ML20155G133

Exam Rept 50-528/OL-86-01,for Units 1 & 2,on 860311-21.Exam results:24 Candidates Passed Written exam,23 Candidates Passed Operating Exam & 1 Failed Plant Walkthrough Portion of Exam

ML20155G133
Person / Time
Site:	Palo Verde
Issue date:	04/22/1986
From:	Elin J, Johnston G, Obrien J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20155G121	List: ML20155G118 ML20155G133
References
50-528-OL-86-01, 50-528-OL-86-1, NUDOCS 8604300152
Download: ML20155G133 (159)

Text

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-9 Examination' Report No. 50-528/0L-86-01 Docket No. 50-528/529 Licensee: Arizona Public-Service Company P. O. Box 21666 Phoenix, Arizona 85036 Facility Name: Palo-Verde I and 2 Examination at: Wintersburg, Arizona Examination conducted: March 11-21, 1986 Chief Examiner:

J. O'Brien, Chief Examiner

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.Dat'e Signed j

Examiner: -

22 J/hnst anilner DateAigned Approved by: [p y0.~Elin,SectionChief Dage pgned Summary: ,

Examinations were conducted from March 11 to March-21, 1986. The Written Exam was administered on March 11, 1986 to six SRO and eighteen RO candidates.

Operating exams were conducted from March 12 to March 21, 1986 to five'SRO-and nineteen RO candidates. All candidates who took the written exams- i passed. All SRO candidates passed the. operating exams, and all but one R0 '

candidate passed the RO operating' exam (he failed the plant walkthru portion).

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71. Persons Examined ,

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R0 Candidates:- , , ,

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Eighteen c'andidates took the' written examination.- Nineteen candidate's-s t took the operating -examination (Simulator and , Oral)... ,.

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-SRO Candidates ,

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Si[ candidates.took'the. written examination. Fi eTcaddidates took th's.( T operating examination (Simulator and Oral). -

2. Examiners J. O'Brien,'RVl(Chief Examiner);_

G. Johnston, RV' ,

J. Upton, PNL" 1 J. Smith, PNL' ~ -

'3. Persons Attending-the Exit Meetings-APS: '

F. Fernow, Training Manager

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+*W.

• J.' Allen, Manager of Operations;

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-+*P. J. Wiley, Training Supervisor, Licensed Operators- '

+ W. Rudolph, Lead Nuclear Instructor. >

+ 0 J. Zerinque, Technica1' Support Manager- ,

~

NRC:  ;  : i

+'J. P. O'Brien. Chief Examiner ./

'+ T. Meadows, Examiner-

+*J. Upton, Examiner '

+*J.--D. Smith,: Examiner

• G. Johnston, Examiner '

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+Present at. Exit on March 14, 1986.

• Present'at Exit'on' . s March'21, 1986,. ..,

-< 4. Written Examination and Facility Review '

- W'ritten examinatifons' were' administered as follows: ~ '

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Eighteen Reactor.. Operator.' Exams. -

_ March:11,21986 U Six Serifor Reactor-Operator Exams March:11,L1986 '

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2 At the conclusion of the exam,'the' facility staff. reviewed-.the exams.

The comments made by the staff at the~ conclusion of'the' review are -

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included in the attachment (1). .The' resolution.of these comments are included in attachment (2). These comments:were discussed with the staff, and' appropriate changes were made to the exam keys' prior to the grading of the exams.

General weaknesses were noted by.the examiners in the accuracy'of.

-reference materials provided by the training st'aff.

5. Operating Examinations i

r Simulator'and oral exans were conducted March 12, 1986 to March 21, -

.1986. The examiners noted a significant improvement'in the operability-and capabilities of the simulator.

I General weaknesses were identified by;the examiners in the area of Radiological procedures cud release points,. plant radiation monitor design, use and operation.

6. Instructor Certifications In a NRC letter dated February ~ 28, 1986 to E. E. 7an Brunt'of ANNP, the NRC authorized ANNP to administer certification examinations to

, 'two instructor, candidates. This action was granted due to the lack of adequate examiner resources during this exam period. The Chief Examiner reviewed ANNP's documentation of their written and operating exams, and finds the exam comparable to a-NRC SRO exam. 1Also, he agrees that both candidates successfully demonstrated SRO level of.

knowledge, and can be allowed to teach. systems.Eintegrated: responses, transients, and simulator courses at the Palo'_ Verde facility. To maintain Instructo'r Certifications the candidates are required to'be enrolled in the NRC approved requalification, program. _The NRC

~

reserves thc right to examine these candidates when resources are available or possibly.at a future requalification program audit.

7 Exit Meeting:

On March ll4, 1986 and March 21, 1986, exit meetings were conducted with the: licensee representatives listed in paragraph 3. Those i candidates who clearly passed the operating exam were identified.

General weaknesses noted.in paragraphs 4 and 5 and procedures for Instructor Certification;were also discussed.-

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-Attachment 1 Following are facility _ comments on the NRC examinations administered on March 11,1986. '

Reactor Operator Examination..

Section 1 -- No comment.

O There~are 2 correct answers listed.in the'provided selections. . Answers b as well as d are correct.

Ref. PGS-15D-10.

2.09-- All valves listed are motor operated valves. The

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loss of electrical power to these valves will cause all valves'to fail as'is. The. reference is based on valve fail positions due to a loss of. central power. .. . .

Therefore'all answers should be AS -IS.

2.10 - Additional answers.not in Key are:

1) Low Lube Oil Press.

2) Iow disc press. shaft driven L.O. pump.

Ref: system description manual " Main Turbine" page PGS-3A-44. ,

Sectica 3 3.09 - b - Answer aa stated is correct but may'also see j

47. of instrument range,.as it is sometimes stated.

c - AMI will only occur if selection switch for ,

neutron flux 'is in the '? AVG" position. - {

i Ref. - Sys.- Descrip. RRS training. Article ,U NS 9C Rev. 1, page NS-9C-7.-

3.10 - The steam flow signal used for.the setpoint genera- i tion and the-quick open signals are' biased.by'PZR-pressure. Some operators may include this.informa- l tion in the answer. _I Ref. - Sys. Desc. " Steam Bypass Control System" l NS-9B, page NS-9B-19..

3.11 - Detector (2) is located at the edge of the pool ~

vs over.the pool.

Ref. - Actual plant location.

3.12 - a -' Correct answer is - SIAS,' CIAS:and MSIS. <

All are received on High. Containment Pressure.

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.3.12 - b - Should be-AFAS.-l'vs:AFAS,2. .But'AFAS.1-is locked-y out by;high Steam Generator" differential pres-?

sure. ::An acceptable; answer should-be: i AFAS- 1 with" a 'high . delta P lockout, or AFAS 1-is notflisted as:it will not occur due -

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.to the high delta P lockout.

I Section 4 - . .

i '4.05 --Question'needs to state ifp power is increasing or p-decreasing to be a valid question..

Also, valve does not fail closed but receives a closed signal at 15% powerJincreasing.

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i- Question"should be invalidated.

4.13 - An'swer should accept.information indicating candidate

! is aware of' panel location,cPanel name and number:

F designation are not required knowledge forLRO's.';

I Section 5 - '

j '5.02 - b - With conversion sh'eet provided~the calculated. ~

, answer is 9652 lbm/hr which is outside answers

' key acceptable range. Widen acceptable ~ band _

, on question..

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5.06 - Another consideration is the change in Beta over

] core life... Beta gets smallerias the core ages ca_us-ing increased' reactor response for a given amount.' .

of positive reactivity.

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5.13.C - Correct answer-should be; The probability

of'a s J

thermal. fission is greater,for Plutonium 239 than ,

Uranium 235 due _ tio Plutonium 239 having a larger - ,

fission cross-section.

5 i

Section 6 - '

6.04 - a .True - Reference 41A012215 states that loss.

of power to PNA-D25 (since'PKA-M41 is lost)

! . cause:r train A, bistables : to trip. (p. 27) - and

' loss of PNC-D27 causes train C.bistables to' trip (p. 16 step 4.2.1). "Therefore'since the.

A and C bistablesLare tripped'the AC matrix '

(: will cause a~Rx trip. ,y -

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Answer is correctLas given..

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!~ c. Answer-is' correct as'given. (CREFAS.will? actuate.

j; but CIAS will not.)..

~, d. ;True'. Same situation as"in partia.7 This'will.

cause the ESFAS "AC"; Matrix to call?for MSIS.'

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<Attrchment 1,(Con't$)

6.06 a2 In the manual mode no setpoint is used. Instead

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~the operator.has direct control over pressure controller output via the manual-adjustment switch Ref. - Sys Desc. NS-9G-7.- PZR Press. Control System.

6'.07 - Valves in a thru e are motor operated.and as-such Valve will fail as is on loss of electrical power.

in part f is solenoid operated and_will fail closed.l ,

The reference is based on valve fail positions due to loss of control power. Therefore answers a.-

e are fail as is. ' <

f-is closed.

However 6.11 - Answer is correct per reference material.

unit controllers were changed out from Foxboro 250 -

series to 270 serie's which have white- backlit indica-3 tion when in auto or Manual mode. ' Expect response to be White.-

6.12 - b - Remove words "and tagged out" from answer key.

No requirement, exists to tag out the power.

1 supplies.

Section 7 -

7.07 - b - When the high CPS alarm is received;(approx.

2000 cps).

Present answer states (approx. 200 cps).

Ref. 410P-12203.

7.08 - c - Answer is false based on the.(about 500 pcm) ~

- The procedure states (about 100.pcm) other than this area'True is the correct-answer.

Ref 410P-12203, 7.11 - Answer c is correct not d'.-

Ref. 4120-1ZZ01 Rx trip pg.,27. ,

Section 8 -

8.10 - Question did not require that positions be specified as.part of answer. : Answer key. reflects'positians-must be specified. ~ Correct number'sh'ould constitute correct answers. 4 sr

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- r ATTACHMENT 2, PARTS 1,:..

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, - F 2itESdLUTIONS TO : FACILITY COMMENTS ' FOR , N ~'

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. THE SENIOR REACTOR' OPERATOR, EXAM:.

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5.02.b. ' -

Corrected calculation uding' number given foric'onversion .

, ' factor on the equation sheeti 1 Acceptable band-remained the- -

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Added the fourth consideration given by the< license'e and ,

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redistributed points. . . Full; credit now requires.4~ .

e considerations.to be given.  ; .

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o'.r 6.04 - Correct a and d t'o read: TRUE; (Reference material ~

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KEY changed to read: "in manual: operator cor:rols direct. *=

output of controller output (re.ference material provided wan' = m .

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Key. corrected: a-e - as is. -

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g. Answer key corrected to , accept;either " white" 'or' " red". ' .

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procedure.-

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RESOLUTIONS TO FACILITY COMMENTS FOR' - , .

' \ --THE REACTOR OPERATOR EXAM , ,

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DAnswers(b[)'and'(d.);aremutually, exclusive....(b.) Tis i wrong because:the swing charger'is normally tied to bus-B..'

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4 No-; change to the answer key.

2.09 "

Loss of electrical power;to the' signal system wil1~ result

in?the failed position, listed in the reference. . Loss of'

' electrical power to the drive' motors:results in "as-is"'

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, positions. Hence,'either the, original answer key or "as-is"'

for each/every valv'e will _be ' accepted. . -

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2.10 - - These two' items will,be added to'the listfof trips in the Answer Key.- .-- ,a 3.09b -

Either ."4% of instrume'nt range or "5% of power;1evel". will' ~

2 be accepted.  ;

3.09c -

"AMI" will be accepted either with or without the qualifier.

3.10 -

Either " steam flowrate" or " steam flowrate biased by:PZR; pressure" will be accepted. ~

3.11 -

" edge of the, pool","or," pool area", or "over th'e pool" ; I will be accepted. -

3.12a The answer'.will be corrected"to read "SIAS, CIAS$ MSIS i

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o' 3- .[+0.33 each]".- '

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3.12b -

The answer will'be'c'orrected to: read "SIAS,-lCIAS,1CSAS,m ,

MSIS and AFAS-1 with a'high. Palockout/or ~no AFAS-1-[+0.2. l each]".'

4.05 .-

Question 4.05'will be deleted. lThe terminology in'the.. .

. S j reference" material, " fail closed",-is indeed mialsading andi H should be corrected.,

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Agreed; inclusion of the panel name and espe' cia'llyynumber, 1 I

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.are not required for full credit.

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U.S. NUCLEAR REGULATORY COMMISSION , -

SENIOR REACTOR OPERATOR LICENSE EXAMINATION - '

. Facility: PALO VERDE 1 AND 2 -

Reactor Type: L,t-FWK '

Date Administered: MARCH 11, 1986 Examiner: J. O'BRIEN Candidate: .~

INSTRUCTIONS TO CANDIDATE:

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

question sheet on top of the answer sheets. Points for each question.are indi-cated in parentheses after the question. The passing grade requires at least  :

70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts. >

% of Category  % of Candidate's Category  !

Value Total Score value Category 25.0 25.0

5. Theory of Nuclear -

Power Plant Operation, .l Fluids, and Thermo -  !

dynamics 25.0 25.0

6. Plant Systems Design, Control, and .

Instrumentation 25.0 25.0  ;

7. Procedures - Normal, Abnormal ~, Emergency, '

and Radiological Control .

25.0 25.0

, 8. Administrative Pro-cedures, Conditions, ,

and Limitations  !

100.0 i Totals  !

Final Grade ~

.-a. .

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-aid.All work done on this examination is my own, I have neither pf"an nor receivedes l .

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i Candidate's St( #e e e. %e-O

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U.S. NUCLEAR REGULATORY COMMISSION .

SENIOR REACTOR OPERATOR LICENSE EXAMINATION -

i . Facility: PALO VERDE 1 AND 2 4

~ Reactor. Type: GL-VWK Date Administered: MARCH ll, 1986 Examiner: J.'0'BRIEN

. Candidate:

INSTRUCTIONS TO CANDIDATE:

Use separati. paper for the answers. ' Write answers on'one side only. Staple question sheet on top of the answer sheets.-. Points;for each' question are indi-cated in parentheses after the question. The passing. grade requires at-least 70% in each category and a final grade of.at.least 80%.< Examination papers will be picked up six (6) hours after the' examination starts.

% of j Category  % of Candidate's '

Category 1

Value Total Score Value ' Category 25.0 25.0 5. Theory of~ Nuclear -

Power Plant Operation, i Fluids, and Thermo-4 dynamics 25.0 25.0 i , ~ 6. Plant Systems Design, Control, and Instrdmentation

. /.

25.0 25.0 7. Procedur7es-' Normal, 4 Abnormal, Emergency, and Radiological

. Control-25.0 25.0 8. , Administrative Pro-cedures, Conditions,

- and Limitations

. 100.0 .

Totals 1 .

Final Grade ,

.\

i All aid, work done on this examination is my own, I have neither given nor received i .

Candidate's Signature s.

l - - _ _ . _ - - - - - . - - . - . - - . - . - _ - - - - . . - .

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• 5

• EQUATION SHEET I* ~*

C v = mg a = v,t + at 2 Cycle efficiency = Net Work (out)

Energy (in)

E = mC a = (vg - v,)/t .

KE = mv vg = v, + a t A = AN A = A,e PE = agh w = 0/t A =.in 2/tg = 0.693/tg ,

W = v4P AE = 931Am t q(eff) = -(t )(ts) i -

(tg+t)b

=$ CAT p -IX ~

1=Ieg ,,_

Q = UAAT I'= I e-UX Pwr = Wf m -XM I=I o 10

. P=P 10 SUR(t)' TVL = 1.3/p t

P=P oe HVL = 0.693/u SUR = 26.06/T -

T = 1.44 DT SCR = S/(1 - K,gg)

SUR = 26 /A*ffp) CR g = S/(1 - K,ggx) .

T = (1 /p ) + .

[(g.- p)/leff] p CR y (1 - K,gg)g = CR 2 (1 - K,gg)2 T = 1*/ (o - D M = 1/(1 - K,gg).= CR g/CR 0

~ ~

! 0 eff M = (1 -- K gg)0/ (1 - K,gg)g 8 " ( eff" }! eff = AK,gg/Keff SDM = (1 - K,gg)/K,gg p= ~

[1*/TKygg ] + [E/(1 + A,ggT )] ,

t,* = 1 x 10 s,econds P = I(V/(3 x 1010)

~

A,gg = 0.1 seconds I = No Idlg=Id22 WATER PARAMETERS Id g =Id22 1 gal. = 8.345 lbm 2 R/hr = (0.5 CE)/d (meters)

I gal. = 3.78 liters R/hr = 6 CE/d (feet)

I ft = 7.48 gal.

MISCELLANEOUS CONVERSIONS .- _.

Density = 62.4 lbm/ft 1 Curie = 3.7 x 1010dps ,,'~^

Density = 1 gm/cm 1 kg = 2.21 lbm Heat of vaiorizations = 970 Etu/lbm 1 hp = 2.54 x 10 BTU /hr

, Heat of fusica = 144 Btu /lbm 1 Mw = 3.41 x 10 Btu /hr k 1 Atm = 14,7 psi = 29.9 in. I'g. 1 Btu = 778 ft-lbf I ft. H O 2 = 0.4333 lbf/in 1 inch.= 2.54 cm

• Fu 9/5 C + 32 C = 5/9 ( F - 32)

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Senior-Operators Exam

. SECTION 5 Theory of Nuclear Power Plant Operation, Fluids, and Thermodynamics ~

5.01 (1.5)

The reactor is deternined to be shutdown by 6%' delta .K/K with indication in the source range of 30 count.s per- second.

a) What is the Keff when the reactor is shutdown- (0.5) by 6% delta K/K?

b) What would the count rate be if Keff is (0. 5 ) -

increased to O.987 c) What would the count rate be if Keff is (0.5) increased to O.99?

5.01 Answer:

. delta K

= 0.06 K

1-K

=

= 0.06 K '

\

1 =K - 0. 06 (K) 1= 1.06(K) ..

a) KEff = 1/1.06 = 0.94 (0.5) 1 - K1 CR2 and 0.06 CR2

-_--__ a _____ ____ n --

1 - K2 CR1 0.02 CR1 j

\

b) CR2 = (30) (3) == 90 cps (0.5)

Conversly CR3 = 0.06 30 0.01'

. c) CR3 = (30)(6) = 180 cps (0.5)

(Rule of thumb, doubles)

Reference:

Glastone and Sesonske, " Nuclear Engineering".

Palo Verde Reactor Theory notes p.8.6-8.11 1

3 4

5.02 (3.0)

The turbine of the Auxiliary Feedwater System (AFS) is rated at 1250 hp with steam supplied at a pressure of 1170 psi. ~

a. When the AFS turbine is being supplied with 1000 psia

, main steam, How much work can be supplied by the turbine? Assume thermodynamica11y ideal operation of the turbine. Express the .. answer in Btu /lbm.

(2.0)

b. If the turbine was receiving 250 Btu /lbm, and- was

. driving the AFS pump with a power of 950 hp, what

~is the steam flowrate through the turbine?

( Express the answer in Ibm /Hr. ) (1.0) 5.02 Answer (3.0)

a. h1 = 1193 Btu /lbm (using CE Steam Tables) (0.5)

Isentropic expansion (0.5) h2 = 905 Btu /lbm (0.5) i Delta h = 288 Btu /lbm (+/-10 Btu /lbm) (0.5) z Sqm* 7 s QIS

b. 950 hp X 2545 Btu /h_r_ = Gr42 M btu /hr (0.5) 2., Yl .3 (L +2 M btu /hr)/(250 Btu /lbm) =

4 ftS2-0,650-Ibm /hr (+/-25 lbm/hr)

(0.5) reference: PGS-11,12,13 and Palo Verde HT&FF notes Section 2, s pp.9-16

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4 9

9

p. 1

-2

4 5.03 (3.0)

Refer to the FIGURE 5.2 that follows this page. . The figure is of " Heat Flux" versus " Temperature Difference between the Clad wall and the Bulk Fluid" for an operating reactor. Note that there are two curves represented for two pressures:

( Assume: P1 < P2 ).

a.-What is the principle type of heat transfer (0.5) that is occuring at pressure P1 and 1.OE4 BTU /Hr-ft*between the wall and the bulk fluid?-

b. What is the principle type of heat transfer (0.5) that is occuring at Pressure P2 and 1.OE4 BTU /Hr-ft*between the wall and the bulk fluid?

c. What pressure will yield a Igwet fuel -(0.5) at 1.OE4 BTU /Hr-ft? i centerline temperature

d. What pressure will yield a highgt (0.5) centerline temperature at3.OESBTU/Hr-i[?fuel e) What type of heat transfer between the wall and (0.5) the bulk fluid is occuring at Pressure El and 3.OE5 BTU /Hr-ft'?

f) Assuming bulk temperature well below (0.5) saturation, Will decreasing the pressure affect the bulk fluid temperature at a heat flux of 3.OE5 BTU /Hr-ft?

5.03 Answers (0.5) for each.

a) Nucleat boiling. ] j l

b) Convection (other terms may be used). )

I c) Pressure 1. j d) Pressure 1.

e) Radiant heat transfer (steam blanketing),

f) The bulk fluid temperature remains constant (is  !

. independent of pressure below saturation.)

I

Reference:

Palo Verde HT&FF notes pp.34-38.

General Physics Nuclear Technology, Section E, Pages 2-144, 2-151, 2-159, and 2-164.

t

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~2 5.04 ( 1. 0 ) -

i l

SELECT'THE BEST ANSWER: .

- As a subcritical reactor nears criticality, the length of time to reach equilibrium count rate after an insertion of a given fixed amount of positive reactivity: (1.0).

a. decreases, primarily dueI to the increased population of delaved neutrons. < .,

b. increases, primarily due to the increased populhtion of delayed nautrons.

c. decreases, because the source neutrons are becoming less .important in relation to the total neutron population.

~

d. increases, because of the larger number of neutron life cycles required to reach equilibrium.

5.04 Answer (1.0) d.

Reference:

Palo Verde - Reactor Theory review , Ch. 8  !

5.05 (2.0) i Answer the f ollowing TRUE or FALSE : ,

' ~

a. The production rate of indirect Xenon from Iodine is .

faster than the decay rate of Xenon to Cesium. (0.5)

b. Slowing the rate of a power decrease, lowers the height:

of the resultant Xenon peak. (0. 5) -

c. TheresultantXenonpeakfromareactor.Triphrom50%-

power is larger than a trip from 100% power. (0.5)

d. During an increase in power from equilibrium' Xenon conditions, Xenon concentration initialy decreases.. - (0. 5 )

5.05 Answer (2.0)

a. True

-b. True

c. False

d. True

Reference:

Palo Verde Reactor Theory Notes Ch. 16.

s i 4

- __ . . _ . _ . ...-. . .- _ _ ..._,; _ _. ._... _ , . . . , . . _ . . . , _ - _ _ . , _ . . .m.. - ,_ . . . . .

. .g s

5.06 (2.0) v Regarding a Main Steam line ruptures a) Why is a rupture of a Main' Steam line at End of (2.0)

Life (EOL) a much more limiting accident' than .

at the Beginning-of Life (BOL)?

5.06 Answers-a) The Moderator. Temperature Coefficient'is less negative' at BOL 193bZ1 .Than at EOL. This difference in magnitude increases'the severity of the addition of positive reactivity. 19t6Z1 due to the sudden cooling of the reactor coolant from the Main Steam- line rupture 19sb21-  % M4 @f18- " M "JA

& f+ / W a ut.- l e. -f 6 (t'

Reference:

Palo Verde Reactor Theory Notes Ch. 12 ,

General Physics, Volume III, Chapter 3, Section 3.

5.07 (1.0)

A tank contains water to a level 40 ft above the bottom of of the tank. A Nitrogen cover gas is at 100 psia. The tank and its contents are at 70 degress F, and.the density of the water is 62.4 lbm/ cubic ft. The pressure at the bottom of the tank is: (1.0)

a. 117 psia

b. 132 psia ,

Y

c. 208 psia

d. 308 psia. e 1

5.07 Answer (1.0)

a. 117 psia (1.0)

Reference:

Palo Verde Thermo Review Sect. 2 pp. 6-9 General Physics HTFF notes pp. 2-115 117 t">

- (

s c.,

b 5.08 ( 1 '. 0 )

If the temperature of the tank in QUESTION 5.07 (container,.

water and gas) were raised and if no water or cover gas was allowed to enter or leave the tank, the pressure at the <

bottom of the tank would . . (1.0)

a. increase, because-the gas volume has decreased and ,

the temperature.of the gas has increased.

b. increase, because of the pressure caused by the  ;

higher water level ( water column effect ), and the-temperature of the gas has increased.

c. decrease, because the water density has decreased.

d. decrease, because the cover-gas density has decreased.

5.08 Answer (1.0)

a. (1.0) i 1

Reference:

See question 1.02 4

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y 5.09 (1.0)

For the following indicate which.is- the mggi- accurate

- statement: -

l.

a. The main condenser functions by: . (0.5)

' 1. Removing the latent heat of vaporization at 4 a constant temperature to allow the steam to condense.

2..Providing a low pressure volume'that allows

-the-steam-to condense.

3. Cooling the steam to the point where it.

{

is at saturation temperature.

4. Lowering the temperature of the steam below ,

j the saturation temperature.

a

b. Condensate depression ist (0.5)-

1. Maintained by adequate backpressure on- the low pressure turbine exhaust.

2. Used to maintain constant temperature I

profile on the condenser tubesheets.

3. Used to maintain adequate Net Positive Suction Head for the condensate pumps.

i a 4. Used to maintain adequate. Net Positive ,

Suction Head for the main feedwater pumps.

i '

5.09 Answer: ,

l i <

t

a. 1. (0.5) l l

b. 3. (0.5)

Reference:

General Physics " Heat) Transfer and Fluid *

. Flow"

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y

-4 5.1l0-(2.0) l l

At Beginning of Life (BOL) the plant nuclear instrumentation count rate will inctgase as plant ~ temperature increases causing a Reactor Coolant System (RCS)- coolant density decrease. -l a) Assuming there is no rod movement or changes in RCS boron concentration. What twg factors contribute to this phenomena? (2.0) 5.10 Answers a.

1. Neutrons travel farther, more leakage. (1.0)

2. Lower boron density, less baron absorption. (1;O) i

Reference:

General Physics, Volume II, Chapter 4, Sections D & D.

