE-BY-CASE BASIS REFERENCE fnp, TS, pp. 3/4 2-14 and 9 3/4 2-2 and 2-5 ANSWER 1.16 ( 1. 0 0

I r. the seccndary system there is a phase change (0.5 pts). A phase chEnge requirer E large delts h. With the lerger delte h of the secorc=*y, +he szt.e hest can be transferred with a lower flow rate (0.5 ptr).

ErERENCE Gene"51 thys10s? -

l TC. SEC+ ion 32 l 1

002/000-5:5.0 /2.'/3.6)_

l l

l

\

1. PPINCIPLES OF HUCLEAR POWER PLANT OPERATION, PAGE 30

~~~~

TEEE566YUdEEC5 ~0E di~TEd55EEE~dU6~ELUf6~FLUU ANS'4EPS -- F ARLEY 162 -06/07/14-NELSON, D.

M'SWEP :.'7 _ (2.00)

c. Nucleate bo: ling creates turbulent flow which p r o it.o t e s more m i::i n g .

CO.52 The coolant picks up le. tent hect of veporinction and carries it to cooler parts of the channel. E0.53

b. In filn, boilins, a filn; of steam costs the clad surface end f ornis an-insulating layer E0.53 which greatly reduces the ' neat transfer I coefficient. E0.52 REFERENCE NUS Nuclest Energy Trns.

o

2. -

DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 31

_ _ _ _ '_' L A N T ANSWERS -- FARLEY 182 -86/07/14-NELSON. [..

ANSWER 2.01 (1.00)

p. -e _r r r. . e_ L c_ r.

FNP, Steam Dump Sys, pp. 46 8 ANSWER 2.02 (1.00) a.

REFERENCE 10CFR50.46 ANSWER 2.03 (1.00) b REFERENCE FNP. Ma:n and Reheat Steam Sys, p.8 ANSWER 2.04 (1.00)

J REFERENCE FNP, LP: SGHLC ANSL!EF 2.05 (1.00) b.

RECERENCE CNo, Lo* A U :< . Fd. Sys.

ANSWER 2,06 (1.00) b.

e

i

2. PLAN 1 DESIGN INCLUDING SAFETY AND EP.EPGE!1CY E'lSlEh5 PAGE 32 2; ANSWERS -- FARLEY 1&2 -86/07/14-NELSON, D. '.

=2 REFERENCC ,

FNP9 LP: CVCS 7 ANSWER 2.07 (1.50) a) True (+.5 ea) b) Felse :

c) False . ,l .

REFERENCE -

FNPo LP: RCS m

ANSWER 2.08 (1.00) ..

s.

a. normal

b. reversed ,

REFERENCE ','

West. PWR Sys. Maquzl }

!, ~

ANSWER 2.09 ' 5 . 0 0 !'

Normal - Emergency 600V LC-1D to 600V MCC-1A to Inverter 1A which , ,

supplies 120V to Vitel AC-1A instrument distribution pane' '

Backup - Emergency 600V LC-1D to Battery Charger 1A (or 1C) which -

supplier 125 V to DC bus 1A to Inverter 1A (and as above) ,.

Alternate - Emergency 600V LC-1D to 600V MCC1B to 600V/208V ,

transformer to'SOLA 1A which supplies 120V.to Res AC-1A {

to Vital AC-1A instrument distribution panel i C1.0 each2 (3.0) L PEFERENCE F a r l e y '_ e s s o - C'r: Volume a, 'eb 1, Fig. 2 ? Tab 3 rig. 2 GLJ 243 _

s 5

9 4

=

2. F ; A t!' DESICE 7MLUSING SAFETY AND EMERGENCY SYSTEMS PAGE 33 cNSWEPS- C U"_. E Y ;F2 -86/07/14-NELSONr D.

ANSWER i'.10 (1.50)

1. Orifice . ;_ o l . re:vec (3:40 A. 9 C) ruct be chut in crder to o ,o e n er close either LCV-c5c or -460.

2. If either LCV-059 or -460 closes, the orifice isol. valves will close.

3. PZR level of 15% decreasing results in closing LCV-459 and -460.

M m /s53 ,f k vtp p E0.52 each REFERENCE CNP, LD: CVCS ANSWER 2.11 (2.00)

1. maximum letdown

2. meximum sec1 water return

2. excess l e t d o u r, 4 not:nel ' lev fror one reactor takeup' water pump '. 0 , 5 2 each R.ur=r:

- - .- . . t., . .:

END, LP: CVCS ANSWER 2.12 (1.50)

a. 600 V MCC U

b. 600 V MCC V

c. c160 V Bus G r. 0 . 5 ] each REFERENCE l rNF. LP: RHO )

1 1

ANSWER 2.;2 (; .00T A:; .Other corporents ta! , svetion fron high elevstions in the tank. CO.52 No out,enstir cct;cns oreve . EO.52 l l

REFERENCE l FNP. LD* AFU

2 rLAN' OCSIGN INCLUDI+JG SAFETY AND EMERGENCY SYSTEMC PAGE 34 ANSWERS -- FARLEY 1&2 -86/07/14-NELSON, D.

ANSWER 2.14 (1.50) onp ' h c' T :- s Loop 7e RTD's Loops C e. A wide range pressure Pressuriner pressure Referenca jonction bo>< RTD Core enit thermocouples CO.252 each REFERENCE FNPr License Retrns, Subcooling Monitor ANSWER 2.15 (1.00)

Available in normal and abnormal conditions in all modes except refueling.

RECERENCE r N t' . icense Petrns, ICCMS ANEVEP 2.16 (1.00) cold; last pump REFERENCE West. PWR Sys. Manual ANSWER 2.17 (1.00) dreefrb arc or b,drerfrbrarc REFERENCE West.*JR Sys. Manual ANEVER 2.18 (2.00) fj g w b r5

M J

a. _?tch housing ant rod travel housing (0.25 ea.)

L :. ::m i n g '0-5) JJed

c. The voltage on the coils is reduced to prevent overheatin3 cf the stctionary coils which could cause demase to the insulation. (1.0)

2. PLAN 1 DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 25 ANSWERS -- CARLEY 1&2 -86/07/14-NELSON, D.

REFERENCE FNPr LP FCSi RDCNTRL ANSWER 2.10 (2.00)

The voltage regulator nul? meter. compares the output of the base adjuster and the automatic sections of the voltase regulator.. E1.02 If the null meter indicates O the outputs are matched and a bump l ess transfer from manuel to auto may be made. E1.02 REFERENCE FNPrLicense Retrnst Main Gen. and Aux.

ANSWER 2.20 (1.00)

Each individual detector ( both upper and icwer ) CO.53 is compared to the Everese of the sum of the upper ( or lower) detectors. [0.52 REFERENCE FtP r LP: E::c o r e NI AN3WER 2.21 (1.00)

Automatic operation of PZR spray by manual operation of PZR heaters.

REFERENCE INP-1-50P-2.3 ANSWER 2.22 (2.00)

1. Provides a .ufficient delta T signal to sive an indication of voiding wher: the heeted junction is uncover ed.

2. Reduces hester power when the heated junctio~ is uncovered t o . p r e v e r: t heste* bur r:out .

E1.02 ete5 RETERENCE TNFr License Retrns

.2. INSTRUMENTS AND CONTROLS FAGE 36 ANSWERS -- FARLEY 1&2 -86/07/14-NELSON, 9.

ANSWER 3.01 (1.00)

REFERENCE FNPr Reactor Protection Sys. p. 50 ANSWER 3.02 (1.00) c.

REFERENCE FNPr LP: PZR Press. and LVL Cont.

ANSWER 3.03 (1.00)

C i RE:ERENCE Nuclest 5 3uer ' lent Instrumentation Sys. M a r.u a l - Ch. 4 ANSWER 3.04 (1,00) c REFERENCE FNPr Gross Failed Fuel Detector, p. 7 ANSWER 2.05 (2.00)

a. 1. No low-low Tavs (P-12) [0.23

2. Steer. dump byptes i n t e r l o c+ switch not in Orc RESET [0.23

2. Condenser available (C-9) ( C o r.d e n s e r vaccur, and CW pump breaker ct 0.1 pts, ecch elrc acceptibleT' E0.22

< b. Positier.ing the MODE select sv".tch E0.22 t o t' E S E T . CO.22

c. Difference between avet. hi s' 1 r vs end ine-]ord(54' ) EO.42

d. 1. Lesd rejection (C-7' E0.2]

2. Turbine trip (C-8) CO.22

3. Mode switch in STEAM PRESSURE [0.23 l

I l

l l

1 I

3. ~INClRUMEN1G AND CONTROLO PAGE 37 AN'iWER S -- CARLEY 1E2 -86/07/14-NELSON, D.

REFERENCE FNPr Steam Dump Sys. pp. 17,22r25,27,28 ANSWER ?.06 (1.00) 5.

REFERENCE FNPr LP: RPS ANSWER 3.07 (1.00)

d. g A,

^EFERENCE FNP, LP: Steam Dump Cont. Sys.

AN SL!ER 3.08 (3.00)

a. reds cut E0.25] 1re* v 11 be ma:: so Tave/ Tref mismatch and NI/lvrbine power r.isng;ch v: . both give a rodt out sient2 E0.753

b. rods in E0.252 Loop i Tas.e increases and auctioneered high Tave also increases. Tave/Tret mismatch sives a rods in sisnci E0.753

c. rods out EO.25] the power mismatch circuit of the reactor control unit responds only to rate of chanse of deviation between turbine and nuclest power but rod motion will occur due to the Tave - Tref difference. [0.752 REFERENCE FNP. LF: Red Cont. Sys.