5.11 (1.0)

Of the following operations, which gns will have a negative effect on the available Net Positive Suction Head (NPSH) for a given centrifugal pump: (1.0)

a. Throttling open the pump's suction valve.

b. Throttling open the pump's discharge valve.

, c. Decreasing the pump's speed.

d. Decreasing the temperature of the fluid being pumped. r 5.11 Answer (1.0) b (1.0) ,,

Reference Palo Verve Thermo Review of L.D. pp.21-28 e

e B

e , , - , - -- - .e ,,,--,y --, . . - o a ,,a ee

3 5.12~ (3.0)

a. Explain how the energy of nuclear fission is transfered to the reactor coolant. (2.0)

b. Explain why it is necessary to continue cooling the core following a . shutdown from power .

generation. .(1.0) 5.12 Answer (3.0)

a. Most of the fission energy (93%) is in the form of-fission product kinetic energy. The fission particles are slowed down in the fuel matrix -by inelastic collisions, and this hec .s up the fuel matrix (1.0).

The fuel heats up the clad by conduction and/or convection (0.5); which in turn heats up the coolant (0.5)

6. the heat generated by fission product decay in-the fuel matrix - decay heat - (0.5) is sufficent to damage the '

core if cooling is not continued (0.5) i

Reference:

General Physics Vol II, Ch 4.

J 5.13 (2.0)

a. Describe what reaction produces Plutonium-239 in the core. (1.0)

b. Explain how and why having Plutonium-239 in the core affects the core's reactivity?. (0.5) i

c. As a fuel, what is the major difference between Plutonium-239's and Uranium-235's fission cross-section 7 y (0.5)  !

5.13 Answer (2.0)

a. Pu-239 results from the neutron-capture reactioh with U-238 (1.0). (will also accept equation)

b. Pu-239 is fissionable, and therefore adds positive reactivity to the core. (0.5)

-c. U-235 predominantly fissions from an absorbtion of a neutron of thermal energy. Where as Pu-239 is more likely to fission due to a interaction with neut-on of epithermal or higher energy. (0. 5) M Pg 277 /uu. cgru.gM

Reference:

General Physics - Reactor Theory Notes Vol.II Ch.2-4 P/"&

i 9

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.- . s 4

5.14 (1.5)

The following pairs of terms sound similar, but are in fact different. Explain for each, the difference betweens

a. Fast and Prompt Neutrons.. (0.5) ,

b. Slow and delayed Neutrons (0.5)

c. Activity and Reactivity (0.5) 5.14 Answer (1.5)

a. Fast neutrons refer to those at high energies. (0.25)

Prompt neutrons are those. released at the. instant of the fission. (0.25)

b. Slow neutrons are those at low energy levels (0.25)

Delayed neutrons are those which appear at some time after the fission event due to the decay of fission products. (0.25)

c. Activity refers to the rate of radioactive decay (0.25) reactivity e:< presses the deviation of a reactor core from the critical conditions (0.25)

Reference:

General Physics - Reactor Theory Notes Vol.II, Ch.2 s

>b e

4 END OF SECTION 5 10

. . s t

SECTION 6

, PLANT SYSTEMS, DESIGN, CONTROL, AND INSTRUMENTATION 6.01 (3.0)

Operating and Bistable Trip. Channel bypasses are used in the Reactor Protection System.

a. List two of the three Operating Bypasses and ('2. 0 )

briefly describe at least one plant condition when you would expect to use each of these bypasses.

b. How does placing one bistable trip channel in (0.5) bypass for maintenance affect the Reactor Protection System 2 out of 4 logic ?

c. What will automatically occur if an operator (0.5) attempts to bypass the same paraceter in more than one channel at the same time ?

4 4

6.01 Answer

a. (1.0 each for any two of the following)

Low pressurizer pressure bypass system tests at low power or low temperature or heatup or cooldown of the RCS.

High log power level bypass: during reactor startup or above 10-4% power.

DNBR/LPD trip bypass: at low power levels or below 1% power. '

b. Converts RPS logic to 2 out of 3 logic (0.5)

, c. the lowest priority channel will return to (0.5) normal (interlock prevents two channels in -

bypass at the same time)

Reference:

NS 6-25 & 26, Reactor Protection System 4

1

l . .s l

l 6.02 (4.0)

Assume- that the Plant is at 75 % power, -control systems are 'in automatic, and the selector switches f or Tave, and Nuclear power control channels selected to " average". Consider each of the f ollowing situations one at a time.. ,

i

a. What will be the instrument response if a Tc (1.0) '

RTD cable is cut ?

(ies fail high, fail low, or stay the same)

b. With a Th RTD failed high list four of the- (1.0) five alarms or indications you would expect.

c. How will the CEDMCS respond in " auto- ( 1. 0) -

sequential" if the single TLI which is selected for control f ails low ?

d. How does the instrument indication fail and (1.0) what two alarm do you expect if one of the nuclear power control channel detectors

. shorts out 7 6.02 Answer (a) Fails high (open circuit) (1.0)

(b) (0.25 each for any four of the following) (1.0)

Tave-Tref Hi-Lo alarm, RCLoops Hi AWP, AMI, Channel deviation (on RRSTD), pressurizer level at or approaching maximum (c) CEAs will insert or an insert signal will be (1.0) I sent to CEAs (d) SU and ONT CH TRBL Alarm, AMI'and h nnel (1.0) 9 indication Z? Deviation LED on the RRSTD /(the g f ails high) s References NS-9C, 14-20, Reactor Regulating System 41AO-12Z34, RRS Malfunction I

2 j

l 6.03 (3.0)

a. Why must the Emergency Diesel Generator (D/G) lube oil circulating pump be run for a minimum of 30 minutes after shutting down a hot D/G7 (1.0)

b. List four (4) of the actuating signals that shoulE automatically start the essential cooling water System.

(2.0) 6.03 Answer (3.0)

.-r . o .

a. to dissipate heat (0.5) from the turbocharger Bearings.

(0.5)

b. (any 4, 0.5 each)

1. Loss of Offsite Power or (LOP)

2. CREFAS

3. SIAS

4. CRVIAS ,

5. AFAS-1

6. AFAS-2

7. CSAS References SD: Emergency Diesel Generator Sect.3.28, pp.10 SD: Essential Cooling Water Sys. Sect. 4.1.2 , pp.2 s

O j

e 3

q uo 6.04 (2.0) 125 VDC IE bus PKC-M4J has just been donergized due to a fmult.

Alternate power is being lined up to the associated 125 VAC IE distribution pannel PNC-D17 while the DC bus is being examined to decide wheather to reenergite the bus or to shutdown the plant.

State wheather the following are true or false.

a. A loss of power on 125 VDC bum PKA-M41 before (0. 5) -

PNC-D27 is reenergized will SCRAM the plant.

b. A loss of power on 125 VDC bus PKD-M44 before (0.5)

PNC-D27 is reenergized will cause a SIAS to occur.

c. A loss of power on 125 VAC Di stribution (0.5)

Pannel PNB-D26 after PNC-D27 is reenergized will cause CREFAS and CIAS actuation.

l d. A loss of power on 125 VAC Distribution (0.5)

Pannel PNA-D25 before PNC-D27 is rr.energi z ed will cause main steam and feedwater isolation.

6.04 Answer (0.5 each)

u) wa s z n F l' d "

(b) True I

(c) False ,

(d) F-a l s e M I M" *

,_; -~ =:::r+"- t -i pt- (1.0)

Reference:

PGS-15D, 125 VDC Class IE Power PGS-15E, 125 VAC Class IE Power 41AO-12Z17, Loss of 125 VDC Class IE Electrical Power 41AO-1ZZ15, Loss of 125 VAC Class IE Electrical Power I

4

6.05 (3.0)

Assumo' that. the controlling pressure- transmitter for the.

Pressurizer Pressure Control-System has failed high while in service. .

a. What two alar. , would you expect to receive- (1.0) in the control room ?

b. What will the spray valve (s) and the heaters (1.0) do 7

c. If the operator does not take action how will (1.0) the plant be affected ?

6.05 Answer (a) PZR TRBL and PZR PRESS HI-LO (1.0)

(b) Spray valve (s) open and all heaters will be (1.0) 044.

(c) Plant pressure decreases until SIAS and (1.0) reactor trip occur.

Reference:

NS-9G-13, Pressuri:er Pressure Control System 3

e

.6 e

5

d 6.06 (2.0)

The pressurizer pressure controller provides singular control of heaters and sprays.

a. How is the output of the pressurizer pressure master controller determine when in:

~

1. AUTOMATIC (0.5)

2. MANUAL (0.5)

b. As the controller output increases from 0% to 67% ,

What takes place to control pressure? (0.5)

( assume controller in AUTO )

c. As the controller output increases from 67% to 100%,

What takes place to control pressure? (0.5)

( assume controller in AUTO )

6.06 Answer (2.O) , , j a-Difference .= -

between thE s _etpoint an[ the measured a.1 i pressure of the Pressurizer. (0.5) a.2 a A .p u. n e. M ~ t d 4 A .3 4 J Difference between the setpoint and the value set by the operator. (0.5)

b. The power supplied to the proportional heaters will go from full power to zero power. (0.5)

c. ( the output of the spray valve controller will increase from 0% to 100%) The spray valve goes from full closed to full open. (0.5)

, References SD: NS-G9 PSZR Pressure Control Sys.'

l a

O e

6

4 6.07 (2.0)

As a ' result of a loss of electrical power, what is the failed position for each of the following valves associated with the

. Safety Injection System. (OPEN, CLOSED, or AS-IS)

.a. Low pressure header isolation valve (s) (0.33)

b. HPSI pump mini flow line isolation valve (s) (0.33)

c. Containment sump line isolation valve (s) (0.33)

~

d. LPSI pump suction isolation valve (s) (0.33)

e. Hot-leg injection isolation valve (s) (0.33)

f. SI Tank vent valve (s) (0.33) 6.07 Answer (0.33 for each of the following)

(a) As-is p/, g MA ,

(b) Closed i

O'f Jc (c) Open (d) Closed (e) As-is g

. (f) As-is UM s

t Reference NS-3A, Safety Injection and Shutdown Cooling System,.

page NS-3A-25

t e

l 7

, 3 i

6.08 (2.0)

Draw a one-line diagram of the auxiliary feedwater system that contains the following components. Label the components on -the diagram.

• Essential pump

• Turbine-driven essential pump

• Non-essential pump

• Line providing hydrizene addition

• Line providing ammonia addition

• Water sources for the three pumps

• The eight (8) cross-over valves

• Line from the main feedwater supply

• Valves conten11ed by the FWCS

• MSIS valves

• Steam generators 6.08 Answer See Figure 6.1 (2.0)

Reference:

PGS-11, Auxiliary Feedwater System, pages 2 & 3 i

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4 Palo Verde 1 & 21 March 11,-1986 1

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Fipure 6.1 (ANSWER)

-Section 4.0 Continued on Next Page-9

. . r

(

&;09 (1. O)

L' l Answer' True. or False. to- each of- the fallowing statements

- concerning control of each -atmospheric , dump valve.  ;

" ~

a. A nitrogen gas accumulator ic provided 'to (0.5) ensure operability for at least eight hours fallowing the loss of- the Instrument and Service Air and Nitrogen Gas' systems.

b. One P/I converter is installed to convert the (0.5) pneumatic signal of 3 to 30 psig to an electrical signal, which is used_to control the position of the dump valve.

6.09 Answer (a) True (0.5)

(b) False (0.5) i

Reference:

PGS-1A, Main Steam System, pages 10 & 11 [

l l

i 1

I

$ 1 a r 1

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9

4 6.10 (1.0)

Listed below as "a" .through "d" are components of the Fire Detection and Suppression System for the Emergency Diesel Generators. Also listed below as "1" through "4" are f our possible results from the components exposed to a fire. Match the. correct result with each listed component. Note that one result may occur for more than one component. Listing the letters "a" th' rough "d" with the correct number (s) "1", "2", "3", or "4" is an adequate answer.

a. Sprinkler system fusible link (0.25)

) Fumahle. link on the chain for the rollup fire door (0.25)

c. Ionization smoke detector in the DG local control room (0.*25)

d. Ultra-violet flame detector in the DG engine room (0.25)

1. Activates water flow to the sprinkler heads in the DG engine room.

12. Activatus water flow to the sprinkler heads heated in the silencer room.

3. Activates an alarm indication only.

4. Isolates the DG engine room from the DG local control room.

6.10 Answer (a) 1 (0.25)

(b) 4 ,' (0.25)

(c) 25 (0.25)

(d) 1 (0.25)

Referenco Diesel Generator and Auxiliaries Handbook, pages 7 & B 10

6.11 (1.0)

A Foxborn control station is provided in the control room for positioning the condensate domineralizer bypass valve. Answer 'the -

following parts of this question which refer to the control of the

~

' ,I Cndensate System.

a. When in the " Manual" mode, what color of (0.25) indicating light is illuminated 7

b. When in the " Auto" mode, what signal varies (0.25) valve position 7

c. When in the " Auto" mode the valven position- (0.5) will vary automatically only if three

• conditions are met. State two of these three conditions.

6.11 Answer h p b W o g.7c (a)

(WW red y.)gestf (O.25)

(b) demineralizer differential pressure (0.25)

(c) (0.25 each for any two of the following)

Demineralizer inlet valvo is fully open Demineralizer outlet valve is fully open Demineralizer differential pressure less than 35 psi

Reference:

PGS-9A, Condensate Syste'm, pages 21, "" m, t< 3 0 4

.8 m

O 11

t.

6.12 (1.0) i Four Safety Injection Tanks (SIT) rapidly flood the core following an inadvertent depressurization_ of the Reactor Coolant i System (RCS).

a. How are the tanks isolated from the RCS ? (0.5)

b. How is it ensured that the tanks will not become isolated from the RCS during normal at power operation. (0.5) 6.12 Answer (1.0)

a. Isolated from the RCS by means of a motor operated valve and two check valves. (0.5)

b. Motor Operated valvo is locked open and power to the motor is removed and t2;;rd rut. (0.5)

$h

Reference:

SD: NS-3A Safety Injection

-s f

s O

END DF SECTION 6 CONTINUE TO'SECTIDN 7 12

~

.. s G

a SECTION 7 PROCEDURES: NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL CONTROL 7.01 (3.0) ,f The plant is at'4 % reactor power and the secondary systems are 3

being warmed up per 41DP-12ZO4, Plant Start-up Mode 2 to Mode 1.

All surveillances are complete and systems appear to be operating a normally. Without any warning an MSIS occurs !

Respond to the f ollowing questions as though you wLee following up this MSIS per 41AO-1ZZ31, Inadvertent MSIS.

I

a. Besides the closed indications for MSIVs, (1.0)

FIVs, and S/G blowdown isolation valves there are three alarms and two direct indications

of an MSIS. List four of these alarms and or j indications and their locations. ,

e I b. In order to evaluate wheather a MSIS was (1.0) required or not the procedure directs the -

operator to look at three recorders which j would indicate plant conditions just prior to l the MSIS. List the three recorded plant i

conditions.

c. Having determined that the MSIS should be (1.0) 1 reset the operator must reset the 4 Actuation i Trip Path Channels. Briefly describe how and ,

where this is done. 5 1

l 6.01 Answer l (a) (0.25 each for any four of the following) (1.0)

1) MSIS A channel alarm on the main control board, 2) MSIS B channel alarm on the main

{ control board, 3) MSIS alarm (5CO3A) on the

. main control board, 4) Trip path indicators

! de-energized on the PPS remote operators .

. modules on B05, 5) Trip path indicators de-energized on'the indication pannel above the i PPS cabinets (behind the main control board) j (b) (0.30 each) (1.0)

Steam generator pressure, Containment pressure, and Steam generator level (c) Turn the Actuation Trip path Reset Key and at (1.0) the same timo push the MSIS pushbutton at the i

PPS cabinets.

1

Reference:

41AO-1ZZ31, Inadvertent MSIS, pages 4, 5, & 6.

1 l

. .s 4

7.02 (3.0)

During a normal plant heat-up a bubble is being established in the pressurizer per procedure 410P-1ZZO1, Cold Shutdown to Hot Standby. Heaters are on and the pressurizer will be vented to the Reactor Coolant Drain Tank (RDT) to vent the nitrogen bubble and to maintain pressure control.

a. Che of the precautions for the procedure (1.0) states that while a nitrogen bubble-exists in the pressurizer the pressurizer temperature (not Pressure) should be used to determine the correct level instrument ~ compensation ,

curve (indicated vs actual level). Briefly state why this precaution is necessary.

b. What initial RDT level and pressure (1.0) indications do you expect to see during the venting ?

c. What RDT level and pressure indications do (1.0) you expect to see after all of the nitrogen gas has been vented from the pressurizer 7 7.02 Answer (a) The compensation is necessary due to the (1.0) density change in water (and or steam) in the pressuri:er which changes with temperature.

and/or Density of water in pressuri=er is controlled by pressure when a steam bubble is present, nitrogen pressure has a very small effect on density.

v (b) Initially pressure will rise and level Jwill (1.0) stay the same (constant temperature).

(c) After nitrogen has been vented, pressure will rise slowly,and stabli=e, (and level will. ,

begin to increase, temperature will increase as seen on B03,ERFDADS).

References 410P-1ZZO1, Cold Shutdown to Hot Standby, pages 10 & 63 NS-9G-11, Pressurizer Pressure Control System 2

4 7.03'(2.0) l In the left-hand column below is a list (itemized numerically) of

" symptoms" from .the procedure " Shutdown Dutside Control Room".

The right-hand column is a list (itemized alphabetically) of "causes". Choose one (1) item from the cause list for each item in the symptom list. -Use each item from the cause list for one (1) item f rom the symptom list. (0.4 each) 1 EYUEIQd GBUSE l 1. S/G 1evel less than O % WR A. Condensible voids collecting j Tcold increasing above in RCS flowpath j Tsat of S/G (That follows) ~,

2. Tculd and S/G pressure B. Inadequate RCS inventory-increasing

! 3. RCS loop delta T increasing C. Inadequate secondary steam j and abnormal PLCS response flow I

i 4. That increasing and low D. Non-condensible voids

' pressurizer level collecting in RCS flowpath I i or physical blockage of j RCS flowpaths i

5. RCS loop delta T increasing E. Inadequate secondary l and indication of no delta water inventory

! Press. across the RCP's.

l 7.03 Answer (0.4 each) i ,

1. E .<

I 2. C a 3. A t 4. B

5. D ..

Reference:

41AO-12Z27, Shutdown Outside Control Room, page 50 1

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7.04 (1.0)

According to 41AO-1ZZ13, Natural Circulation Cooldown, which of the following is agt a proper indication of natual circulation 7

m. RCS cold leg temperature at s.aturation for S/G pressyre and is controlable.

b. Core exit thermocouples indication are stable and trending down.

c. QSPDS indicates the RCS is becomming more subcooled.

d. Core delta temperature (That - Tcold) is more than 57 degress F and is increasing.

7.04 Answer D. (1.0)

Reference:

41AO-1ZZ13, Natural Circulation Cooldown, page 4 7.05 (1.0)

During natural circulation with the potential for void formation in the RCS. The Natural Circulation Cooldown procedure 41AO-1ZZ13 cautions against the use of the pressurizer Invel control system in the automatic mode. Why would operation of the pressurizer level control system in " automatic" in this situation be a problem 7 ,

r 7.05 Answer (1.0)

If a void formed in the RCS, letdown will automatically increase ,

in response to pressurizer level thorby depleting RCS inventory. l

Reference:

41AO-12213, Natural Circulation Cooldown, page 6 e

1 4

5

. . . . . -~

~

11 p-7

4. k' 7.06 (3.0)

a. List the' administrative control limits for (1.0) whole body exposure.

Weekly _ mrem .

Quarterly rem Yearly _ rem

' ~

b. Which two individuals, by titir, imust approve (1.0) a request for exceeding the normal quarterly administrative control limits ?

c. What is the quarterly exposJre lidit for (1.0) individuals who. do not have their complete lifetime exposure history and who have received undocumented radiation exposure. in the present quarter?

,)

arem

7.06 Answer

• 1 (a) (0.33 each) (1.0)

Weekly 300 mrem Quarterly 1.0 rem Yearly 4.0 rem l 1

l (b) Radiation Protection Supervisor and -the (1.0)

, Radiological Services Manager l

(c) 900 mrem $ (1.0)

/

Reference:

75AC-9ZZO1, Radiation Exposure and Access Control, page 25 f.

L .

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5

i 7.07 (3.0)

A normal reactor start-up is being conducted in accordance with 410P-1ZZO3, Reactor Start-up.

a. Verification of proper overlap between . (1.0) startup and log safety NI channels is required prior to turning off the high voltage to the startup channels. Over what range of each instrument should overlap occur 7 (CPS and % power)

b. When is the high voltage to the startup (0.5) channels turned off ?

c. Why is the high voltage turned off 7 (0.5)

d. Below the point of adding heat what (1.0) indication do you expect when the reactor becomes critical ?

e. If the reactor is critical how can you tell (1.0) you are at the point of adding heat 7 7.07 Answer (a) 20 to 2000 CPS = 2x10-8 to 2x10-6 % power (1.0)

(b) When the high CPS alarm is recieved (Approx. (0.5) 2OOOCPS) i (c) To prolong the life of the startup detectors (0.5) l (d) Criticality is indicated when a controls. rod (1.0)  !

withdrawl results in a sustained positive startup rate which maintains after rod motion has stopped.

(e) The reactor is adding heat when the effects , (1.0) of moderator feedback become evident and/or when reactivity addition results in a positive startup rate which slowly tapers off. (about 10-3 % power)

Reference:

410P-1ZZO3, Reactor Start-up, pages 5, 8, 13, and 15 6

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i 7.08 (2.0)

Indicate .whether each of the following statements are TRUE or '

FALSE as they pertain to the Reactor Start-up procedure.

a. If a required boron concentration change is -

(0.5) greater than 50 ppm then pressurizer. spray should be initiated to equalize pressurizer and RCS boron concentrations within 20 ppm.

b. After criticality is achieved, reactor power (0.5) ,

is increased at a rate of 1 DPM to 10-4 %  ;

power to log data.  ;

c. If group 5 has been withdrawn.to 135 inches (0.5) ,

and the reactor is not critical, insert group 5 a sufficient amount (about 500 pcm) to ensure the reactor remains subtritical, then dilute the RCS boron concentration to make up for rod insertion prior to continuing the startup.

d. If the reactor is not critical upon reaching (0.5) _

ECC + 500 pcm the regulating rod should be ..

inserted to ECC - 500 and Reacto Engineering should be contacted.

7.08 Answer (0.5 each)

(a) False (b) True s (c) True f 800 4 @

(d) False .

Reference:

410P-1ZZO3, Reactor Startup, page 10, 13, 14, and 15 e

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7.09 (4.0)

While the plant is operating at steady state power a minor steam generator tube leak occurs.

(a) List four of the six symptoms of the steam , (1.0) generator tube leak AO procedure.

(b) Briefly, -how you can determine the size of (1.0) the leak 7 (c) If it is not obvious how would you determine (1.0) which steam generator is leaking.

(d) What are two of the three reasons to_ keep the (1.0) steam generator 100 psi below RCS pressure 7 7.09 Answer (a) (0.25 each for any four of the following) (1.0)

Main steam line high radiation Condenser off-gas high radiation S/G blowdown high radiation Charging / letdown flow mismatch Decreasing VCT level Secondary side sample activity high (b) With the plant stablized: Measure the VCT (1.0) level change over a known time to calculate the leak rate.

(c) The S/G with the highest activity is the (1.0) leaker.

(d) (0.5 each for any two of the following) ) (1.0)

Prevent baron dilution Prevent S/G chemicals in the RCS Ensure adequate RCS subcooling

Reference:

41AO-1ZZOB, Steam Generator Tube Leak, pages 4, 5, 7, and 9

)

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7.10 (1.0)

Following a LOCA-plant conditions appear under control, SI flow has been stopped, and SIAS has been reset.

What two conditions would require re-initiation of full'SI. flow 7 7.10 Answer

a. Subcooling is lost (0.5)

b. Pressurizer level is lost (0.5)

Reference:

41CP-1ZZO1, Emergency Operations, page 88 7.11 (1.0)

Which gne of the following is ngt one of the four criteria that must be met before safety injection flow may be throttled 7

a. Reactor vessel level indicates void restricted to upper head

b. RCS subcooled more than 28 F

c. Core delta T is less than 10 F 2

d. Pressurizer Invel greater than 33  %, and controllable. 3 7.11 Answer

(,,. (1.0)

Reference:

41RO-1ZZO1, Reactor Trip,.page 27 4

9

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7.12 (1.0)

Match the following (numerical) emergency classifications with the appropriate (alphabetical) description. (An answer consisting of the numbers "1" through "4" and the qcorrect associated letters is adequate.) . .

1. Unusual Event

2. Alert

3. Site Area Emergency

4. General Emergency (a) Consists of events which are in progress or have occured which involve actual or likely major failures of plant functions needed for the protection of the public. Any releases are not expected to exceed Environmental Protection Agency Guideline expos _tre levels beyond ths site boundary.

(b) This classification applies to events which are in progress or have occured which indicate a potential degradation of the level of safety of the plant. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.

(c) Consists of events which are in progress or have occured which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity.'

Releases can be reasonably expected to exceed Environmental Protection Agency Protective Action Guideline exposure levels offsite for more than the immediate site area.

(d) This classification consists of events which are in progress or have occured which involve an actual or potential substantial degradation of the level of safety of the plant.

Any releases are expected to be limited to small fractions of the Environmental Protection . Agency Protective' Action Guideline exposure levels.

7.12 Answer (0.25 each)

1. B

2. D

3. A

4. C

Reference:

EP1P-02, Emergency Classification, page 5 END OF SECTION 7 CONTINUE TO SECTION 8 10

. a y_

f SECTION 8 ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 8.01 (2.C)

In accordance with 10 CFR 20, " Standards for Protection Against Radiation":

a. What is a RADIATION AREA ? (1.0)

b. What is a HIGH RADIATION AREA 7 O (1.')

8.01 Answer (2.0)

a. An area (accessible to personnel) where a major poriton of the body could receive (greater than):

5 Mrem in one hour (0.5) or 100 Mrem in five (consecutive) days. (0.5)

b. An area (accessible to personnel) where a major portion of the body could receive (greater than):

100 Mrem in one hour. (1.0)

Reference:

10 CFR 20. 203 (b) (2) and (b) (3)

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

. .- . 3-(

8.02 (3.0)

In accordance with 10 CFR 55, " Operators' Licenses":

a. As -definid in 10 CFR 55, when is an individual '

deemed to be operating the controls of a nuclear facility ? (1.0) '

b. What are the " controls" as defined in 10 CFR 55 ? (1.0)

c. According to the " Exemptions from License" provisions of 10 CFR 55 which individuals are allowed to operate the the reactor controls without a license ? (1.0) ,

8.02 Answer (3.0)

a. An individual is deemedEto operate the controls of 4 a nuclear facility if he directly manipulates the controls or directs another to manipulate the controls. (1.0)  ;

b. " Controls" means apparatus and mechanisms the manipulation of which directly affect the reactivity or power level of the reactor. (1.0)

c. An individual may manipulate the controls as a .;

part of his training to qualify for an operator  !

license under the direction _and in the presence of.

a licensed operator or senior operator.- (1.0) l

> ^i

Reference:

10 CFR 55.4(d) t< (f) and 55.9 (b) i

< l

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s.03 (3.0)

Refer to 10 CFR 50.72 and or 10 CFR-20.403 ( ATTACHMENT). Which of the following events require iomediate notification (within a period of one hour or sooner) to the NRC Operations Center .via

'- the Emerge.1cy Notification System ? (0.5 each) ,

a. Personnel exposure to an. individuals hands of 200 Rems of radiation while performing steam-generator tube repairs.

b. Personnel exposure to an individual's whole body of 30 Rem of radiation while working in -a steam generator.

c.- Declaration of an " Unusual Event" at PVNGS.

d. Taking the plant from mode 1 to mode 3 to comply with Technical Specification requirements.

e. Taking the plant from mode 3 to mode 5 to comply with Technical Specification requirements.

f. Discovery in mode 1 that two safety injection level instruments were improperly calibrated and that the levels have been above or below Technical Specification limits for two weeks.

] 8.03 Answer (3.0)

• B, C, D, and F require notification within one 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. (0.5 each)

• • A and E do not require notification within; one hour. (0.5 each) y,

Reference:

10 CFR-20.403(a) and 50.72(a)(1)i & 50.72(b)-

l I

i l.

3

_. . _ , _ - . _ _ - . - - . . - _ --. . i ' _ _. _, . - _. _ _

- ....m.....t...m.....,

i 8 50.72-

" s b ] : 1n er R:gulatory Co,nenission 8 i l $3 B5 y a r t 2 d(g ,

ti:n must be made as follows: der l 50.21(b) er i 50.22 of this part ( A) In en unanalyzed to compromia.c.V plants

[that -

shall notify the NRC Operations significantly i

5 28.seJ ; Notificatione er incidents. (1) Licensecs that have an installed .

"' " UU in ys em shall Center via the Emergency Notification safety; (a) immedsafe nof(/scaffon. Each li. make 1 n ut ed by para. System of

(B) In a condition that 1: outside the -

censee shall immediately report anY h 8'C D (i) The declaration of any of the design basis of the plant; or esents insolving byproduct, source, or eN e rd- Emergency Classes specified in the li. (C) In a condition not covered by the 4

specill nuclear material possessed by ance with 5 7 o h h[In a #" * "*"*""*" U'"E'"'I E

1. $', that may have caused or (2) All other licensees shal a'ke the " #'(lg"[l[Anynatural (11) Of those non-Emergency events phenomenon or (1) Exposure of'the whole body of (b of hi eti y be sphs(a)and specified in paragraph (b) of this sec. other external condition that poses an

, gny individual to 25 rems or more of telegram. mallgram, or fa simi e to t 11 n. actual threat to the safety of Ilic nu.

r

dictirn; erposure of the skin of the Administrator of th (2) H th Emmency Nomeadon clear power plant or significantly ham.

vthole body of any individual of 150 Regional Office liste in A pend x D System is inoperative, the licensee pers site personnel in the performance j rems er more or radiation; or exposure of this part* shall make the required notifications of dulles necessary for the safe oper.

, af thz feet, ankles, hands or forearms _. via commercial telephone service, ation of the plant.

Cf cny individual to 375 rems or more other dedicated telephone system, or tiv) Any event that results or should

, sf rtdlition;or any ofher method which will ensure have resulted in Emergency Core j (2)The release of radioactive materi. that a report is made as soon as practi. ! Cooling System (ECCS) discharge into j clin concentrations which,if averaged cal to the NRC Operations Center.'  ; the reactor coolant system as a rtsult

1. eser a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. would (3) The licensee shall notify the of a valid signal-

exceed 5.000 times the limits specified NRC immediately after notification of (v) Any event that results in a major

frr such materials in Appendix B' the appropriate State or local agencies Tchia If of this part;or and not later than one hour after the loss ity, of emergency offsite responseassessmentcapableity, orcapabil.

com-(3) A loss of one working week or time the licensee declares one of the munications capability ( e.g., signili.

m:re cf the operation of any facilities Emergency Classes. cant portion of control room indica.

. (iffected; or (4) When making a report under tion. Emergency Notification System. ,

" (4) Damage to property in exc'ess of paragraph (ax3) of this section, the 11 or ofIsite notification system).

{ $200.000. censee shall identify: tvil Any event that poses an actual

) (b) Tioenfy-four hour noit/icaffon (1) The Emergency Class declared; or threat to the safety of the nuclear Each licensee shall within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of (11) Either paragraph (bMI), "One. power plant or significantly hampers '

i dise:vtry of the event, report any Hour Report " or paragraph (bx2),

event involving licensed material pos. "Four.Ilour Report." as the paragraph site dutiespersonnel necessary it; the performance for the of -

safe operation  ?'

4 sessed by the licensee that may have of ihls section requiring notification of of the nuclear power plant including j c:used er threatens to cause; the Non #mergency Event.

l . (1) Exposure of the whole body of (b) Non-cmergency evenis-(1) One. fires, toxic gas releases, or radioacthe 1 ' gny individual to 5 rems or more of ra. hour reports. If not reported as a dec. releases. (2) Four hour :poris. If not rt; port.

. dillion; exposure of the skin of the laration of an Emergency Class under ed under paragraphs (a) or (bMit of

. while body of any individual to 30 paragraph (a) of this section, the 11 i rems or more of radiation; or exposure censee shall notify the NRC as soon as this section, the licensee shall notify 4 Cf th2 feet, ankles, hands, or forearms the NRC as soon as practical and in all

. Wi$ r:ms or more of radiation; or practical hour of the occurrence of any of the rencewithin and in all cases within one cases. four hours of the occur.

(2) Th2 release of radioactive matert. following

of any of the following:

tl in concentrations which. if averaged tim A) The initiation of any nuclear (i) Any event, found while the rene.

t .. cvtr o period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, would *', plant shutdown required by the forfound is shut down. that, had it been while the reactor was in oper.

thceed 500 times the limits speelfled plant's Technical Specifications.

} for such materials in Appendix B* (D) Any deviation from the plant's ation, would have resulted in the nu.

] T;bl2 II of this part; or Technical Speel!! cations authorized c! car power plant, including its princi-(3) A loss of one day or more 6f the pursuant to i 50.54(x) of this .;;.-* pal safety barriers..being seriously dc.

t 4 operr. tion of any facilities affected; or !I 5n32 Immediate notificati"" 8" ire- (11) Any event or condition durina arraded ,.- being in an unanalyzed con-j $2.000.

(4) Damage to property in excess of meats for oPcrating nuclear power re. operation that results in the condition ditiv.. that significantly compromkes I g of the nuclear powerplant, including . plant safety. .

]

(c) Any report filed with the Com.

8 its principal safety barriers, being seri- (ii) Any event or condition that re.

nW requiremenfs.' (I) Each j mission pursuant to this section shall

• nu(r)nr power reactor licensee licensed ously degraded; or results in the miele. sults in manual or automatic actuation

be prepared so that names of Individ. ar power plant being

of any Engineered Safety Featurc u;is who have received exposure to ra. (ESF). Including the Reactor Protre-j di; tion will be stated in a separate 'Other requirements for immediat ~

tion System (RPS). Ilowetcr. actu-

!-- partif the report. ition of the NRC hy hcensed operatir e e. 8These Emcreeney Classes are addressed ation of an ESP. Including the ItPS. ,

', , car rmurr rractors are rnntained the- in.Appendtu M of this part.commerciai icicohone that results number from or the and is part of the i

'"' '" this chapter. in pa, t ituta,*

N RC Operat tons Center is (202) 951 0550. preplanned sequence dueinn testing or .

i 256 "20 2n5. 20 403. 50 36. and 73.73. ' '

q ,

l 457 t

- - _ _ . , - ,..m_.,__---__._-__ _ _ ,_ _ , - _,,_._ - _ _ _ _ _--._._ _ . _ _ - _ _ - - - _ _ _ _ _ _ _ _ _ _

. . )

I 8.04 (1.0) 10 CFR 50.54(x) allows a licensee to take reasonable action that departs. from a license condition or Technical Specification in an emergency when this action is immediatly needed to protect the public health and safety and no action consistent with -license conditions and Technical Specifications that can provide adequate or equavilent protection is immediatly apparent.

Licensee action permitted by this paragraph shall be approved, as a minimum, by _ prior to taking the action.

Which response best completes the above statement 7

a. the senior licensed operator present (RO)

b. the Plant Superintendent

c. any licensed control operator (RO)

d. any licensed Senior Reactor Operator (SRO) 8.04 Answer (1.0) d

Reference:

10 CFR 50.54 (x) and (y) e i

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8.05'(2.0)

Answer ~each of the following TRUE or FALSE:

a. UNIDENTIFIED LEAKAGE shall'be all leakage which does not constitute either IDENTIFIED LEAKAGE or .

reactor coolant pump controlled bleed-of f flow. (0.5)

b. The local DNB ratio (DNBR), is defined as the ratio of the actual heat flux at a particular core location to that of the predicted DNB heat flux at ,

same location. (0.5) l i c. The peak linear heat rate (adjusted for fuel rod :l dynamics) of the fuel shall be maintained'less l d' than or equal to 21 kw/f t. - (0.5)

d. The Steam Generator Low Level trip in the RPS is j

- is designed to protect the Reactor Coolant System

.from overpressure. (0.5) l 8.05 Answer (2.0) i i

j a. True j

b. False  ;

)

i c. True

d. True ,

Reference:

Tech. Spec's 1.0.2.1.1, 2.1.1.2, and-2.2.1 ,

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8.06 (1.0)

In accordance' with PVNGS station tagging and clearance requirements, which one of the following statements is inggttest 7

a. More than one (red) dangsr tag may be attached to one device at the same time,

b. More .than one (blue) men at work tag may be attached to one device at the same time.

c. A (yellow) caution tag and a (red) danger tag may be attached to one device at the same time.

d. A (blue) men at work tag may not hang on one-device at the same time with a (r,ed ) danger tag.

8.06 Answer (1.0) b, is the only incorrect statement (1.0)

Reference:

PVNGS Procedure 40AC-9ZZ15, Station Tagging and Clearance, page 19 8.07 (1.0)

a. What is the maximum time a RCP can be operated after a loss of nuclear cooling water, with seal injection flow established? (0.5)

b. What period of time may a RCP be operated-without.

seal injection flow, assuming nuclear' cooling water is still. supplied? (0.5) 8.07 Answer (1.0)

a. Ten (10) minutes (0.5)

b. An indefinite period of time (0.5)

Reherence: 410P-1CHO3 RCP Seal Injection System.

6

. .8 .

(

8.08 (3.0)

.The PVNGS Technical Specification require.that shutdown margin be greater than 6 % while in modes 1.through 4 and greater than 4 %

while in mode 5.

t .

a. What.is the definition of' shutdown margin 7 (1.0)~

b. What are the most restrictive accident conditions (ie. Technical Specification Bases) which require the 6 %.SDM limit 7 (1.0)

c. What ard"the'most restrictive accident conditions (ie. Technical Specification Bases) which require the 4 % SDM limit 7 (1.0) 8.08 Answer (3.0)

a. SDM is the amount of reactivity by which the reactor is, or would be, subtritical from its present condition assuming no change in part-length CEA position and all other CEAs are fully inserted except the single assembly of highest worth which is fully withdrawn. (1.0)

b. The 6 % SDM limit is based on controling the reactivity transient associated' with an uncontrolled RCS cooldown caused by a steam line break at end of life and T(c) at no load operating temperature. (1.0)

> c. The 4 % SDM limit is based on ensuring that reactivity transients resulting from a single CEA withdrawl event are minimal. (1.0) v v

Reference:

l PVNGS Technical Specification, Definition 1.28, and Bases'3/4.1.1 l l

l l

l 1

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. 1 7

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f 8.09 (2.0)

Based on. the PVNGS Technical Specifications, with regard to Minimum Temperature for Criticality, with the plant in mode 1:

a. What is the minimum operating loop temperature +

(1.0) ti.e. LCD for T(c)] 7

b. In the event T(c) decreased below its LCD what actions required by the Technical Specification Action statement (s) must be taken (including time limits) within one hour 7 (1.0)

A :-;8.09 Answer (2.O)

a. 552 degrees F (1.0)

b. Restore T(c) to meet the LCO in 15 minutes (0.5) or be in hot shutdown in the following 15 minutes. (0.5)

Reference:

PVNGS Technical Specifications, 3.1.1.4 l

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B.10 (2.O)

Complete the minimum shift crew composition for the modes listed

-below. (Do not consider health physics, chemistry or IEC personnel.)

PERSONNEL MODE 4 (1.0) MODE 5 (1.0)

SROs _ _ _ ___

. ROs _ ___

Others(AO's, ___ _

STA*s) i 8.10 Answer (2.0)

PERSONNEL MODE 4 (1.0) MODE 5 (1.0)

SROs 2 -1 ROs 2 1 Others(specify (AO), (STA) 1h0) i positions) f i

Reference:

PVNGS Technical Specifications, Table 6.2-1 1 Y

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B.11 (2.0)

During the absence of the Shift Supervisor from the control. room

. for mgdes 1 through 6 , who may be designated to ar,sume the control room command function ? - .

8.11 Answer (2.0)

• When in modes 1 through 4 an individual, other (1.0) than the STA, with a SRO license may assume the Control Room command function.

• • While in modes 5 and 6 an individual with a RO or (1.0)

SRO license may assume the Control Room command function.

Reference:

PVNGS Technical Specifications Table 6.2-1 4

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I B.12 (3.0) l i

The plant.has been operating at 90 % power for about three weeks.

The Non-Essential

• auxiliary feed pump (AFN-PO1) was placed out of service the previous day due to a bad bearing. While testing the (GOO 1) Diesel Generator, the Di.esel Governor fails to control frequency and the Diesel must be shutdown for' repairs. The associated' Technical Specifications are attached- for. your reference and the time of various events are. listed below.

1100 14 November auxiliary feedpump placed out of service 1000 15 November diesel generator placed out of service 1015 15 November current time

a. What surveillances or verifications must be completed within the next two hours 7 (1.0) i b. When must the plant be placed in hot standby if neither the diesel nor the auxiliary feed pump  :

can be returned to service 7 (1.0) l

c. If the steam dr:amn auxiliary feedpump were found inoperable at 1130 15 November when would the plant have to be in hot standby ? (1.0) 8.12 Answer (3.0)

(a) Each independent offsite power supply must be (.25) checked operable. .

The remaining diesel generator must be checked (.25) operable.

All required systems, subsystems, trains, (.25) components, and devices that depend on ( the '

remaining diesel generator for power must be checked for operability.

The steam driven auxiliary feed pump must be (.25) ,

operable.

(b) 1700 17 November (1.0)

(c) 1730 15 November (1.0)

Reference:

PVNGS Technical Specifications,.3.7.1.2 and 3.8.1.1

. [

END OF THE' EXAM l

4 11

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,uuT SvSEOR 1NFORMATION'ONLY AUXILIARY FEEDWATER SYSTEM

~

LIMITING CONDITION FOR OPERATION s

3.7.1.2 At least three independent steam generator auxiliary feeddater pumps .

and associated flow paths shall be OPERABLE with:

..a. Two fe dwater pumps, each capable of being powered from separate

• OPERABLE emergency busses, and

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

APPLICABILITY:' MODES 1, 2, 3, and 4*.

l ACTION:

a.

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

With two auxiliary feedwater pumps inoperable be in at least HOT STANDBY 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following

)((

4 c.

With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to DPERABL' status as soon as possible.

SURVEILLANCE REOUIREMENTS 4.7.1.2 E.ach auxiliary. feedwater* pump shall ,be demonstrated OPERABLE:

a.

At least once per 31 days on a STAGGERED TEST BAS'IS by: -

1.

- Testing the turbine-driven pump and both motor-driven pumps pursuant to Specification 4.0.5. The provisions of Specifica-

- tion 4.0.4 entry into MODE 3.

are'not applicable for the turbine-driven pump for 2.

Verifying that each valve (manual, power operated, or automatic) in the flow path that is not locked, sealed, o'r otherwise secured in position, is in its correct position.

3.

Verifying that all manual valves in the suction lines from the primary AFW supply tank (condensate storage tank CTE-T01) to ,

each essential AFW pump, and the manual discharge line valve of *

\ each AFW pump are locked, sealed or otherwise secured in the .

open position.

• Until the steam generators are no longer required for heat removal.

PALO VERDE - UNIT 1 3/4 7-4 o

.....,? .

FORJNFORMATION'O.NLY e PLANT SYSTEMS F 1

.\ s

.b sN SURVEILLANCE REOUIRFMENTS (Continued)

\

b. At least once per 18 months during shutdown by:

1. Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an auxiliary feedwater actuation test signal.

2. Verifying that each pump that starts automatically upon receipt -

, of an auxiliary feedwater actuation test signal will start

.' automatically upon receipt of an auxiliary feedwater actuatio,n test signal.

c. Prior to startup following any refueling shutdown or cold shutdown of 30 days or longer, by verifying on a STAGGERED TEST _ BASIS (by means of a flow test) that the normal flow path from the condensate storage tank to each of the steam generators through one of the essential auxiliary feedwater pumps delivers at least 750 gpm at 1270 psia or eqpivalent.

d. Th'e provisions of Specification 4.0.4 are not applicaole for entry into MODE 3 or MODE 4 for the turbine-driven pump.

\

f-Q

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,' " PALO VERDE - IINIT 1 ~4/A 7 8i

.. s j 3/4.8 EL _

.I

/ 3/4.8.1 A.C. SOURCES OPERATING '-

LIMITING CONDITION FOR OPERATION i

3. 8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

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 IE distribution system, and

b. Two separate and independent diesel g,enerators, each with:

1. '

Separate day fuel tank with a minimum level of 2.75 feet

(550 gallons of fuel), and ,

.._ 2. A separate fuel storage system'with a minimum level of 80%

(71,500 galJons of fuel), and

3. A separate fuel transfer pump.

APPLICABILITY: MODI'S 1, 2, 3, and 4.

ACTION:

. a. With either an offsite circuit or diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.la. and 4.8.1.1.2a.4 within I hour and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter; restore at least two offsite circuits and two diesel generators to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

- or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD

. SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />,

b. With one offsite circuit and one diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance

.Requizaments 4.8.1.1.la. and 4.8.-121.2a.4. within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter; restore at least one of the inoperable sources to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Restore at least two offsite circuits and two diesel generators to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from the time of initial lo:s or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c. With one diesel generator inoperable in addition to ACTION a. or b.

above, verify that:

1.

- All required systems, subsystems, trains, components, and devices that depend on the remaining OPERABLE diesel generator as a

• source of emergency power are also OPERABLE, and

2. When i.n MODE 1, 2, 3, or 4*, the steam-driven auxiliary feed pump is OPERABLE.

- If't'hese conditions are not satisfied within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, be in at least

- HOT STANDBY following within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

~

"Until the steam generator is no longer required for heat removal.

PALO VERDE - UNIT 1 3/4 8-1

F m - -- - - - ----r--J

.. . .q s ELECTRICpLQ{e jYSTEMSRFDMATION 05RY

[ LIMITING CONDITION FOR OPERATION (Continued) s ..

fu A

ACTION (Continued) d.

With two of the above required offsite A.C. circuits inoperable, demonstrate the OPERABILITY of two diesel generators by performing Surveillance Requirement 4.8.1.1.2a.4. within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter, unless the diesel generators are already operating;-restore at least one of the inoperable offsite sources to OPERABLE the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT STANDBY w '

With 'only one offsite source restored, restore at least two offsite circuits to.0PERABLE sta)us within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from time of initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

e.

With two of the above required diesel generators inoperable, demonstrate the OPERABILITY of two offsite A.C. circuits by performing Surveillance Requirement 4.8.1.1.la. within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter; restore at least one of the inoperable diesel generators to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and,in COLD SHUTDOWN

' within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Restore at-least two diesel generators to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from time of initial loss or be in at least within HOT STANDBY the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHU

.

• I SURVEILLANCE REOUIREMENTS ,

. l 4.8.1.1.1 Each of the above required independent circuit's between the offsite transmission network and the onsite Class 1E distribution system shall be: {

a.

Determined OPERABLE at least once per 7 days by verifying correct breaker alignments, indicated power availablity, and .

1 b.. '

Demonstrated OPERABLE at least once per 18 months during shutdownI transferring (manually) unit power supply from the normal circuit to the alternate circuit. l' 4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:

. a.

In accordance with the frequency specified in Table 4.8-1 on a STAGGERED TEST BASIS by: .

1.

Verifying the fuel level in the day tank, 2.

- ~~

Verifying the fuel level in the fuel storage tank,

3. '

Verifying the fuel transfer pump can be started and transfers I fuel from the storage system to the day tank, 6

, PALO VERDE - UNIT 1 3/4 8-2

.(s FOR'INEORNATION ONW ELECTRICAL POWER SYSTEMS i

SURVEILLANCE REQUIREMENTS (Continued)

• c) A flash point equal to or greater than 125*F, and d)

A clear and bright appearance with proper color when tested in accordance with ASTM D4176-82.

2. By verifying within 31 days of obtaining the sample that the other properties specified in Table 1 of ASTM D975-81 are met )

when tested in accordance with ASTM D975-81 except that the j analysis fon sulfur may be performed in accordance with ASTM .

. D1552-79 or ASTM D2622-82.

d. At least once every 31 days by obtaining a sample of fuel oil from the storage tanks in accordance with ASTM D2276-78, and verifying that total particulate contamination is less than 10 mg/ liter when checked in accordance with ASTM D2276-78, Method A. -

o

(

O .

, p. -

e J

o 9 o .

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

y p. Tjk Ujik

.y

,j j .

• 1 U.S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION Facility: Palo Verde _1 and_2 (50-528 & 529)

Reactor Type: PWR-cE-Date Administered: March 11. 1986 j Examiner: Joe Unton._ John Smith

~

Candidate: Answer Kev _

1

> INSTRUCTIONS TO CANDIDATE

' Print your name on the line above marked " Candidate." The grade points available for each question are indicated within parentheses after.each ques-tion. The passing grade is 70% in each of the four (4) . categories and is 80%

, for the total grade. Use separate paper for your answers and write on only-

one (1) side of the paper, unless a specific question instructs you otherwise.

Staple this question package to your answer sheets. The examination questions and answers will be picked up six (6) hours after the examination was started.

Read the statement at the bottom of this page. When you have finished this  :

examination, affirm the statement by signing your name.

l l Category  % of Candidate's . % of

j. Value Total. score cat. Value- Cateanrv 25 25 1. Principles of Nuclear .,

j ' Power Plant Operation, j

)

Thermodynamies, Heat ';

j ., Transfer and. Fluid Flow l

1 25- 25 2. Plant Design Including- . 1 Safety and Emergency _;

Systems.

25 25 3. Instruments and Controls  !

1 25 25 4. . Procedures - Normal,

. Abnormal, Emergency and  !

Radiological Control 100 TOTALS- ,

p  :

Final Grade  %

• i  !

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

aid.

J

)

Candidate's Signature -

i 5

t

t. , _ _ _ _ , , . . . . . .--...u.._ _ _ _ -_ -.;_,_._ _ _ - . . _ _ . _ . _ _ . , . ---.. __

b egg - .

FURTHER INSTRUCTIONS TO CANDIDATE ~

1. At the end of the written examination package.is a copy of a part of the Core Data Book for Unit-1 in Cycle 1. Use the tables and curves as. .

1 appropriate. ,.

2. At the end of the written examination package is a reference page containing equations, formulas, and constants. Use them as necessary.--

3. Use the " Steam Tables" as necessary.

i 4

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l i

4 l

'N _f .

.s; ,,

Page 1 'Palo Verde 1 and 2

March 11, 1986 Points Available- ,

a

1.0 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. THERMODYNAMICS.

. HEAT TRANSFER AND FLUID FLOW _(25.0)

LQUESTIONS 1.01 through 1.06 are " multiple-choice" questions. For i each question soecify the letter designation for the phrase that 1

provides the most correct statement.

QUESTION 1.01 A tank contains water to a level of 40 ft above the. bottom of.

the tank. A nitrogen cover gas is at 100 psia. The tank and its contents are at 70'F and the density of the water is 62.4 lbm/fts. The pressure at the bottom of the tank -is- (1.0) j (a.) 117 psia.

3

(b.) 132 psia. -

r'

,,5-l (c.) 208 psia. , /i -

i

(#

~

(d.)308 psia. .

+

a ANSWER 1.01 (a.) [+1.0] -

l  :

! Reference (s) 1.01 1 1. Palo Verde 1 and 2: Thermohydraulics Review for Licensed j onerators, Section 2, pp. 6-9.

i l 2. Generic: Academic Program for Nuclear Power Plant

! Personnel, Volume III, " Nuclear Power Plant Technology",

1973, General Physics Corporation, pp. 2-115 117.

3. St. Lucie 1 and 2: Thermodynamics and Heat Transfer -

4 Module 4, " Thermodynamics", pp.-3-11.

i l -Section 1.0 Continued on Next Page-

-) 'g Page 2 ' Pale. Verde 1 and 2 MarchEll, 1986

' Points Available

[

OUESTION 1.02

{; If.the temperature of the tank in QUESTION 1.01 (container, water and cover gas) were raised and if no water or cover gas was allowed to enter or leave the. tank,'the pressure at the bottom of the tank would -(1.0).