ANSWEt 3.09 (1.00)

A v e t i o n e e r e <- .;gh delta T REFERENCE CNPr LP: Rod Cent. S y .E .

d

3. INSTRUMENTE AND CONTROLS PAGE 30 ANSWERS -- FARLEY if2 -06/07/14-NELSON, D.

ANSWER 3.10 (3.00) 1 543 des : 2< 3

+

S.P.

' ~

,-. e . ._;.wi v u mi c ' CP

2. 75% 2/? on 1/2 SG's S.P.

3. Variable: 1/2 >= S.P. in 2/3 loops 40% for P -:20%

increases linearly 40 - 140?. for P>20% e in%

Tave < P-12 E0.52 each st. Pt. I E0.52 each coine.

RECERENCE Fr!Pr License Rettans, SG Protect.

ANSWER 3.11 (2.00)

2. high PZR level

5. low DZR pressure E1.03 erch RECEFENCE CHFr ' P: RPSr RCS, CVCS, RODCNTRL ANSWER 3.12 (3.00)

s. 1 DZR lo press. trip

2. ~'R le press. SI

3. PZR hi press. trip

4. P-11 or SI block pern. 8 PORV interlock 4 eg'd E0.331 each

b. 1. PZR herte- group control

2. C'ZR spray valve control

3. PORV cortrel 3 reg'd CO.332 ecch

c. 1. Beckup hette- control 2 eg'- :0.33: erer

2. CVCS and heater i n t e r l o c e. t RECEPENCE h P W hi (cA '

F N r' r LF; FZR Precs. and Level Cont. Sys.

3. :NSTRUMENTS AND Cor ROLS ACE 3?

A N E 'w 2 R S -- FARLEY 1!2 -Bd/07/14-NE_ SON. D.

ANSWER 3.23 (3.00)

Low cWF suc.:on pr* nure (30 psig or 1, z e t. ) st:'-

ttEndby condensate pung EO.52. Lou rWP tuc t i c . pr et : .'c p r' e s s u r e (300 psig for 30 seci causes FWP ti ip :0. 5_ . (: .0)

b. Velve eute clos es on FWP trip signc1 f r o r. both punps with control switch in AUTO. (0.5)

<. T 'T. ol< sc5 M D TT S/G level - c ie(r r e t: by A V k fM cd4m;t r i p br e ti: e r s

c. 1. Hi-P closing reactor

2. SI - cleered by closing reactor trip breakers

3. Le Teve E R-o - cleered by menvel reset button (1./2) on the MCB tc rese+ La Tave signal CO.5 each] (1.5)

REFERENCE C arley Lesson Plans Volume c, Tab 7, Tab 11, pp 28,29,36 E Volone 6, pp 18,19 v.

u__ , < ; c.

ANSWEP ? 1." (2.001

e. Manus 1 signal calling for red .ovecent Rods move IN. ". 0 . 5 re.2

b. Tave-Tref deviction calling for rod movement however with 'D' selected rods DO NOT move.

c. Manual signal calling for rod movement however reds DO NOT move t' . Teve-Tref deviatior, calling fc- rod movement hovever M steps blocks r.cvement, rods DO NOT nove. Ecaf2 Sto REFERENCE FNP, LP: Rod Cont. Sys.

ANSWER 2.15 (1.00)

'he ind: cation would re:u 5;;a.

RErERENCE CNP. LP: OZO ' -ess. a n c. LV'_ Cont. Sys.

. .. =

3. INSTRUMENTS AND EONTROL5 PAGE 40 ANSWERS -- FARLEY 1E2 -86/07/14-NELSON, D.

ANSWER 3.16 (3.00)

P u n.p s A E C; controlled f om MCB or HS' :0.253 MCB - STOP/AUT0/ START :0.253

- Operable only when remote selected at HSP E0.252 Auto start:

1. ESF sequencer if B pump bkr open EO 253

2. LOSP sequencer E0.252

3. B pump fault trip EO.25~

Pump B controlled from MCB .or HSP [0.252 MCB STOP/AUT0/ START [0.253 one for each train E0.252

- Operable only when remote selected at HSP CO.252 Auto start:

1. Selected train charging pump trips C O . 25'!

2. L O S t' c- EEr sequencer with telected i tr::n chareir.3 pur.p bkr racket out LO4252 RE ERENCE FNP, License Retrns, ECCS ANSWER 3.17 (1.00)

A derivative function in the control circuit provides anticipatory valve response, or PID controller with shaping function generator.

REFERENCE FNP, License Retrne, DCN s 9

i

. i i

l

4. 0 0CESURES - NORt'ALr ABNORMAL, EMERGENCY AND PAGE 41 ,