~

(a.). increase because the. gas volume has decreased and-the tem-l perature of the gas has increased.

! .(b.) increase because the pressure due to the water would rise t

and the temperature of the gas has increased.

(c.) decrease because the water density has decreased.

(d.) decrease because the cover-gas density has decreased. i l

7 J ANSWER 1.02 1 l I (a.) [+1.0] .

i. '

Reference (s) 1.02

1. Generic: Academic Program for Nuclear Power Plant Personnel, Volume III, " Nuclear Power Plant Technology", -

! 1973, General Physics Corporation, pp. '2 2-10, 2.115 .'

i j

4

- 2.117.

l l 2. St. Lucie 1 and 2:' Thermodynamics and Heat Transfer -

Module 4, " Thermodynamics", pp. 3-11, 38-40.  !

i  :

, l

! - .1 i  :

i

-Section 1.0 Continued on Next Page-  ;

a

. . - ,.L. m. -_. .. . . . . . . _ , - - . _ - _ . - , _ , . . , _ _ . - _ . - . . . , , . _ . . . , . .-_..s. ., , _ . _ , . /,..,,.,,.

~-

Page 3 Palo Verde 1 and 2 March 11, 1986 4 Points Available .

QUESTION 1.03 Delayed neutrons arc emitted l (1.0)

[ (a.) through the radioactive-decay chain of. various fission-fragment nuclei and comprise less than 1% of the total number of neutrons released in the fission process.

(b.) through the radioactive-decay chain of various fission-fragment nuclei and appear, on the average, with more energy than the average energy associated with the neutrons pro-duced immediately in the fission process.

(c.) by delayed-neutron precursor nuclei and appear, on the average, with less energy than the average energy associated with the neutrons produced immediately in the fission process.

(d.) through the radioactive-decay chain of various fission-fragment nuclei and comprise between 6 to 7% of the total number of neutrons released in the fission process.

l ANSWER 1.03 l

l (a.) [+ 1.0]

Reference (sl 1.03

1. Palo Verde 1 and 2: Palo Verde Licensed Operator Training, Reactor Theory, Pre-License Review Course NLC55, pp. 4.6, 5.2, and Figure 4.12.

2. St. Lucie 1 and 2: Reactor Physics Training % nual,

" Reactor Physics", pp. 7.2 7.2-8.

-Section 1.0 Continued on Next Page-

+--

Vl.

m+ ,. .

.Page.4-Palo Verde 1 an'd'2-March-11, 1986-

- Pof n' ts -

Available OUESTION 1.04 '

'_ J)p; Upon shutting down a reactor from full power by the. instantaneous'- .

insertion of negative. reactivity, it is impossible to obtain -

'an arbitrarily'small magnitude for the stable reactor period (the decay period several minutes after the.. insertion of the-negative. reactivity) by inserting a larger amount of: negative reactivity because (1.0)

(a.) .the decay rate associated with the reduction of the temperature of the. fuel determines the long-term decay.

rate for the. neutron flux.

(b.) most of the neutrons are " prompt" and h'ence the prompt

' neutron diffusion time determines the long-term decay ~

rate .for the neutron flux. .

(c.) the delayed neutron emitters with the shortest. half-life

• determine the long-tem decay rate for the neutron flux.

(d.) the delayed neutron emitters with the longest half-life determine the long-term decay rate for the neutron flux., .

i ANSWER 1.04 '~

s r .(d.) [+1.0] 'i -

Reference (s) 1.04

1. 7Palo Verde 1.and 2: Palo Verde Licensed Operator Training, Rea6 tor Theory, . Pre-License Review Course NLC55, p.11.5 and Figure 11.5.

2. GA: Svilabus & Triga Traint4_..'mual, " Reactor Kinetics" .

GA Technologies Inc., p. Sit, T

l

,4 ,

.1 5

'f', -Sectiond,.0, Continued on Next Page-

E

.,9' ,

f'

.g

6 e-Page 5 Pal'o Verde 1 and 2 March 11, 1986 Points Available: ._

t OUESTION 1.05 The nuclear reactors at Palo Verde are called " thermal reactors" because

-(1.0)

(a.) the reactors provide the thermal-energy input for the plant.

(b.) the'themal power produced by the reactors equals' the power removed by the secondary-side of the plant (when.in the normal full-power operation).

(c.) on the average the neutrons produced by the. fissioning process are at an energy level that corresponds to the temperature of the surrounding material.

-(d.) on the average, the neutrons causing fission are at an energy level that corresponds to the temperature of the surrounding materials.

ANSWER 1.05 (d.) [+1.0]

Reference (s) 1.05

1. Palo Verde 1 and 2: Palo Verde Licensed Operator Training, Reactor Theory, Pre-License -Review Course NLC55, pp. 3.3, 5.2, and 6.3.

2. St. Lucie 1 and 2: Reactor' Physics Trainina Manual,

" Reactor Physics", p. 7.1-2.

L

-Section 1.0 Continued on-Next Page-

.,o Page 6- . Palo Verde ~1-and 2.

March 11, 1986 Points

.Available ._

5 OUESTION 1.06-The-isotope of plutonium, Pu'", is' found in the nuclear reactor core of Palo Verde Unit 1 because (1.0)

(a.) Pu'" is found in significant quantitie's- (percentage wise) in pitchblend (the uranium material that is mined).

(b.) of the non-fjgston absorption of a thermal or epithermal neutron by U nuclei.

(c.) of the fissioning of U*" nuclei by fast neutrons.

(d.)o[uthe non-fission absorption of a thermal neutron by U nuclei.

ANSWER 1.06 (b.) [+1.0]

Reference (s) 1.06

1. Palo Verde 1 and 2: Palo Verde Licensed Operator Training, Reactor Theory, Pre-License Review Course NLC55, pp. 4.3

, - 4.6. -

r 1

-Section 1.0 Continued on Next Page-

Page 7- Palo Verde 1:and 2 March 11, 1986

, Points Available ,

Ol]ESTION 1.07 In the left-hand column below is a list'.of elements that are significant to the operation ~ of nuclear-reactor power plants..

In the right-hand column below-is a list of element symbols

.(not all of which are . correct element. symbols). For each element in the left-hand column, choose its symbol from those in.the  :

4 right-hand column. .(1.5)

Element " Symbol

1. Zirconium Pl-

2. Antimony 'Fe

3. Beryllium Pb

4. Sodium N

5. Plutonium 0

6. Boron Ni 4

7. Nitrogen Pu

8. Bromine Sm i 9. Carbon Na

10. Chlorine U

11. Helium Zr

12. Samarium Xe

13. Iron Sb

14. Lead Be

15. Cobalt C

.' B' -

^

'Ar Br Cl .l He 1 Cu

! H Co Nt I

l l

l l

-Section 1.0 Continued on Next'Page--

~

'4

[4, Page Palo-Verde 1 and'2 March.11, 1986 Points. -

Available .

ANSWER 1.07

1. Zr-

2. Sb

3. Be

4. Na

15. Pu

6. B

7. N

8. Br [+0.1each]

-9.

C

10. C1

11. He

12. JPJ'-

13. Fe

14. Pb

15. Co Reference (s) 1.07

. 1. Palo Verde 1 and 2: Palo Verde Licensed Operator Training, Reactor Theory, Pre-License Review Course NLC55, .p.1.2.

i  ; .

-Section 1.0 Continued on Next Page- '

,= - - ~, m. < , -e em rw , m , - >

, ' ,c_

Page 9 Palo Verde 1"and-2 March 11,.-1986-4.

, Points -

Available .

j OUESTION 1.08 The turbine of the Auxiliary Feedwater System-(AFS).is -

rated at 1250 hp with steam supplied at a pressure of 1170 psig. .

'a. bentheAFSturbine:isbeingsuppliedwith1000 psia main steam, han much work can be supplied by the turbine? ' Assume thermodynamically ideal' operation -

of the turbine. Express the answer in Btu /lbm. - (2.0) . 4

b. If the turb1.ie was receiving 250 Btu /lba and was driving the AFS pump' with a power-of 950 ~hp,

what is the steam flowrate through the turbine? (1.0)~

ANSWER 1.08 j

a. 51 = 1193 Btu /lbm [+0.5]

i isentropic expansion [+0.5] .

l h2 = 905 Btu /lbm [+0.5]

i

! Ah = 288 Btu /lbm [+0.5 for 280 - 296 Btu /lbe]

Bt hr ~

-- b. 950 hp x 2545 = 2.42 M; Btu /hr [+0.5 for conversion] <

2 "

Bt = 9,680 lbm/hr [+0.5 for 9,180 - 10,180 lbm/hr]

Reference (sl 1.08 ,

1. Palo Verde 1 and 2: System Descriotions,'" Auxiliary' Feed-water System", Training Article PGS-11, pp. PSG-11-6,l12,E >>

13.

2. Palo Verde 1 and 2: Thermohydraulics Review for Licensed Ooerators, Section 2, pp. 9-16.

J i

-Section 1.0 Continued on Next Page-  ;

i

4 Page 10 Palo Verde 1 and 2 March 11, 1986 Points Available ,

s.

QUESTION 1.09 What effect (INCREASE, DECREASE, REMAIN-THE-SAME) would each of the actions below have on the available Net Positive Suction Head (NPSH) to a centrifugal' pump?.

a. Raising the pump elevation to be closer to the' surge tank that feeds it (Neglect friction. losses in the pipe.) (0.5)-

b. Decreasing the inlet pipe diameter- (0.5)

c. Cooling the fluid upstream of the pump .(0.5)

ANSWER 1.09 2

a. DECREASE [+0.5]

b. DECREASE [+0.5]

c. INCREASE [+0.5]

Reference (s) 1.09

1. Palo Verde 1 and 2: Thermohydraulics Review for Licensed

., Ooerators, Section 2, pp. 21, 28, 29. ,

T i

l I

I

-Section 1.0 Continued on Next Page-i

1 Page 11' Palo Verde 1-and 2

- March 11, 1986 4

4-Points  :

Available _

i OUESTION 1.10 Assume that Figure 1.10 (QUESTION) is a graphical illustration

• i of the neutron population during one-(1) neutron generation in -

nuclear-reactor core at Palo Verde. Unit 1; not illustrated are the neutrons that " leak-out" of.the core. Assume that the core i.

is at MOL (middle-of-life). Complete the'following' parts of

this QUESTION by " filling-in the blanks."

a. If the fraction of neutrons (include neutrons of all' speeds) 1-that " leak-out" in a neutron generation was 0.2, the core

. would be (SUBCRITICAL, CRITICAL, 1 SUPERCRITICAL). (0.5)

b. If CEAs were inserted further into this core, the primary affect on reactivity would be to .

(INCREASE or DECREASE) the - (e, p, f, or rf) term. (1.0)

c. If the temperature of the core were increased, the."p"

, term would INCREASE or DECREASE), the -

"f" term would INCREASE or DECREASE) and the- .

total leakage of neutrons fraction of neutrons per gen- '

eration) would

~

INCREASE or DECREASE). -(1.5)' -

l ANSWER 1.10 l a. SUBCRITICAL [+0.5] ^ '

b. DECREASE '+0. 5'

f l+0.5 1

c. DECREASE '+0. 5' '

INCREASE '+0.5'

INCREASE '+0.5l

, r i

Reference (sl 1.10

1. Palo Verde I and 2

Palo Verde Licensed Operator Training, '

Reactor Theory, Pre-License Review Course NLC55, p. 6.2 -

l 7.2.

L

{

Section 1.0' Continued on Next Page-

_____.__________.____.____________._____________f___________.__._____A

Page 12 Palo Verde 1 and 2 March 11, 1986 Points Available .

NEUTRON GENERATION IN A REACTOR OF INFINITE SIZE.

FAST ENERGY

' ~ I gffggy'  : THERMAt. ENERGY, '- $T E ERW "'**

I i l a

i I

@  ?- U235 j l pV "

TN O

l *

" wur l i

-l 1

~

I

= - - -

l l

@ > U235

"  ! imp 1 '

i gL I n U l ,  ; .

n X U235 FAST FISSION e

I RESONANCEk

• U236 7

CAPTURE l

? n ,;- #

l l U235 @

. l l WM '

=Q

' ~

e

=s g 3, 1

t ,

V  %-

L- ,

I n

@ . ', l [ l . .

1

1. 1 6  : p  : f a g

7)

-3 ,

Figure 1.10 (OUESTION) j l

1

-Section 1.0 Continued on Next Page- 1 I

1

.e Page 13 Palo Verde 1,and 2 March 11, 1986 Points l .. Available _,

OUESTION 1.11 I

Answer the following parts of this QUESTION concerning the -

operation of the Palo Verde Unit'1 power plant with 208 EFPD.

a. If the plant had been.oper'ating at 100% of full power for'~

5 days, the magnitude of the xenon worth would be

! pcm. -- (0.5) ? ..

.! b. If the plant had been operating at 100% of full power a for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> (the plant had been in HOT STANDBY for 15 days before FULL-POWER operation) the estimated magnitude of the xenon worth would be- pcm. (0.5)-

~

i c. If the plant tripped while in the conditon:of part "a.",

the magnitude of the peak xenon worth would be-pcm.

-(0.5)

d. If the plant tripped while in the condition of part "b.",

t

. the estimated magnitude of the peak xenon worth would be 7

pcm. (0,5)

, e. When, in part "c."hsthe power level instantaneously dropped;

the "burnup" of Xe (INCREASED, STgED-THEgAME or DECREASED) while the production of the

Te and I

.' DECREASED) and while(INCREASED,STAYEg the rate-of-decay of I ,THE-SAME or '

(INCREASED, STAYED-THE-SAMEdrDECREASED). (1.5) i i

i ANSWER 1.11

a. Using Curve 2.4.1, Table 2.4.2 or Curve 2.4.2, 2720~ 20 pcm [+0.5]

4

b. 1400 t 400 pcm [+0.5]

c. Using Table 1.1, 5470 50,pcm [+0.5]

j d. 3500 600 pcm [+0.5]

4

e. DECREASED '+0.5' l DECREASED '+0.5'

, STAYED-THE-SA'ME ' [+0.5]

i

-Section 1.0 Continued on Next Page-

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

Page.14 - Palo Verde 1 and 2 March 11, 1986 ,

! .. -Points

. Available .

L Reference (s) 1.11

1. Palo Verde 1 and 2: .Palo Verde Licensed Operator Training,

-Reactor Theory, Pre-License Review Course NLC55, p. 16.2

- 16.10. '

2. Palo Verde 1 and 2: core Data Book,. Unit 1, Cycle 1, Revision 001, Table 1.1, Curve 2.4.1, Table 2.4.2, and

Curve 2.4.2, pp. 1, 32, 33,_34.-

a 1

i 4

i

\ .:  ;

1 l

1 4

l

'l

i i l

, i

\

i 1 i

-Section 1.0 Continued on Next Page- ,

Page 15 Palo Verde 1 and 2-March 11, 1986

.. Points -

Available .

OtJESTION 1.12 The Unit 1 power plant is in the final stage of primary system heatup. The power -level is 0%.

a. If the reactor was at BOC with 0 EFPD and 1500 ppm boron, soecify the magnitude of the change in reactivity '(in pcm) as the plant changes -temperature from 550*F to 565'F? (0.5)

b. As the plant changes temperature from 550'F to 565'F, soecify whether the reactivity of the core has INCREASED or DECREASED and orovide an explanation for the direction of change of the reactivity. (2.0)

ANSWER 1.12

a. Using Curve 2.2.1 of the Core Data Book, Ffinal = 0 pcm  ;

Fi nitial = -11 pcm Ap = 11 pcm. [+0.5 for 1111 pcm]

., b. The reactivity has INCREASED [+0.5] as the temperature has increased. '

This is due to the fact that as the temperature increased from 550 to 565'F, the moderator density decreased [+0.5].

With a lower density, there is less boron in the core, which provides an increase in reactivity and less moderating capability, which provides a decrease in reactivity [+0.5].

With a concentration of 1500 ppm of boron, the reactivity effect of the loss of boron is greater than that due to the decrease in moderating capability [+0.5].

-Section 1.0 Continued on Next Page-

, - Page 16 Palo Verde 1 an'd 2 March 11, 1986

" Points

! Available .

m

3 Reference (s) 1.12 j- 1. Palo Verde -1 and 2:- Palo Verde Licensed Operator Training,

• Reactor Theory, Pre-License Review Course NLC55, . pp.10.5

- 10.6.

j

[ 2. Palo Verde 1 and 2: . core-Data Book, Unit 1, Cycle 1,

' Revision 001,' Curve'2.2.1, p. 8.

a DUESTION 1.13 The Unit i nuclear reactor at BOC had just been taken critical t

! with the CEAs in the manual-sequential (MS) mode.- The boron t

' concentration is 1250 ppa. The CEAs are at 90 inches on Group-4 and T., is 505'F. Determine the value.of the boron concentration to which the primary coolant must be taken to reach ARO (all rods out). (1.5) '

i ANSWER 1.13 ,

From Curve 2.5.6, '

e j Ap = 400 pcm. [+0.5, +0.25 for E0C value of 700]

From Table 2.3.4,

boron worth = 11.75 pcm/ ppm [+0.5]

1

) Hence, I

A boron worth = 1h5-34.04 ppm ,

i new concentration =- 1250 + 34 = 1284 ppm [+0.5]

4 t f Reference (s) 1.13 i 1. Palo Verde 1 and 2: Palo Verde Licensed Operator Training, Reactor Theory, Pre-License Review Course NLC55, p.10.8

- 10.9. ,

i .

2. Palo Verde 1 and 2: Core Data Book, Unit'1, Cycle 1,

. Revision 001, curve 2.5.6, and Figure 2.3.4, pp. 23, 48. r i '

-Section 1.0 Continued on Next Page-j

- _ --- ----- --- = 1 2 / - i -

s ..

Page 17 Palo Verde.1 and 2-March 11, 1986 Points Available ,

OUESTION 1.14 Answer the following parts of this QUESTION by calculating the answer or by " filling-in the blanks."

~

a. The heat transfer in the Steam Generators involves a com--

bination of types of heat transfer. Heat transfer from the bulk of the reactor coolant to the tube:. surface is by

, through the tube is by , and from 4

the tube surface to the bulk of the -feedwater/ steam is by .

(0.75)

b. Assume the following conditions for a Steam Generator:

a feedwater temperature = 400'F 7

feedwater flowrate = 1.0 x 10 lbm/hr

• steam pressure = 1000 psia 7

steam flowrate = 1.0 x 10 lbm/hr.

Beg much thermal energy in MW is being transferred to the 4

secondary fluid in the Steam Generator? (Showyour calculations.) (1.75)

, b .

J ANSWER 1.14 1

, a. convection '+0.25' '

4 conduction '+0.25' convection l+0.25 I

1 It 1

-Section 1.0 Continued on Next Page- )

l

.~

Page 18  : Palo? Verde 1 and 2 ~

March 11', 1986 Points-Available ,

• 4 ANSWER 1.~14 (continued)

I

b. .h=4(h steam -hfeedwater) [+0.5]

~

= 1.0 x 10 7lha(1192IANlba hr - 375.1 iba EAN) .

= 817 x 10 7EfE [+0.75]

i 817 x 10 7 IDE hr 6 Btu 3.41 x 10 y ,

j = 2396 MW -[+0.5]

Reference (s) 1.14

1. Generic: Academic Proaram for Nuclear Power Plant Personnel,

, Volume III, " Nuclear Power Plant Technology",-1973 ' General' Physics-l Corporation, pp. 2-141 148.

,~ -

! -End of Section 1.0 -

t-t i

I a

9 i

i f

f

Page 19 Palo Verde 1 and 2 March 11, 1986 Points Available ,

s 2.0 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS (25.0)

QUESTIONS 2.01 through 2.05 are " multiple-choice" questions. For each question specify the letter designation for the phrase that provides the most correct statement.

OUESTION 2.01

, TheCirculatingWater(CW)Systemfunctionstoprovidecooling water to the main condenser; it consists of four (4) CW pumps with (1.0)

(a.) each pump powered from a separate 13.8 kV bus, each pump with a motor-operated valve in its output line, and each pump feeding one (1) of four (4) independent tube bundles in each of the three (3) condenser sections.

(b.) each pump powered from a separate 13.8 kV bus, each pump with a motor-operated valve in its output line, and which combine their flows into two (2 flow paths with each path feeding one (1) of two (2)) independent tube bundles '

in each of the three (3) condenser sec.tions.

i (c.) each pair of pumps powered from a separate 13.8 kV bus, each pump with a motor-operated valve in its output line,

.' and which combine their flows into two (2 flow paths -

with each path feeding one (1) of two -(2)) Independent -

4 tube bundles in each of the three (3) condenstr sections.

, (d.) each pair of pumps powered from a separate 13.3 kV bus,  !

which combine their flows into three -(3) flow paths with I each  !

one (path containing

1) of the three (3)a condenser motor-operated valve and feeding sections.  !

I ANSWER 2.01 (c.) [+1.0]  ;

I l

Reference (s1 2.01

1. Palo Verde 1 and 2: System Descriotions, " Circulating Water System", Training Article PGS-7A, pp. PGS-7A PGS-7A-8. >

, -Section 2.0 Continued on Next Page- i

/

.c .

.Page 20 Palo Verde 1.and 2 March 11, 1986 Points Available _

OUESTION 2.02 Each Emergency Diesel Generator can be supplied diesel fuel oil to continuously operate at 100% load for up to (1.0)

(a.) I hour with the day tank and 7 days with the storage tank.

(b.) 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> with the day tank and 7 days with the storage tank.

(c.) I hour with the day tank and 2 days with the storage tank.

(d.) 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> with the day tank and 2 days with the storage tank.

ANSWER 2.02 (a.) [+1.0]

Reference (s) 2.02

1. Palo Verde 1 and 2: Diesel Generator and Auxiliaries Handbook, pp. 16-18.

s .

-Section 2.0 Continued on Next Page-

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

+

1 i Page 21 Palo Verde 1 and 2 ~

March,11, 1986 Points -

'Available ,

~

l OUESTION 2.03 i

[ To protectthe condition, the Steam main steam Generators (S/Gs) system contains from (20 twenty an safety 'overp)ressure 1- valves that -(1,0).

l.

(a.) have a total relieving capacity of 100% of. full steam i= flow and have sequential relief settings starting at 1255 ,

j psig and ending with 1315 psig. '

l (b.) must-be combined'with the atmospheric dump valves to'have 4

a combined total relieving capacity of 100% of full steam-flow and have sequential relief settings starting at 1255 psig and ending with 1315 psig.

j (c.) have a total relieving capacity of 100% of full steam I flow and have sequential relief settings starting at 1070 j psig and ending with 1130 psig.

i (d.)mustbecombinedwiththeatmosphericdumpvalvestohave

! a combined total relieving capacity of 100% of full steam flow and have sequential relief settings starting at 1070 l J psig and ending with 1130 psig.

j i ANSWER 2.03 2 (a.) [+1.0] l 4 ,

Reference (s) 2.03 -

f 1. Palo Verde 1 and 2: system Descrintion, " Main Steam

~  :

} System", Training Artic1e PGS-1A, pp. PGS-1A PGS-1A-10. 1 i -

p  !

i f

(

i i i i

-Section 2.0 Continued on'Next Page- [

l f- i

~~

i Page 22 Palo: Verde 1 7.nd 2 .

March _11, 1986

~ Points

~

Available .

QUESTION 2.04 The four (4) motor-operated valves.for the safet'y injection '

tanks (SITS)-are ' (1.0)

(a.) closed during normal Mode l operation to isolate the SITS i- from the RCS but receive an open signal on a SIAS.

(b.) open'during normal Mode 1 operation but receive an open signal on a SIAS.

(c.) interlocked with the Pressurizer pressure.to automatically open if the RCS pressure decreases below-500 psig. '

(d.) interlocked with the Pressurizer pressure to ensure that they are closed if the RCS pressure increases above 500 psig.

ANSWER 2.04 I (b.) [+1.0]

Reference (s) 2.04

., 1. Palo Verde 1 and 2: System Descriotions, " Safety Injection ' .

and Shutdown Cooling System", Training Article NS-3A, ,

Section 2.1.3, pp. NS-3A-13, NS-3A-14.

1  ;

1 8

4 i ,

j i

a Section 2.0 Continued on Next Page-

s e l

l C Page 23 .Palo Verde 1'and~2

~

March 11,.-1986

. Points.

Available- ,

5

OUESTION 2.05 .;

i The 125 VDC Class IE. Power System consists of four 125 VDC control centers, A, B, C, D such that. (1.0)

(a.) A .is- tied to the A battery charger and' normally to the A-C backup battery charger and supplies DC loads including a 125 VDC to 480 VAC inverter for the shutdown-cooling j isolation valves.

l (b.) B is tied to the B battery charger and nonnally is NOT }

tied to the B-D backup battery charger and supplies DC '

i loads including a 125 VDC to 120 VAC inverter for Class IE

120 VAC loads.

(c.) C is tied to the C battery charger and normally to the '

, A-C backup batter charger and supplies DC loads including j a 125 VDC to 120 VAC inverter for Class IE 120 VAC loads.

(d.) D is tied to the D battery charger and normally is NOT.

i tied to the B-D backup battery charger and supplis DC

] loads including a 125 VDC to 480 VAC' inverter.for.the >

j shutdown-cooling isolation valves.

ANSWER 2.05

) (d.) [+1.0]

Reference (s) 2.05 1

, 1. Palo Verde 1 and 2: System Descriotions,'"125 VDC Class i i IE Power System", Training Article PGS-150, pp. PGS-150-7  ;

through PGS-15D-10. '

i

q l

s .

$ l I

l-

! I

-Section 2.0 Continued on Next Page-l 1 .

! l

Page 24 Palo Verde 1 and 2 March 11, 1986

. Points Available , .

OUESTION 2.06 Answer TRUE or FALSE.

A reactor coolant pump may be operated without seal injection' water for an indefinite period of time as long as nuclear .

cooling water is still supplied. (0.5)

ANSWER 2.06 TRUE [+0.5]

Reference (sl 2.06

1. Palo Verde 1 and 2: Palo Verde Nuclear Generatina Station Manual, " Reactor Coolant Pump Operation", 410P-1RC01, Section 3.17, p. 8.

QUESTION 2.07 WhymusttheEmergencyDieselGenerator(D/G)lubeoil circulating pump be run for a minimum of 30 minutes after shutting down a hot D/G7 (1,0)

ANSWER 2.07 to dissipate heat [+0.5] from the turbocharger bearings [+0.5]

Reference (sl 2.07

1. Palo Verde 1 and 2: Palo Verde Nuclear Generatina' Station

~

Manual, " Emergency Diesel Generator", 410P-1DG01, Section.

3.28, p. 10. ,

-Section 2.0 Continued on Next Page-

1 ,

Page 25 Palo Verde 1 and 2 March 11, 1986 Points Available ,

00ESTION 2.08 List four (4) of the actuating signals that should automatically start the Essential Cooling Water System. (2.0) r., . 4, ANSWER 2.08

1. Loss of Offsite Power (LOP)

2. Control Room Essential Filtration Actuation Signal (CREFAS)

3. Safety Injection Actuation Signal (SIAS)

4. Control Room Ventilation Isolation Actuation Signal (CRVIAS) '

5. Auxiliary Feedwater Actuation Signal 1 (AFAS-1)

6. Auxiliary Feedwater Actuation Signal 2 (AFAS-2)

7. Containment Spray Actuation Signal (CSAS) l [+0.5 each, +2.0 max]

j j Reference (sl 2.08 i

1. Palo Verde 1 and 2: Palo Verde Nuclear Generating Station '

Manual, " Essential Cooling Water System", 410P-1EW01, Section 4.1.2, p. 7.

I i

l

)

l 1 I i

4

-Section 2.0 Continued on Next Page- 'l i

)

Page 26 Palo Verde 1 and 2 March 11, 1986 Points Available _

00ESTION 2.09 As a result of loss of electrical power, what is the failed position (OPEN, CLOSED, AS-IS) for each of the following actuator-operated valves associated with the Safety _ Injection System.

a. Low-pressure header isolation valve (0.5) i i b. HPSI pump mini-flow line isolation valve' (0.5)

c. Sump line isolation valve (0.5)

! d. LPSI pump suction isolation valve (0,5)

e. Hot-leg injection isolation valve (0.5)

, ANSWER 2.09

a. AS-IS [+0.5]

b. CLOSED [+0.5]

l c. OPEN [+0.5]

l ., d. CLOSED [+0.5] _.

1

e. AS-IS [+0.5]

Reference (s) 2.09 ,

1. Palo Verde 1 and 2: System Descriotions, " Safety Injection j and Shutdown Cooling System", Training Article NS-3A, Table j NS-3-I, p. NS-3A-25.

-Section 2.0 Continued on Next Page- ,

1 .

i

.- l 4 'Page'27' Palo Verde 1 and 2' l

March 11', 1986 Points -

t Available ..

OUESTION 2.10 Other than an Overspeed Trip, lisi six (6) trips that will '

initiate closure of the main turbine steam admission valves. (3. 0)-

ANSWER 2.10 j a.- Main condenser low vacuum

b. Excessive thrust bearing wear

.l' j c. Reactor trip

d. Generator / reactor initiated trip -

e. Master trip from control room ,

'1

f. Manual trip using lever located at the turbine

g. Excessive vibration

} h. High exhaust hood temperature i

1. Moisture separator high-high level ,

i j j. Prolonged loss of stator coolant ". -

'~ ". " "

, .. r -

" ' " ' ~ ~ '

i k. Low hydraulic fluid pressure ' , f, - ~ #

a- .,. . .

i 1. Loss of EHC 125 VDC control power

. [+0.5 each, +3.0 max] . ,

Reference (s) 2.10

1. Palo Verde 1 and 2: system Descriotions, " Main Turbine System", Training Article PGS-3A, Section 4.3, p. PGS-3A-43.

i

-Section2.0ContinuedonlNextPage-l

Page 28 Palo Verde 1 and 2

,. March 11, 1986 Points Available ,

OUESTION 2.11 lihat is the function of the CVCS back pressure control valve? (1.0)

. ANSWER 2.11 1

Maintains upstream pressure high enough to prevent letdown flow from flashing to steam in the letdown line. [+1.0]

t

Reference (sl 2.11 I

! 1. Palo Verde 1 and 2: system Descrintions, "CVCS-1", Training

! Article NS-2A, Section 1.3.1, p. NS-2A-7.

i i

i 1

1 .

i i -- .,

i i

l 4

I 1

i I

i

)

i

-Section 2.0 Continued on Next Page-i

Page 29 Palo Verde'l and 2 March 11, 1986

. Points' Available ,

00ESTION 2.12 Draw a one-line diagram of the auxiliary feedwater system that

• l contains the following components. Label the components on the diagram. (2.5).

• essential pump i

• turbine-driven essential pump e non-essential pump i

• line providing hydrozine addition

• line providing ammonia addition water sources for the three pumps i

• the eight (8) cross over valves I line from main feedwater supply valves controlled by the FWCS

• MSIS valves

• steam generators -

ANSWER 2.12 See Figure 2.12 (ANSWERI [+2.5]

Reference (sl 2.12 ,

1. Palo Verde 1 and 2: System Descriotions, " Auxiliary '

Feedwater System", Training Article PGS-11, Figures PGS-11-2 and PGS-11-3.

-Section 2.0 Continued on Next Page-i I

o. .

a-4 Page 30 Palo Verde 1 and 2 ~

March 11, 1986 Points -

Available ,

+

I

_ _ 3,f.4.M.

,, , m , .

c. m 9,y c-4 -..

e g,

.. e- ,

> R cp

~

,,,, ,y ,

f .

-,,3,.... ..

j ,,4 >_ A u s U N hN j~

ys -- s Hr:5...J t.o; qj g vv....e / -

\0 mew (n / gaessovtA O h

• Va6vC3 Q

t. ... . O -

^ -

4, ,

7

h. e

ll.'." d

($18. flak ee .J";"0 3

4 i

e.t.a.s..

./ _ ,

t e.n ey , ,

g O r... ,r

" ' * ' ' ,'\" /.

d .

~ .

4 s 83s.t. f.t44 r .-

. \/  ;

V 4

i, ,

t

[+0.25 each connection 1-10]

4 Flaure 2.12 (ANSWER)

-Section 2.0 Continued on Next Page-1 1

r 4

___.______.__________._.__m_____ _ - . _

s ,- ,

Page 31 Palo Verde 1 and 2 March 11, 1986 Points Available ., ,

OUESTION 2.13 Answer the following parts of this QUESTION which pertains to -

the Essential Spray Pond (SP) and the Essential Spray Pond .

System (SPS). ,

a. According to the Safety Design Basis, up to what length of time can the SP and SPS be operated without water makeup?

(Answer 1 hr,1 day,1 month or 1 year.) (0.5)

b. What are the normal (preferred) and the backup water supplies for the SPS? (0.5)

ANSWER 2.13 i

a. 1 month [+0.5]

b. Domestic Water System (preferred source)  ;+0.25 I station reservoir (non-preferred source) ,+0.25;.

Reference (s) 2.13

1. Palo Verde 1 and 2: System Descrintions, " Essential Spray Pond System", Training Article PGS-8A, Section 1.1, 'l

., pp. PGS-8A PSG-8A-8. -

l l

-Section 2.0 Continued on Next Page-l

e '

Page 32 Palo Verde 1 and 2 4

March 11, 1986-Points Available ,

OUESTION 2.14 Wha.t operational consideration was used to design the minimum '

volumetric capacity of the CVCS volume control tank? (1.0)

ANSWER 2.14 1

l <

sufficient capacity to accommodate the inventory change with 'g no makeup system operation [+0.5] from a full to zero power m decrease [+0.5]

Reference (s) 2.14 -

1. Palo Verde 1 and 2: System Descriotions, "CVCS-I", Training Article NS-2A, Section 1.2.b2, p. NS-2A-4. f

.N OUESTION 2.15 Answer TRUE or FALSE to 'the following parts of this QUESTION $

l which refers to thd operation of the Main Turbine, 4 f , i t

n l a. The Main Turbine shall not be operated above 2/3-rated-speed (1200 rpm) with a condenser pressure above a given 4

maximum value. 5

.s < (0.5) -

7-3

b. Upon loss of sealing steam, this maximum value for the condenser pressure (required for operation above 1200 i";

rpm) is reduced; i.e., a better vacuum is required. (0.5) '

4 s-

-5 ANSWER 2.15  %

a. TRUE [+0.5]

si

b. FALSE [+0.5] n Reference (s) 2.15, -=

. s r

1. PaloVerde1,$nd2: Palo Verde Nuclear Generatina Station Manual, " Main Turbine", 410P-1MT02, Section 3.2.3, p. 10.

3 g-E

-Section 2.0 Continued on Next Page- Y Y

y

=

~

....s .

Page 33 Palo Verde 1 and 2 March 11,-1986 Points Available .

l OUESTION 2.16 i What is the purpose of the six (6) core stops that are welded to the inside wall of the reactor vessel? (1.0)

ANSWER 2.16 Limits the downward drop [+0.5] of the core support vessel

[+0.5] in the event of a flange failure Reference (s) 2.16

1. Palo Verde 1 and 2: System Descrintions, " Core Stops",

Training Article NS-1C, Section 6.2.12, p. NS-1C-24.

OUESTION 2.17

. What are the two (2) reasons for designing the nuclear fuel pellets to be slightly dished inward on both ends. (1.0)

ANSWER 2.17

.' 1. Allow greater axial expansion at the center of the pellet '

during pellet heatup. [+0.5]- '

2. Reduce the effect of radiation induced swelling of stack height. [+0.5]

Reference (s) 2.17

1. Pald Verde 1 and 2: System Descrintions, " Fuel Rods",

Training Article NS-1C, Section 2.2.3.2, p. NS-1C-9.

-Section 2.0 Continued on Next Page-

+

Page 34 Palo Verde l' and 2 March.11, 1986

. Points Available- . .

1

OUESTION 2.18 l' ,

Answer this QUESTION by " filling-in-the-blanks" in;the followingL statement.  :

-The minimum RCS flowrate of:111,'400 gpm ensures capabilities in all modes of operation and the -

maximum flow rate of 129,225 gpm limits the generated on the core. -(1.0)-

l ANSWER 2.18 l heat removal [+0.5]

uplift forces [+0.5]

Reference (s) 2.18 i

1. Palo Verde 1 and 2: System Descriotions, " Reactor Coolant

! System", Training Article NS-1A,' Section 1.2G, p. NS-1A-2..

L -+ . :

t 4

]

-Section 2.0. Continued on Next Page-

- i t

9

-- - --.,.----,e , - , - ~ , =v,., e, , e4 -

m vao , , . ,-y., -,, ,e,* ~

y r- ge n--- e yJ , ~

Page 35 Palo Verde 1 and 2 March 11-~ 1986 Points -

Available .

QUESTION 2.19 Consider the situation in which a steam-line break has occured in Unit 1. '

a. What will initially happen'to T,y9 ? Why does T av9 change in this= manner? (1.0)

b. What will initially happen to the Pressurizer level? (0.5)

ANSWER 2.19

a. T,yg will decrease [+0.5]. The increase in steam flow-rate will increase the heat rate between the primary and secondary [+0.5].

b. The Pressurizer level will drop (due to RCS shrinkage).

[+0.5]

Reference (sl 2.19

1. Generic: Academic Program fDr Nuclear Power Plant Parsonnel, Volume III, General Physics Corporation, p. 3-37.

L = .

-End of Section 2.0-4

)

e

, -- e, r .nn- - w r-

Page 36 _Palo Verde 1:and 2- ~

March 11, 1986

. Points- -

Available 3.0 INSTRUMENTS AND CONTROLS .(25.0). ,

QUESTIONS 3.01 and.3.02 are " multiple-choice" questions. .For each .

question soecify the letter designation for the phrase that provides the most correct statement.

i OUESTION 3.01 .

A reactor-vessel seal alarm would be received inLthe control-room if (1.0) 4 (a.) the inner. seal cannot' hold the primary system pressure and the pressure between the two seals exceeds 1500 psig.

(b.) the outer seal cannot hold the primary-system pressure and the pressure between the two seals drops below 500 psig.

(c.) the outer seal cannot hold the primary system pressure and the pressure between the seals exceeds 1000 psig.

(d.) both seals fail and the primary system pressure drops below 2100 psig.

ANSWER 3.01 '

~' "

(a.) [+1.0]

J t i

Reference (s) 3.01

1. Palo Verde 1 and 2: System Descriotions, " Reactor Coolant System", Training Article NS-1, Revision 1, August 1985, pp,.NS-1A-5,NS-1A-23. .

l 2. Palo Verde 1 and 2: Nuclear Generating ~ Station Manual,

" Excessive RCS Leakage",-41AO-1ZZ14, Appendix C,-p. 2. ,

4 6

-Section 3.0 Continued on Next'Page-

[

_ . . . . , ,,u_.._. _ _ , _ . _ _ _ . , . _ _ , . _ . . . _ _ _ _ _ _ . , . . . . _ _ _ . _ . ,:._.- .._2.. _. . _

Page 37 Palo. Verde 1 and 2 March 11, 1986 Points Available ,

OUESTION 3.02 ,

After. the receipt of. an MSIS and the closure of all four (4)'

MSIVs, the operator could control the temperature and pressure in the Steam Generators by (1.0) ,

(a.) changing the steam flowrate to the main feedwater pump turbine. ,

(b.) changing the steam flowrate to the auxiliary feedwater pump turbine.

(c.) taking manual control of the steam-bypass control valves.

(d.) opening or closing the valves from the main-steam lines to the upstream condensate drain line.

ANSWER 3.02 (b.) [+1.0]

.l Reference (L1 3.02

1. Palo Verde 1 and 2: System Descriotions, " Main Steam ,l

., System", Training Article PGS _1A, pp. PGS-1A PGS-1A-6. . j i

e I

A I l

I l

-Section 3.0 Continued on Next Page- ll 1

m .

' Page 38 Palo Verde-1 and.2 l e ~ March 11, 1986.

Points Available' . 'i QUESTION ~3.03 .l l

Answer TRUE or FALSE to each of the following statements con- '

i cerning the control of each atmospheric dump valve.

a. A nitrogen gas acci'="lator'is provided to ensure operability .

for at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the loss of. the Instrument -

! and Service Air and Nitrogen-Gas Systems. (0.5)

b. One (1) solenoid-operated-isolation valve is installed.in' the nitrogen gas supply line.which can be opened.from the-control room or the remote-shutdown _ panel to unisolate

-l the backup pneumatic supply (nitrogen gas) if instrument i and service air is lost.

(0.5) .1 l

i

c. One (1) P/I converter is installed to convert the. pneumatic- .

I signal of 3 to 30 psig to an electrical signal, which is used to control the position of the dump valve. (0.5).

d. The.two (2) safety related, solenoid-operated, three-way-valves, installed in series'in the air supply line to the actuator, must hoth be opened to vent the accumulator to the atmosphere. '(0.5) .

ANSWER 3.03. ,

a. TRUE [+0.5] -l 1

b. FALSE [+0.5]

c. FALSE [+0.5]

d.

~

FALSE [+0.5]

Reference (s) 3.03

1. Palo Verde 1 and 2: System Descriotions, " Main Steam'  :

System", Training Article.PGS-1A,' pp. PGS-1A PGS-1A-11.

s 4

-Section-3.0 Continued on.Next Page-

Page 39 Palo Verde 1 and 2 ^

March 11, 1986 Points .

Available _

i OUESTION 3.04 Shown.as Figure 3.04 (OUESTION) is the EW/NC X-TIE VALVES LOGIC DIAGRAM. Answer the following questions concerning logic diagrams in general and the control of the.EW System in particular.. , . - a .

a. Region-1 on Figure 3.04 (00ESYLOS). represents whitt type of logic connection (AND,-OR, NOR, NOT, NAND)? -(0.5)

b. Region ~2 on Figure 3.04 (00ESTIbN) represents what type of logic connection (AND, OR, NOR, NOT, NAND)? (0.5)

c. A typical logic wiring diagram is also shown on Figure 3.04 (00ESTION). Does this logic wiring diagram provide an alternate schematic for the logic represented by Region 1 or Region 27 (0.5)

d. If there were a LOP signal and a SIAS, whitt would be the condition of the cross tie (OPEN, CLOSED)? (0.5)

e. Describe the operator manipulation of the control-room switch to CLOSE UV-145 after it has been automatically OPENED.

(0.5)

ANSWER 3.04 -

a. OR [+0.5]

b. AND [+0.5]

c. Region-2 [+0.5]

I

d. CLOSED [+0.5]

e. (Followingautomaticopening,thecontrol-room-operator-can close the valve by) momentarily moving the control switch to the OPEN position and then to CLOSE. [+0.5]

Reference (s) 3.04 ,

1. Palo Verde 1 and 2: System Descriotions, " Essential Cooling Water System", Training Article PGS-88,.pp. PGS-88-17, PGS-88-18, and Figure PGS-88-9.

-Section 3.0 Continued on Next Page-i

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

Page 40 Palo Verde 1 and 2 ~

March 11, 1986 Points Available _

Region-1

-- y --- .

~

' Region-2 ,MCC CR OPEN m/ o,CR o\

' ' ^

. V OPEN f

LOP EW LOOP j / A TO NW X-TIE UV-145 ,

MCC CR N ' N /

SIAS 1 .

/ N -' \

CLOSE

. O EW LOOP

, V A TO NW CR CLOSE

- X-TIE UV-145 SEAS TOROUE PROTECTION I

. STOP VALVE ELECTRICAL PROTECTION ._. _.. . ' s -- -.- -.

s. _- '- '

SEIS EW/NC X-TIE VALVES LOGIC DIAGRAM -

l Typical logic wiring diagram for Part "c".

Input relay contacts A #1~

~

_I__

- ,p - r- ,

1. 2 Output relay I

Figure 3.04 (00ESTIONI

-Section 3.0 Continued on Next Page--

~

.- m  :

l i

Page 41 Palo Verde 1 and 2 March 11,-1986 Points' Available _

OUESTION 3.05.

. Listed below as "a." through "e." are components of the Fire -

, Detection System for the Emergency _ Diesel Generator System.

Listed below as "1." through "4." are four.(4) possible actions of the components when they respond to the conditions of a -

fire. For each-fire-control component, give the one -(1) .

associated design action. (Any action item may be.used for more than one fire-control component or not at ~all.) (2.5) -

a. fixed-temperature detector above the day tank

b. sprinkler-system fusible link

c. fusible link on the chain for the rollup fire door j d. ionization smoke detector in the DG local-control room i

! e. ultraviolet flame detector in the DG engine room activates water flow to the sprinkler heads located in 1.

the DG engine rocm ,

2. activates water flow to the sprinkler heads located in the

., silencer room. .

3. activates an alarm indication ONLY f i

4. 1solates the DG engine room from the DG local-control room '

4 ANSWER 3.05

-i

a. 1. '+0.5' *' i

b. 1. '+0.5'

c. 4. '+0.5'

d. 3. '+0. 5'

e. 1.- I'+0. 5 Reference (sl 3.05

1. Palo Verde 1 and 2: Diesel Generator and Auxiliaries .

4 Handbook, pp. 7-8.

I h

-Section 3.0' Continued on Next Page-9 i

. , _ - - ._...__.--.n. _a _. -

+ . .

. s .

Page 42 - Palo Verde 1 and 2 March 11,11986 Points

, Available , ,

OUESTION 3.06 Refer to. Figure 3.06 (00ESTION) which shows a typical level-measuring system for such closed tanks as the Pressurizer and -

the Steam Generator. Answer the following parts to this QUESTION L by choosing the correct response, by " filling-in the blanks", -

or by completing the sentence.

a. The output of the D/P cell is PR - Py. This output is equal 4

to the water density times . (ht , h2, h 3, . h4 , or h5 ) -(0.5)

b. If the water level decreased, hon would the output of the D/P cell (P (INCREASE, DECREASE, or STAY-THE-SAME) R - Py) change? (0.5)> ,

c. If the reference leg broke and some of the waterLin the

. reference leg drained out, the output of the D/P cell (Pg - Py) would (INCREASE, DECREASE, .or STAY-THE-SAME) and the indicated level would .

, (INCREASE, DECREASE, or STAY-THE-SAME). (1.0)

, d. If there were a large RCS or steam line break within

• i, containment that resulted in elevated temperatures; why.

• would the accuracy of the level-measuring system be  !

effected?

(1.0)  ;

~~

, ANSWER 3.qh '

a. h [+0.5]

{ 2

b. INCREASE [+0.5]

_i

c. DECREASE '+0. 5' 3 INCREASE '.+0.5  ;

d. The ambient-temperature of the reference leg would increase, l which would decrease the density of the water in the -;

reference leg, which would decrease PR , which.would increase

• the indicated tank level. [+1.0]

4 Reference (s) 3.06 I

1. Generic: C-E Trainino Document, " Controllers and Process.- -I j Instrumentation", pp. 969(92W3)/ds 30_.

s 1

i b l -Section 3.0 Continued on NextlPagea

.,-Qg y ,,---e-.+ 4..ym.-p9 ,--g --,-.m.t" ' * " '-

Page 43 Palo Verde 1 and 2 March 11, 1986 Points j Available .

COND NSING VARIABLE l LEG REFERENCE LEG

_ ==) -

PRESSURE XL ,

(P3) i  !

hz' ,

U *  ;

=_ - :_ - s :  :-- i I

hi >I .

I i bf ,

v

'~

V i A  : ,

i I

l i h3 l

V D/P y CELL P P V R l

Fiaure 3.06 (00ESTIOtll' l!I i

-Section 3.0 Continued on Next Page-l

I- e Page 44 Palo Verde ~1.and 2 March 11,.1986

, Points

- Available _

OUESTION 3.07

$h .

Answer the follosiing partr of this QUESTION _which pertain to the control of the Condenser Air Removal System.-

a. A two-position (ST4RT/STOP) control switch is provided in the control room /or the A, B, and C vacuum pumps. List -

the three (3) rr.sults of momentarily placing the control-switch for the A-pump in .the START position.. (Do not list any light indications or that the switch w1_ll . return to the neutral-position when released. Do list such actions i as pump X receives a start signal, valve Y receives a-close-permissive signal, etc.)- (1.5)

b. A three-position.(OPEN/CLOSE/AUT0) control switch is pro-vided in the control room for the isolation valve in the suction line from the A section of the main condensor to the standby vacuum pump. When in the AUTO position, the

valve will automatically OPEN if two (2) conditions are

! met. LLs1 these two (2) conditions. (1.0) c.

~/. Ylwre- os /

From the exhaust header are three 3(. ) flowpaths to the

' atmosphereyone (1) radiation monitor,(and one (1) fil-tering unit. If the radiation monitor is triggered on 4 a high-radiation condition, describe the flowpaths by completing the table below. (1.5)

Flowoath-1 Flowcath-2 Flowoath-3

Is the filter this flowpath?

(YES or NO)

Isolation valye position?

(OPEN, CLOSED,

or STAY-THE- .

SAME)

I

-Section 3.0 Continued on Next Page-

_-__ ___ _. _ _ . - _ _ _ _ 7 . ._ . _ .

lPage 45 Palo Verde 1 an'd 2?

March 11, 1986 Points' --

Availabl e' ,,

ANSWER 3.07 a..

• the vacuum pump A will receive a start signal.

. an open-permissive is.sent to the A pump's suction

-isolation valve

~

• associated seal-water recirculation pump will receive a start signal .

[+0.5 each] ,

b. . a standby vacuum pump running signal is received /

• the pressure in sectiton A exceeds its setpoint 'N (3.7 in. HgA)

[+0.5 each]

c. flowoath-1 flowcath-2 flownath-3 NO YES NO CLOSED OPEN STAY-THE-SAME % j eccar .p

[+ 1.5] c,oy e p ;

fi, f , pw & ~

t -

Reference (s) 3.07 . y ,(J/ A es.oi[V

1. Palo Verde 1.and 2: System Descriotions, " Condenser Air 4 -4 5~ M n Removal-System", Training Article PGS-98, pp. PGS-9B

• k . 7p* .

PGS-98-22.

-Section 3.0 Continued on Next Page-

Page 46 Palo: Verde 1 an'd'2~-

March 11,;1986 Points Available ..

QUESTION 3.08 Answer the following parts of this QUESTION which refer to the -

-control of the condensate System.

a. The Condenser hotwell makeup control valves LV-81 and LV-82 will automatically close upon receiving a signal Indicating what condition? (0.5)

b. When in the AUTO mode, the condensate polishing demineralizer bypass valve position is varied in response to what sional? (0.5)

c. When in the AUTO mode, the condensate polishing-demineralizer bypass valve matica11y only if three (3) position conditions willarevarymet. auto-Lt.s1 these three (3) conditions. (1.0) i d. Snecify the buses which normally power each of the.

conden' s ate pumps (A, B, and C).

(1.0)-

, ANSWER 3.08

a. Condensate storage tank level decreases to 28 ft actual 1 eve 1 [+0.5] ,,e J .- .!

[v.; - .

b. demineralizer differential pr' essure [+0.5]

i

, c. .

the deminalerizer input valve is fully.0 PEN [+0.25]l _

y-

. u --

. the deminalerizer output valve is fully OPEN [+0.25] .' ' ^A " "'

+

the differential pressure is less than 35 psi -[+0.5]'

d.

• A, B on S01 ~[+0.5]: 2

, r , ', '

• C on S02 [+0.5] ' -

Reference (s1 3.08 l 1. Palo Verde 1 and 2: Svstem Descriotions, " Condensate System", Training Article PGS-9A,:pp. PGS-9A-21, PGS-9A-22,  ;

, PGS-30.

. l

-Section 3.0 Continued on Next Page- j

+----t-y r y 7-f4 yn A -y9 --ya,-- y---a-'# -gye f y'Ww- e'- -=P=* -F 4 *=

• g-+- "-7' "?Pt* * *NG r- W r -

Page 47 Palo. Verde 1 and 2 March 11, 1986 Points Available. _. .