~~~~~~ ~ ~~ '~~~ ~~~~~~

~~~IEU5b[UGEUk F CUUUU ANSWERE -- FARLEY 1&2 -86/07/14-NELSON, D.

A N S 'JIF 1.0; (2.00; A. 1. 150 spm.

2. 350 spt..

3. 700 spm. CO.5 ea.3 (1.5)

B. Service water does not n.e e t secondary makeup specifiestions. (0,5)

REFERENCE FNP-1-S0t'-22.0 ANSWER 4.02 (1.00) e REFEPENCE 10 CF; 20.101 ANSWE; 4.03 (1.00) c REFERENCE 10 CFR 20.

ANSWER c.04 ( .50)

FALSE REFERENCE CNb-1-UDF-1.1.

ANSWE; ' . ': 5 '3.00) l t, 175,576r~

t. Ir: 7 2 r 4 r '. E0.3 each3 (3.0)

REFERENCE

E0P-0 GLJ 160 4

1 i

i 4

)

c. PROCEDURES - NORMALr ABNORMALr EMERGENCY AND DACE 42

~~~~RdDE6E555C E E5sTR5E-~~~~~--------~~~-------

ANSWERS -- FARLEY 1E2 -96/07/14-NELSON, D.

ANEWER 4.06 (3.00)

a. RCS boreted to at least the cold shutdown concentration (0.5)

b. 350 to 425 psis (0.5)

c. Meatup 100 F/hri cooldown 200 F/hr (1.0)

d. High E0.52 due to measured les density srecter then when hot (1.0)

REFERENCE FNP UDP 1.1 i TS ANSWER o.07 (1.00)

. Io

2. 1A

2. 1C No partial credit.

REFERENCE FNP-1-ESP-0.2.

ANSWER 4.08 ( .50) 100 mret./ week.

' REFERENCE CNP-0-M-001.

AMSL!ER 4.09 (1.00) 0 0.25 poir.t each*

. Stop/ confine spil'..

2. Clear unnecessary personnel from ares.

?. Notify Control Roon.

4. Notify PP Of# ice.

REFERENCE FNP-0-M-001.

1 DROCEDURES - NORMAL, ABNORMAL. EMERGENCY AND PACE 43 .-

~~Ed656[UU5C5E~CUUTEUE~~~~~~~~~~~'~~~~~~~~~~~'

kNSWERS -- FARLEY 1&2 -06/07/14-NELSONr D. -

?