QUESTION 3.09 The Reactor Regulating System generates a'"pn signal" and a -

"pnDeviation" indication.

a. What two (2) signals are used.to. produce the "pn signal" and the "pn Deviation" indication? (0.5) '

b. What is the acceptable relationship between the two (2) signals that would prevent-the indication of "pn Deviation"? (0.5)

c. What signal would be sent to CEDMCS if there is a "pn Deviation"? (0.5)-

ANSWER 3.09

a. excore neutron ietectors in control channels 1 and 2 [+0.5]

b. difference in the signals must be within 5% [+0.5] .

c. AMI [+0.5]

Reference (s) 3.09

1. Palo Verde 1 and 2: System bescriotions, " Reactor Regulating System", Training Article NS-9C, Revision 1, .

October 1985, p. NS-9C-7.

I 1

-Section 3.0 Continued on Next Page-L

l .

Page 48 Palo Verde'l'and 2

. March:11,:1986 i

Points .

Available ,

I

- 00ESTION 3.10 Answer the' QUESTION by " filling-in-the blanks."- -

The Steam Bypass' control System.(SBCS) controls the Steam Bypass Control valves by using the output of a PID controller which has two inputs.. One signal' input is the.

which is compared to a setpoint which is a function of the.

input signal. Actuation of a quick-opening.

sequence is based on the rate-of-change of the input signal. (1.5)

ANSWER 3.10 steam-header pressure [+0.5]

steam flowrate [+0.5]

! steam flowrate [+0.5]

Reference (s) 3mlQ

1. Palo Verde 1 and 2: System Descriotions, " Steam Bypass

,1 Control System", Training Article NS-98, p. NS-9B-19..

's -

l .

i T

k 2

-Section 3.0 Continued on Next Page-

.w, ~

- - - , , ,- ..r-- ,-..,,-m ._,,.,--,,.,m-,,,., , , s- ~. , , . w, ,,,-y , e n ,4v...,,.m.q

Page 49 Palo Verde 1 and 2 .

March 11, 1986 Points Available _

QUESTION 3.11 The Fuel inputs Building from ) Essential two (2 radiation Ventilation monitors System (FBEVS) uses for ESF actuation. Soecify the location for the detectors of these

~ ~

two (2) radiation monitors. (1.0)

ANSWER 3.11

1. exhaust duct [+0.5]

2. over the fuel pool [+0.5]

Reference (s) 3.11

1. Palo Verde 1 and 2: System Descriotions, " Engineered Safety Features Actuation System 1/2", Training Article NS-78, pp. NS-78-15, NS-78-16.

A -

e t

i J

-Section 3.0 Continued on Next Page-

e Page 50 Palo Verde.1 and'2 .

March.11, 1986 Points Available a .

OUESTION 3'.12 If the following conditions existed at Unit 1, lis.t the ESF S: .

actuation signals that should have been generated. (2.0)

a. Pressurizer pressure = 1650 psig (no bypass)

AND .

containment pressure = 4 psig AND Steam Generator 1 & 2 pressures = 940 psia AND Steam Generator 1 & 2 level = 57% t

b. Pressurizer pressure = 2200 psig (no bypar,s)

> AND containment pressure = 11 psig AND Steam Generator 1 pressure = 700 psia AND

. Steam Generator 2 pressure = 950 "

. AND 1

Steam Generator 1 level = 20%

AND Steam Denerator 2 level = 50% '

(ANSWER 3.12 g

y, s y 3 { j sf ,, ,. .

a. SIAS [t0:5]

b. SIAS -

CSAS CIAS d MSIS MA 'A' ^'= 4 ,tGlN ;-~ ,_ , , ,ees-t i

[+0.-3-eachr-+1v5-max.] 1'+Aa 'e-c4]

Reference (si 3.12

1. Palo Verde 1 and 2

System Descriotions, " Engineered Safety Features Actuation Syr, tem 2/4*, Training Article '

NS-7A,-pn. NS-7A-32 .NS-7A-35.

-End of'Secticn 3.0-

. - - , ,-*-cr y ---, ----r-, m--- rw - c -m - - , - , - , e- -- e -e-- & , -,,m- ww--- -

Page 51 Palo Verde 1 and 2.

March 11, 1986

, Points Available ,

4.0 PROCEDURES. NORMAL. ABNORMAL. EMERGENCY AND RADIOLOGICAL-CONTROL (25.0)- .

QUESTIONS 4.01 and 4.03 are " multiple-choice" questions. For: each I

question soecify the letter designati.on for the phrase that provides .

the most correct statement.

QUESTION 4.01 ,

On-coming shift personnel shall review shift logs (1.0)

(a.) back to the last shift worked or the previous 3 days of '

! logs, whichever is the shortest, prior to relieving the -

off-goir.g shift and after relief back to the last shift  :

4 worked. i t

(b.) of the last shift prior to relieving the off-going shift-and after relief back to the last shift worked or the previous 3 days of logs whichever is the shortest...

1 a 4

5 (c.) of the last 3 days prior to relieving the off-going. shift and after relief the Shift Turnover comment Sheet.

(d.) back to the last shift worked prior to' relieving the off--

going shift and after relief the Shift Turnover Comment '

i ., Sheet. ,..

i ANSWER 4.01 4

(a.) [+1.0]

5

, Reference (s1 4.01 1

i 1. Palo Verde 1 and 2: Palo Verde Nuclear Generatina' Station Manual, " Shift Turnover", 40AC-9ZZ16, ' Appendix D,'p. 3. .

}

T

-Section 4.0 Continued on Next Page-i-

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

~ Page 52 Palo Verde-1 and 2 March 11, 1986

. Points

Available-OUESTION 4.02

~

A change in T,yg;is taken into account in the calculation of.

estimated _ critical rod position by (1.0);

(a.) making a Ap(T,yg) calculaiton if T,yg (projected) and T,yg (previous) are both 2 565'F.~

(b.) making a Ap(T,yg) calculation if T,yg (projected) and 4

, T,yg (previous) are both i 565'F.

(c.) making a Ap(power) calculation if power (projected) and power (previous) are both 2 0%.

(d.) making a Ap(power) calculation if Tav9 (pr jected) and l T,yg (previous) are both s' 565'F.

l ANSWER 4.02 (b.) [+1.0]  !

Reference (s) 4.02 l

~

-- 1. Palo Verde 1 and 2: Palo Verde Nuclear Generatina Station -r

! Manual, " Calculation of Estimated Critical Condition", ,

720P-9RX01, pp. 2-7.  ;

i .

E

i i

-t

-Section 4.0 Continued on Next Page-i A n y a

,y .,-..o,, ..v., - , _ - e--,, ,, , _ _ _ . . , , . . , , _ 4,.. - , m, -,.,__n,) ,,.4'4.. ,.,.L,.h,,. , ,,y

1 Page 53 Palo Verde IL and 2 March 11. 1986 Points Available _

4 QUESTION 4.03

. With Unit 1 in Mode 1 and one or more CEAs determined to be -

inoperable (1.0)

(a.) it is assumed that the shu~tdown margin is less than 6%

Ak/k if. one or more inoperable CEAs are immovable /

untrippable.

(b.) it is assumed that the shutdown margin is_less than 6%

Ak/k if two or more inoperable CEAs are immovable /.

untrippable and the shutdown margin is greater than 6%

Ak/k if only one is immovable /untrippable.-

(c.) boration of 40 gpm or greater of a solution of 4000 ppm or greater boron is immediately required if the. inoperable CEAs are still trippable after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

(d.) boration of 40 gpm or greater of.a solution of 4000. ppm or greater boron is immediately required if one of the inoperable CEAs is the most reactive rod.- ,

1 ANSWER 4.03 (a.) [+1.0]

t Reference (si 4.03

1. Palo Verde 1 and 2: Palo Verde Nuclear Generatino' Station Manual, " Shutdown Margin", 72ST-1RX09, pp. 5-7.

-Section 4.0 Continued on Hext Page--

Page 54 Palo Verde 1 and 2 '

March 11, 1986 Points Available .

QUESTION 4.04 Answer TRUE or FALSE. ~

Following a LOCA, the hydrogen monitors must be placed in service 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the event. (0.5)

ANSWER 4.04 FALSE [+0.5] in service within 30' minutes Reference (s) 4.04

1. Palo Verde 1 and 2: Palo Verde Nuclear Generatina Station Manual, " Loss of Coolant Accident", 41R0-1ZZ07, Revision 1, Section 3.4, p. 6.

'00ESTION 4.05 N /

Answer TRUE or FAL3E.

,e h At approximately 15% power the steam generator downcomers will t

c) fail closed and all feed control will be with the economizer. (0.5)

ANSWER 4.05

/ TRUE [+0.5],

Reference (s) 4.05

1. Palo Verde 1 and 2: Palo Verde Nuclear Generatina Station Manual, " Plant Startup Mode 2 to Mode 1", 410P-1Z204, Section 3.0, p. 23.  ;

t

-Section 4.0 Continued on Next Page-

l i l

i 1

Page 55 Palo Verde 1 and 2 l March 11, 1986 l Points Available .

QUESTION 4.06  ;

Answer TRUE or FALSE.

The spray-pond pump motor, rated at 4160 volts, should be limited to a maximum of four (4) consecutive attempts-to-start with the motor initially at ambient temperature. (0.5)

ANSWER 4.06 FALSE [+0.5] only two (2) consecutive starts Reference (s) 4.06 f

1. Palo Verde 1 and 2: Palo Verde Nuclear Generatina Station Manual, " Essential Spray Pond - Train A", 410P-1SP01, Revision 3, p. 11.

-Section 4.0 Continued on Next Page-

! Page 56- Palo Verde l'and 2

. March.11, 1986 .

, - Points Available ,

! OUESTION 4.07 In the left-hand column below is a list (itemized numerically)~ -

of " symptoms" from the procedure " Shutdown Outside Control Room". The right-hand column is a list (itemized alphabetically) of "causes". Choose one (1) item from the cause list for each item in the symptom list. Ilia each item from the cause-list .

i for one (1) item from the symptom list. (2.5)

Symotom Cause 1

l' 1. S/G level less than 0% WR- A. Condensible voids, col-T cold increasing above T sat lecting in RCS flowpath of S/G (T hat foll ws)

2. T cold and S/G pressure B. Inadequate RCS inventory j increasing

3. RCS Loop Delta T increasing - C. Inadequate secondary i and abnormal PLCS response steam flow I

4. T hat increasing and low D. Non-condensible voids i

l Pressurizer level collecting in RCS flow-path or physical blockage of RCS flowpaths i . ,

5. RCS Loop Delta T increasing, 'E . Inadequate secondary -

f no other indication at RCPs water inventory f

4 ANSWER 4.07 ,

l 1. E i 2. C -

'~

3. A

4. B

5. D t

Reference (s) 4.07 l 1. Palo Verde 1 and 2: Palo Verde Nuclear Generating Station Manual, " Shutdown Outside control Room", 41AO-1ZZ27, Revision 1, p. 58.

Section 4.0 Continued on Next Page-4 v- - -- , - - ,

• v-- -

4 e ,# . w y------w----- ,r,3.-,t e -,-m,--.., - ww , ---,- .m-.-.,. 4 -w, w - - - - * , * , , o.e m -w y n w ,w w -

Page 57 Palo Verde 1 and 2 ~

March 11, 1986 4'

Points- -

Available _

i OUESTION 4.08 Snecify when each of the tags listed below should be utilized; -

i.e., what does hanging that tag indicate?

a. -red tag (1.0)

b. yellow tag (1.0)

f. c. blue tag (1.0)

ANSWER 4.08

a. Red Danger Tag (Appendix D)

}. A red danger tag shall be placed on equipment (mechanical-or electrical) which, if'the equipment was operated, would endanger personnel or damage equipment. [+1.0]

b. Yellow Caution Tag (Appendix E)

The yellow caution tag or sticker is utilized where no I

danger to personnel is. involved. It shall state specific

. information which shall be understood and observed before operating the equipment. [+1.0]

j c. Blue Men-at-Work Tag (Appendik F) '

A blue tag is placed on system components'to indicate l

} that the equipment is to be onerated only by the direction of the individual to whom the tag is issued, after that individual has obtained permission from the Responsible .

Supervisor in order to ensure operation.of the equipment  !

does not affect the system's operation. .A blue men-at-work l tag is to be used used only for trouble shootina or testina. ~~

l l maintenance work /renair is not allowed. [+1.0]~ i i  :

Referenceisi 4.08

\ +y?'

1. Palo Verde 1 and 2: Palo Verde Nuclear Generating Station Manual, " Station Tagging and Clearance", 40AC-9ZZ15, p. 6.

i  :

/

j -Section 4.0 Continued on Next Page- l l

A

,m- - - , e--_, , . , , . _ . - . . , , , . ,-m , , , , . . .s , - . .,, . , - - ~ , , . , , , , . , , + ~ . . - + -, . , .

Page 58 Palo Verde 1 and 2 March 11, 1986 Points Available _

-i OUESTION 4.09 According to Procedure No 41RO-1ZZ05, " Loss of Feedwater," there -

is a preferred order or priority for which Auxiliary Feedwater Pumps are utilized - assuming.that there is no MSIS present.

List the preferred order as indicated below~. (1.5)

1. (pump to be utilized first)

2. .

3. (last pump in priority)

ANSWER 4.09

1. startup auxiliary-feed pump i 2. essential electric auxiliary-feed pump

3. essential steam-driven auxiliary-feed pump i

[+0.5 each] 3 ,

Reference (s) 4.09

1. Palo Verde 1 and 2: ,Palo Verde Nuclear Generatina Station 2

~

Manual, " Loss of Feedwater", 41RO-1ZZ05, Revision 1, Section 4.1, Note,'p. 9.

4 i

-Section 4.0 Continued on Next Page-1

'e Page 59 . Palo Verde 1 and 2 '

March 11, 1986 Points Available ,

QUESTION 4.10 List the four (4) criteria that must be met before safety -

injection (SI) flow may be throttled. (2.0)

ANSWER 4.10

1. RCS subcooled ) 28'F

2. Rx vessel level indicates void restricted to upper head

3. Pzr level ) 33% and controllable-

4. one (1) S/G capable of maintaining heat removal

[+0.5 each]

Reference (s) 4.10

1. Palo Verde 1 and 2: Palo -Verde Nuclear Generating Station Manual, " Reactor Trip", 41R0-1ZZ01, Section 3.0, p. 27.

'.s a.

1

-Section 4.0 Continued on Next Page-4.-

E

4 $

Page 60 Palo Verde 1 and 2

, March 11, 1986 ~

Points Available ,

QUESTION 4.11 Match the following (numerical) bergency classifications with the appropriate (alphabetical) description. (2.0)  :

c 1. Unusual Event A. Consists of e' vents which are in progress or have occurred which involve actual cr likely major -

failures of plant functions needed .

for'the protection of the public. $

(

Any releases are not expected to exceed Environmental Protection {

Agency Guideline exposure levels beyond the site boundary.

~

2. Alert B. This classification applies to events which are in progress or have occurred which indicate a potent'ial degradation of the {

b.

level of safety of the plant.

No releases of radioactive material i

requiring offsite response or monitoring are expected unless further degradation of safety systems occur.

3. Site Area Emergency C. , Consists of events which are in -

, progress or have occurred which "

l involve actual or imminent sub-stantial core degradation or melting with potential _ for loss of containment integrity. Releases can be reasonably expected to exceed Environmental Protection Agency Protective Action Guideline exposure levels offsite for more 4 than the immediate site area.

1 h

l g ...ct,on .. _ , _ d o _ ..g..

t U

b.

=.

lk~

Page 61' Palo Verde 1 and 2 .

March 11,~1986 Points-Available ,_ .

4. General Emergency D. This classific'ation consists of-events which~ are in progress or have occurred which involve an

• actual or potential substantial degradation o.f the level ofl safety of the plant.. Any releases'are expected to be limited to small -

fractions of the Environmental Protection Agency Protective-Action Guideline exposurelevels.

ANSWER 4.11

1. B

2. D

3. A

4. C 1

[+0.5 each] j Reference (s) 4.11

\

1. Palo Verde 1 and 2: Palo Verde Nuclear Generatina Station '

Manual, " Emergency Classifications". .EPIP-02, Revision 4, Section 3.2, p. 5. ~

~

-Section 4.0 Continued on Next Page-I

i Page 62 Palo Verde 1 and 2 .

March 11, 1986 Po'ints Available ,

OUESTION 4.12' If, during reactor startup and approach to criticality, the baron concentration were to be changed by more than 50 ppm, what action should the operator.take with the Pressurizer system? (1.0)-

ANSWER 4.12 Initialize Pressurizer spray to equalize RCS and pressurizer boron concentration. [+1.0]

Reference (s) 4.12

1. Palo Verde 1 and 2: Palo Verde Nuclear Generatina Station Manual, " Reactor Startup", 410P-1ZZO3, Revision 3, Section 4.3.12 Caution, p. 10.

OUESTION 4.13 Procedure No. 410P-1ZZ01, " Cold Shutdown to Hot Standby", directs you to reset both inadvertent-dilution alarms after the startup channel high voltage is turned off. Where are the controls to' '

accomplish this located? (1.0) ,

ANSWER 4.13 Behind B04 in the Miscellaneous'Eautoment Cabinet, IJ-ZJN-C06

[+1.0]

Reference (s) 4.13

1. Palo Verde 1 and 2: Palo Verde Nuclear Generatina Station Manual, " Cold Shutdown to Hot Standby", 410P-1ZZ01, Revision 2, p. 59.

-Section 4.0 Continued on Next Page-

Page 63 Palo Verde 1 and 2- .

March 11, 1986

~

Points Available ,

00ESTION 4.14 Explain why the differential temperature between the spray -

water temperature and the Pressurizer temperature should be maintained less than 200'F. (1.0).

ANSWER 4.14 to avoid unnecessary spray-nozzle fatigue [+1.0]

Reference (s) 4.14

. 1. Palo Verde 1 and 2: Palo Verde Nuclear Generating Station Manual, " Cold Shutdown to Hot Standby", 410P-1ZZ01, Revision 2, p. 49.

QUESTION 4.15 Eire the administrative control limits for whole body exposure. (1.5)

Weekly Quarterly i

Yearly " -

ANSWER 4.15

~

weekly -

300 mrem [+0.5]

quarterly - 1.0 rem [+0.5]

yearly -

4.0 rem [+0.5]

Reference (s) 4.15

1. Palo Verde 1 and 2: Palo Verde Nuclear Generating Station Manual, " Radiation Exposure and Access Control", 75AC-9ZZ01, Revision 4, p. 25.

-Section 4.0 Continued on Next Page-

Page 64 Palo Verde 1 and 2 March 11, 1986

. Points.-

Available _

QUESTION 4.16 Gire the beta-gamma area contamination limits for the following

1. contamination area (0.5)

2. high contamination area (0.5)

ANSWER 4.16

1. > 1000 dpm/100 cm2 -[+0.5]

2. > 50,000 dpm/100 cm2 [+0.5]

Reference (s) 4.16

1. Palo Verde 1 and 2: Palo Verde Nuclear Generatina Station Manual, " Radioactive Contamination Control", 75AC-9ZZ03, Revision 3, p. 11.

OUESTION 4.17

a. As specified by Tech-Specs., what is the minimum allowable  ;

, value for Te when operating in Modes 1 and 27 Neglect

• any special g}gt exceptions.' (0.5) b'

. lion much time is allotted to correct T cold befcre the reactor power level must be reduced? (0.5) ,

i ANSWER 4.17 1

a. 552*F [+0.5] l l

, b. 15 minutes [+0.5]

Reference (s) 4.17

1. Palo Verde 1 and 2: Technical Soecifications, p. 3/41-6. .;

i l

-Section 4.0 Continued on Next Page-  !

, -p, fy - ., -.7 ..,..--,

m

,. -r-i i

Page 65- Palo Verde 1 and 2:

. March 11, 1986 Points -

Available-

'0UESTION 4i18~

. According ;to the Unit 1 Tech-Specs., two (2)'out of three -(3) -

boron injection flow paths shall be OPERABLE when in Modes 1,

'2, 3 and 4. List these three (3) flow paths._ -

(1.5) 2

-ANSWER 4.18

, 1. A gravity feed flow path from either the refueling water tank or the spent fuel pool:through CH-536. (RWT Gravity Feed Isolation Valve) and a charging pump to the Reactor-Coolant System. [+0.5]-

i

2. A gravity feed flow path from the refueling water tank through CH-327 (RWT Gravity Feed / Safety Injection System Isolation Valve) and a charging pump to the Reactor Coolant'  ;

System. [+0.5]

3. A flow. path from either the refueling water tank or the

~

spent fuel pool through CH-164 (Boric Acid Filter Bypass Valve), utilizing gravity feed and a charging pump to the

! Reactor Coolant System. [+0.5]

Reference (si 4.18-

1. Palo Verde 1 and 2: Technical Snecifications, p. 3/41-8. '

J i

i

I i

~

1 4

1

-Section 4.0 Continued on Next Page- ,

. - _ _ , _ . . - - - - - , ..m ~;. _ . ._._. _ ._ ,_ . .~.,,_._. ...__,__ _ -.-_... . _ .- .. -_, - __ , ,.

i

• Page 66 Palo Verde 1 and 2 March 11,-1986 Points Available .

QUESTION 4.19 i List.five (5) pieces of information that would be found on a Radiation Exposure Permit-(REP) at PVNGS. (1.5).

ANSWER 4.19 .

1. Job description

2. radiological conditions at job location

3. required protective clothing

4. required dosimetry

5. required respiratory equipment

6. Job classification

7. dates and times of validation

8. special job instructions

[+0.3 each, +1.5 max].

Reference (s) 4.19 ,

1. Palo Verde 1 and 2: Palo Verde Nuclear Generating Station Manual, " Radiation Exposure and Access Control", 75AC-9ZZ01, l Rev,ision 4, p. 14.

-End of Section 4.0- .

-End of Exam-l 1

0

)

• , . . _ - . . , , - - _ _ . -_,.y -~ ,

.Y

. N CORE DATA BOOK

• J UNIT 1 CYCLE 1 PVNCS TRA.1NING-OPERAT RCVISION 001 SEPTEMBER 19 r

TABLE OF CONTENTS FIGURE REV. TITLE PAGE TABLE 1.1 00i' 'TTPICAL VRLUES" FOR QUICK REFERENCE TABLE 2.1.1 001 POWER DEFECT VS POWER LEVEL: BOC MOC. EOC 1

CURVE 2.1.1 001 5 TABLE 2.2.1 POWER DEFECT VS POWER LEVEL: SOC MOC. EOC 6 001 ISOTHERMAL TEMP DEFECT VS T-RVG: 550-565 F. BOC CURVE 2.2.1 001 7 TABLE 2.2.2 001 ISOTHERMAL TEMP DEFECT VS T-RVG: 550-565 F. BOC B CURVE 2.2.2 001 ISOTHERNRL TEMP DEFECT VS T-RVG: 550-565 F. MCC 9 TRBLE 2.2.3 001 I5OTHERMAL TEMP DEFECT VS T-RVG: 550-565 F. MOC 10 CURVE 2.2.3 001 ISOTHERMAL TEMP DEFECT VS T-RYG: 550-565 F. EOC 11 TRBLE 2.3.1 000 ISOTHERMAL TEMP DEFECT VS T-RVG: 550-565 F. ECC 12 BORON WORTH VS T-RYG. 505-615 F. 100% P. BOC 13 CURVE 2.3.1 000 BORON WORTH VS T-AVG. 505-615 F. 100% P. BOC TABLE 2.3.2 000 15 CURVE 2.3.2 BORON WORTH VS T-RYG. 505-615 F. 100% P..MOC 16 000 BORON WORTH VS T-RVG. 505-615 F. 100% P. MOC TRBLE 2.3.3 000 18 BORON WORTH VS T-RVG. 505-615 F. 100% P. EOC 19 CURVE 2.3.3 000 BORON WORTH VS T-RYG. 505-615 F. 100% P. EOC TABLE 2.3.4 000 21 '

CURVE 2.3.4 BORON WORTH VS T-RVG.70-565 F. 0% P. BOC 22 l 000 BORON WORTH VS T-RVG,70-565 F. 0% P. BOC TABLE 2.3.5 000 . -

24 l CURVE 2.3.5 BORON WORTH VS T-RYG.70-565 F. 0% P. MOC 25 000 BORON WORTH VS T-RVG.70-565 F. 0% P. MCC i

TABLE 2.3.6 000 - 27 1 CURVE 2.3.6 000 BORON WORTH VS T-RVG.70-565 F. 0% P. ECC 2B  !

TRBLE 2.4.1 000 BORON WORTH VS T-RYG.70-565 F. 0% P. EOC 30 EO. XENON WORTH VS BURNUP. HFP l CURVE 2.4.1 000 31 EQ. XENON WORTH VS BURNUP HFP -

32 TRBLE 2.4.2 000 EO. XENON WORTH VS POWER LEVEL. BOC MOC. EOC 1

CURVE 2.4.2 000 33 i TRELE 2.5.1 000 EQ. XENON WORTH VS POWER LEVEL. 80C. MOC. ESC 34  !

CURVE 2.5.1 000 REG GROUP 5 WORTH VS WITHORRWAL. HFP. BOC. MOC. EOC 35 TRBLE 2.5.2 000 REG GROUP 5 WORTH VS WITHORRWRL..HFP. BOC. MOC. EOC 37 CURVE 2.5.3 000 REG CER WORTH IOVERLRP) VS WITHORRWAL. 50% P. BOC. MOC. EOC 38 CURVE 2.5.4 000 REG CER WORTH IOVERLRP) VS WITHORAWRL. 50% P. BOC 40 CURVE 2.5.5 000 REG CER WORTH IOVERLRP) VS WITHORAWRL. 50% P. MOC 41 TABLE 2.5.6 001 REG CER WORTH IOVERLAPI VS WITHORRWRL. 50% P. EOC 42 CURVE 2.5.6 001 REG CER WORTH (OVERLRP) VS WITHORRWRL. HZP. BOC. EOC 43 TRELE 2.6.1 000 REG CER WORTH (OVERLRP) VS WITHORRWRL. HIP. BOC. EOC 4B PLCER WORTH VS WITHORAWRL HFP. BOC 50 CURVE 2.6.1 000 .

PLCER WORTH VS WITHORAWRL HFP. BOC TABLE 2.6.2 000 52 PLCER WORTH VS WITHORAWRL. HFP. MOC 53

- CURVE 2.6.2 000 PLCER WORTH VS WITHORRWRL. HFP. MOC TABLE 2.6.3 000 55 PLCER WORTH VS WITHORAWRL. HFP. EOC .

56 CURVE 2.6.3 000 PLCER WORTH VS WITHORRWRL. HFP. EOC TABLE 2.6.4 000 q 58 PLCER WORTH VS WITHORAWRL. 50% P. BOC .59 CURVE 2.6.4 000 PLCER WORTH VS WITHORRWRL. 50% P. BOC TABLE 2.6.5 000 . 61 PLCER WORTH VS WITHORRWRL. 50% P. MOC CURVE 2.6.5 000 PLCER WORTH VS WITHORRWRL. 50% P. MOC- ( 6264 TRBLE 2.6.6 000 PLCER WORTH VS WITHORRWAL. 50% P. EOC CUMVE 2.6.6 000 65-PLCER WORTH VS WITHORRWRL. 50% P. EOC 67 1ABLE 2.6.7' 000 PLCER WORTH VS WITHORRWAL. HZP CURVE 2.6.7 000 68 PLCER WORTH VS WITHORAWRL. HZP 70 TABLE 2.7.1 000 EQUILIBRIUM SHAPE INDEX VS BURNUP RT 100% POWER CURVE 2.7.1 000 70R EQUILIBRIUM SHRPE INDEX VS BURNUP RT 100% POWER 70B TRDLE 2.B.1 000 DELTR RSI RT B0%. 50%. 20% POWER.0 TO 208 EFPO CURVE 2.8.1 000 70C 1ABLE 2.9.] 000 DELTR RSI RT B0%. 50%. 20% POWER.0 TO 208 EFPD 700 CUhVE 2.9.1 000 EUUILIBRIUM IODINE VS POWER LEVEL.BOC THROUGH EOC 70E EQUILIBRIUM IODINE VS POWER LEVEL.BOC'THROUGH EOC 70F d

ASSIGNED COPY

__ PVNGSLA-7A i

.. ( P Wn ";*c-3d^).)

J . ..s v. ]' O/ '*p af 8A%"T"3 . ':.

e i CORE UNIT 1 CYCLE 1 DATA B DK),[h(Ql h [ h[g,2

, , PAGE 1 SEPTEMBER 24,.1985 PVNGS/ NUCLEAR SECTION REVISION 001 TABLE 1.1 (PACE ^ 1 0F 4)

(s._ _ " TYPICAL VALUES" FOR QUICK REFERENCE l l

CAUTION: These values are presented for a quick reference. The  ;

appropriate values from Section 2 or 3, or from the Operator

{

Assistance ProDrams should be used for calculations affecting i Reactor Operation.

1. Reactivity Effects of Temperatuve and Powee EE,f,960M M24M  ;

~

Overall power defect, O ) 100% power -1380 PCM -1950 PCM Moderator Temp. Defect, 0 ) 100% - 150 PCM - 700 PCM Fuel Temp. (Doppler) Defect, 0 ) 200% -1230 PCM - 820 PCM Power Redistribution Effects, 0 ) 100% 0 PCM - 430 PCM Reactivity defect from Hot Zero Power ) 210F +1660 PCM +2130 PCM l PCM Reactivity defect from 210F ) 80F +1950 PCM +2650 PCM PCM

(. MTC,.at 100% power PCM/F 5 PCM/F -25 PCM '

Fuel Temp. Coef., at 100% power - 1.3 PCM/F - 1.5PCM/F PCM/F

. Fuel Temp. at 100% power (effective)- 1342 F 1084 F The values above do not include effects of xenon change or PPM change.

(

2. Soluble Boron Worth ,

[

Hot full power -11 PCM/ PPM -12 PCM/ PPM Hot zero power -12 PCM/ PPM -13 PCM/ PPM 210F, PPM =2000 -13 PCM/ PPM -14 PCM/ PPM 210F, PPM =500 -15 PCM/ PPM -16 PCM/ PPM

, 80F, PPM =2000 -13 PCM/FPM -14 PCM/ PPM

3. Xenon Worth Equilibrium full power xenon worth ~- 2680 PCM -2860 PCM Peak xenon worth after tr.ip from 100% -5380 PCM -5560 PCM Time of xenon peak after trip 8.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 8.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> PLRNT CONDITIONS REFERENCE SOURCE OF Dana

0. - 1007. POWER V-CE-19010 0-455.7 EFPD 85-NA-hTM-528 COMMENTS: T121Z1A APPROVED.Fd$i'USE

~w-B $".98"

- - _ u ne s e s w me na_ __-

C*$"I 47'K _-

vane un en.uwwn pga 2 F '*

i UNIT 1 CYC.LEs 1.s . .. g .. E NUAR) 11,t934 i PVHGS/NUCLEJR, SECTION_y ). , =h . 3[ 3, Jyp=,k.

. RIVISION 000 :

TAB -1.1 PA GE 2 OF 4')

" TYPICAL VALUES" FOR QUICK R5'FERENCE. .

h.QL.- EL-

4. Control Red uerths -

Total Rod Worth, hot full power -18500 pcm -19600'pem Total Rod Worth, hot =ero power -17270 pcm A!!YTTME n! CYCLE Minimum Total Rod Worth- , hot =ero power -16400

, 210F -11200

, 80F ~10500 Minimum Rod Worth- with Worct Rod out,

, hot =ero power - 9810 h , 210F -

5920

, SOF . -

5510 E9k.* E2.'v ' -

Individual Rod Banic Worths, Hot full power SD Groups A & B (Reg.Grps 1-5 & P ARI) -13900 pcm -14800 pc.?

, pLCEA Group P (Reg. Grps 1-5 ARI) -

510 '550 Reg. Group 1 (Reg.Grps 2-5 ARI) -

1400 -

1540 .

j 2 (Reg.Grps 3-5 ARI) -

980 -

920 3 (Reg.Grps 4-5 ARI) -

960 -

1100 4 (Reg. Grp 5 ARI) -

480 -

440 5 (All other rods out) -

300 -

320 PLCHA Group P (all other rods out) -

240 -

270 Indiv:. dual Red Bani: Worths, hot =ero power

. SD Groups At B -13140 "

PLCEA Group P (Reg. Grps. 1-5 ARI) -

330 Reg. Group 1 -

1270 g 2 -

990 3 -

850 ,

4 -

' 420

• 5 -

260.

PLCIA Group P (all other rods out) -

230

• These rod worths were csiculated at critical boron concentrations for each rod coniaguration; this method causes the total Rod Worth to be somewhat high. Qp.,not p for Shutdown Margin calculations.

4 Then.w naninum Rod Wort.h values ri?. n21 include the worth of the PLCEAs  ;

c a pprc::: ma tel y 200-400 pcm). l l

l PLANT CONDITIONS RE~ERENCE SOURCE OF DATA j b 0 - 100T. POWER V-CE-19010 0-455.7 EFPD .J

, COMMENTS: T121218 l

• ^

J il u kst. A > 0 a v J L.s

uni s - 1 UTCLt.1 '1 Ji.NUARY 11. 1951

PVHGS/ NUCLEAR SECTION]'L]"j"C(N ] Q(PA TABLE"'171" Ki[ C "'Q GE 3]0F"W) .

REVISION 000

" TYPICAL VALUES" FOR QUICK REFERENCE l

5. -Rasetivitv Unit = and A-Effectiva BOL (M =n.00728) EOL r S =0.0052)  !

I pcm = 0. 001?: 4. p = 0.1394 1 pcm = 0.001% A p = 0.1925 l 1000 pcm s l.000: Ap = 139 $ 1000 pcm = 1.000% A p = 192?  !

716 pcm = 0 . 7 1 8 n A p = 1 0 0 $.- 520 pcm = 0.520% Ap = 100?'

- )

1 1

6. Stsrtur Rate .

P = Po

• e (t /T) T" '

A +ff ~8-

• Af SUR=Startup Rate (decades / min.

k = effective delayed neutro decay constant-(1/sec) l I

i SUR = 2j..OG o= 6 aff T ' (1+ A T) T= Reactor Period (seconds)

For BOL: g eif = 0.00713 K= 0.07754 see -1 i S"~ (O?!!) Period (zec) PC!' g d 1.0 26.1 237.7 0.2377 33.1 0.3 32.6 203.6 0.2036 28.4 j o.6 43.4 C 0.4 65.2 164.4 116.6 ,

0.1644 0.1136 22.9 16.5 j

0.2 130.3 64.7 0.0647 9.0

-C.2 130.3 -73.9 -0.0789

, -11.0

7. Beretion/ Dilution I Gallons of RMW to Gallons of Refuel 4ng Reduce PPM by 1 PPM Water to Increase' PPM ppm

• bv 1 PPM 100 836 19 500 153 21 -

1000 76 24 PLANT CONDITIONS RErERENCE SOURCE OF DATA 0 - 100T. POWER '

V-CE-19010 0-455.7 EFPD COMMENTS: T12121C '

~

• '~..

% w,L \ f ubu

.1 G W a au

UNF,1 CYCLE 11 ~

. JANUARY 11,1984 REVISION 000 PVNGS/ NUCLEAR ' SECTIOky]D.f*1"fTPA F GE 4'0,U}_)) K 7G ] p "4 TABLE .

" TYPICAL VALUES" FOR QUICK REFERENCE a

8. Dcc v %g Rate of Pr, duction ci decay heat (e.llowing shutdown from.iull power.

Tine Aft +r ShuMoun f: ef Full Power 1 second 6 1 minute 4.5 30 minutes 2.0 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1.6 8 .5ours 1.0 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> .7 48 heurs .6 1 morth- .2 4 months .02 W

9. Core *urnut ifn f tz ,

I-1 F.FPD = 38.4 MUD /7

10. Appro:-:imate relationship of change in Power Operating Limit (POL) to -

% changes in cortain core parameters.

A (?: POL) = 0.5

• A ( ?: D)bR)

A ( ?: POL) =-1.0 -

A (?: Pedial Peaking Factor)

. A r*: POL) = 0.7

• A ( *' Ocre Flow)

A (f: PCL) = 0.03

• A (Pritrary Pressure, psia)

. A '. !: POL) =-0.5 -

A ' Tin. F) "

( .

I

!r j

i i 4

PLANT CONDITIONS REFERENCE SOURCE OF DATA 0 - 100T. POWER V-CE-19010 0-455.7 EFPD COMMENTS: T1Z121D

( *

~

LT @w.J7 M YM "S t. 9 wwP . -E, dbe  !

W ld a@ ~'3*

JANUARY 27

? .= %

UNh 1h,[CLE'AR _SECTIOJ),byh % [ bDC'. -

==

REVISION 000 PVNGS/pUqLE TABLE 2.1.1

. POWER DEFECT VS POWER LEVEL

~~' ' (BOC, MOC, EOC)

POWER DEFECT (PCM)

- MOC EOC 80C 455.7 EFPO POWER 0 EFPO 208.3 EFPO L E V E L (*/ 1 O

O O

O m

-100 -90 5

-85

-196 -180 10

-170

-285 -270 15

-252

-365 -362 20

-331

-439 -452 'l 25

-408 '

-509 -540 30 -482

-579 -631' 35 -555

-653 -718 40 -625 ,

-734 -805-45 -695 .

-816 -896 '

9 50 -762

( ,

-893 -390 -

55 -829 b

-SGS -1084 60 -894

-1045 -1183 6!i -959

-1104 -1286 ,

70 -1021

-1172 -1398 ,

75 -1083

-1240 -150d 80 -1143

-1309 -1610 85 -1203

-1376 -1720 l

~

90 -1260 "1444

-1836 95 -1319

-1511 -1949 a 100 -1377 l

REFERENCE SOURCE OF. DATA PLANT CON 0iT10NS V-CE-19010 0 TO 100% P0hTR 0 TO 455.7 EFPD COMMENTS: T22121 A, IDENTICAL TO TABLE 3.1.1 h APPROVE 0 FOR USE BY:

DM ham -

h2 r/fY DATE 1 NUCLEAR SUPERVISOR ,

COPHRCi. LED B'f USER

^" ^"

GS/NUCLEA. SEC50{- ]'hl l I

.h h -

REVISIO CONVE 2.1.1

_ POWER DEFECT VS POWER LEVEL (BOC, MOC, EOC) J

-2000 ' '

' ' ' , e i 1 i ~ '

-1750i /

l l

/ (

/

-1500;' _"

./ ...

-1250i

/ \

[

,/ ,.

/ '.. . ' , ,/y

< '>y P -1000" ' '

i

l. ,.* '/ .

-750i 3

./

7 j/ ,[,

i ,

C.yf

-500"

' g 7

x 5

/ W:

5 ff

-250' j'7

. . /

/ -

0-

g. .g. ,j. . . . ,

.j.

.g. ,- .g 0 10 20 30 40 50 60 7C 80 90 )00

% FULL POWER LEGENO: EFPO +-+-+ 0 0 o-e 208. 3 *-a-6 455.7 PLANT CONDITIONS 0 TO 1007. POWER REFERENCE SOURCE OF DATA 000 - 455.7 ETPD V-CE- 19010 COMMENTS: C22121, IDENTICAL TO CURVE 3.1.1 APPROVED FOR USE GY: D M dea. k7'? /M NOCLEAR SUPEMISOR DATE k

e p% @w. n 3 a a J,- .J O e

, L .:s

.., PVNGS T 1 CYCLE "( '

• AUGUST 15196 REVISION 001 TABLE 2.2.1

(: 1SOTHERMAL TEMPERATURE DEFECT VS T-AVG

. (550 -565 F, 0% P, BOC) ., .

150THEP. MAL TEMPERATURE DEFECT (PCMI T-avg (F)

~

BORON-CONCENTRATION 600 PPM 1000 PPM 1500 PPM 2000 PPM 550 119.0 55.4

-11.3 -72.4' ,

551 111.3 51.8 -10.5' -67.6 . '

552 103.5 48.1 -9.8 -63.0

• 553 95.7 44.5 -9.1-

-58.2 . I 554 87.9 40.9 -8.3 -53.4

~~

555 .

80.0 37.2 '

-7.6 -48.7

556- 72.2 33.6 -6.8 -43.8 557 64.3 30.0 -6.0 -39.0

.i -

558 56.3 26.2 -5.3 -34.2 559 48.4 22.5 -4.5 -29.3 560  !

40.3 18.8 -3.3 -24.5 561- 32.4 15.1 -3.0 -19.6 562 24.3 11.3 -2.3 -14.8

$(

563 16.2 7.6 -1.5 -9.8 564 8.1 3.7 -0.8 -5.0 565 0.0 0.0 0.0 0.0 i

j . g 4

e t

I 2 PLANT CONO:TIONS i 07. POWER REERENCE SOURCC OF DATA 000 EFPD CALC. FILE- NFMNA-85001 l COMMENTE: T22221A, IDENT: CAL TO TABLE 3.5.1 wPPROVED FOTI USE BY: k-27"8T I

NUCLEAR SUPERVISOR .DATE ITC87

. - CONTROLLED 3Y USEi

I UN T 1 CYC 5 e  : a  ; . PAGE 8 l

'. PVNGS . AUGUST 15.1985

)

~ '

REY!SION 001 ,

CURVE 2.2.1 .

ISOTHERMAL TEMPERATURE DEFECT VS T-AVG (550 -

565 F, 0% P BOC) s 125l' ;j >

x ,

8 '

N '

y '

100,' N i' e

x ,

- i x 1 x

75' -

N

-  % \

N f. 1 x i

~

N x -

l CO'

• - l -

NA N m t N. , , ,

^w x

)

Nw x , e

'N. xi ,

25 * - "N K' ' ' ' '

'  % i N t '

i P '

A i

' N C NL M

' 6X 1 N N '  !

O ,' i hk.

, , sr i

~

% - ' .f ,  ;

, r

-25' ' # ~

' s

7 .

' )

, -- \

-50" -

2'_ --

/

A/~

- y

- p '

gg r l

J l

-100' .-' 1

)

4 550 552 554 556 558 560 562 564 556 i

T-HVG (DEGREES FI i LEGEND: PPM 600 *-e-+ 1000 -

1500 +-e-e 2000 P'. ANT CONDITIONS

• REFERENCE SOURCf OF DATA 000 EFPD CALC. FILE -NFAINA-65001 CC).!!.fENTS: 722221A, IDENTICAL 70 CURVE 3.5.1

( APPROVED FOR USE BY: bw/ bok NUC1 EAR SUPERVISOR T-27~ 6 DATE

. .Gw 3 . a - ..

..p w - uE .s w-

. _- s <.$. - . .

e

UN T CYCLE m e -

AUGUST 15 85 l

- REVISION 00* i TABLE 2.2.2

( ISOTHERMAL TEMPERATURE DEFECT VS T-AVC (550 -

565 F, 0% P, MOC) .. -

ISOTHEP. MAL TEMPERATURE DEFECT (PCMI BORON CONCENTRRTION T-AVG [F1 0 PPM 400 PPM 800 PPM 1200 PPM -

550 202.2 110.7 60.0 '11.1 551 189.2 103.7 56.4 10.8 552 176.0 96.6 52.7 10.3

. 553 162.8 89.5 49.0 9.8 .

554

149.6 82.4 45.3 9.3 555 -136.3 75.1 41.4 8.7 556 123.0 '67.9 37.5- 8.1 557 109.6 60.6 33.6 7. 5 ,

i 558 96.0 53.2 29.5

• 6.7 i

559 82.5 45.8 25.5 5.9 l 560 68.9 38.2 21.4 5.0 561 55.3 30.7 17.2 4.1 562 41.5 23.1 13.0 3.1 563 27.8 15.5 8.7 2.2 564 13.9 7.7 4.4 1.1 565 0.0 0.0 0.0 0.0 .

l i

4

~

?

PLANT CONDITIONS 04 POWER RETE *4ENCE SOURCE OF DATA 208 EFPO M C. FILE NFAINA-85001 CC!.U.!ENTS: 722022A, IDENT! CAL TO TABLE 3.5.2 APPROVED FOR USE By: bd 4-T7-d7 l NUCLIAR SUPERYlSOR DATE f

..CONTROLL:-D 3Y USE.R' ,

d

..PVNGS

[uhiS*CtEr '

. CONTRollID BY USIR .

,,,,,,g,;; - c l I' ' REVISION 00*

l CURVE 2.2.2  ;

i ISOTHERMAL TEMPERATURE DEFECT VS T-AVG \

(550 - 565 F, 0% P, MOC) , . -

.l 225 ", -

se i- . , , , , e ,

I l

200 "3

N '

A 4

(

\

175: ,

s -

j - ,

x -

w 1

• 150; ,

hN <

l -

x X ,

!. N 125' '.

.s p ,

i C "

\

i

' M

'N ^

100" N \

N N s

i '

N N- A

!s 7t w 75 " \ '

j-

-x s

,t s i .

% s

~% x \

i 50' N \

j -

w '

m 2 - ' N ,

(

% u N I -

N( l'( \ q i '% . N \

(

25" -

' '- A A -

'  % ,k N .

'N  % N byy ZhQ4 3 - -o --

4 r-O . . r r ., -

o- ymmia==qqhtf-m t,ii M "%4 550 552 554 556 558 i 550 562 564 566 i

1 1

T-RVG (DEGREES F1 j LEGENO: PPM 0 +-+-+ 400 - 800 +-+-e 1200 j PLANT CCNDITIONS i on POWER RETERENCE SOURCE OF DATA i

208 EFPD CALC. FILE NP.vNA-es001 -

CCtfifENTS: C22222, IDENTICAL TO CURVE 3.5.2 i

1 l % APPRGVED FOR USE BY: b k '27 4J' NUCIIAR SUPERVISOR DATE 1TC6 -

..CONTROLi3D STUSEPJ

CORE DATA BOOK

., . UNIT 1 CYCLE 1

,(i hh i h $ -

PAGE 11 PVNGS AUGUST 15.1SC5 REVIS!ON 001

.i -

TABLE 2.2.3 '

• ( -

ISOTHERMAL TEMPERATURE DEFECT VS T-AVG (550 -

565 F, 0% P, EOC) .. .

150THERMRL TEMPERATURE DEFECT (PCMI 8080N CONCENTRATION f T-AVG IF) ~0 PPM 400 PPM 800 PPM 1200 PPM

, 550 240.7 144.5 68.1 3.7 .

551 225.3 135.4 '64.0 3.8 . -- -

552 . 209.6 126.1 59.7 3.8 '

553 194.0 116.8 55.5 3.8 i 4

554 178.3 107.4 51.2 3.8

~~

555 162.5 97.9 46.8 3.6

^

556 146.6 88.5 42.4 3.5 i

557 130.7 78.9 38.0 3.3 558 114.6 69.2 33.4 3.0  !

559 98.5 59.6 28.8 2.8 i

560 82.3 49.8 24.1 2.4 l

561 66.0 40.0 19.4 2.0 I

562 49.6 30.0 14.6 1.5  !

t 563 33.2 20.1 . 9.8 1.1 564 16.6 10.0 4.9 0.5 '

t 565 0. 0 0.0 0.0 0.0 4

i

[-

t PLANT CONDITIONS 04 POWER REFERENCE SOURCE Or DATA 456 ETPO CALC FILE NF.\lNA-65001

, CC!.!! TENTS-. 722223A. IDENTICAL TO TABLE 3.5.3 APPROVID TOR USE BY: /PW k~U"I t NUCLEAR SUPERVISOR DATE

. , m ,$. n e =. e L,'.3h. hG d 8 V,2. C h m /Cqeae='s.dDS1

4

CCRE DATA BOOK

• UNIT 1 CYCLE 1

,'. OPUR'G.EE.Ff UER

• AUGUST 15,1925-Act 2  :

PYHGS

. REVISION 001 CURVE 2.2.3

~

i. ISOTHERMAL TEMPERATURE DEFECT VS T-AVG (550 -

565" F, 0% P, EOC) .

250 1 .

j N*

. . j X __

i 225 (

h

, 200 N. "

e j  %

i x i i

j 175 \ -

N 6

I 16 ,

xi e e 150

-s.

N

, x j i N ',  %

i 4 i 'N # N. .

• P 125 ** '

.' x' x 1 C

s. , ,

1 t

M x w .

lk 3 100 '

s '. s ik x j 'A li s 75 ,

,q s s

.g ,

j i 'N i L A

! e 6 .N '

N (

- 50 ' '

i i

~w w

l '.- g

-g s ,

4 w s x -.

x x. n i

25 '

x w

x .'

-s x

w w 1

3 02 ". ,-

,",*"

• E 8 ..

ET::=i-t D I .;

l

550 552 554 556 556 560 562 564 566 s

T-RYG (OEGREES F) l LEGEN0: PPM D +-*-* 400 - 800

• 1 m 1200 i

i i PL ANT CCNDITIONS RETE:tENCE SOURCE OF DATA-I 07. POWER CALC FILE t

456 ETPD NF.\tNA-85001 CCtU.fESTE C2:223. IDE:U!OAL TO CURVE 3.5.3 4

4 APPROVED FOR USE BY: 4W i NUCLEAR SUPERVfSOR DATE

ITC7 i

f,f)d

._.-___a... - . _ . _ _ _ . . . ,

J [h IDOf t

nu.l*3 2,I a['$v 3v$ wa

ow.u waia ovvn ~

PACL

. UNIT,1 CYCLE h q M m J 74 3*/ "UMA "It* *"3 . JANUARY 27,19i

' ' :PVNJS/NUCWE;CIlQNA '

. - REVISION O(

TABLE 2.3.1' (PAGE :1 OF 2) s

' BORON WORTH VS T-AVC

.(505 - 615 F,100% P, BOC) ,

BORON WORTH (PCM/ PPM) 80RON CONCENTRRT10N-T-RVG IF) 600 PPM 800 PPM 900 PPM 1000 PPM- 1200 PF 505 -12.63 -12.44 -12.35 -12.26'- -12.0E 510 -12.57 -12.38 -12.29 '-12.'20 -12.02

~515 -12.50 -12.31 -12.23 -12.13 -11.95 520 -12.44 -12.25 -12.26 -12.07 -11.90 525 -12.37- -12.19 -12.10 -12.01 -11.84 530 -12.31 -12.12.. -12.04 -11.94 -11.78' 535 -12.24 -12.05 -11.97 -11.88

~

-11.72 540 -12.17 -11.99 -11.91 -11.82 -11.65 545 -12.10 -11.92 -11.84 -11.75. -11.59 550 -12.02 -11.85 -11.77 -11.69 -1 '2 555 -11.94 -11.78

-11.70. -11.62 -11.46 560 -11.86 -11.70 -11.62 -11.55 -11.3S 565 -11.78 -11.63 -11.54 -11.47 ,11.32 570 -11.70 -11.55 -11.47~ -11.40 -11.24 575 -11.62

-11.47 -11.38 -11.31 -11.18 580 i

-11.55 -11.38 -11.30 -11.23 '

-11.08 585 -11.44 -11.29

-11.21 -11.14. -10.99 590 -11.34 -11.20 ' -11.12 -11.05 -10.90 595 -11.24 -11.10 -11.02 -10.95' -10.81 600 -11.14 -11.00 -10.92 -10.851 -10.72 PLANT CONOTTIONS REFERENCE SOURCE OF OtTA.

.100Y. POWER 0 EFPD - V-CE ' '10 COMMENTS: T223Z1 A, IDENTICAL TO TABLE 3.8.1

• APPROVED FOR USE BY:  % ..../ h ys,s._ da.7/rf NUCLEAR SUPERVi5OR DATE

. .rm

't"'O Q j ;' ,j f 5z% s s '= r~.) % L $ ' ' + C %v  % f 4 K)O $A

,, ., , , . . ........-.--e--

.- . CORE DATA 900K - PAGE 20

- UNIT'1'CYCLEli 3811 T]' :(( lj*$~p} -

- JANUARY 27,1984

' REVISION 000 PVNGS/ NUCLEAR SECTION*=

TABLE 2.3.3- (PAGE 2 OF 2)

, BORON WORTH VS T-AVG (505 -

615 F,100% P, EOC)

BORON' WORTH (PCM/ PPM)

BORON CONCENTRRTION T-AVG IF) O PPM 100 PPM 200 PPM 400 PPM 605 -11.79 -11.71 -11.64 -11.48 610 -11.S6 -11.58 -11.51 -11.35 615 -11.52 -11.44 -11.37 -11.22 i

t t

.o . '

. l PLANT CONDITIONS REFERENCE SOURCE OF DATA l

% ~ ,

1007. PonTR V-CE-19010 l 455.7 EF'PD .

1 COMMENTS: T2Z3239, IDENTICAL TO TABLE 3.8.3 ja y . % 7

• 9 ,. % l 'j ' ;.%, ,p r y g!7 j

- M T1MSA'dFh:".d d :'-- LJ' Sh O----.-- .--m~ -- -

wiu e i wew 6 .