EE:

- 4.10 (2 00' -

1

. Place tain feed res valve in manva'. and maintain SC proerv 'evel.CO.52 '

. Place 10 sprey valve in nenuc1 cnd ent.ure it's fully clot.cd. [0.52 EFERENCE

NP-ADP-4.0. .

SWER a,11 (1,00) -

h rcint : ch ,'ll 4'k 'S

. Place rods in msnual. ,

. Place turbine DEH in hold. g

. Insert control bank 9 to return Tave to program.

EFERENCE ,

NP-1-ADP-19.0. i <

t RS'!ER

4.1C (1.00) 0.25 points each: i _

i. Verify R, Trip and Bypass breakers opened. [. '

. Verify Turbine tripped. 2

. Check RCS isoleted. .

. Check total AFW >377 gym. ,

i .

EFERENCE NP-l-ECR-0.0 E ERP Rezual Training.

s 4,12 (3,003 '

4 SUER 1 To r:1r urter #ct tr,-11 g . t' .

. r:cwpeth cligned 'ro- "W E ' t c. CVCS 2 e" d v- /v:thout e::c ec er :nz 130 Spm t h r '.' '_T ON HX) ent:~_ boren concentrat on is e z u c '. 1; ?-

srecter thrn RCS beror- c o n o c o t t a t i o r, . (;.0 -

. Increase 0.5) t' . RCS overpressure protecticn provided by RHP inlet relief vtives. (1.0) .

4

4. PROCEDURES - NORMALr ' Et 0 R ri A L . EMERGENCY AND PAGE 44

~~~~~~~~~~~~~~~~~~ ~~~~

~~~~EED5bLU55C5E~CUEEUL ANSWERS -- FARLEY 1&2 -86/07/14-NELSON, D.

REFERENCE ENo S3P ~.0 0~ J 163 ANSWER 4.14 (1.00)

Reduce makei.ip flow. en, b c REFERENCE FNP-EEP-2.

ANSWER 4.15 (1.00) 31 spm.

REFERENCE FNP-1-UDP-1.1.

M'S'vER 4.16 ('.00)

Curve ?co: . . g gi, d.d O REFERENCE FNP-1-UDP-1.1.

A N Sk'E R 4.17 (1.00)

When charging pumps ca not maintain pressuriner level.

REFERENCE vl [eb pt 7 T/f [i d .

FNP-1-ADP-1.0.

M:E L'E P 4 10 *: .00' Rem a in on Its-iar with 'he ___cr -. ic .3 as you rensin +be ectort.

R~;ERENCE NP-0-EIe-10.

h

T o .. -

c. "iCCEDURES - i O ~t A L , ADNORPA.. EFERGENCY AND PAGE 45

~~~~E5556[UU5C5[~UUFM 6[~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- FARLEY 182 -86/07/14-NELSONr D.

f, *: S U E F 4,1c (;.00 Start / Star chargins pumpc.

REFERENCE FNP-1-AOP-6.0.

ANSWER 4.20 (1.00) b A

> apsis in containment CO.52 OR > cr = 10E5 in containment CO.52.

REFERENCE FNP-EEP-0, Fold Out Page.

ANSWEP 4.21 (1.50) 0 0.5 poir.ts each'

. By 15= CPG ditertion.

2. ;y ed reath occurs.

I. Any hic:,t r priority ortnse pcth occurs.

REFE dbE FNP ERP Rec;ut! Training.

ANSWEP 4.22 ( .50) 500 p e r. .

REFERENCE rNP-1-UDP-1.3.

ANSUE' a 20 (2.00:

1 cir t a:r:: crid trent's- putp

2. Oper t erger+cy berate to charging pump valve ( M C '.'- 010 4 )

. veri'y one ehtesing eur.p r unning and t her sins pur.ip suction h e :.: d e r is v. i vcives are open. (MOV-8131A 8 E)

4. Incrette letdown flow (to 120 spm).

I i

e l

l

r 1

...o 4, :- v r " < r. !cEs - Notr:<_, ABNORMAL, E M E.F G E N C Y AND r AC: i

~~~~RE5iBE55iEEE~E5F6E5E-~~~~~~~----~~~~-~~~~~~-

ANSWERS -- cARLE1 182 -86/07/14-NELSON, D.

p..r r--r ; c_ h. c. c-

"'*0 g %

f P 9' g i mg & y

ML20215N174

Text

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