REV!!'ON 00C

]

. PVNGS/ NUCLEAR SECTION e wem A 1 1 ,= % . y 7 3 3 ,, . '

~ ~ J iLtC-UR VE 2.3.340 .

BORON WORTH' VS T-AVG: .

(505 * - 615

• F, 100% P, EO C)

( . , , , , , ,.

- 14 . 0 0 .' ' ..

N

-13.75-

. N

- y '..N w, . w .

} .,7,\

-13.50, , ,,

. N s, - ,N

-13.25- X N.'

.'s

. N . J-nN

-13.00' N',*,

- -N N m s p . g , i' .N C -12.75- N 'J N M . N -

J .N  ;

'I'D P -12.50' - \ \' 'N E

R .

- N N' \ -

-12.25- N '

N l 1 P . l N N;..Ni P .

{ N,, N 'A i M -12.00, y,y

. . A NNA

-11.75- A ' '- \

's x;.. X C -11.50

- \ ' '

\ '.' m

=

l Q h

'% 1

-11.25- t t i i I

.i

" I

' I I I I

-11.00 * .. ..

510 520 530 540 550 560 570 580 590 600 610 620 500 T-RYG (DEOREE5 F1 LE0END: PPM +-+-+ 0 n.e-a 100 .*-di-* 200 m 4d0 I

PLANT CONDITIONS REFERENCE SOURCC OF DATA 100T. Pon'ER V-CE-19010 455.7 EFPD

. COMMENTS: C22323. IDENTICAL TO CURVE 3.8.3 APPROVED FOR USE BY: Deed & - A7/IV DATE -'

/

HUCLEAR SUPERVISOR Q.

.. . . . . . y 3 ~ .

. b .1. 4 .L u n i .e e ; O. .2. w i. 1 2p _ . 34h. -.,._ , . . ., _ .

' CORE DATA BOOK '

PAGE 22 '

UNIT 1' CYCLE ll (99 l ~:"=h -(( h, .i=k, ~ JANUARY 27. 1984

' PVNGS/ NUCLEAR SECTION" REVISION 000 TABLE 2.3.4 (PAGE 10F 2)

, BORON WORTH' VS T-A VG (70 - 565 F, 0% P, BOC)

BORON YORTH (PCM/ PPM) 80RON CONCENTRATION T-AVG (F) 600 PPM 1000 PPM 1500 PPM 2000 PPM 70 -15.02 -14.35 -13.68 -13.18 80 -14.99 -14.32 -13.65 -13.15 100 -14.92 -14.26 -13.59 -13.10 120 -14.85 -14.19 -13.53 -13.05 140 -14.78 -14.13 -13.47 -13.00 160 , -14.71 -14.06 -13.41 -12.95 180 -14.63 -13.99 -13.35 -12i89 200 -14.55 -13.91 -13.28 -12.83

220 -14.46 -13.83 -13.21 -12.76 240 -14.37 -13.75 -13.13 -12.69

', , 260 -14.26

-13.65 -13.04 -12.61 280 -14.15 -13.56 -12.95 -12.52 300 -14.04 -13.46 -12.86 -12.4.3 320 -13.92 -13.35 -12.76 -12.33 t

340 -13.79 -13.24 -12.65 -12.'23 L

360 -13.66 -13.12 -12.54 -12.12 380 -13.52 -12.99 -12.43

, -12.01 I - * * '

400 -13.38 -12.86 -12.30 -11.89

. 420 -13.22 -12.72 -12.17 -11.76 440 -13.06 -12.57 -12.03 -11.62 PLANT CON 0!TIONS REFERENCE SOURCE OF Ot.TA j OT.' POWER 0 EFPD V-CE-19010

( COMMENTS

T223Z4A, IDENTICAL TO TABLE 3.8.4 APPROVED TOR USE BY: D dw 3 [s.7 /r V ,

NUCLEAR SUPERYlSOR DATE  !

.s.* '

Ya* p.- . - h,,ub";^y A L*:'d*

R *;* % .d.W f l W '

_ .. n e.m n. .m -~ ----~.  ? ' .'l'"*

- . 2Y.*. :h.1--~~.-.s r- -- m r mW ".' ~~ _ _ _

n. c CORE DATA BOOK F AGE 23

,.' UNIT *1 CYCLE 1' O.TN l 3'[ dh - JANUARY 27,'1984

- PVNGS/ NUCLEAR SECTION** REVISION 000

~

TABLE 2.3.4 (PACE 2 OF 2)

BORON WORTH VS T-AVG

( (70 -

565 F, 0% P, BOC)

BORON WORTH (PCM/ PPM)

BORON CONCENTRRTION T-RVG IF) 600 PPM 1000 PPM 1500 PPM 2000 PPM 460 -12.88 -12.41 -11.88 -11.48 480 -12.69 -12.23- -11.72 -11.33 500 -12.48 -12.04 -11.55 -11.16 520 -12.24 -11.83 -11.36 -i0.98 540 -11.98 -11.60 -11.16 -10.79 560 -11.70 -11.35 -10.95 -10.58 ^

~

565 -11.63 -11.29 -10.89 -10.53

(

0 PLRNT CONDITIONS

('.

i REFERENCE SOURCE OF DA.  !

0% POWER 0 EFPD V-CE-19010 2 COMMENTS: T2Z3Z48, IDENTICAL TO TABLE 3.8.4 k 1

......;. P A - "'.}

p"'gAI'7'$

wd a ";7 'l 7["'I'

.',._._" " ' ".jhD

- - i.----- . - -- -- o- ve v --

n.... . 4

UNil 1 CYCLE 1 FAbh. 2 JANUARY 27,198, TVNGS/ NUCLEAR SECJION mm . n 1 m V f 7 M ". REVISION 00(

wfU Y-CURVE *2.3t4"Q ,

' BORON WORTH VS T-AVC:

(70 * - 565

• F, 0% P, B O C)

-15.0

s'N m l

'N i l s

-14.5-u-

\ l N

-14.0 N l

,g, a s_ ', ,

N

-13.5- '" '

s,.

N-u .,,

N .

p C

m, .

y N -13.0

~N' ',

' N

- ,N w m '. s ,'

I_ Ds l

{

N N-(

R -12.5- '

N L l N 's ,

N. ,  % -

f l p

N s l - ,

'Nj

~

M -12.0 '

sN sx .

-11.5-N 's, N.c 1 l A -

\ A., '.l

\

s b

-11. 0 ' .

I \l h,

-10.5- NJ l l l i l ,.P 50 100 150 200 250 300 350 400 450 500 ESO T-AVO (OEGREES F) i g LEGEN0: PPM .*- +-+ 6 0 0 o-o-e 1000 e-6 -6 1500 m 2000 PLRNT CONDITtONS

07. PO NER REFERENCE SOURCE Or DATA 000 EFPD V-CE-19010 COMMENTS: C22324. IDENTICAL TO CURVE 3.8.4.

APPROVED FOR USE BY: D kN- 2 /17 //V

}4UCLEAR SUPERVISOR DATE L -

ff s* .g "'

1. 'mi ' .I i g f.ss ,

? hh'. ~. , **.k '* -[.?.?N,I ,,d% * 'g M '12.*.*$. . t*Lh. .- J !. 's 2.*b,# ., me -. o.e-m ~= e . - ~y.e---.~3 -r*

ow.u. mia,Ivun

.}'[ )) {' ]

• ' ' , . UNIT 11 CYCLE-1).p N 1 PAGE 21-PVNGS/ NUCLEAR 'SECTIONg= JANUARY 27.198<

.' REVISION OOC

~-

TABLE 2.3.5 (PAGE 1- OF 2)

BORON WORTH VS T-AVG

( (70 -

565 F, 0% P, NOC)

BORON WORTH (PCM/ PPM)

T-RVG IF) 80RON CONCENTRATION O PPM 400 PPM 800 PPM 1200 PPM 70 -16.29 -15.52 -14.85 -14.28 80 -16.23 -15.47 -14.81 -14.24 100 -16.11 -15.37 -14.72 -14.16 120 -15.99 -15.26 -14.63 -14.08 140 -15.87 -15.16 -14.54 -14.00 160 -15.75 -15.05 -14.44 -13.92 180 -15.62 -14.93 -14.34 -13.~83 200- -15.48 -14.81 -14.23 -13.73 220 -15.33 -14.68 ~~

-14.12 -13.63 240 -15.17 -14.54 -14.00 -13.51 260 -15.01

-14.40 -13.86 -13.39 280 -14.83 -14.24 -13.72 -13.26 300 -14.65 -14.08 -13.57

- -13.14 320 -14.46 -13.91 -13.42 -12.99 340 -14.26 4

-13.73 -13.26 -12.84 360 i

-14.06 -13.55 -13.09 -12!69 380 -13.85 -13.36 -12.92 -12.53 400 -13.63 -13.16 -12.74. 12.36 420 -13.40' -12.95 -12.55 -12.18 i 440 -13.16 .

-12.73 -12.35 -11.99 PLANT COHOITIONS

_ REFERENCE SOURCE Or .;ATA 0;; POWER 208.3 ETPD V-CE-190

{. COMMENTSi T2Z3Z5A, IDENTICAL TO TABLE 3.8.5 APPROVED FOR USE BY: DM 0777=

NUCLEAR SUPERylSOR ML7/tf DATE

... .- i rn m ,.e 'i =' 9 '- T,C %'S .

G,%hs.i il .' b ;'f p, ~L.u. .-D---- : ~' e- .~"~';"~n

. ... . . . . . .-. . . - - - --- - --- ~%.%u . "'. '

' _ UNIT 1 CYCLE 1 PAGE 30 JANUARY 27 1984 PVNGS/ NUCLEAR SECTION 4"7' wh (^j l 'M [l$h* REVISi$H00C

• CUR VE 2. 3]. W^*

• BORON WORTH VS T-A VG:

(70 -

565

• F, 0% P, EO C)

-20 ' ' '

i i e 6 s

-19.' '

. I I I

• 6

! I l

-18 : ' ' '

N N '

s G- N

,.. x u l ,

p . . . w. . i i 6 C

M

'"' \ ' ' '

i -16' N - ' '" '- ' ' ' '

p -

'm -

. N , i i ,' i E

'^

N

"'- N ' ' '

R "- N '

N -

s r, , x i p .gg - N '

'_ '~'. \ l l

p - '

N s . x

! g -

A 'N **. A

.N_ w , . N m, ( x N

_gy-q x - x i -

x -s '

y ,

tN  %. t #N' e i x iN 6J Ys , .

N. _ i N, F.! Ni i #.

! -13' N!

- - E i \ '- '

- N s #.

x N 1. . 6N

. . N N es t N Ni,

, -12? \ \ C i

- <x w t 4

N 9 N

i 60 120 100 240 300 .160 420 480 540 T-RVG (DEGREES F1 LEGEN0: PPM + - - + 0 0 -EF-G 400 800 6-6 -6 m 1200 L PLANT CON 01TIONS 0". POWER REFERENCE SOURCE O~ DATA 455.7 EFPD V-CE-19010 COMMENTS: C22326. IDENTICAL TO CURVE 3.8.6 APPROVED TOR USE BY: hw b n.- 2/2-)/W NUCLEAR SUPERVISOR DATE

b N 'r%1'" s 2 .1 :*5,fJ

,,, l' wwG=J 3 TD ' '*2) ,,y.&

/ 5 '4. :.=..DQ'73 ._. .. --. .--~ ~~~ _ . -- _ _ .

~

. . uni [1 dhblEZ , bL m PJNGS/NUCLEMJEQIlQN c.*

"7 JANUAR"h; 27=[ [dD'L RWisidN

?

TABLE 2.4.1 EQ. XENON; WORTH VS BURNUP (HFP)

( BURNUP (EFPD) 0.0

XENON WORTH (PCM) 0 1.3 l -2680 13.0 -2800 26.0 -2780 *-

52.1 -2750 78.1 -2740 104.2 -2720 h

130.2 g

-2730 156.3 * -2730 .

182.3 -2720 208.3 -2720 234.4

( "

260.4 ,

-2740

-2750 286.5 -2770 312.5 -2790

, 338.5 -2800 364.6 -2820 9 390.6 -2840  ?

416.7 -2860 429.7 -2860 442.7 -2860 455.7 -2860 PLANT CONDITIONS REFERENCE SOURCE OF D 1005 POWER

{. COMMENTS: T2242M, IDENTICAL TO TABLE 3.11.1 V-C 38.4 MWD /T = & El

'S APPROVED FOR USE BY: M Onf.w .

s/L7/( 8(

NUCI.f.AR SUPERVISOR DATE

.: w re , , r mm;.s._137.,

w.4 ws 3"i; x ;P- f.ic.:7

~

2Ta__.wm+a

--- - - .~

UNIT 1 CYCLE 1 JANUARY 27, is

~ PVNGS/ NUCLEAR SECTION . ,,. . ) 3.  % . p REVISION O

'w Y J .I S ' @@R5[ &,-4',Df*

• EQUILIBRIUM XENON WORTH VS BURNUP (HFF

-3000 ' '

I II_

M

_ m .i i l l

-27504 / ,' E :_ I -

gl

, I ll l lllll lll ll

-2500' I II li i 11 ll11 l l-l

-2250- ,

l ll l 1 1 I I l

~

-2000 -

!III III l l l l

!! Ilill 1

-1750-l llllllll l

~

l 1

'llllllll P -1500  !  ! '! II III

'C i l l l l llllllllllllll 1 I I I l' l 1lllllllll

-1250- 1,ll

)

l l s -1000  !  !!!II i l i 111lllll l I iiliIIII

. -750- '

i l llll ll 1 '

~

L lilli ll

-500 -

II

,1

-250-II yl l

O '. . I I  !

O Sb 120 180 2b0 3b0 3b0 yhD yhD

, BURNUP (EFPD)

PLANT CONDITIONS REFERENCE SOURCE OF DA' 2007. POWER V-CE- 19C COMMENTS: C2Z4Z1, IDENTICAL TO CURVE 3.11.1 38.4 MWD /T=1 EF APPROVED FOR USE BY: -

'DL r ND+ w 2/27 /P86 NOCLEAR SUPERVISOR ' DATE L

---

.--.itriMJ-,;i"'%")rF.l^.7 8  : i."

- . . . . . . - . . . . . . . i Cr ' *.1 D 4-

' a:-hr l/. ) %@.*-~<-

u. .~

- --m- -- --

.. . CORE DATA BOOK . ppgg 33 UNIT 1 CYCLE 1* 9 811 3 5f.]3Q *, JANUARY 27,1984 PVNGS/ NUCLEAR SECTION .

4 h34as Sd REVISION 000 TABLE 2.4.2 '

a' EQUILIBRIUM XENON WORTH VS POWER LEVEL

~ --

' EQUILIBRIUM XENON WORTH POWER BOC HOC 'EOC LEVEL t%) O EFPD 208.3 EFPD 455.7 EFPD-0 0 0 0 5 -650 ' 665

-685 10 -1040 -1100 -1165 15 -1330 -1415 -1510 l 20 -1540 -1630 -1730 25 -1700 -1800 -1900 l

30 -1830 -1940 -2065 35 -1930 -2060 -2200 _

40 -2070 -2160 -2290

~

45 *

-2160 -

-2245 -2385 50 ' -2250 -2315 -2460 55 -2310 -2385- -2520 60 -2370 -2440 '2590

. 65 -2440 -2500 -2640 70 -2540' -2690'

-2490 l

l 75 -2530 -2570 -2130 l 80 -2560 .-2620

-2}70 85 -2595 -2555' ,-2810

. , , . 90 -2620 --2680 -2840 95 -2645 -2710 -2850 100 -2670 -2725' -2865-PLANT CONDITIONS REFERENCE SOURCE OF DA7' 0-455.7 EFPD

'h V-CE-190 i

COMMENTS: T22422A, IDENTICAL TO TABLE 3.11.2 '

APPROVED FOR USE BY: L ben-NUCLEAR SUPERYlSOR d27[fY DATE.

y g-e a p m =.

= .# g m- w, ==%

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

.'? .L ~ ' --0*- - -* m-~- m~" ' '~

uma e civLt i JANUARY 27, 1984 REV:SION 000

,O k,u.,.. ave .ey .2.w.2B3d PVNGS/ NUCLEAR SECIIO4 s v N ax .

~ EQUILIBRIUM XENON WORTH VS POWER LEVEL: BOC, MOC, EOC

-3000 .

e l

~

_ .e-----t

-2750- -

e' l _,.m t

,. s*" ,, . . .y - *} s

_ggon ' /

• 1. ..n-$ d f

/ ,.a: ~' /'

~

^

/ ,

,.*- /

-2250- ,. ,

f

/ ,P'/ ,3 l

~ ' '

-2000 'p.- j ,

i

,. ,. y

-1750- jj'/ l ltll l

P -1500' ' i i l-C .

/,k'/ l iy

-1250- ///

W,

-1000' d

~

Il

-750- g

. .  ?

-500

~

l L I

l (

-250- /

l  ;

~

0  : l l

.- r .- ... ., .

0 10 20 30 40 50 60 70 80 90 100

% FULL POWER LEGEND: EFPO -.-+ 0 o-o a 208.3 c-o-a 4 55. 7 PLANT CONDITIONS REFERENCE SOURCE OF DATA 2007. Pon'ER - -V-CE-19010 COMMENTS: C224Z2, IDENTICAL TO CURVE 3.11.2 APPROVED FOR USE BY: h /[v fe 7,fg /pif 4

NUCLEAR' SUPERVISOR 'DATE L .

A;

'> = - :' i a A ~ -, +' i- a OL

CURE DATA BOOK i .

UNIT 1 CYCLE 1,=9 8 ] . PAGI 35 JANUARY 27,1984 '

PVNGS/ NUCLEAR SECTION =-] g "Qy ]']["%

REVISION 000 =

t

\ ,.

TABLE 2.5.1 (PACE 1 OF 2)

REC GROUP 5 WORTH VS WITHDRAWAL (HFP)  %

( REG GROUP 5 POSITION BOC REG GROUP 5 WORTH (PCM)

MOC EOC 3 -"

(INCHES WITHDRAWN) 0 EFPD 208.3 EFPD .

455.7 EFPD Z O.0 -303.8 -295.1

-319.9 3 5.0 -302.1 _

-293.5 -318.3 10.0 -298.5 -290.1 -314.5 g 15.0 -293.0 -284.9 -308.9 J 20.0 m

-285.5 -278.0 -302.0 9 25.0 -276.3 M

-269.5 -294.y c 30.0 -265.6 35.0.

-259.9 -286.6 $m

-253.8 -249.6 -278.9 40.0 -241.1 -238.7 -271.2 mm 45.0 -227.7 -227.3 -263.5 g

50.0 -213.8 -215.7 -255.8 9 C 55.0 -199.7 -203.9 -247.9 i

q 60.0 -192.0

~

-185.5 -239.8 N 65.0

-171.2 -179.9 -231.4 -

~ ,

2 70.0 75.0

-156.9 -167.7 -222.5 39

-142.7 -155.3 -213!3 9 80.0 -128.6 -142.8

-203.5 85.0

-114.8 -130.1 -1.93.2 90.0 -101.3 -117.3

-182.1 95.0 -88.1 -

-104.4 -170.1 _

100.0 -75.3 -91.5 -157.1 PLANT CONDITIONS 100% POWER REFERENCE SOURCE OF DATA 0-455.7 EPD' V-CE-19010 3

- ALL RODS OUT

.( COMMENT 51 T225Z1 A, IDENTICAL TO TABLE 3.12.1 -

APPROVED. FOR USE BY: DM bee e ---- "L-[27[l'f NUCLEAR SUPERvlSOR DATE ^

" * ~\ 1" J 38 % .3 V.Of ,1 Z TI.'3

%..n %R ) 5 '-$'.*v:.YAi.:"k ,'. l h:i %..p.{Ji&?.: ------ -

CORE DATA B00X COMTROLLED 3Y USER '

PACE 48 UNIT 1 CYCLE 1 JULY 29,1985 PVNCS/NUOLEAR SECil0N REVISION C01

_ CURVE 2.5.6 REC. CEA WORTH (OVERLAP) VS WITHDRAWAL (HZP, BOC, EOC)

-95x uiii.6, ,,, , , ,

1 tao /J t i

\i -l 1 i i '

i I

)

-4000 l ,

6_1 w t ' ' '

I Mt t i t t  % e 1

-3500 l ,- -

,4 ., W 4 4

6i is i  % 6 4 6 6 4t t 4, e ' >

s t #

-300,0 ee s \,6 se eeei

, es

,, %, y %g ,

r , , ,, ,,

e, ti6 m , u 66 il ' 6 ' ,

4il ti P, h 6 l 'l 'i 1 i 4

!ie i ii  %)) .

ti 6 i 4 4 -

R ie4 , e . % 'm 6A- ei '

O -2500 ,, , , , ,I , iy gest y , , , , , , ,,

, ,', , ,t I

D i i t ' 6 '

i -

iN i 4 I i i *

( . 'N T e i $

W 6 k i 4 T

- 2000 , ,

t I:CC 5

\ 1 4

l i

l H --

i '

e ls .

6 i , o

( l A I -150"* '\ '-

i, e, i ei a i , i6 i P 6t 81 it *6 %'t '

6'6i t l

C * *i ' 8 'i$ '

• a ' ' * **it i i

- M i eei 6' i s e t is t

't . 4 o sl l 6 i l

-1000<-- l ,'" ,l ' '

w , g N 17A 1' I "4m i

'% 0 g

-500 ' ' ' '

N '

I .% r -  %

i t #  % 4  %

i g

l ti N . It T.

I '

ie %t i N 6 i e6 l 4i % 1

0. CROUD 1 . . . CROUP 3 . . . CROUP 5

60. 90 120 1$0 0 30 60 so 120 150 0 30 0- 30 '60 '90 '120'150 OROUP 2 GROUP 4' d 3U 66 90 120 150 d 35 60 90 120 150 CEA POSITION (INCHES WITHDR.WN)

FLONT CONDITIONS REFUENG SOURCE F M TA HOT ZERO POTER 84-NFM-205 & V-CE-19010 BOC = 000 EPPD. ECC = 455.7 ETPD COMMENTS: C225Z5. IDENTICAL To CURVE 3.12.6

( APPRCVED FOR USE BY: ~

NUCLEAR' SUPERY150R DATE

. -me -< . .q y-= >g .

b. t a .i ; . $.$ ; .=_e p q }g '7. t.

• g,)q]}

d d 2 sw. .

y Vi ; ]" "jhy*\]

ja= e *\

w s ud] ~",",7

hV lU?'o uDe.J d t a -

CORE DATA BOOK UNIT 1 CYCLE 1 PAGE 49 JANUARY 27,1984 PVNGS/ NUCLEAR SECTION

(

REVISION 000 CUR VE 2.5. 9 REG CEA WORTH (CVERLAP) VS WITHDRA WAL (HZP, EOC)

-2400 --

g

\

$ 1

-2000 L

\

\

\, 1

-1600 '

R O t  %( *

. D .

l 3 (

y -1200 o

\

• g R ht l-T H N l (I -800 -

%  % l

[

M h l Nl

-400 ~'

s% . ,

N . (

\

0

. GROUP. 3 N

, GRO,UP 5 ,

0 30 60 90 120 150 0 30 60 90 120 -150

. CROUP 2 .

CROUP 4. .

90 120 150 0- 30 60 90 120 150 CEA POSITION (INCHES WITHDRAWN)

PLRNT CCNDITIONS

- HOT ZERO POWER REFERENCE SOURCE OF'0ATA 455.7 EFPD V-CE-19010 COMMENTS: C22529, IDENTICAL TO CURVE 3.12.9

\

APPROVED FOR USE BY: h-/ b#ff/ W h27 /F'f NUCLEAR SOPERVISOR 'DATT.

v, .s i

- .~ y, %.

. j ," i : J { v ,5sdfa m.

' g :UJ"

• r ;d.l)3* *=5 "3 h-

k' e

, s EQUATION FORMULA AND PARAMETER SHEET -

Where mi = m2

( den s i ty)1 ( v el oci ty)1 ( a rea )1 = ( den si ty) 2 ( vel oci ty ) 2 ( a'rea ) 2 KE = mv2 PE = mgh PEi +KEl +P1Yi = PE +KE where Y = specific 2 +P2 Y22 "li volume P = Pressure Q = mcp (Tout-Tin) Q = UA (T,y,-Tstm l - 0 " 5Ihl-h2)

P = Po10(SUR)(t) P = Po et/T SUR = 26.06 T T = (8p - oi Eg = j( Ay - mi 1 0.08 sec- 1

, ___________________________________________..__..e.f_f__= -_-------------------

delta K '= (K,f f-1) CR 1 (1-Keffi)=CRj(1-Keff2)-

, CR = S/(1-Keff)-

M = (1-K,ffi) . SDM = (1-K,ff) x 100%

I Il-Keff2) K eff-decay constant = in (2) = 0.693 A =.Ag e-(decay constant)x(t) 1

t t 1/2 1/2 q.

f ' .

Water Parameters Miscellaneous Conversions 1 gallon = 8.345 lbs 1 Curie = 3.7 x 1010'dps i

1 gallon = 3.78 liters .1 tg = 2.21.lbs 1 ft3 = 7.48 gallons. I hp = 2.54 x 10 3. Btu /hr 3

' Density =62.4lbg/ft 1 MW = 3.41 x 106 Btu /hr Density = 1 gm/cm 1. Btu = 778 ft-lbf Heat of Vaporization = 970 Btu /lbm Degrees F = (1.8 x Degrees C) + 32.

Heat of Fusion = 144 Btu /lbm 1 inch = 2.54 centimeters 1 Atm = 14.7 psia =. 29.9 in. Hg g = 32.174 ft-lbm/lbf-sec 2-

, 1 ft H 2O = 0.4335.lbf/in.2 t

, v.

4 y- ,.---,.,--mm -y.-,, .,-,--,.nr-w.-..--eweeww --

e

• e -e i o w w -* + - -t--'~+1 e w e *-v 9 * *"w- "